mercoledì 27 novembre 2013

Lettera aperta all'Ordine dei Giornalisti ed ai Giornalisti figli di un dio minore


Lettera aperta all'Ordine dei Giornalisti
ed ai Giornalisti figli di un dio minore

Egregio Presidente del Consiglio dell'Ordine Nazionale dei Giornalisti, Enzo Iacopino,
Egregio Presidente del Consiglio dell'Ordine Regionale dei Giornalisti, Mimmo Sammartino,
Stimati Colleghi,
la sentenza di primo grado che ci condanna alla reclusione ed all'ammenda per diffamazione a mezzo stampa costituisce, a nostro avviso, una questione molto seria da affrontare e giudicare. Non si tratta di “tirare dentro” nostre vicende personali (tali sono le responsabilità penali e tali devono restare) i colleghi e l'Ordine dei Giornalisti.
Ci siamo sempre difesi da soli, con l'aiuto determinante dell'impareggiabile avvocato Leonardo Pinto (che ci difende gratis, diversamente avremmo dovuto ricorrere all'avvocato d'ufficio poiché, come ben sapete, di giornalismo difficilmente si vive e noi stessi abbiamo scelto di scrivere senza alcun compenso) e di una pattuglia di altri professionisti che, come lui, ha scelto di sostenere un ideale e non solo degli imputati. Cito fra tutti l'avv. Alessandro Sisto e l'avv. Franco Iuele. Da soli continueremo a difenderci, negli oltre cento procedimenti penali che ci hanno visto indagati, imputati, persone offese o parti civili. Solo la settimana prossima, abbiamo tre udienze a Catanzaro il 4, 5 e 6 dicembre ed una a Roma, il 7 dicembre. E pensare che non possiamo portare in parlamento nessuno dei nostri avvocati, sic!
Non vi chiediamo, quindi, di difenderci e non riteniamo nemmeno opportuno che assumiate posizioni di sostegno di categoria. Non si può e non si deve generalizzare mai, nessuno è innocente o colpevole a priori, perché è “dei nostri” o “dei loro”.
Meno ancora, gradiremmo che si manifestassero posizioni “garantiste”, del tipo “è solo il primo grado, quindi sono innocenti sino alla sentenza passata in giudicato”. La diffamazione è un reato grave e, per un giornalista, il solo fatto che si possa sussurrare “è un diffamatore”, come fanno da qualche giorno alcuni personaggi da sempre proni ad ogni sopruso, costituisce tortura insopportabile.
Invece, vorremmo che si approfondisse la vicenda specifica nel merito. Gradiremmo un vero e proprio giudizio sul nostro operato di giornalisti: che l'Ordine valutasse l'articolo per cui siamo stati condannati, pronunciandosi nel merito giornalistico e deontologico.
Se abbiamo commesso diffamazione meritiamo un provvedimento disciplinare e non in sede di “verità giudiziaria” che, come si sa, non sempre coincide con la verità storica e fattuale e, quasi sempre, nemmeno arriva in tempo utile per essere pronunciata.
Noi abbiamo bisogno di comprendere se davvero abbiamo diffamato le persone di cui scrivemmo, perché, se lo avessimo fatto, dovremmo rispondere prima di tutto alla nostra coscienza, poi all'Ordine dei Giornalisti e, infine, alle Leggi dello Stato.
Diversamente, però, se nell'articolo non vi è diffamazione, se la frase che il PM ha posto a fondamento della richiesta di condanna, non esiste e nemmeno è sottintesa (sì, vi sembra assurdo ma è così e non abbiamo difficoltà a documentarlo perché i giornali sono stampati con inchiostro indelebile!). Allora vuol dire che in gioco non è la posizione processuale di due giornalisti figli di un dio minore (quello dei Pubblicisti), ma la libertà garantita dall'articolo 21 della Costituzione.
In questo caso, ed è proprio questo il caso, cari colleghi e cari rappresentanti dell'Ordine dei Giornalisti, quella che siamo tutti tenuti a combattere è una battaglia epocale che va ben oltre la difesa della libertà di stampa ma è collocata alla estrema difesa della democrazia stessa.
Per questo motivo, chiediamo di poter riferire con la massima urgenza al Consiglio dell'Ordine Regionale e Nazionale i dati specifici del Processo che ci ha visti condannare, affinché si proceda all'assunzione delle responsabilità e dei provvedimenti conseguenti: sia che si tratti di provvedimenti a nostro carico, sia che si tratti delle doverose azioni di tutela della democrazia in Italia.
Nino Grilli e Nicola Piccenna

venerdì 15 novembre 2013

Filippo Bubbico: prendeva il pizzo più alto del mondo, l'hanno fatto viceministro dell'Interno

25/6/1994 Bubbico Filippo, presidente del Consorzio Seta Italia (dove la Camera di Commercio lo chiamava Bibbibo!) e del Consorzio Seta Basilicata, intascava il 75% dei compensi del Dr. X, agronomo. Erano soldi finanziati dalla Comunità Europea per il progetto di gelsibachicoltura che i Consorzi avrebbero realizzato in Italia (mai visto un filo di seta, miliardi di lire buttati).
Un giornale pubblicò le ricevute della vergogna il 26/11/2006 e tutti hanno potuto leggere e conoscere.
 
 
Filippo Bubbico ha fatto carriera, è stata premiata la sua abilità nell'incassare un "pizzo" record.
Nemmeno la 'ndrangheta riesce a farsi pagare il 75% di un appalto!
Filippo Bubbico  oggi è Viceministro agli Interni
 
Riproponiamo le ricevute della vergogna, affinché nessuno possa dire di non sapere, nemmeno quei magistrati che lessero l'articolo pubblicato in data 26/11/2006
 







 
 

martedì 12 novembre 2013

Lucania: triste, indolente y final

Lucania: triste, indolente y final

Verrebbe da abbandonarla questa Lucania. Questi Lucani pigri, indolenti, accattoni, infingardi, schiavi di un qualche privilegio, foss'anche quello di anticipare una visita specialistica di quindici giorni e, per quello, pronti a sostenere una classe politica inetta, pigra, indolente, infingarda, schiava del potere che le consente un accesso al denaro facile, senza lavoro, senza controllo, senza ritegno.

Verrebbe da lasciare questa terra scuotendo la polvere dai sandali, se non avessimo qui affetti, storia e cuore. La retorica della gente umile, lavoratrice e onesta, continua stancamente a mulinare nell'aria immobile dei contadini che seminavano a novembre per raccogliere a giugno ed hanno sostituito questa (già lenta) pratica con il set-a-side: non seminano nemmeno e, a giugno, raccolgono l'elemosina dell'Europa che li trasforma da contadini pigri in accattoni nullafacenti. I borghesi, poi, quelli che qualche lume d'intelletto dovrebbero averlo, si guardano bene dall'azzardare la formulazione d'idee o progetti, una qualche attività illuminata e costruttiva del domani. Emergono solo i più pigri, i più indolenti, i più infingardi, i più inetti; coloro che aggiungono a tali eccellenze la destrezza e la furbizia. Questi sono la classe dirigente, coloro che dirigono la nave Lucana negli sprofondi di ogni classifica di merito e sulla cresta di tutte le elencazioni di disvalore sociale, economico e culturale.

É accertato che il vertice della Procura Generale presso la Corte d'Appello di Potenza, Vincenzo Tufano, Modestino Roca, Gaetano Bonomi (ed una imprecisabile schiera di funzionari apicali delle forze dell'ordine) meritano una richiesta di rinvio a giudizio per associazione per delinquere finalizzata alla costituzione di una setta segreta.

É acclarato che ad un Sostituto Procuratore presso la Procura della Repubblica di Matera, Annunziata Cazzetta, viene consentito per anni di occuparsi (quasi in regime di monopolio) di decine di procedimenti penali a carico di un giornalista dopo averlo ripetutamente querelato. Salvo poi condannare la sventurata alla censura mitigata dall'attenuante (falsa) di aver tenuto un solo procedimento! E Annunziata Cazzetta continua a restare a Matera, monito per quanti volessero illudersi che ai magistrati infedeli possa capitare un qualche dispiacere.

É proprio vero che un politico di alto profilo, Bubbico Filippo (architetto), intascava il 75% delle progettazioni finanziate da fondi europei per la realizzazione di impianti di gelsibachicoltura (anche questi finanziati per miliardi di lire) presso le aziende agricole di suo padre, di suo suocero e di altri cittadini italiani che non hanno prodotto un solo filo di seta.

É emerso dagli atti giudiziari che il Procuratore Capo presso la Procura della Repubblica di Matera, Celestina Gravina, ha nascosto una querela per calunnia contro Emilio Nicola Buccico infilandola in un procedimento penale che era già avviato all'archiviazione (ben prima che la querela fosse depositata) e, non contenta, ha trasmesso l'opposizione a quella stessa archiviazione solo dopo che il giudice aveva assunto la decisione di archiviare. Un episodio di favoreggiamento spudorato e gravissimo.

Sono agli atti del Consiglio dei Ministri del 13 novembre 2003, le dichiarazioni virgolettate del Ministro Altero Matteoli e del Sottosegretario Gianni Letta, in cui Filippo Bubbico viene indicato come consenziente (ma non entusiasta) della scelta di ubicare a Scanzano il Deposito Unico delle scorie nucleari Italiane. Bubbico, in difesa del suo onore violato, ha querelato il Ministro Carlo Amedeo Giovanardi ed i giornalisti che quel verbale resero pubblico. Come se avesse un onorabilità da difendere uno che strozza l'agronomo pretendendo il “ritorno” del 75% del suo fatturato! Ma in Italia, questi signori li fanno vice-ministri agli interni, così che possano industriarsi con dovizia di mezzi nell'affinamento delle qualità già mostrate.

É sotto gli occhi di tutti il villaggio “Marinagri”, finanziato con decine di milioni di euro di soldi pubblici e costruito nell'alveo di piena del fiume Agri. Occorre aggiungere spiegazioni o commenti?

É stato reso formalmente noto alle Procure della Repubblica di Potenza e Napoli, che dal più grande giacimento petrolifero dell'Europa continentale, sotto il suolo Lucano, insieme col petrolio, vengono estratte decine di migliaia di tonnellate di “gas-stream” di cui non si conosce il destino. Ma nella terra dei misteri, non esiste nemmeno il contatore che misura le quantità di greggio estratto, queste domande non meritano nemmeno una risposta.

Migliaia di giovani hanno scelto di abbandonare questa terra ed i genitori coscienziosi è questo che consigliano ai propri figli: partire per non tornare.

Allora bisogna chiedersi se vale ancora la pena di restare, se vale la pena di scrivere quando tanti tacciono o, peggio, prestano la penna al valvassore di turno che gli lancerà un tozzo di pane per sopravvivere (e nemmeno tanto grande!).

Certo, ad evitare entusiasmi da “scampato pericolo” di siffatti cialtroni, siamo consapevoli che non potremo lasciare la Lucania, abbandonare il campo, prima di aver terminato quanto abbiamo iniziato. Poiché siamo cavalieri d'antico e indomito cuore e non ci ritireremo se non quando i nostri figli potranno tornare e ricostruire dove loro hanno depredato, distrutto e cancellato. Ma non ci vorrà ancora molto.
Nicola Piccenna, alias Filippo de Lubac

mercoledì 28 agosto 2013

Uguaglianza dei cittadini Italiani davanti alla Legge: convocazione straordinaria sez. Disciplinare CSM del 5/9/2013


In Italia vi sono alcuni cittadini di serie “B” che hanno bisogno di leggi speciali, contorcimenti giuridici, protezioni parlamentari e via dicendo. Poi c'è la maggioranza, la serie “A”, che tra mille difficoltà e sacrifici tira a campare, onestamente e senza nemmeno lamentarsi. Occorre battersi in favore dei cittadini di serie “B”, affinché anche per loro la giustizia sia fruibile senza dover ricorrere a sovvertimenti costituzionali. Non è giusto discriminarli e costringerli ad una vita piena di ingiustizie ad personam.
Il caso del povero Berlusconi, che vede Annibale Marini costretto agli straordinari e del privilegiato Nicola Piccenna che invece è trattato nei modi e nei tempi consueti.
Viva l'Italia che si salverà grazie ai cittadini di serie “A”.
 
(questo il testo dell'esposto inviato in vista della convocazione della sezione disciplinare del CSM)
Consiglio Superiore della Magistratura c.a. Presidente On. Giorgio Napolitano,
Suprema Corte di Cassazione c.a. Procuratore Generale Dr. Gianfranco Ciani,
Ministero della Giustizia c.a. Ministro Prof.ssa Annamaria Cancellieri,
Procura della Repubblica presso il Tribunale di Catanzaro c.a. Procuratore Capo Dr. Vincenzo Antonio Lombardo,

Ecc.mi signori,
in data 21 giugno 2013, la sezione disciplinare del CSM ha pronunciato la sentenza n. 86 a carico della D.ssa Annunziata Cazzetta, sostituto procuratore presso il Tribunale di Matera, infliggendole la sanzione disciplinare della censura.
Secondo gli accertamenti effettuati, Cazzetta, pur avendo più volte querelato l'odierno scrivente manteneva la titolarità del procedimento penale 1184/07 RGNR Matera.
Non rileva il relatore e omette di rilevarlo la Procura Generale, che paradossalmente chiedeva l'assoluzione di Cazzetta per insussistenza dell'illecito (!), che l'incolpata era titolare di circa venti procedimenti penali in cui l'odierno istante era parte: indagato oppure parte offesa.
Omette, la Procura Generale, di istruire il procedimento adeguatamente, inserendo cioè le numerose querele di cui aveva formalmente notizia e che riguardavano gli abusi reiterati di Cazzetta che sin dal marzo 2007 (data della sua prima querela contro il sottoscritto) operava illegittimamente aprendo e mantenendo la titolarità decine di procedimenti penali e arrivando a disporre sette mesi di intercettazioni telefoniche ininterrotte ed un provvedimento di perquisizione e sequestro domiciliare.
Omettono il relatore nel procedimento disciplinare, Prof. Annibale Marini, ed il sostituto procuratore generale Dr. Antonio Gialanella, di leggere la ponderosa documentazione inviata dallo scrivente in cui erano dettagliati i comportamenti illeciti e delittuosi della D.ssa Annunziata Cazzetta reiterati nel corso degli anni e resi possibili dall'inerzia della Procura di Catanzaro e della stessa Procura Generale presso la Corte di Cassazione che di questi delitti, non avendoli impediti, sono correi.
Omettono, Marini e Gialanella, di rilevare che nel procedimento 1184/07, durante l'udienza del 28/11/2008, davanti al GUP D.ssa Rosa Bia, invitata formalmente ad astenersi perché nello status di grave inimicizia verso l'imputato, qui ricorrente, mentiva spudoratamente dichiarando di “non aver mai querelato il Piccenna” (del fatto sono state fornite persino le registrazioni audio dell'udienza).
Ebbene, ecc.mi signori dei piani alti,
CHIEDO
che nell'udienza straordinaria del 5 settembre 2013, convocata da Annibale Marini insieme con la Procura Generale presso la Corte di Cassazione, si prenda atto che vi è un'ingiustizia ben più urgente e grave delle esternazioni del giudice di Cassazione Antonio Esposito sulla sentenza a carico di Silvio Berlusconi. Circostanza che merita l'attenzione disciplinare degli organi preposti alla sorveglianza del corretto uso dei poteri conferiti ai magistrati e l'intervento cautelare della magistratura penale per evitare il reiterarsi del reato da parte di Annunziata Cazzetta.
Con tutto il rispetto per il Prof. Annibale Marini, per il Dr. Antonio Gialanella e quanti si stracciano le vesti per una pseudo intervista che non si capisce in cosa abbia leso i diritti o la sensibilità del cittadino Silvio Berlusconi, condannato definitivamente per frode fiscale ai danni del Paese di cui era Presidente del Consiglio, il sottoscritto ritiene sia più urgente e doveroso intervenire per impedire ad Annunziata Cazzetta di amministrare la giustizia inquirente dopo aver abusato per oltre sette anni della sua funzione in danno del cittadini Nicola Piccenna, incensurato.
Si allegano nuovamente i documenti che il solerte (per Esposito) Marini e il prostrato (alla casta dei magistrati) Gialanella hanno omesso di valutare, confidando che qualcuno si sforzi di mantenere all'Italia una sufficiente dignità istituzionale.
Viva l'Italia
Nicola Piccenna

 
p.s. Si allegano:
 
1. sentenza sez. Disciplinare CSM N.86 del 21/6/2013
 
2. Querele relative al mendacio di Cazzetta nell'udienza del 28/11/2008 (proc. Pen. 1184/07 RGNR Matera, oggetto della censura inflitta alla Cazzetta): 9/12/2008; 16/12/2008; 18/1/2010; 17/11/2010; 26/11/2010; 11/8/2011;20/10/2011;
 
3. Compendio delle querele e delle istanze inoltrate tra il 2007 ed il 2013 per gli abusi reiterati e tollerati di Annunziata Cazzetta. (840 pagine distinte in tre file in formato pdf)




domenica 18 agosto 2013

Materiali radioattivi del Centro Itrec di Rotondella: Lettera aperta al Presidente degli Stati Uniti d'America, Mr. Barack Obama

Cari concittadini Lucani, Italiani e cittadini degli Stati Uniti d'America,
una esigenza di trasparenza su vicende delicatissime, ci ha indotto a proporre l'invio della seguente lettera aperta al Presidente degli Stati Uniti d'America Mr. Barack Obama.
Riteniamo fondamentale un suo intervento per conoscere dove è stato portato il carico di materiale radioattivo prelevato il 29 luglio scorso dal centro Itrec-Enea di Rotondella (Mt - Italia) e consegnato agli Stati Uniti d'America, secondo la dichiarazione resa da autorevoli fonti governative Italiane.
Interloquire col Presidente degli Stati Uniti d'America è semplice, basta compilare un "form" sul sito ufficiale della Casa Bianca. Chi vorrà potrà, quindi, chiedere direttamente al Presidente Obama le notizie che, sino ad oggi, nessun ha potuto o voluto fornire.
Di seguito, riportiamo un testo sintetico sulla vicenda, ovviamente, nulla vieta di modificarlo o sostituirlo con altro di gradimento di ciascuno scrivente.
Confidiamo che siano numerosi i Lucani (ma anche gli Italiani e gli Americani) che vorranno personalmente ricercare chiarimenti e spiegazioni su una vicenda, ad oggi, piena di punti oscuri.
di Nicola Piccenna
 
 
inizio testo lettera aperta al Presidente Barack Obama
 
Illustrious President of the United States of America,
On 29 July 2013, a shipment of nuclear material has left Italy and was delivered to the United States of America.
The news is confirmed by influential members of the Italian Government and in particular by the deputy minister of the Interior, Filippo Bubbico.
We are journalists and we have documented the transport and write in a local weekly magazine: "L'Independente Lucano" but, for the occasion, we published an article in a newspaper prestigious nationally: the weekly magazine "Oggi" of the publishing group RCS (the first publishing group of newspapers in Italy).
Many people in Italy is concerned to know what happened to the radioactive material taken from Basilicata and destined for the United States.
We think it is also in the interest of the people of the United States know where this material and what precautions have been taken to its transport into U.S. territory.
Confident in the liberal and democratic tradition of the United States, we then interviewed your agency "U.S. NRC" through a series of emails and receives answer in which they confirm “no records” related to nuclear material shipment from Italy.
U.S. NRC suggest to ask at NNSA and today we sent email to them..
Always known the respect and protection that the United States of America reserves the free press, we can not explain why U.S. Agencies had difficults to answer at our simple questions..
For this reason, Distinguished Mr. President Obama, we ask for your intervention so that we can know when, where and by what security guarantees, the radioactive cargo from Italy has passed in the United States.
Below the emails that until today are devoid of response.
Respectful Regards
 
fine testo lettera aperta al Presidente Barack Obama
 
 
 
p.s. per inviare la lettera al Presidente degli Stati Uniti d'America, Mr. Barack Obama, seguire nell'ordine:
 
1) selezionare e copiare (ctrl c) il testo di questo "post" sino alle parole "Respectful Regards" comprese;
 
 
3) compilare tutti i campi del "form" contrassegnati da asterisco;
 
4) cliccare nello spazio dove è previsto l'inserimento del messaggio;
 
5) "incollare" il testo della lettera (ctrl V);
 
6) cliccare sul pulsante "submit"
 
un messaggio inviato al vostro indirizzo email, confermerà l'inoltro della lettera al Presidente Obama.
 

Domande chiare sul trasporto di materiale radioattivo del 29 luglio scorso dall'Itrec di Rotondella

Egregi Signori e Autorità deputate alla gestione dei materiali radioattivi,
il giorno 29 luglio 2013 un carico di materiale radioattivo ha transitato per le strade italiane.
Partito dal centro Itrec situato nel comune di Rotondella (Provincia di Matera, Italy), esso è stato scortato sino all'aeroporto militare di Gioia del Colle (provincia di Bari, Italy).
Le autorità italiane hanno dichiarato che si è trattato di circa 1,2 Kg di Biossido di Uranio arricchito al 91%. Essi hanno specificato che la destinazione finale del carico erano gli Stati Uniti d'America e che il materiale è stato consegnato alla destinazione finale.
Il trasporto, dicono le autorità italiane era coperto da segreto di Stato. Per questo motivo, le autorità locali e la popolazione non sono stati avvisati.
Noi abbiamo verificato attentamente senza riuscire ad avere conferme. In particolare, noi vorremmo sapere in quale sito/città si trova oggi il materiale trasportato e quale percorso ha seguito all'interno degli Stati Uniti.
Poniamo queste domande poiché sappiamo che il biossido di uranio è un materiale debolmente radioattivo e abbiamo verificato che le agenzie statunitensi: U.S. NRC (United States Nuclear Regulatory Commission), per il settore "civile", e NNSA (National Nuclear Security Administration) appartenente al DOE (Department Of Energy), per il settore "militare"; non hanno mai avuto difficoltà a fornire informazioni su questo genere di spedizioni.
In effetti, sui siti pubblici delle due agenzie, ci sono molte comunicazioni che riguardano programmi, spedizioni e gestione di uranio arricchito.
In verità, U.S. NRC ha già risposto alla nostra domanda, confermando che a loro non risulta alcun trasporto di materiale nucleare dall'Italia.
Noi non possiamo dubitare delle dichiarazioni del Governo Italiano, perciò sarà NNSA a conoscere i dettagli di questo trasporto insieme con EURATOM (EURATOM SUPPLY AGENCY) e IAEA (International Atomic Energy Agency) che sono gli organismi internazionali che sorvegliano e disciplinano la gestione ed il trasporto dei materiali radioattivi strategici quale è l'uranio 235 arricchito.
Noi riteniamo che i cittadini Italiani e Statunitensi abbiano diritto di conoscere esattamente cosa è avvenuto nel trasporto del 29 luglio scorso e riteniamo che le autorità competenti non abbiano nessuna difficoltà a fornire tutti gli elementi necessari per tranquillizzare le popolazioni interessate.
Noi confidiamo che le autorità interpellate, vorranno rispondere celermente nonostante il periodo di vacanze, poiché trattasi di questioni estremamente delicate per la salute e la sicurezza dei cittadini.
Noi apprezziamo molto la politica di trasparenza sempre confermata dagli Stati Uniti d'America e la grande apertura sempre mostrata verso la stampa libera da quel grande Paese.
di Nicola Piccenna e Ivano Farina, giornalisti

Indirizzi utili:
U.S. NRC (United States Nuclear Regulatory Commission)
Addresses: U.S. Nuclear Regulatory Commission - Washington, DC 20555-0001 – Tel: (+1) 1-800-368-5642
Email: NRCExecSec@nrc.gov
Website: http://www.nrc.gov
english version:
Dear Sirs and authority delegated to the management of radioactive materials,
the day July 29, 2013 a shipment of radioactive material has passed through the Italian streets.
Party Itrec center located in the town of Rotondella (Province of Matera, Italy), it was escorted up to the military airport of Gioia del Colle (province of Bari, Italy).
The Italian authorities have stated that it was about 1.2 kg of uranium dioxide enriched to 91%. They specified that the final destination of the cargo was the United States of America and that the material was delivered to the final destination.
The transport, Italian authorities say was covered by state secrecy. For this reason, local authorities and the population were not warned.
We have carefully checked without being able to have confirmation. In particular, we would like to know which site / city is today the transported material and which path followed within the United States.
Ask these questions because we know that the uranium dioxide is a weakly radioactive material and we have verified that the U.S. agencies: U.S. NRC (United States Nuclear Regulatory Commission), for the "civil" and NNSA (National Nuclear Security Administration) belonging to the DOE (Department of Energy), for the "military" have never had difficulty in providing information on these kinds of shipments.
In fact, on public sites of the two agencies, there are many programs that affect communication, shipping and handling of enriched uranium.
In truth, U.S. NRC has already replied to our question, confirming that they do not is no transport of nuclear material from Italy.
We can not doubt the statements of the Italian Government, therefore, will NNSA to know the details of this transport together with EURATOM (EURATOM SUPPLY AGENCY) and IAEA (International Atomic Energy Agency) which are international bodies that oversee and govern the management and transport of radioactive material which is strategic enriched uranium 235.
We believe that the Italian and U.S. citizens have a right to know exactly what happened in the transport of 29 July last year and we believe that the competent authorities have no difficulty in providing all the elements necessary to reassure the populations concerned.
We are confident that as the authorities will want to respond quickly despite the holiday period, since these are highly sensitive issues for the health and safety of citizens.
We greatly appreciate the policy of transparency always confirmed by the United States of America and the grand opening always shown towards the free press from that great country.
by Nicola Piccenna and Ivano Farina, journalists
Useful addresses:

venerdì 16 agosto 2013

Il sostituto procuratore Annunziata Cazzetta: 6 anni di abusi e resta al suo posto. Viva l'Italia

 
Bubbico viceministro dell'Interno, Cazzetta Ministro della Giustizia?

 
Vi è un pubblico ministero, D.ssa Annunziata Cazzetta da Matera che, pur avendo denunciato un cittadino italiano (marzo 2007) resta titolare di nove procedimenti a suo carico. Quel magistrato, dopo averlo denunciato, dispone che tutte le utenze telefoniche del cittadino vengano intercettate per sette mesi consecutivi. Quel magistrato, dopo averlo denunciato, dispone la perquisizione dell'abitazione del malcapitato, dell'ufficio della testata giornalistica per cui scrive e della abitazione dei suoi genitori. Quel magistrato, dopo aver depositato tre istanze querelatorie contro il cittadino italiano, invitato formalmente ad astenersi durante una udienza del Giudice per l'Udienza Preliminare, mente dichiarando di non aver mai presentato querela contro il nostro.
Vi è un cittadino Italiano che querela un magistrato Italiano che vìola le precise disposizioni del codice di procedura penale in materia di obblighi di astensione. Vi è un cittadino Italiano che informa costantemente del protrarsi degli abusi le Procure della Repubblica competenti, la Procura presso la Suprema Corte di Cassazione, il Consiglio Superiore della Magistratura, il Presidente della Repubblica nella sua veste di Presidente del CSM. Costantemente, dal 2007 ad oggi. Oltre duemila pagine di querele, esposti, atti documentati.
Vi è un membro del CSM di nomina PdL (sì perché alcuni membri del CSM sono di nomina politica in Italia), Prof. Annibale Marini, chiamato a fare il relatore nell'unico procedimento disciplinare a carico di Annunziata Cazzetta che giunge a definizione presso la sezione disciplinare. Il prof. Marini, nell'udienza del 21/6/2013 relaziona: “va condivisa la valutazione del Procuratore Generale che ha dichiarato irrilevante e quindi inammissibile la voluminosa documentazione depositata dal Piccenna il giorno stesso dell'udienza dibattimentale”. Documentare che il magistrato giudicato ha reiterato i comportamenti illeciti per anni, che vi sono prove documentali del mendacio e della ostinata negligenza e neghittosità con cui CSM, Procura della Cassazione (che in udienza chiede l'assoluzione per insussistenza dell'illecito) e almeno tre Procure della Repubblica e cinque Procure Generali, lo stesso Presidente della Repubblica, per anni hanno “coperto” illeciti evidenti e abusi gravissimi è irrilevante, così, senza un motivo, senza un cenno di spiegazione. Irrilevante, dice il Prof. Marini. E Annunziata Cazzetta, viene condannata alla “censura”. Ha abusato della sua funzione di magistrato per anni e, su un qualche foglio di carta ci sarà scritto: “Censura”. Sei anni e duemila pagine bellamente ignorate per scrivere “censura”.
Vi è un membro del CSM, di nomina PdL (sì perché alcuni membri del CSM sono di nomina politica in Italia), Prof. Annibale Marini, che dispone la convocazione di una seduta straordinaria della commissione disciplinare per il 5 settembre 2013 in relazione all’intervista del presidente della sezione feriale della Cassazione, Dr. Antonio Esposito, pubblicata su Il Mattino il 6 agosto 2013. Un cittadino Italiano è stato condannato con sentenza definitiva alla perdita di alcuni diritti civili ed alla pena detentiva. Dopo tre gradi di giudizio. Dopo aver condizionato l'attività parlamentare per produrre leggi “ad personam” che gli evitassero il processo, che eliminassero i reati, dopo aver subito altra condanna in primo grado per sfruttamento della prostituzione minorile. Quel cittadino merita l'urgenza di un procedimento disciplinare a carico di uno dei magistrati componente del collegio che l'ha condannato. L'altro cittadino, dice il Prof. Marini (di nomina PdL), non merita nemmeno una motivazione per escludere mille pagine di atti probatori a carico di Annunziata Cazzetta.
Vi è un'Italia che è più forte delle pastoie in cui una classe dirigente corrotta dentro vuole tenerla.
Vi sono cittadini Italiani che continuano ad utilizzare i mezzi e le leggi vigenti anche quanto il Prof. Marini ed i suoi degni compari fanno di tutto per scoraggiarli, dissuaderli, vincerli.
Vi sono cittadini Italiani che salveranno l'Italia, checché ne pensino Marini e giù a scendere sino a Bubbico, viceministro agli Interni che incassava il 75% del fatturato di un onesto agronomo e non è mai stato indagato per questo. Anzi, oggi è viceministro dell'Interno.
 
Bubbico viceministro dell'Interno, Cazzetta Ministro della Giustizia?

di seguito, la sentenza di censura per Annunziata Cazzetta
 




 

venerdì 9 agosto 2013

Non siamo Stato noi: le domande sull'Itrec di Rotondella a cui lo Stato deve rispondere

Non siamo Stato noi: le domande sull'Itrec di Rotondella a cui lo Stato deve rispondere (mentre Bubbico non può farlo)
 
Nel centro Enea-Itrec di Rotondella (Mt) vi sono materiali radioattivi. Dichiaratamente: Uranio 235 arricchito oltre il 90%, Torio, tracce infinitesimali di Plutonio. Poi, vi è del combustibile irraggiato, molto radioattivo (emette i penetranti raggi gamma), 64 "barre" e una parte di combustibile irraggiato già lavorato secondo le finalità originarie (dichiarate) dell'Itrec (acronimo di Impianto di Trattamento e Rifabbricazione Elementi di Combustibile). In pratica, si sarebbero dovute estrarre, dal combustibile nucleare "esausto", le sostanze radioattive ancora utilizzabili per (ri)fabbricare nuovi elementi di combustibile nucleare da utilizzare nelle centrali attive. Il materiale giunse a Rotondella proveniente dalla centrale nucleare di Elk River negli Stati Uniti d'America che operava con una miscela di Uranio e Torio, tali erano le barre che Itrec avrebbe (ri)costruito.
Peccato che quanto entrò in funzione tutta questa giostra, la tecnologia Uranio-Torio era già stata abbandonata in tutto il mondo e la centrale di Elk River era stata chiusa e, da lì a poco, sarebbe stata smantellata e l'area su cui sorgeva completamente bonificata. A vederla oggi su internet, sembra il paradiso terrestre.
Comunque, Itrec lavorò, partendo dalle prime 20 barre: sminuzzate, sciolte in acido nitrico ad altissima concentrazione, centrifugate. Ottenendo quelle che ci spiegano come le tre fasi liquide: residui ad altissima (radio)attività; residui a bassa attività; miscela uranio-torio da separare per ottenere i prodotti "pregiati" da destinare al nuovo combustibile.
Ma, come si era detto, nessuno più voleva combustibile uranio-torio e le cose rimasero ferme. Ferme si fa per dire, poiché quando si ha a che fare con i materiali radioattivi, il concetto di fermo non esiste. E, da quarant'anni, i tecnici dell'Enea, ma anche dell'Itrec, ma anche della Sogin SpA (società controllata dal Tesoro) badano che tutto sia tenuto sotto controllo, raffreddato, conservato in modo da evitare o quantomeno ridurre i rischi per l'ambiente ed i danni per la popolazione.
Evitare non è stato possibile, qualcosa è uscito, qualcosa è stato reso noto, qualcosa è trapelato. Insomma gli incidenti non sono mancati. A giocare col fuoco, prima o poi ci si scotta.
Dai dati ufficiali diffusi dalla Sogin, emerge che altri 18,15 Kg di uranio 235 arricchito al 91% si trovavano all'interno dell'Itrec al 31.12.2012. Certamente giunti dagli Stati Uniti negli anni 70, così conferma l'attuale responsabile del centro Itrec, Dr. Edoardo Petagna.
Il 29 luglio scorso, un trasporto supersegreto, ha trasferito poco più di un chilogrammo di uranio 235 arricchito al 91% negli Stati Uniti passando per l'Aeroporto Militare di Gioia del Colle (Ba), lo afferma un comunicato della Sogin S.p.A. Più generiche le dichiarazioni rese dalla politica e dal Governo, ma la sostanza, la verità ufficiale è questa. Il viceministro all'Interno, Filippo Bubbico, ha dichiarato che si tratta di uno dei viaggi concordati con gli Stati Uniti d'America e che porteranno al progressivo svuotamento del centro Itrec con contestuale restituzione ai legittimi proprietari di quelle sostanze pericolose: 64 barre residue; miscela uranio-torio in acido nitrico; uranio 235 arricchito e (si spera) anche le centinaia di "barili" di materiali inquinati da radioattività "prodotti" dalla gestione di quell'ospite sgradito e scomodo giunto da Elk River.
Nessuno ha prodotto evidenza di quell'accordo, mentre tutti i documenti accessibili comprese delicate corrispondenze tra il Governo Italiano e l'ambasciatore Statunitense Dr. Spogli, affermano il contrario.
Sorge subito una domanda: Perché si parte restituendo 1 Kg di Uranio 235, il cui valore commerciale supera abbondantemente i tre milioni di euro e non dalle 64 barre + 20 disciolte che ci costa mantenere al prezzo venale di diversi milioni di euro all'anno, per non parlare dei rischi per la salute di intere comunità?
Bubbico ha affermato che esiste un programma concordato di restituzioni, aggiungendo che la notizia del trasporto (avvenuto con dispiegamento di uomini e mezzi stile "Apocalipse Now") avrebbe turbato gli americani al punto da rendere probabile una loro rinuncia agli impegni di reimportazione del materiale radioattivo. Ebbene, dalle informazioni acquisite durante una visita presso il Centro Itrec, si apprende che mancano i contenitori per "stivare" le 64 barre residue e non è stata effettuata la cementificazione delle 20 barre "lavorate parzialmente". Queste due operazioni sono preliminari e propedeutiche alla stipula di accordi di sorta, giacché, le norme internazionali che regolano questi delicati trasporti, prevedono che questi possano essere assunti solo quando il "materiale" è impacchettato e pronto per il trasporto. Bubbico, vorrà precisare qualcosa? Diversamente, sembrerebbe che Bubbico abbia voluto scaricare sui giornalisti il fallimento di un accordo mai stipulato, seminando sentimenti di odio e disinformazione forieri di pericolo anche per l'incolumità personale. Stai a vedere che dopo quarant'anni, la politica negligente e neghittosa (ad esser buoni) attribuisce la mancata “partenza” del materiale radioattivo a di due sfortunati giornalisti che vedono partire su strada pubblica (ed arrivare da strada pubblica) un corteo “nucleare” ed informano i cittadini, cioè fanno il loro mestiere/dovere.
Ora, gli Stati Uniti di combustibile esausto ed irraggiato, ne custodiscono centinaia se non migliaia di barre. Sono dotati di un sito unico nazionale di stoccaggio e numerosi siti di passaggio e/o stoccaggio intermedi, cosa costerebbe aggiungere le 64 barre dell'Itrec di Rotondella?
E, dulcis in fundo, la domanda che non spiega ma lascia capire molto: è notizia ufficiale che nell'Itrec di Rotondella vi sono (erano) 18,15 Kg di Uranio 235 e tutto il resto, perché il trasporto ed i documenti di questa ed altre vicende legate all'Itrec sono coperti da “segreto di Stao”? Perché notizie sensibili, perché se un terrorista... Non è così, da Saluggia (Vc), il 12 marzo scorso, è partito un intero treno di barre di combustibile irraggiato. Si conosceva con molto anticipo cosa sarebbe avvenuto, quello che si trasportava, quale percorso avrebbe seguito e la destinazione finale (Francia). Lì non c'era segreto di Stato, non c'erano terroristi appostati, non c'era allarme per la popolazione. Lì c'era uno Stato trasparente che rispetta i suoi cittadini e non ha nulla da nascondere. Qui al Sud, lo Stato è rappresentato da un tale Bubbico e suoi degni compari, espressione di abitanti sudditi abituati a pietire col cappello in mano anche i diritti fondamentali e inalienabili garantiti dalla Costituzione Repubblicana.
di Nicola Piccenna

giovedì 8 agosto 2013

Quanto importa ai Governi ed ai politici del materiale radioattivo nell'Itrec di Rotondella?

Tra i documenti resi noti da Wikileaks, qualcuno riguarda i rifiuti nucleari custoditi presso il centro Enea-Itrec di Rotondella (Mt). Contiene informazioni interessanti da conoscere per comprendere fatti e accadimenti del passato recente e del futuro prossimo. Il presente, invece, è fatto dei silenzi ostentati dell'intera classe politica nazionale e regionale. Quanto importi a Codesti signori della gente comune, lo leggiamo nella corrispondenza che riportiamo di seguito. Il problema della barre di combustibile irraggiato e del materiale radioattivo tenuto presso l'Itrec è affrontato solo dal punto di vista delle convenienze elettorali. A nessuno importa della salute dei cittadini e dei luoghi che ospitano questi materiali non si sa perché anche se si intuisce per chi.


Disposition Of Elk River Spent Nuclear Fuel: Letter From U/s Of The Council Of Ministers Gianni Letta To Ambassador

Mon, 13 Feb 2006 17:09 UTC

C O N F I D E N T I A L SECTION 01 OF 02 ROME 000439
DEPT. FOR ISN/NESS (WHAMMACK, EUR/WE (FETCHKO), AND EUR/PRA
(JCONLON), DOE FOR NNSA/NA-212 (TANNO)
E.O. 12958: DECL: 02/13/2015

SUBJECT: DISPOSITION OF ELK RIVER SPENT NUCLEAR FUEL:
LETTER FROM U/S OF THE COUNCIL OF MINISTERS GIANNI LETTA TO AMBASSADOR
Classificato da: Ambasciatore Ronald P. Spogli, motivi 1.4 (B) e (D)
1. (C) SINTESI E AZIONE RICHIESTA: Il 6 febbraio, l'ambasciatore ha ricevuto una lettera da Gianni Letta, Sottosegretario al Consiglio dei ministri chiedendo assistenza per il rientro negli Stati Uniti di 64 barre di di combustibile nucleare esaurito (SNF) torio-uranio, proveniente da Elk River MN (e custodito presso il centro Enea-Itrec di Rotondella – Mt, ndr). Letta ha chiesto una risposta entro la fine di febbraio. Il sottosegretario ha fatto la stessa richiesta direttamente al Dipartimento dell'Energia (DOE) nel mese di agosto 2004 (reftel), ma nel gennaio 2005, l'ex Segretario di Energia Abraham ha risposto che, dopo la revisione tecnica e giuridica approfondita, il DOE non poteva accettare il combustibile nucleare esaurito nel quadro dei programmi esistenti. La ragione per cui il governo italiano ha ripetuto la richiesta in via d'urgenza è che la questione è politicamente delicata per la coalizione di centro-destra del primo ministro Berlusconi che sta affrontando una dura battaglia per la rielezione nel mese di aprile. Con le elezioni alle porte, Letta scrive, il governo italiano sarà "costretto" a trasferire il SNF di Elk River alla Russia, se non potesse essere rimpatriato negli Stati Uniti. Una traduzione non ufficiale della lettera di Letta (par. 2-5), insieme con la risposta proposta dell'ambasciatore (par. 9-11) seguono.
RICHIESTA DI AZIONE: rivedere la lettera di Letta e rispondere a firma dell'ambasciatore secondo replica suggerita nei paragrafi da 9 ad 11 di seguito riportati. Inviare risposta entro 17 febbraio.
FINE SOMMARIO E RICHIESTA DI AZIONE.
. (C) Egregio Signor Ambasciatore, Vi scrivo per richiamare la vostra attenzione su una questione su cui l'ambasciata è probabilmente già informata. Si tratta di una questione molto importante per il governo (italiano) anche dal punto di vista psicologico. Avvieremo presto il trasferimento dei nostri rifiuti nucleari, attualmente conservati in Piemonte e in Emilia Romagna, in Francia. Il combustibile esaurito resterà in Francia fino al 2025 almeno, quando l'Italia dovrebbe avere un proprio sito di stoccaggio di scorie nucleari. Questa notizia sta causando le proteste nel sud (Italia). Infatti, sessantaquattro barre di torio-uranio di combustibile esaurito sono custodite nel sito ITREC (Trisaia Research Center) a Rotondella (in provincia di Matera), e queste barre non possono essere riprocessate in Europa. Si tratta di combustibile arrivato in Italia dall'impianto di Elk River (Stati Uniti) nel 1970, nel quadro di un progetto condiviso, che fu poi abbandonato, tra CNEN (oggi ENEA - l'Ente per le Nuove tecnologie, l'Energia e l'Ambiente) e l'AEC (Atomic Energy Commission - USA) (ora DOE). Centinaia di barre simili, che contengono l'uranio-235 al 93% e addirittura fino al 95% combinato con il torio, sono attualmente conservate presso il sito americano fiume Savannah.
3. (C) Abbiamo chiesto alle autorità statunitensi (il Presidente e il DOE) di trasferire in quello stesso sito anche le poche barre conservate a Rotondella, che possono essere contenute in due soli “cask” conformi alle esigenze degli Stati Uniti per lo stoccaggio e il trasporto. Italia fornirà il finanziamento per l'operazione. Come si può vedere, si tratta di una piccola cosa. Ma se il problema pratico è poco significativo, l'impatto psicologico e quindi politico è l'opposto. Il problema è già cavalcato dalle forze di opposizione locali e nazionali, che sostengono che il governo Berlusconi sostiene il nord più che il sud.
4. (C) In data 9 novembre 2005, il presidente della Sogin (l'Agenzia per la gestione degli impianti nucleari) ha ripetuto la stessa richiesta al Dipartimento di Energia e di altri rappresentanti dell'ambasciata, sottolineando che l'Italia aveva bisogno di una risposta entro la fine di febbraio. In assenza di risposte, per ragioni di rilevante opportunità - in pratica, al fine di evitare le manifestazioni - saremmo costretti a trasferire i due “cask” in Russia per circa 50 anni. Una società di Rosatom (ex Minatom) ha già mostrato interesse per la conservazione delle barre per un prezzo decisamente modesto. Dato il rilievo di livello militare dell'uranio arricchito, presente nelle barre Itrec, l'operazione potrebbe essere inclusa nel quadro dell'iniziativa Global Threat Reduction, dal momento che, come ho detto, il combustibile esausto è venuto da un impianto statunitense. SOGIN è a disposizione per fornire tutti i dettagli. La decisione degli Stati Uniti potrebbe essere resa nota in un incontro pubblico a cui vorrei invitare le autorità regionali e locali delle realtà limitrofe e i media.
5. (C) Siamo certi del vostro interessamento in questa materia, Vi ringrazio in anticipo per conto del governo, e sottolineo ancora una volta che abbiamo bisogno di una risposta entro la fine di febbraio. Con cordiali saluti,
Gianni Letta
 
RISPOSTA PROPOSTA
9. (C / divulgabile in Italia) Caro sottosegretario Letta: Grazie per la sua lettera del 6 febbraio scorso
10. (C / divulgabile in Italia) Le posso assicurare che abbiamo capito che il governo Italiano ha necessità di premere per trovare un percorso risolutivo, e abbiamo interessato su questo tema il Dipartimento dell'Energia (DOE). L'Ufficio del Global Nuclear Material per la riduzione dei rischi nucleari del Nuclear Nazional Security Administration (NNSA), ha attentamente esaminato la sua richiesta. I funzionari DOE hanno redatto una nuova recensione al riguardo, dopo che i rappresentanti NNSA avevano incontrato il Presidente della SOGIN nello scorso novembre. Purtroppo, non vi è stato alcun cambiamento nella posizione del Department Of Energy che non può accettare il materiale di Elk River.
11. (C / divulgabile in Italia) Come sapete, gli Stati Uniti apprezzano molto la solida partnership con l'Italia in materia di cooperazione alla non-proliferazione ed allo sviluppo della scienza e della tecnologia nucleare per scopi pacifici. Abbiamo cercato di essere utili, nei limiti possibili per la difficile questione dello smaltimento dei rifiuti nucleari. In questo caso, seppure con una risposta negativa, volevo garantire di rispondere rapidamente, in modo che possiate procedere, se necessario, con i rimedi alternativi descritti nella tua lettera.
FINE TESTO PROPOSTO. SPOGLI
 
 
 

martedì 6 agosto 2013

Rischi e i lati oscuri della vicenda "nucleare" in Basilicata: un'inchiesta della Direzione Distrettuale Antimafia di Potenza

Inchiesta della Procura Distrettuale Antimafia di Potenza sulla gestione del nucleare: archiviata per mancanza di soldi!

Un approfondimento specifico, circa i rischi e i lati oscuri della vicenda "nucleare" in Basilicata, si può acquisire piegandosi al lavoro di lettura di un documento giudiziario che il blog: www.attigiudiziari.blogspot.com pubblica integralmente.
Lunedì scorso, avevamo invitato gli oltre 700 giornalisti pubblicisti lucani ed i parlamentari eletti in Basilicata a venire con noi per vedere da vicino e con i nostri occhi lo stato e l'organizzazione della struttura (Itrec-Enea di Rotondella - Mt) che controlla e gestisce i materiali radioattivi arrivati negli anni 70 dagli Stati Uniti.
C'erano cinque giornalisti, due cittadini lucani e basta.
Molti sono in ferie e molti altri sono certamente impegnati in attività non rimandaili. Poi, molti altri, semplicemente hanno preferito non venire, ed è giusto che facciano ciò che credono meglio per loro.
Alcuni, però, almeno alcuni è importante che sappiano, che approfondiscano, che conoscano quello che succede e quello che non accade.
Solo per questi, pubblichiamo integralmente gli atti conclusivi di una importante inchiesta sul "nucleare lucano", una parte delle tante inchieste sul medesimo argomento.
La proponiamo affinchè si sappia come vanno certe cose e come sono andate certe altre. Troverete tante storie di nostri concittadini che sapevano e non hanno detto, che potevano e non hanno fatto, che dovevano e hanno fatto altro.
Conosciamo almeno trenta lucani che non guardano solo alla propria ristretta sfera d'interessi, che si interessano di un vicino in difficoltà, che guardano alla cosa pubblica con la stessa attenzione con cui guardano alla propria famiglia. Ma forse ce ne sono molti di più, solo che non li abbiamo mai conosciuti o non hanno avuto mai la possibilità di conoscere e capire per potersi impegnare in un battaglia civica per il riscatto della Lucania.
Questa pubblicazione è per loro.
Molti parlano, alcuni scrivono, pochi leggono e si documentano.

Itrec di Rotondella: uranio che va, uranio che viene ed i "segreti di Stato"


L'Itrec di Rotondella, il viceministro Bubbico: e il materiale radioattivo

Ad una settimana esatta dal trasporto di materiale radioattivo del 29 luglio scorso, siamo tornati al centro Itrec-Enea di Rotondella. Un carico di derivati dell'uranio era partito nottetempo dal centro Enea di Rotondella, 2500 anime in provincia di Matera, per essere consegnato all'aeroporto militare di Gioia del Colle alle prime luci dell'alba. Questa volta ci è stato consentito di visitare il sito e, in un clima di collaborazione cordiale nel rispetto delle funzioni e delle esigenze di ciascuno, approfondire alcune questioni di sicuro interesse pubblico. Già il 29 luglio, subito dopo il trasporto, avevamo chiesto di parlare con gli “attori” principali della misteriosa vicenda. Il comandante dell'aeroporto non era in sede o, comunque, non ci ha ricevuto dopo un'attesa di quasi 3 ore. Il viceministro dell'Interno, Filippo Bubbico, lucano di Montescaglioso, non rispondeva alle domande più volte poste alla sua segreteria. Il responsabile dell'Itrec ci indirizzò al responsabile dell'ufficio stampa della Sogin S.p.A. (società a partecipazione statale che ha in carico la gestione e la bonifica di tutto quanto è nucleare in Italia), il quale rispose di non sapere nulla del trasporto.

Questa volta ci è andata meglio, il Dr. Edoardo Petagna, l'Ing. Salvatore Bruno e, alternandosi, altri dirigenti e funzionari che lavorano nel centro Enea di Rotondella hanno potuto riceverci dedicandoci una intera giornata del loro tempo. “Potuto” e non “voluto”, poiché come ci hanno spiegato, il giorno del trasporto e sino all'arrivo del carico alla destinazione finale, erano sottoposti al vincolo del segreto di Stato.

Tra l'11 ed il 15 marzo scorso, dal deposito Avogadro di Saluggia (Vc), è stato trasportato combustibile nucleare irraggiato alla volta della centrale nucleare di La Hague (Francia). Quel trasporto non era coperto da segreto di Stato. La popolazione e gli organismi di tutela e protezione civile erano allertati da giorni e sul sito dell'Agenzia di tutela dell'ambiente piemontese (Arpa Piemonte) è possibile leggere un dettagliato rapporto.

Nel caso Lucano, invece, è stato trasportato biossido di uranio arricchito al 91%, poco più di un chilo. Occupava lo spazio di un grosso barattolo di birra, inserito in un guscio protettivo tipo matrioska, intabardato con tiranti ad una base quadrata. Il tutto fissato al centro del cassone di un grosso autotreno che, dalle immagini, appare pressoché vuoto.

Il biossido di uranio può essere maneggiato anche con le mani e le sue emissioni, per intensità, non sono paragonabili a quelle del combustibile irraggiato. Perché il segreto di Stato? Perché un carico molto più pericoloso e consistente (si parla di vagoni rispetto ad un tir semivuoto) viene “esposto” alla mercé di ogni possibile catastrofica previsione (incidenti, attentati, sabotaggi) e la nostra lattina di birra è un segreto di Stato?

Viene da pensare che il “segreto” non serva per garantire sicurezza e security al trasporto ma per l'imbarazzo di dover spiegare come, quando e perché il biossido di uranio arricchito è arrivato in Italia. Domande per la cui risposta già altri hanno lavorato ed alcuni, oggi, non ci sono più.

Circa, invece, le altre domande a cui Filippo Bubbico comunque non risponde, cioè quelle relative alla sicurezza dell'impianto Itrec, la visita di ieri è stata davvero esaustiva. Una struttura in cui l'ordine, l'organizzazione e dotazioni tecniche all'avanguardia assicurano mantenimento e controlli iper-sicuri. Fra tutti gli aspetti, quello più tranquillizzante è che tutti i dirigenti vivono con le loro famiglie nella zona e, poiché non sono né folli né incoscienti, ci permettiamo di considerarli un fattore di valutazione molto più significativo delle api.

Una contaminazione ambientale non rilevata è impossibile.
di Nicola Piccenna

lunedì 5 agosto 2013

I doveri/poteri ispettivi dei cittadini: guardiamo dentro l'Itrec di Rotondella



Egr. Dr. Edoardo Petagna
Responsabile del Deposito Materiali Radioattivi
tenuti dall'ITREC presso il centro ENEA
Loc. Trisaia di Rotondella (Mt)

Egregio Direttore,

le persone oggi presenti nelle rispettive e specifiche qualità di seguito specificate e, comunque, nella preminente qualità di cittadini italiani residenti nei territori limitrofi al centro Itrec in indirizzo, non registrando alcuna risposta alle domande poste da molti giornalisti e, più recentemente, da Ivano Farina e Nicola Piccenna in diversi articoli pubblicati su testate internet, testate giornalistiche locali e testate giornalistiche nazionali,

CHIEDONO

di poter accedere con ogni precauzione e presidio di sicurezza ed accompagnati da personale tecnico a conoscenza della dislocazione, delle caratteristiche di pericolosità e delle norme di sicurezza relative ai materiali radioattivi conservati presso codesta struttura, ai siti di stoccaggio e visionare lo stato degli impianti e la cura ed il rispetto delle prescrizioni di sicurezza.

Tanto si richiede con somma urgenza, dovendo i cittadini di cui innanzi e di cui si allega elenco nominativo con firma apposta in presenza degli addetti alla vigilanza procedere immediatamente alle constatazioni di cui innanzi, costituendo la circostanza esercizio dei diritti costituzionalmente garantiti e fra tutti del diritto alla tutela della salute.

Trisaia di Rotondella, 5 agosto 2013

domenica 4 agosto 2013

TRANSPORTATION OF RADIOACTIVE MATERIAL: negli U.S.A. funziona così


PART 71—PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL


Part Index

Subpart A—General Provisions


71.0 Purpose and scope.

71.1 Communications and records.

71.2 Interpretations.

71.3 Requirement for license.

71.4 Definitions.

71.5 Transportation of licensed material.

71.6 Information collection requirements: OMB approval.

71.7 Completeness and accuracy of information.

71.8 Deliberate misconduct.

71.9 Employee Protection.

71.10 Public Inspection of application.

71.11 Protection of Safeguards Information

Subpart B—Exemptions


71.12 Specific exemptions.

71.13 Exemption of physicians.

71.14 Exemption for low-level materials.

71.15 Exemption from classification as fissile material.

71.16 [Reserved]

Subpart C—General Licenses


71.17 General license: NRC-approved package.

71.18 [Reserved]

71.19 Previously approved package.

71.21 General license: Use of foreign approved package.

71.22 General license: Fissile material.

71.23 General license: Plutonium-beryllium special form material.

71.24 [Reserved]

71.25 [Reserved]

Subpart D—Application for Package Approval


71.31 Contents of application.

71.33 Package description.

71.35 Package evaluation.

71.37 Quality assurance.

71.38 Renewal of a certificate of compliance or quality assurance program approval.

71.39 Requirement for additional information.

Subpart E—Package Approval Standards


71.41 Demonstration of compliance.

71.43 General standards for all packages.

71.45 Lifting and tie-down standards for all packages.

71.47 External radiation standards for all packages.

71.51 Additional requirements for Type B packages.

71.53 [Reserved]

71.55 General requirements for fissile material packages.

71.57 [Reserved]

71.59 Standards for arrays of fissile material packages.

71.61 Special requirements for Type B packages containing more than 105A2.

71.63 Special requirements for plutonium shipments.

71.64 Special requirements for plutonium air shipments.

71.65 Additional requirements.

Subpart F—Package, Special Form, and LSA-III Tests


71.71 Normal conditions of transport.

71.73 Hypothetical accident conditions.

71.74 Accident conditions for air transport of plutonium.

71.75 Qualification of special form radioactive material.

71.77 Qualification of LSA-III Material.

Subpart G—Operating Controls and Procedures


71.81 Applicability of operating controls and procedures.

71.83 Assumptions as to unknown properties.

71.85 Preliminary determinations.

71.87 Routine determinations.

71.88 Air transport of plutonium.

71.89 Opening instructions.

71.91 Records.

71.93 Inspection and tests.

71.95 Reports.

71.97 Advance notification of shipment of irradiated reactor fuel and nuclear waste.

71.99 Violations.

71.100 Criminal penalties.

Subpart H—Quality Assurance


71.101 Quality assurance requirements.

71.103 Quality assurance organization.

71.105 Quality assurance program.

71.107 Package design control.

71.109 Procurement document control.

71.111 Instructions, procedures, and drawings.

71.113 Document control.

71.115 Control of purchased material, equipment, and services.

71.117 Identification and control of materials, parts, and components.

71.119 Control of special processes.

71.121 Internal inspection.

71.123 Test control.

71.125 Control of measuring and test equipment.

71.127 Handling, storage, and shipping control.

71.129 Inspection, test, and operating status.

71.131 Nonconforming materials, parts, or components.

71.133 Corrective action.

71.135 Quality assurance records.

71.137 Audits.

Appendix A to Part 71—Determination of A1 and A2

Authority: Atomic Energy Act secs. 53, 57, 62, 63, 81, 161, 182, 183, 223, 234, 1701 (42 U.S.C. 2073, 2077, 2092, 2093, 2111, 2201, 2232, 2233, 2273, 2282, 2297f); Energy Reorganization Act secs. 201, 202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste Policy Act sec. 180 (42 U.S.C. 10175); Government Paperwork Elimination Act sec. 1704 (44 U.S.C. 3504 note); Energy Policy Act of 2005, Pub. L. 109–58, 119 Stat. 594 (2005). Section 71.97 also issued under sec. 301, Pub. L. 96–295, 94 Stat. 789 –790.

Source: 60 FR 50264, Sept. 28, 1995, unless otherwise noted.

[72 FR 63974, Nov. 14, 2007; 73 FR 63572, Oct. 24, 2008; 77 FR 39908, Jul. 6, 2012; 77 FR 34204, Jun. 11, 2012]

Subpart A--General Provisions




Source: 69 FR 3786, Jan. 26, 2004, unless otherwise noted.

§ 71.0 Purpose and scope.


(a) This part establishes--

(1) Requirements for packaging, preparation for shipment, and transportation of licensed material; and

(2) Procedures and standards for NRC approval of packaging and shipping procedures for fissile material and for a quantity of other licensed material in excess of a Type A quantity.

(b) The packaging and transport of licensed material are also subject to other parts of this chapter (e.g., 10 CFR parts 20, 21, 30, 40, 70, and 73) and to the regulations of other agencies (e.g., the U.S. Department of Transportation (DOT) and the U.S. Postal Service)1 having jurisdiction over means of transport. The requirements of this part are in addition to, and not in substitution for, other requirements.

(c) The regulations in this part apply to any licensee authorized by specific or general license issued by the Commission to receive, possess, use, or transfer licensed material, if the licensee delivers that material to a carrier for transport, transports the material outside the site of usage as specified in the NRC license, or transports that material on public highways. No provision of this part authorizes possession of licensed material.

(d)(1) Exemptions from the requirement for license in § 71.3 are specified in § 71.14. General licenses for which no NRC package approval is required are issued in §§ 71.20 through 71.23. The general license in § 71.17 requires that an NRC certificate of compliance or other package approval be issued for the package to be used under this general license.

(2) Application for package approval must be completed in accordance with subpart D of this part, demonstrating that the design of the package to be used satisfies the package approval standards contained in subpart E of this part, as related to the tests of subpart F of this part.

(3) A licensee transporting licensed material, or delivering licensed material to a carrier for transport, shall comply with the operating control requirements of subpart G of this part; the quality assurance requirements of subpart H of this part; and the general provisions of subpart A of this part, including DOT regulations referenced in § 71.5.

(e) The regulations of this part apply to any person holding, or applying for, a certificate of compliance, issued pursuant to this part, for a package intended for the transportation of radioactive material, outside the confines of a licensee's facility or authorized place of use.

(f) The regulations in this part apply to any person required to obtain a certificate of compliance, or an approved compliance plan, pursuant to part 76 of this chapter, if the person delivers radioactive material to a common or contract carrier for transport or transports the material outside the confines of the person's plant or other authorized place of use.

(g) This part also gives notice to all persons who knowingly provide to any licensee, certificate holder, quality assurance program approval holder, applicant for a license, certificate, or quality assurance program approval, or to a contractor, or subcontractor of any of them, components, equipment, materials, or other goods or services, that relate to a licensee's, certificate holder's, quality assurance program approval holder's, or applicant's activities subject to this part, that they may be individually subject to NRC enforcement action for violation of § 71.8.

1 Postal Service Manual (Domestic Mail Manual), section 124, which is incorporated by reference at 39 CFR 111.1.

§ 71.1 Communications and records.




(a) Except where otherwise specified, all communications and reports concerning the regulations in this part and applications filed under them should be sent by mail addressed: ATTN: Document Control Desk, Director, Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, by hand delivery to the NRC's offices at 11555 Rockville Pike, Rockville, Maryland; or, where practicable, by electronic submission, for example, via Electronic Information Exchange, or CD-ROM. Electronic submissions must be made in a manner that enables the NRC to receive, read, authenticate, distribute, and archive the submission, and process and retrieve it a single page at a time. Detailed guidance on making electronic submissions can be obtained by visiting the NRC’s Web site at http://www.nrc.gov/site-help/e-submittals.html; by e-mail to MSHD.Resource@nrc.gov; or by writing the Office of Information Services, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001. The guidance discusses, among other topics, the formats the NRC can accept, the use of electronic signatures, and the treatment of nonpublic information. If the submission date falls on a Saturday, Sunday, or a Federal holiday, the next Federal working day becomes the official due date.

(b) Each record required by this part must be legible throughout the retention period specified by each Commission regulation. The record may be the original or a reproduced copy or a microform provided that the copy or microform is authenticated by authorized personnel and that the microform is capable of producing a clear copy throughout the required retention period. The record may also be stored in electronic media with the capability for producing legible, accurate, and complete records during the required retention period. Records such as letters, drawings, and specifications must include all pertinent information such as stamps, initials, and signatures. The licensee shall maintain adequate safeguards against tampering with and loss of records.

[69 FR 3786, Jan. 26, 2004; 69 FR 58038, Sept. 29, 2004; 70 FR 69421, Nov. 16, 2005; 72 FR 33386, Jun. 18, 2007; 74 FR 62683, Dec. 1, 2009; 75 FR 73945, Nov. 30, 2010]

§ 71.2 Interpretations.




Except as specifically authorized by the Commission in writing, no interpretation of the meaning of the regulations in this part by any officer or employee of the Commission, other than a written interpretation by the General Counsel, will be recognized to be binding upon the Commission.

§ 71.3 Requirement for license.




Except as authorized in a general license or a specific license issued by the Commission, or as exempted in this part, no licensee may--

(a) Deliver licensed material to a carrier for transport; or

(b) Transport licensed material.

§ 71.4 Definitions.




The following terms are as defined here for the purpose of this part. To ensure compatibility with international transportation standards, all limits in this part are given in terms of dual units: The International System of Units (SI) followed or preceded by U.S. standard or customary units. The U.S. customary units are not exact equivalents but are rounded to a convenient value, providing a functionally equivalent unit. For the purpose of this part, either unit may be used.

A1 means the maximum activity of special form radioactive material permitted in a Type A package. This value is either listed in Appendix A, Table A-1, of this part, or may be derived in accordance with the procedures prescribed in Appendix A of this part.

A2 means the maximum activity of radioactive material, other than special form material, LSA, and SCO material, permitted in a Type A package. This value is either listed in Appendix A, Table A-1, of this part, or may be derived in accordance with the procedures prescribed in Appendix A of this part.

Carrier means a person engaged in the transportation of passengers or property by land or water as a common, contract, or private carrier, or by civil aircraft.

Certificate holder means a person who has been issued a certificate of compliance or other package approval by the Commission.

Certificate of Compliance (CoC) means the certificate issued by the Commission under subpart D of this part which approves the design of a package for the transportation of radioactive material.

Close reflection by water means immediate contact by water of sufficient thickness for maximum reflection of neutrons.

Consignment means each shipment of a package or groups of packages or load of radioactive material offered by a shipper for transport.

Containment system means the assembly of components of the packaging intended to retain the radioactive material during transport.

Conveyance means:

(1) For transport by public highway or rail any transport vehicle or large freight container;

(2) For transport by water any vessel, or any hold, compartment, or defined deck area of a vessel including any transport vehicle on board the vessel; and

(3) For transport by any aircraft.

Criticality Safety Index (CSI) means the dimensionless number (rounded up to the next tenth) assigned to and placed on the label of a fissile material package, to designate the degree of control of accumulation of packages containing fissile material during transportation. Determination of the criticality safety index is described in §§ 71.22, 71.23, and 71.59.

Deuterium means, for the purposes of §§ 71.15 and 71.22, deuterium and any deuterium compounds, including heavy water, in which the ratio of deuterium atoms to hydrogen atoms exceeds 1:5000.

DOT means the U.S. Department of Transportation.

Exclusive use means the sole use by a single consignor of a conveyance for which all initial, intermediate, and final loading and unloading are carried out in accordance with the direction of the consignor or consignee. The consignor and the carrier must ensure that any loading or unloading is performed by personnel having radiological training and resources appropriate for safe handling of the consignment. The consignor must issue specific instructions, in writing, for maintenance of exclusive use shipment controls, and include them with the shipping paper information provided to the carrier by the consignor.

Fissile material means the radionuclides uranium-233, uranium-235, plutonium-239, and plutonium-241, or any combination of these radionuclides. Fissile material means the fissile nuclides themselves, not material containing fissile nuclides. Unirradiated natural uranium and depleted uranium and natural uranium or depleted uranium, that has been irradiated in thermal reactors only, are not included in this definition. Certain exclusions from fissile material controls are provided in §71.15.

Graphite means, for the purposes of §§ 71.15 and 71.22, graphite with a boron equivalent content less than 5 parts per million and density greater than 1.5 grams per cubic centimeter.

Indian tribe means an Indian or Alaska Native tribe, band, nation, pueblo, village, or community that the Secretary of the Interior acknowledges to exist as an Indian tribe pursuant to the Federally Recognized Indian Tribe List Act of 1994, 25 U.S.C. 479a.

Licensed material means byproduct, source, or special nuclear material received, possessed, used, or transferred under a general or specific license issued by the Commission pursuant to the regulations in this chapter.

Low Specific Activity (LSA) material means radioactive material with limited specific activity which is nonfissile or is excepted under §71.15, and which satisfies the descriptions and limits set forth below. Shielding materials surrounding the LSA material may not be considered in determining the estimated average specific activity of the package contents. LSA material must be in one of three groups:

(1) LSA—I.

(i) Uranium and thorium ores, concentrates of uranium and thorium ores, and other ores containing naturally occurring radioactive radionuclides which are not intended to be processed for the use of these radionuclides;

(ii) Solid unirradiated natural uranium or depleted uranium or natural thorium or their solid or liquid compounds or mixtures;

(iii) Radioactive material for which the A2 value is unlimited; or

(iv) Other radioactive material in which the activity is distributed throughout and the estimated average specific activity does not exceed 30 times the value for exempt material activity concentration determined in accordance with Appendix A.

(2) LSA—II.

(i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/liter); or

(ii) Other material in which the activity is distributed throughout and the average specific activity does not exceed 10 –4A2/g for solids and gases, and 10–5A2/g for liquids.

(3) LSA—III. Solids (e.g., consolidated wastes, activated materials), excluding powders, that satisfy the requirements of § 71.77, in which:

(i) The radioactive material is distributed throughout a solid or a collection of solid objects, or is essentially uniformly distributed in a solid compact binding agent (such as concrete, bitumen, ceramic, etc.);

(ii) The radioactive material is relatively insoluble, or it is intrinsically contained in a relatively insoluble material, so that even under loss of packaging, the loss of radioactive material per package by leaching, when placed in water for 7 days, would not exceed 0.1 A2; and

(iii) The estimated average specific activity of the solid does not exceed 2 x 10–3A2/g.

Low toxicity alpha emitters means natural uranium, depleted uranium, natural thorium; uranium-235, uranium-238, thorium-232, thorium-228 or thorium-230 when contained in ores or physical or chemical concentrates or tailings; or alpha emitters with a half-life of less than 10 days.

Maximum normal operating pressure means the maximum gauge pressure that would develop in the containment system in a period of 1 year under the heat condition specified in §71.71(c)(1), in the absence of venting, external cooling by an ancillary system, or operational controls during transport.

Natural thorium means thorium with the naturally occurring distribution of thorium isotopes (essentially 100 weight percent thorium-232).

Normal form radioactive material means radioactive material that has not been demonstrated to qualify as "special form radioactive material."

Optimum interspersed hydrogenous moderation means the presence of hydrogenous material between packages to such an extent that the maximum nuclear reactivity results.

Package means the packaging together with its radioactive contents as presented for transport.

(1) Fissile material package or Type AF package, Type BF package, Type B(U)F package, or Type B(M)F package means a fissile material packaging together with its fissile material contents.

(2) Type A package means a Type A packaging together with its radioactive contents. A Type A package is defined and must comply with the DOT regulations in 49 CFR part 173.

(3) Type B package means a Type B packaging together with its radioactive contents. On approval, a Type B package design is designated by NRC as B(U) unless the package has a maximum normal operating pressure of more than 700 kPa (100 lbs/in2) gauge or a pressure relief device that would allow the release of radioactive material to the environment under the tests specified in §71.73 (hypothetical accident conditions), in which case it will receive a designation B(M). B(U) refers to the need for unilateral approval of international shipments; B(M) refers to the need for multilateral approval of international shipments. There is no distinction made in how packages with these designations may be used in domestic transportation. To determine their distinction for international transportation, see DOT regulations in 49 CFR Part 173. A Type B package approved before September 6, 1983, was designated only as Type B. Limitations on its use are specified in §71.19.

Packaging means the assembly of components necessary to ensure compliance with the packaging requirements of this part. It may consist of one or more receptacles, absorbent materials, spacing structures, thermal insulation, radiation shielding, and devices for cooling or absorbing mechanical shocks. The vehicle, tie-down system, and auxiliary equipment may be designated as part of the packaging.

Special form radioactive material means radioactive material that satisfies the following conditions:

(1) It is either a single solid piece or is contained in a sealed capsule that can be opened only by destroying the capsule;

(2) The piece or capsule has at least one dimension not less than 5 mm (0.2 in); and

(3) It satisfies the requirements of §71.75. A special form encapsulation designed in accordance with the requirements of §71.4 in effect on June 30, 1983 (see 10 CFR part 71, revised as of January 1, 1983), and constructed before July 1, 1985, and a special form encapsulation designed in accordance with the requirements of §71.4 in effect on March 31, 1996 (see 10 CFR part 71, revised as of January 1, 1983), and constructed before April 1, 1998, may continue to be used. Any other special form encapsulation must meet the specifications of this definition.

Specific activity of a radionuclide means the radioactivity of the radionuclide per unit mass of that nuclide. The specific activity of a material in which the radionuclide is essentially uniformly distributed is the radioactivity per unit mass of the material.

Spent nuclear fuel or Spent fuel means fuel that has been withdrawn from a nuclear reactor following irradiation, has undergone at least 1 year's decay since being used as a source of energy in a power reactor, and has not been chemically separated into its constituent elements by reprocessing. Spent fuel includes the special nuclear material, byproduct material, source material, and other radioactive materials associated with fuel assemblies.

State means a State of the United States, the District of Columbia, the Commonwealth of Puerto Rico, the Virgin Islands, Guam, American Samoa, and the Commonwealth of the Northern Mariana Islands.

Surface Contaminated Object (SCO) means a solid object that is not itself classed as radioactive material, but which has radioactive material distributed on any of its surfaces. SCO must be in one of two groups with surface activity not exceeding the following limits:

(1) SCO-I: A solid object on which:

(i) The nonfixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 Bq/cm2 (10–4 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 0.4 Bq/cm2 (10–5 microcurie/cm2) for all other alpha emitters;

(ii) The fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 x 104 Bq/cm2 (1.0 microcurie/cm2) for beta and gamma and low toxicity alpha emitters, or 4 x 103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha emitters; and

(iii) The nonfixed contamination plus the fixed contamination on the inaccessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 4 x 104 Bq/cm2 (1 microcurie/cm2 ) for beta and gamma and low toxicity alpha emitters, or 4 x 103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha emitters.

(2) SCO-II: A solid object on which the limits for SCO-I are exceeded and on which:

(i) The nonfixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 400 Bq/cm2 (10–2 microcurie/cm2) for beta and gamma and low toxicity alpha emitters or 40 Bq/cm2 (10–3 microcurie/cm2) for all other alpha emitters;

(ii) The fixed contamination on the accessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 8 x 105 Bq/cm2 (20 microcuries/cm2) for beta and gamma and low toxicity alpha emitters, or 8 x 104 Bq/cm2 (2 microcuries/cm2) for all other alpha emitters; and

(iii) The nonfixed contamination plus the fixed contamination on the inaccessible surface averaged over 300 cm2 (or the area of the surface if less than 300 cm2) does not exceed 8 x 105 Bq/cm2 (20 microcuries/cm2) for beta and gamma and low toxicity alpha emitters, or 8 x 104 Bq/cm2 (2 microcuries/cm2) for all other alpha emitters.

Transport index (TI) means the dimensionless number (rounded up to the next tenth) placed on the label of a package, to designate the degree of control to be exercised by the carrier during transportation. The transport index is the number determined by multiplying the maximum radiation level in millisievert (mSv) per hour at 1 meter (3.3 ft) from the external surface of the package by 100 (equivalent to the maximum radiation level in millirem per hour at 1 meter (3.3 ft)).

Tribal official means the highest ranking individual that represents Tribal leadership, such as the Chief, President, or Tribal Council leadership.

Type A quantity means a quantity of radioactive material, the aggregate radioactivity of which does not exceed A1 for special form radioactive material, or A2, for normal form radioactive material, where A1 and A 2 are given in Table A-1 of this part, or may be determined by procedures described in Appendix A of this part.

Type B quantity means a quantity of radioactive material greater than a Type A quantity.

Unirradiated uranium means uranium containing not more than 2 x 103 Bq of plutonium per gram of uranium-235, not more than 9 x 106 Bq of fission products per gram of uranium-235, and not more than 5 x 10–3 g of uranium-236 per gram of uranium-235.

Uranium—natural, depleted, enriched:

(1) Natural uranium means uranium with the naturally occurring distribution of uranium isotopes (approximately 0.711 weight percent uranium-235, and the remainder by weight essentially uranium-238).

(2) Depleted uranium means uranium containing less uranium-235 than the naturally occurring distribution of uranium isotopes.

(3) Enriched uranium means uranium containing more uranium-235 than the naturally occurring distribution of uranium isotopes.

[69 FR 3787, Jan. 26, 2004; 69 FR 58038, Sep. 29, 2004; 77 FR 34204, Jun. 11, 2012]

§ 71.5 Transportation of licensed material.




(a) Each licensee who transports licensed material outside the site of usage, as specified in the NRC license, or where transport is on public highways, or who delivers licensed material to a carrier for transport, shall comply with the applicable requirements of the DOT regulations in 49 CFR parts 107, 171 through 180, and 390 through 397, appropriate to the mode of transport.

(1) The licensee shall particularly note DOT regulations in the following areas:

(i) Packaging—49 CFR part 173: subparts A, B, and I.

(ii) Marking and labeling—49 CFR part 172: subpart D; and §§ 172.400 through 172.407 and §§ 172.436 through 172.441 of subpart E.

(iii) Placarding—49 CFR part 172: subpart F, especially §§ 172.500 through 172.519 and 172.556; and appendices B and C.

(iv) Accident reporting—49 CFR part 171: §§ 171.15 and 171.16.

(v) Shipping papers and emergency information—49 CFR part 172: subparts C and G.

(vi) Hazardous material employee training—49 CFR part 172: subpart H.

(vii) Security plans—49 CFR part 172: subpart I.

(viii) Hazardous material shipper/carrier registration—49 CFR part 107: subpart G.

(2) The licensee shall also note DOT regulations pertaining to the following modes of transportation:

(i) Rail—49 CFR part 174: subparts A through D and K.

(ii) Air—49 CFR part 175.

(iii) Vessel—49 CFR part 176: subparts A through F and M.

(iv) Public Highway—49 CFR part 177 and parts 390 through 397.

(b) If DOT regulations are not applicable to a shipment of licensed material, the licensee shall conform to the standards and requirements of the DOT specified in paragraph (a) of this section to the same extent as if the shipment or transportation were subject to DOT regulations. A request for modification, waiver, or exemption from those requirements, and any notification referred to in those requirements, must be filed with, or made to, the Director, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.

§ 71.6 Information collection requirements: OMB approval.




(a) The Nuclear Regulatory Commission has submitted the information collection requirements contained in this part to the Office of Management and Budget (OMB) for approval as required by the Paperwork Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number. OMB has approved the information collection requirements contained in this part under control number 3150-0008.

(b) The approved information collection requirements contained in this part appear in §§ 71.5, 71.7, 71.9, 71.12, 71.17, 71.19, 71.22, 71.23, 71.31, 71.33, 71.35, 71.37, 71.38, 71.39, 71.41, 71.47, 71.85, 71.87, 71.89, 71.91, 71.93, 71.95, 71.97, 71.101, 71.103, 71.105, 71.107, 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 71.123, 71.125, 71.127, 71.129, 71.131, 71.133, 71.135, 71.137, and Appendix A, Paragraph II.

[75 FR 73945, Nov. 30, 2010]

§ 71.7 Completeness and accuracy of information.




(a) Information provided to the Commission by a licensee, certificate holder, or an applicant for a license or CoC; or information required by statute or by the Commission's regulations, orders, license or CoC conditions, to be maintained by the licensee or certificate holder, must be complete and accurate in all material respects.

(b) Each licensee, certificate holder, or applicant for a license or CoC must notify the Commission of information identified by the licensee, certificate holder, or applicant for a license or CoC as having, for the regulated activity, a significant implication for public health and safety or common defense and security. A licensee, certificate holder, or an applicant for a license or CoC violates this paragraph only if the licensee, certificate holder, or applicant for a license or CoC fails to notify the Commission of information that the licensee, certificate holder, or applicant for a license or CoC has identified as having a significant implication for public health and safety or common defense and security. Notification must be provided to the Administrator of the appropriate Regional Office within 2 working days of identifying the information. This requirement is not applicable to information which is already required to be provided to the Commission by other reporting or updating requirements.

§ 71.8 Deliberate misconduct.




(a) This section applies to any--

(1) Licensee;

(2) Certificate holder;

(3) Quality assurance program approval holder;

(4) Applicant for a license, certificate, or quality assurance program approval;

(5) Contractor (including a supplier or consultant) or subcontractor, to any person identified in paragraph (a)(4) of this section; or

(6) Employees of any person identified in paragraphs (a)(1) through (a)(5) of this section.

(b) A person identified in paragraph (a) of this section who knowingly provides to any entity, listed in paragraphs (a)(1) through (a)(5) of this section, any components, materials, or other goods or services that relate to a licensee's, certificate holder's, quality assurance program approval holder's, or applicant's activities subject to this part may not:

(1) Engage in deliberate misconduct that causes or would have caused, if not detected, a licensee, certificate holder, quality assurance program approval holder, or any applicant to be in violation of any rule, regulation, or order; or any term, condition or limitation of any license, certificate, or approval issued by the Commission; or

(2) Deliberately submit to the NRC, a licensee, a certificate holder, quality assurance program approval holder, an applicant for a license, certificate or quality assurance program approval, or a licensee's, applicant's, certificate holder's, or quality assurance program approval holder's contractor or subcontractor, information that the person submitting the information knows to be incomplete or inaccurate in some respect material to the NRC.

(c) A person who violates paragraph (b)(1) or (b)(2) of this section may be subject to enforcement action in accordance with the procedures in 10 CFR part 2, subpart B.

(d) For the purposes of paragraph (b)(1) of this section, deliberate misconduct by a person means an intentional act or omission that the person knows:

(1) Would cause a licensee, certificate holder, quality assurance program approval holder, or applicant for a license, certificate, or quality assurance program approval to be in violation of any rule, regulation, or order; or any term, condition, or limitation of any license or certificate issued by the Commission; or

(2) Constitutes a violation of a requirement, procedure, instruction, contract, purchase order, or policy of a licensee, certificate holder, quality assurance program approval holder, applicant, or the contractor or subcontractor of any of them.

§ 71.9 Employee protection.




(a) Discrimination by a Commission licensee, certificate holder, an applicant for a Commission license or a CoC, or a contractor or subcontractor of any of these, against an employee for engaging in certain protected activities, is prohibited. Discrimination includes discharge and other actions that relate to compensation, terms, conditions, or privileges of employment. The protected activities are established in section 211 of the Energy Reorganization Act of 1974, as amended, and in general are related to the administration or enforcement of a requirement imposed under the Atomic Energy Act of 1954, as amended, or the Energy Reorganization Act of 1974, as amended.

(1) The protected activities include, but are not limited to:

(i) Providing the Commission or his or her employer information about alleged violations of either of the statutes named in paragraph (a) of this section or possible violations of requirements imposed under either of those statutes;

(ii) Refusing to engage in any practice made unlawful under either of the statutes named in paragraph (a) of this section or under these requirements if the employee has identified the alleged illegality to the employer;

(iii) Requesting the Commission to institute action against his or her employer for the administration or enforcement of these requirements;

(iv) Testifying in any Commission proceeding, or before Congress, or at any Federal or State proceeding regarding any provision (or proposed provision) of either of the statutes named in paragraph (a) of this section; and

(v) Assisting or participating in, or is about to assist or participate in, these activities.

(2) These activities are protected even if no formal proceeding is actually initiated as a result of the employee's assistance or participation.

(3) This section has no application to any employee alleging discrimination prohibited by this section who, acting without direction from his or her employer (or the employer's agent), deliberately causes a violation of any requirement of the Energy Reorganization Act of 1974, as amended, or the Atomic Energy Act of 1954, as amended.

(b) Any employee who believes that he or she has been discharged or otherwise discriminated against by any person for engaging in protected activities specified in paragraph (a)(1) of this section may seek a remedy for the discharge or discrimination through an administrative proceeding in the Department of Labor. The administrative proceeding must be initiated within 180 days after an alleged violation occurs. The employee may do this by filing a complaint alleging the violation with the Department of Labor, Employment Standards Administration, Wage and Hour Division. The Department of Labor may order reinstatement, back pay, and compensatory damages.

(c) A violation of paragraph (a), (e), or (f) of this section by a Commission licensee, certificate holder, applicant for a Commission license or a CoC, or a contractor or subcontractor of any of these may be grounds for:

(1) Denial, revocation, or suspension of the license or the CoC;

(2) Imposition of a civil penalty on the licensee, applicant, or a contractor or subcontractor of the licensee or applicant; or

(3) Other enforcement action.

(d) Actions taken by an employer, or others, which adversely affect an employee may be predicated upon nondiscriminatory grounds. The prohibition applies when the adverse action occurs because the employee has engaged in protected activities. An employee's engagement in protected activities does not automatically render him or her immune from discharge or discipline for legitimate reasons or from adverse action dictated by nonprohibited considerations.

(e)(1) Each licensee, certificate holder, and applicant for a license or CoC must prominently post the current revision of NRC Form 3, "Notice to Employees," referenced in §19.11(c) of this chapter. This form must be posted at locations sufficient to permit employees protected by this section to observe a copy on the way to or from their place of work. The premises must be posted not later than 30 days after an application is docketed and remain posted while the application is pending before the Commission, during the term of the license or CoC, and for 30 days following license or CoC termination.

(2) Copies of NRC Form 3 may be obtained by writing to the Regional Administrator of the appropriate U.S. Nuclear Regulatory Commission Regional Office listed in Appendix D to part 20 of this chapter or by calling the NRC Publishing Services Branch at 301-415-5877 begin_of_the_skype_highlighting GRATIS 301-415-5877 end_of_the_skype_highlighting.

(f) No agreement affecting the compensation, terms, conditions, or privileges of employment, including an agreement to settle a complaint filed by an employee with the Department of Labor pursuant to section 211 of the Energy Reorganization Act of 1974, as amended, may contain any provision which would prohibit, restrict, or otherwise discourage an employee from participating in a protected activity as defined in paragraph (a)(1) of this section including, but not limited to, providing information to the NRC or to his or her employer on potential violations or other matters within NRC's regulatory responsibilities.

[72 FR 63975, Nov. 14, 2007]

§ 71.10 Public inspection of application.




Applications for approval of a package design under this part, which are submitted to the Commission, may be made available for public inspection, in accordance with provisions of parts 2 and 9 of this chapter. This includes an application to amend or revise an existing package design, any associated documents and drawings submitted with the application, and any responses to NRC requests for additional information.

§ 71.11 Protection of Safeguards Information




Each licensee, certificate holder, or applicant for a Certificate of Compliance for a transportation package for transport of irradiated reactor fuel, strategic special nuclear material, a critical mass of special nuclear material, or byproduct material in quantities determined by the Commission through order or regulation to be significant to the public health and safety or the common defense and security, shall protect Safeguards Information against unauthorized disclosure in accordance with the requirements in § 73.21 and the requirements of § 73.22 or § 73.23 of this chapter, as applicable.

[73 FR 63572, Oct. 24, 2008]

Subpart B--Exemptions




Source: 69 FR 3786, Jan. 26, 2004, unless otherwise noted.

§ 71.12 Specific exemptions.


On application of any interested person or on its own initiative, the Commission may grant any exemption from the requirements of the regulations in this part that it determines is authorized by law and will not endanger life or property nor the common defense and security.

§ 71.13 Exemption of physicians.




Any physician licensed by a State to dispense drugs in the practice of medicine is exempt from § 71.5 with respect to transport by the physician of licensed material for use in the practice of medicine. However, any physician operating under this exemption must be licensed under 10 CFR part 35 or the equivalent Agreement State regulations.

§ 71.14 Exemption for low-level materials.




(a) A licensee is exempt from all the requirements of this part with respect to shipment or carriage of the following low-level materials:

(1) Natural material and ores containing naturally occurring radionuclides that are not intended to be processed for use of these radionuclides, provided the activity concentration of the material does not exceed 10 times the values specified in Appendix A, Table A-2, of this part.

(2) Materials for which the activity concentration is not greater than the activity concentration values specified in Appendix A, Table A-2 of this part, or for which the consignment activity is not greater than the limit for an exempt consignment found in Appendix A, Table A-2, of this part.

(b) A licensee is exempt from all the requirements of this part, other than §§ 71.5 and 71.88, with respect to shipment or carriage of the following packages, provided the packages do not contain any fissile material, or the material is exempt from classification as fissile material under § 71.15:

(1) A package that contains no more than a Type A quantity of radioactive material;

(2) A package transported within the United States that contains no more than 0.74 TBq (20 Ci) of special form plutonium-244; or

(3) The package contains only LSA or SCO radioactive material, provided--

(i) That the LSA or SCO material has an external radiation dose of less than or equal to 10 mSv/h (1 rem/h), at a distance of 3 m from the unshielded material; or

(ii) That the package contains only LSA-I or SCO-I material.

§ 71.15 Exemption from classification as fissile material.




Fissile material meeting the requirements of at least one of the paragraphs (a) through (f) of this section are exempt from classification as fissile material and from the fissile material package standards of §§ 71.55 and 71.59, but are subject to all other requirements of this part, except as noted.

(a) Individual package containing 2 grams or less fissile material.

(b) Individual or bulk packaging containing 15 grams or less of fissile material provided the package has at least 200 grams of solid nonfissile material for every gram of fissile material. Lead, beryllium, graphite, and hydrogenous material enriched in deuterium may be present in the package but must not be included in determining the required mass for solid nonfissile material.

(c)(1) Low concentrations of solid fissile material commingled with solid nonfissile material, provided that:

(i) There is at least 2000 grams of solid nonfissile material for every gram of fissile material, and

(ii) There is no more than 180 grams of fissile material distributed within 360 kg of contiguous nonfissile material.

(2) Lead, beryllium, graphite, and hydrogenous material enriched in deuterium may be present in the package but must not be included in determining the required mass of solid nonfissile material.

(d) Uranium enriched in uranium-235 to a maximum of 1 percent by weight, and with total plutonium and uranium-233 content of up to 1 percent of the mass of uranium-235, provided that the mass of any beryllium, graphite, and hydrogenous material enriched in deuterium constitutes less than 5 percent of the uranium mass.

(e) Liquid solutions of uranyl nitrate enriched in uranium-235 to a maximum of 2 percent by mass, with a total plutonium and uranium-233 content not exceeding 0.002 percent of the mass of uranium, and with a minimum nitrogen to uranium atomic ratio (N/U) of 2. The material must be contained in at least a DOT Type A package.

(f) Packages containing, individually, a total plutonium mass of not more than 1000 grams, of which not more than 20 percent by mass may consist of plutonium-239, plutonium-241, or any combination of these radionuclides.

§ 71.16 [Reserved]




Subpart C--General Licenses




Source: 69 FR 3792, Jan. 26, 2004, unless otherwise noted.

§ 71.17 General license: NRC-approved package.


(a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package for which a license, certificate of compliance (CoC), or other approval has been issued by the NRC.

(b) This general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.

(c) This general license applies only to a licensee who--

(1) Has a copy of the CoC, or other approval of the package, and has the drawings and other documents referenced in the approval relating to the use and maintenance of the packaging and to the actions to be taken before shipment;

(2) Complies with the terms and conditions of the license, certificate, or other approval, as applicable, and the applicable requirements of subparts A, G, and H of this part; and

(3) Before the licensee's first use of the package, submits in writing to: ATTN: Document Control Desk, Director, Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards, using an appropriate method listed in § 71.1(a), the licensee's name and license number and the package identification number specified in the package approval.

(d) This general license applies only when the package approval authorizes use of the package under this general license.

(e) For a Type B or fissile material package, the design of which was approved by NRC before April 1, 1996, the general license is subject to the additional restrictions of § 71.19.

[75 FR 73945, Nov. 30, 2010]

§ 71.18 [Reserved]




§ 71.19 Previously approved package.




(a) [Reserved]

(b) A Type B(U) package, a Type B(M) package, or a fissile material package, previously approved by the NRC but without the designation "- 85" in the identification number of the NRC CoC, may be used under the general license of § 71.17 with the following additional conditions:

(1) Fabrication of the package is satisfactorily completed by April 1, 1999, as demonstrated by application of its model number in accordance with § 71.85(c);

(2) A package used for a shipment to a location outside the United States is subject to multilateral approval as defined in DOT regulations at 49 CFR 173.403; and

(3) A serial number which uniquely identifies each packaging which conforms to the approved design is assigned to and legibly and durably marked on the outside of each packaging.

(c) A Type B(U) package, a Type B(M) package, or a fissile material package previously approved by the NRC with the designation "-85" in the identification number of the NRC CoC, may be used under the general license of § 71.17 with the following additional conditions:

(1) Fabrication of the package must be satisfactorily completed by December 31, 2006, as demonstrated by application of its model number in accordance with § 71.85(c); and

(2) After December 31, 2003, a package used for a shipment to a location outside the United States is subject to multilateral approval as defined in DOT regulations at 49 CFR 173.403.

(d) NRC will approve modifications to the design and authorized contents of a Type B package, or a fissile material package, previously approved by NRC, provided--

(1) The modifications of a Type B package are not significant with respect to the design, operating characteristics, or safe performance of the containment system, when the package is subjected to the tests specified in §§ 71.71 and 71.73;

(2) The modifications of a fissile material package are not significant, with respect to the prevention of criticality, when the package is subjected to the tests specified in §§ 71.71 and 71.73; and

(3) The modifications to the package satisfy the requirements of this part.

(e) NRC will revise the package identification number to designate previously approved package designs as B, BF, AF, B(U), B(M), B(U)F, B(M)F, B(U)-85, B(U)F-85, B(M)-85, B(M)F-85, or AF-85 as appropriate, and with the identification number suffix "-96" after receipt of an application demonstrating that the design meets the requirements of this part.

§ 71.20 General license: DOT specification container.




(a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a specification container for fissile material or for a Type B quantity of radioactive material as specified in DOT regulations at 49 CFR parts 173 and 178.

(b) This general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.

(c) This general license applies only to a licensee who--

(1) Has a copy of the specification; and

(2) Complies with the terms and conditions of the specification and the applicable requirements of subparts A, G, and H of this part.

(d) This general license is subject to the limitation that the specification container may not be used for a shipment to a location outside the United States, except by multilateral approval, as defined in DOT regulations at 49 CFR 173.403.

(e) This section expires October 1, 2008.

§ 71.21 General license: Use of foreign approved package.




(a) A general license is issued to any licensee of the Commission to transport, or to deliver to a carrier for transport, licensed material in a package, the design of which has been approved in a foreign national competent authority certificate, that has been revalidated by DOT as meeting the applicable requirements of 49 CFR 171.12.

(b) Except as otherwise provided in this section, the general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the applicable provisions of subpart H of this part.

(c) This general license applies only to shipments made to or from locations outside the United States.

(d) This general license applies only to a licensee who--

(1) Has a copy of the applicable certificate, the revalidation, and the drawings and other documents referenced in the certificate, relating to the use and maintenance of the packaging and to the actions to be taken before shipment; and

(2) Complies with the terms and conditions of the certificate and revalidation, and with the applicable requirements of subparts A, G, and H of this part. With respect to the quality assurance provisions of subpart H of this part, the licensee is exempt from design, construction, and fabrication considerations.

§ 71.22 General license: Fissile material.




(a) A general license is issued to any licensee of the Commission to transport fissile material, or to deliver fissile material to a carrier for transport, if the material is shipped in accordance with this section. The fissile material need not be contained in a package which meets the standards of subparts E and F of this part; however, the material must be contained in a Type A package. The Type A package must also meet the DOT requirements of 49 CFR 173.417(a).

(b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.

(c) The general license applies only when a package's contents:

(1) Contain no more than a Type A quantity of radioactive material; and

(2) Contain less than 500 total grams of beryllium, graphite, or hydrogenous material enriched in deuterium.

(d) The general license applies only to packages containing fissile material that are labeled with a CSI which:

(1) Has been determined in accordance with paragraph (e) of this section;

(2) Has a value less than or equal to 10; and

(3) For a shipment of multiple packages containing fissile material, the sum of the CSIs must be less than or equal to 50 (for shipment on a nonexclusive use conveyance) and less than or equal to 100 (for shipment on an exclusive use conveyance).

(e)(1) The value for the CSI must be greater than or equal to the number calculated by the following equation:


(2) The calculated CSI must be rounded up to the first decimal place;

(3) The values of X, Y, and Z used in the CSI equation must be taken from Tables 71-1 or 71-2, as appropriate;

(4) If Table 71-2 is used to obtain the value of X, then the values for the terms in the equation for uranium-233 and plutonium must be assumed to be zero; and

(5) Table 71-1 values for X, Y, and Z must be used to determine the CSI if:

(i) Uranium-233 is present in the package;

(ii) The mass of plutonium exceeds 1 percent of the mass of uranium-235;

(iii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight percent enrichment; or

(iv) Substances having a moderating effectiveness (i.e., an average hydrogen density greater than H2O) (e.g., certain hydrocarbon oils or plastics) are present in any form, except as polyethylene used for packing or wrapping.

Table 71-1. Mass Limits for General License Packages Containing Mixed Quantities of Fissile Material or Uranium-235 of Unknown Enrichment per § 71.22(e)

Fissile material Fissile material mass mixed with moderating substances having an average hydrogen density less than or equal to H2O (grams) Fissile material mass mixed with moderating substances having an average hydrogen density greater than H2Oa (grams)
235U (X)
60
38
233U (Y)
43
27
239Pu or 241Pu (Z)
37
24

a When mixtures of moderating substances are present, the lower mass limits shall be used if more than 15 percent of the moderating substance has an average hydrogen density greater than H2O.

Table 71-2. Mass Limits for General License Packages Containing Uranium-235 of Known Enrichment per § 71.22(e)

Uranium enrichment in weight percent of 235U not exceeding Fissile material mass of 235U (X) (grams)
24
60
20
63
15
67
11
72
10
76
9.5
78
9
81
8.5
82
8
85
7.5
88
7
90
6.5
93
6
97
5.5
102
5
108
4.5
114
4
120
3.5
132
3
150
2.5
180
2
246
1.5
408
1.35
480
1
1,020
0.92
1,800

[69 FR 3786, Jan. 26, 2004; 69 FR 58038, Sept. 29, 2004]

§ 71.23 General license: Plutonium-beryllium special form material.




(a) A general license is issued to any licensee of the Commission to transport fissile material in the form of plutonium-beryllium (Pu-Be) special form sealed sources, or to deliver Pu-Be sealed sources to a carrier for transport, if the material is shipped in accordance with this section. This material need not be contained in a package which meets the standards of subparts E and F of this part; however, the material must be contained in a Type A package. The Type A package must also meet the DOT requirements of 49 CFR 173.417(a).

(b) The general license applies only to a licensee who has a quality assurance program approved by the Commission as satisfying the provisions of subpart H of this part.

(c) The general license applies only when a package's contents:

(1) Contain no more than a Type A quantity of radioactive material; and

(2) Contain less than 1000 g of plutonium, provided that: plutonium-239, plutonium-241, or any combination of these
radionuclides, constitutes less than 240 g of the total quantity of plutonium in the package.

(d) The general license applies only to packages labeled with a CSI which:

(1) Has been determined in accordance with paragraph (e) of this section;

(2) Has a value less than or equal to 100; and

(3) For a shipment of multiple packages containing Pu-Be sealed sources, the sum of the CSIs must be less than or equal to 50 (for shipment on a nonexclusive use conveyance) and less than or equal to 100 (for shipment on an exclusive use conveyance).

(e)(1) The value for the CSI must be greater than or equal to the number calculated by the following equation:


(2) The calculated CSI must be rounded up to the first decimal place.

§ 71.24 [Reserved]




§ 71.25 [Reserved]




Subpart D--Application for Package Approval




§ 71.31 Contents of application.


(a) An application for an approval under this part must include, for each proposed packaging design, the following information:

(1) A package description as required by § 71.33;

(2) A package evaluation as required by § 71.35; and

(3) A quality assurance program description, as required by § 71.37, or a reference to a previously approved quality assurance program.

(b) Except as provided in § 71.13, an application for modification of a package design, whether for modification of the packaging or authorized contents, must include sufficient information to demonstrate that the proposed design satisfies the package standards in effect at the time the application is filed.

(c) The applicant shall identify any established codes and standards proposed for use in package design, fabrication, assembly, testing, maintenance, and use. In the absence of any codes and standards, the applicant shall describe and justify the basis and rationale used to formulate the package quality assurance program.

§ 71.33 Package description.




The application must include a description of the proposed package in sufficient detail to identify the package accurately and provide a sufficient basis for evaluation of the package. The description must include --

(a) With respect to the packaging --

(1) Classification as Type B(U), Type B(M), or fissile material packaging;

(2) Gross weight;

(3) Model number;

(4) Identification of the containment system;

(5) Specific materials of construction, weights, dimensions, and fabrication methods of --

(i) Receptacles;

(ii) Materials specifically used as nonfissile neutron absorbers or moderators;

(iii) Internal and external structures supporting or protecting receptacles;

(iv) Valves, sampling ports, lifting devices, and tie-down devices; and

(v) Structural and mechanical means for the transfer and dissipation of heat; and

(6) Identification and volumes of any receptacles containing coolant.

(b) With respect to the contents of the package --

(1) Identification and maximum radioactivity of radioactive constituents;

(2) Identification and maximum quantities of fissile constituents;

(3) Chemical and physical form;

(4) Extent of reflection, the amount and identity of nonfissile materials used as neutron absorbers or moderators, and the atomic ratio of moderator to fissile constituents;

(5) Maximum normal operating pressure;

(6) Maximum weight;

(7) Maximum amount of decay heat; and

(8) Identification and volumes of any coolants.

§ 71.35 Package evaluation.




The application must include the following:

(a) A demonstration that the package satisfies the standards specified in subparts E and F of this part;

(b) For a fissile material package, the allowable number of packages that may be transported in the same vehicle in accordance with § 71.59; and

(c) For a fissile material shipment, any proposed special controls and precautions for transport, loading, unloading, and handling and any proposed special controls in case of an accident or delay.

§ 71.37 Quality assurance.




(a) The applicant shall describe the quality assurance program (see Subpart H of this part) for the design, fabrication, assembly, testing, maintenance, repair, modification, and use of the proposed package.

(b) The applicant shall identify any specific provisions of the quality assurance program that are applicable to the particular package design under consideration, including a description of the leak testing procedures.

§ 71.38 Renewal of a certificate of compliance or quality assurance program approval.




(a) Except as provided in paragraph (b) of this section, each Certificate of Compliance or Quality Assurance Program Approval expires at the end of the day, in the month and year stated in the approval.

(b) In any case in which a person, not less than 30 days before the expiration of an existing Certificate of Compliance or Quality Assurance Program Approval issued pursuant to the part, has filed an application in proper form for renewal of either of those approvals, the existing Certificate of Compliance or Quality Assurance Program Approval for which the renewal application was filed shall not be deemed to have expired until final action on the application for renewal has been taken by the Commission.

(c) In applying for renewal of an existing Certificate of Compliance or Quality Assurance Program Approval, an applicant may be required to submit a consolidated application that incorporates all changes to its program that, are incorporated by reference in the existing approval or certificate, into as few referenceable documents as reasonably achievable.

§ 71.39 Requirement for additional information.




The Commission may at any time require additional information in order to enable it to determine whether a license, certificate of compliance, or other approval should be granted, renewed, denied, modified, suspended, or revoked.

Subpart E--Package Approval Standards




§ 71.41 Demonstration of compliance.


(a) The effects on a package of the tests specified in § 71.71 ("Normal conditions of transport"), and the tests specified in § 71.73 ("Hypothetical accident conditions"), and § 71.61 ("Special requirements for Type B packages containing more than 105 A2"), must be evaluated by subjecting a specimen or scale model to a specific test, or by another method of demonstration acceptable to the Commission, as appropriate for the particular feature being considered.

(b) Taking into account the type of vehicle, the method of securing or attaching the package, and the controls to be exercised by the shipper, the Commission may permit the shipment to be evaluated together with the transporting vehicle.

(c) Environmental and test conditions different from those specified in §§ 71.71 and 71.73 may be approved by the Commission if the controls proposed to be exercised by the shipper are demonstrated to be adequate to provide equivalent safety of the shipment.

(d) Packages for which compliance with the other provisions of these regulations is impracticable shall not be transported except under special package authorization. Provided the applicant demonstrates that compliance with the other provisions of the regulations is impracticable and that the requisite standards of safety established by these regulations have been demonstrated through means alternative to the other provisions, a special package authorization may be approved for one-time shipments. The applicant shall demonstrate that the overall level of safety in transport for these shipments is at least equivalent to that which would be provided if all the applicable requirements had been met.

[60 FR 50264, Sept. 28, 1995 as amended at 69 FR 3794, Jan. 26, 2004]

§ 71.43 General standards for all packages.




(a) The smallest overall dimension of a package may not be less than 10 cm (4 in).

(b) The outside of a package must incorporate a feature, such as a seal, that is not readily breakable and that, while intact, would be evidence that the package has not been opened by unauthorized persons.

(c) Each package must include a containment system securely closed by a positive fastening device that cannot be opened unintentionally or by a pressure that may arise within the package.

(d) A package must be made of materials and construction that assure that there will be no significant chemical, galvanic, or other reaction among the packaging components, among package contents, or between the packaging components and the package contents, including possible reaction resulting from inleakage of water, to the maximum credible extent. Account must be taken of the behavior of materials under irradiation.

(e) A package valve or other device, the failure of which would allow radioactive contents to escape, must be protected against unauthorized operation and, except for a pressure relief device, must be provided with an enclosure to retain any leakage.

(f) A package must be designed, constructed, and prepared for shipment so that under the tests specified in § 71.71 ("Normal conditions of transport") there would be no loss or dispersal of radioactive contents, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging.

(g) A package must be designed, constructed, and prepared for transport so that in still air at 38°C (100°F) and in the shade, no accessible surface of a package would have a temperature exceeding 50°C (122°F) in a nonexclusive use shipment, or 85°C (185°F) in an exclusive use shipment.

(h) A package may not incorporate a feature intended to allow continuous venting during transport.

§ 71.45 Lifting and tie-down standards for all packages.




(a) Any lifting attachment that is a structural part of a package must be designed with a minimum safety factor of three against yielding when used to lift the package in the intended manner, and it must be designed so that failure of any lifting device under excessive load would not impair the ability of the package to meet other requirements of this subpart. Any other structural part of the package that could be used to lift the package must be capable of being rendered inoperable for lifting the package during transport, or must be designed with strength equivalent to that required for lifting attachments.

(b) Tie-down devices:

(1) If there is a system of tie-down devices that is a structural part of the package, the system must be capable of withstanding, without generating stress in any material of the package in excess of its yield strength, a static force applied to the center of gravity of the package having a vertical component of 2 times the weight of the package with its contents, a horizontal component along the direction in which the vehicle travels of 10 times the weight of the package with its contents, and a horizontal component in the transverse direction of 5 times the weight of the package with its contents.

(2) Any other structural part of the package that could be used to tie down the package must be capable of being rendered inoperable for tying down the package during transport, or must be designed with strength equivalent to that required for tie-down devices.

(3) Each tie-down device that is a structural part of a package must be designed so that failure of the device under excessive load would not impair the ability of the package to meet other requirements of this part.

§ 71.47 External radiation standards for all packages.




(a) Except as provided in paragraph (b) of this section, each package of radioactive materials offered for transportation must be designed and prepared for shipment so that under conditions normally incident to transportation the radiation level does not exceed 2 mSv/h (200 mrem/h) at any point on the external surface of the package, and the transport index does not exceed 10.

(b) A package that exceeds the radiation level limits specified in paragraph (a) of this section must be transported by exclusive use shipment only, and the radiation levels for such shipment must not exceed the following during transportation:

(1) 2 mSv/h (200 mrem/h) on the external surface of the package, unless the following conditions are met, in which case the limit is 10 mSv/h (1000 mrem/h):

(i) The shipment is made in a closed transport vehicle;

(ii) The package is secured within the vehicle so that its position remains fixed during transportation; and

(iii) There are no loading or unloading operations between the beginning and end of the transportation;

(2) 2 mSv/h (200 mrem/h) at any point on the outer surface of the vehicle, including the top and underside of the vehicle; or in the case of a flat-bed style vehicle, at any point on the vertical planes projected from the outer edges of the vehicle, on the upper surface of the load or enclosure, if used, and on the lower external surface of the vehicle; and

(3) 0.1 mSv/h (10 mrem/h) at any point 2 meters (80 in) from the outer lateral surfaces of the vehicle (excluding the top and underside of the vehicle); or in the case of a flat-bed style vehicle, at any point 2 meters (6.6 feet) from the vertical planes projected by the outer edges of the vehicle (excluding the top and underside of the vehicle); and

(4) 0.02 mSv/h (2 mrem/h) in any normally occupied space, except that this provision does not apply to private carriers, if exposed personnel under their control wear radiation dosimetry devices in conformance with 10 CFR 20.1502.

(c) For shipments made under the provisions of paragraph (b) of this section, the shipper shall provide specific written instructions to the carrier for maintenance of the exclusive use shipment controls. The instructions must be included with the shipping paper information.

(d) The written instructions required for exclusive use shipments must be sufficient so that, when followed, they will cause the carrier to avoid actions that will unnecessarily delay delivery or unnecessarily result in increased radiation levels or radiation exposures to transport workers or members of the general public.

§ 71.51 Additional requirements for Type B packages.




(a) A Type B package, in addition to satisfying the requirements of §§ 71.41 through 71.47, must be designed, constructed, and prepared for shipment so that under the tests specified in:

(1) Section 71.71 ("Normal conditions of transport"), there would be no loss or dispersal of radioactive contents--as demonstrated to a sensitivity of 10-6 A2 per hour, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging; and

(2) Section 71.73 ("Hypothetical accident conditions"), there would be no escape of krypton-85 exceeding 10 A2 in 1 week, no escape of other radioactive material exceeding a total amount A2 in 1 week, and no external radiation dose rate exceeding 10 mSv/h (1 rem/h) at 1 m (40 in) from the external surface of the package.

(b) Where mixtures of different radionuclides are present, the provisions of appendix A, paragraph IV of this part shall apply, except that for Krypton-85, an effective A2 value equal to 10 A2 may be used.

(c) Compliance with the permitted activity release limits of paragraph (a) of this section may not depend on filters or on a mechanical cooling system.

(d) For packages which contain radioactive contents with activity greater than 105 A2, the requirements of § 71.61 must be met.

[60 FR 50264, Sept. 28, 1995 as amended at 69 FR 3794, Jan. 26, 2004]

§ 71.53 [Reserved]




[62 FR 5913, Feb. 10, 1997; 69 FR 3794, January 26, 2004]

§ 71.55 General requirements for fissile material packages.




(a) A package used for the shipment of fissile material must be designed and constructed in accordance with §§ 71.41 through 71.47. When required by the total amount of radioactive material, a package used for the shipment of fissile material must also be designed and constructed in accordance with § 71.51.

(b) Except as provided in paragraph (c) or (g) of this section, a package used for the shipment of fissile material must be so designed and constructed and its contents so limited that it would be subcritical if water were to leak into the containment system, or liquid contents were to leak out of the containment system so that, under the following conditions, maximum reactivity of the fissile material would be attained:

(1) The most reactive credible configuration consistent with the chemical and physical form of the material;

(2) Moderation by water to the most reactive credible extent; and

(3) Close full reflection of the containment system by water on all sides, or such greater reflection of the containment system as may additionally be provided by the surrounding material of the packaging.

(c) The Commission may approve exceptions to the requirements of paragraph (b) of this section if the package incorporates special design features that ensure that no single packaging error would permit leakage, and if appropriate measures are taken before each shipment to ensure that the containment system does not leak.

(d) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in § 71.71 ("Normal conditions of transport") --

(1) The contents would be subcritical;

(2) The geometric form of the package contents would not be substantially altered;

(3) There would be no leakage of water into the containment system unless, in the evaluation of undamaged packages under § 71.59(a)(1), it has been assumed that moderation is present to such an extent as to cause maximum reactivity consistent with the chemical and physical form of the material; and

(4) There will be no substantial reduction in the effectiveness of the packaging, including:

(i) No more than 5 percent reduction in the total effective volume of the packaging on which nuclear safety is assessed;

(ii) No more than 5 percent reduction in the effective spacing between the fissile contents and the outer surface of the packaging; and

(iii) No occurrence of an aperture in the outer surface of the packaging large enough to permit the entry of a 10 cm (4 in) cube.

(e) A package used for the shipment of fissile material must be so designed and constructed and its contents so limited that under the tests specified in § 71.73 ("Hypothetical accident conditions"), the package would be subcritical. For this determination, it must be assumed that:

(1) The fissile material is in the most reactive credible configuration consistent with the damaged condition of the package and the chemical and physical form of the contents;

(2) Water moderation occurs to the most reactive credible extent consistent with the damaged condition of the package and the chemical and physical form of the contents; and

(3) There is full reflection by water on all sides, as close as is consistent with the damaged condition of the package.

(f) For fissile material package designs to be transported by air:

(1) The package must be designed and constructed, and its contents limited so that it would be subcritical, assuming reflection by 20 cm (7.9 in) of water but no water inleakage, when subjected to sequential application of:

(i) The free drop test in § 71.73(c)(1);

(ii) The crush test in § 71.73(c)(2);

(iii) A puncture test, for packages of 250 kg or more, consisting of a free drop of the specimen through a distance of 3 m (120 in) in a position for which maximum damage is expected at the conclusion of the test sequence, onto the upper end of a solid, vertical, cylindrical, mild steel probe mounted on an essentially unyielding, horizontal surface. The probe must be 20 cm (7.9 in) in diameter, with the striking end forming the frustum of a right circular cone with the dimensions of 30 cm height, 2.5 cm top diameter, and a top edge rounded to a radius of not more than 6 mm (0.25 in). For packages less than 250 kg, the puncture test must be the same, except that a 250 kg probe must be dropped onto the specimen which must be placed on the surface; and

(iv) The thermal test in § 71.73(c)(4), except that the duration of the test must be 60 minutes.

(2) The package must be designed and constructed, and its contents limited, so that it would be subcritical, assuming reflection by 20 cm (7.9 in) of water but no water inleakage, when subjected to an impact on an unyielding surface at a velocity of 90 m/s normal to the surface, at such orientation so as to result in maximum damage. A separate, undamaged specimen can be used for this evaluation.

(3) Allowance may not be made for the special design features in paragraph (c) of this section, unless water leakage into or out of void spaces is prevented following application of the tests in paragraphs (f)(1) and (f)(2) of this section, and subsequent application of the immersion test in § 71.73(c)(5).

(g) Packages containing uranium hexafluoride only are excepted from the requirements of paragraph (b) of this section provided that:

(1) Following the tests specified in § 71.73 ("Hypothetical accident conditions"), there is no physical contact between the valve body and any other component of the packaging, other than at its original point of attachment, and the valve remains leak tight;

(2) There is an adequate quality control in the manufacture, maintenance, and repair of packagings;

(3) Each package is tested to demonstrate closure before each shipment; and

(4) The uranium is enriched to not more than 5 weight percent uranium-235.

[60 FR 50264, Sept. 28, 1995; 61 FR 28724, June 6, 1996; 69 FR 3794, Jan. 26, 2004]

§ 71.57 [Reserved]




§ 71.59 Standards for arrays of fissile material packages.




(a) A fissile material package must be controlled by either the shipper or the carrier during transport to assure that an array of such packages remains subcritical. To enable this control, the designer of a fissile material package shall derive a number "N" based on all the following conditions being satisfied, assuming packages are stacked together in any arrangement and with close full reflection on all sides of the stack by water:

(1) Five times "N" undamaged packages with nothing between the packages would be subcritical;

(2) Two times "N" damaged packages, if each package were subjected to the tests specified in § 71.73 ("Hypothetical accident conditions") would be subcritical with optimum interspersed hydrogenous moderation; and

(3) The value of "N" cannot be less than 0.5.

(b) The CSI must be determined by dividing the number 50 by the value of "N" derived using the procedures specified in paragraph (a) of this section. The value of the CSI may be zero provided that an unlimited number of packages are subcritical, such that the value of "N" is effectively equal to infinity under the procedures specified in paragraph (a) of this section. Any CSI greater than zero must be rounded up to the first decimal place.

(c) For a fissile material package which is assigned a CSI value--

(1) Less than or equal to 50, that package may be shipped by a carrier in a nonexclusive use conveyance, provided the sum of the CSIs is limited to less than or equal to 50.

(2) Less than or equal to 50, that package may be shipped by a carrier in an exclusive use conveyance, provided the sum of the CSIs is limited to less than or equal to 100.

(3) Greater than 50, that package must be shipped by a carrier in an exclusive use conveyance, provided the sum of the CSIs is limited to less than or equal to 100.

[69 FR 3795, Jan. 26, 2004]

§ 71.61 Special requirements for Type B packages containing more than 105A2.




A Type B package containing more than 105A2 must be designed so that its undamaged containment system can withstand an external water pressure of 2 MPa (290 psi) for a period of not less than 1 hour without collapse, buckling, or inleakage of water.

[69 FR 3795, Jan. 26, 2004]

§ 71.63 Special requirement for plutonium shipments.




Shipments containing plutonium must be made with the contents in solid form, if the contents contain greater than 0.74 TBq (20 Ci) of plutonium.

[69 FR 3795, Jan. 26, 2004]

§ 71.64 Special requirements for plutonium air shipments.




(a) A package for the shipment of plutonium by air subject to § 71.88(a)(4), in addition to satisfying the requirements of §§ 71.41 through 71.63, as applicable, must be designed, constructed, and prepared for shipment so that under the tests specified in --

(1) Section 71.74 ("Accident conditions for air transport of plutonium") --

(i) The containment vessel would not be ruptured in its post-tested condition, and the package must provide a sufficient degree of containment to restrict accumulated loss of plutonium contents to not more than an A2 quantity in a period of 1 week;

(ii) The external radiation level would not exceed 10 mSv/h (1 rem/h) at a distance of 1 m (40 in) from the surface of the package in its post-tested condition in air; and

(iii) A single package and an array of packages are demonstrated to be subcritical in accordance with this part, except that the damaged condition of the package must be considered to be that which results from the plutonium accident tests in § 71.74, rather than the hypothetical accident tests in § 71.73; and

(2) Section 71.74(c), there would be no detectable leakage of water into the containment vessel of the package.

(b) With respect to the package requirements of paragraph (a), there must be a demonstration or analytical assessment showing that --

(1) The results of the physical testing for package qualification would not be adversely affected to a significant extent by --

(i) The presence, during the tests, of the actual contents that will be transported in the package; and

(ii) Ambient water temperatures ranging from 0.6°C (+33°F) to 38°C (+100°F) for those qualification tests involving water, and ambient atmospheric temperatures ranging from -40°C (-40°F) to +54°C (+130°F) for the other qualification tests.

(2) The ability of the package to meet the acceptance standards prescribed for the accident condition sequential tests would not be adversely affected if one or more tests in the sequence were deleted.

§ 71.65 Additional requirements.




The Commission may, by rule, regulation, or order, impose requirements on any licensee, in addition to those established in this part, as it deems necessary or appropriate to protect public health or to minimize danger to life or property.

Subpart F--Package, Special Form, and LSA-III Tests2




§ 71.71 Normal conditions of transport.


(a) Evaluation. Evaluation of each package design under normal conditions of transport must include a determination of the effect on that design of the conditions and tests specified in this section. Separate specimens may be used for the free drop test, the compression test, and the penetration test, if each specimen is subjected to the water spray test before being subjected to any of the other tests.

(b) Initial conditions. With respect to the initial conditions for the tests in this section, the demonstration of compliance with the requirements of this part must be based on the ambient temperature preceding and following the tests remaining constant at that value between -29°C (-20°F) and +38°C (+100°F) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be considered to be the maximum normal operating pressure, unless a lower internal pressure consistent with the ambient temperature considered to precede and follow the tests is more unfavorable.

(c) Conditions and tests.

(1) Heat. An ambient temperature of 38°C (100°F) in still air, and insolation according to the following table:

INSOLATION DATA

Form and location of surface Total insolation for a 12-hour period (g cal/cm2)
Flat surfaces transported horizontally;
  Base None
  Other surfaces 800
Flat surfaces not transported horizontally 200
Curved surfaces 400

(2) Cold. An ambient temperature of -40°C (-40°F) in still air and shade.

(3) Reduced external pressure. An external pressure of 25 kPa (3.5 lbf/in2) absolute.

(4) Increased external pressure. An external pressure of 140 kPa (20 lbf/in2) absolute.

(5) Vibration. Vibration normally incident to transport.

(6) Water spray. A water spray that simulates exposure to rainfall of approximately 5 cm/h (2 in/h) for at least 1 hour.

(7) Free drop. Between 1.5 and 2.5 hours after the conclusion of the water spray test, a free drop through the distance specified below onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.

Criteria for Free Drop Test (Weight/Distance)

Package weight Free drop distance
Kilograms
(Pounds)
Meters
(Feet)
Less than 5,000 (Less than 11,000)
1.2
(4)
5,000 to 10,000 (11,000 to 22,000)
0.9
(3)
10,000 to 15,000 (22,000 to 33,100)
0.6
(2)
More than 15,000 (More than 33,100)
0.3
(1)

(8) Corner drop. A free drop onto each corner of the package in succession, or in the case of a cylindrical package onto each quarter of each rim, from a height of 0.3 m (1 ft) onto a flat, essentially unyielding, horizontal surface. This test applies only to fiberboard, wood, or fissile material rectangular packages not exceeding 50 kg (110 lbs) and fiberboard, wood, or fissile material cylindrical packages not exceeding 100 kg (220 lbs).

(9) Compression. For packages weighing up to 5000 kg (11,000 lbs), the package must be subjected, for a period of 24 hours, to a compressive load applied uniformly to the top and bottom of the package in the position in which the package would normally be transported. The compressive load must be the greater of the following:

(i) The equivalent of 5 times the weight of the package; or

(ii) The equivalent of 13 kPa (2 lbf/in2) multiplied by the vertically projected area of the package.

(10) Penetration. Impact of the hemispherical end of a vertical steel cylinder of 3.2 cm (1.25 in) diameter and 6 kg (13 lbs) mass, dropped from a height of 1 m (40 in) onto the exposed surface of the package that is expected to be most vulnerable to puncture. The long axis of the cylinder must be perpendicular to the package surface.

2 The package standards related to the tests in this subpart are contained in subpart E of this part.

§ 71.73 Hypothetical accident conditions.




(a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential application of the tests specified in this section, in the order indicated, to determine their cumulative effect on a package or array of packages. An undamaged specimen may be used for the water immersion tests specified in paragraph (c)(6) of this section.

(b) Test conditions. With respect to the initial conditions for the tests, except for the water immersion tests, to demonstrate compliance with the requirements of this part during testing, the ambient air temperature before and after the tests must remain constant at that value between -29°C (-20°F) and +38°C (+100°F) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be the maximum normal operating pressure, unless a lower internal pressure, consistent with the ambient temperature assumed to precede and follow the tests, is more unfavorable.

(c) Tests. Tests for hypothetical accident conditions must be conducted as follows:

(1) Free Drop. A free drop of the specimen through a distance of 9 m (30 ft) onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.

(2) Crush. Subjection of the specimen to a dynamic crush test by positioning the specimen on a flat, essentially unyielding horizontal surface so as to suffer maximum damage by the drop of a 500-kg (1100-lb) mass from 9 m (30 ft) onto the specimen. The mass must consist of a solid mild steel plate 1 m (40 in) by 1 m (40 in) and must fall in a horizontal attitude. The crush test is required only when the specimen has a mass not greater than 500 kg (1100 lb), an overall density not greater than 1000 kg/m3 (62.4 lb/ft3) based on external dimension, and radioactive contents greater than 1000 A2 not as special form radioactive material. For packages containing fissile material, the radioactive contents greater than 1000 A2 criterion does not apply.

(3) Puncture. A free drop of the specimen through a distance of 1 m (40 in) in a position for which maximum damage is expected, onto the upper end of a solid, vertical, cylindrical, mild steel bar mounted on an essentially unyielding, horizontal surface. The bar must be 15 cm (6 in) in diameter, with the top horizontal and its edge rounded to a radius of not more than 6 mm (0.25 in), and of a length as to cause maximum damage to the package, but not less than 20 cm (8 in) long. The long axis of the bar must be vertical.

(4) Thermal. Exposure of the specimen fully engulfed, except for a simple support system, in a hydrocarbon fuel/air fire of sufficient extent, and in sufficiently quiescent ambient conditions, to provide an average emissivity coefficient of at least 0.9, with an average flame temperature of at least 800°C (1475°F) for a period of 30 minutes, or any other thermal test that provides the equivalent total heat input to the package and which provides a time averaged environmental temperature of 800°C. The fuel source must extend horizontally at least 1 m (40 in), but may not extend more than 3 m (10 ft), beyond any external surface of the specimen, and the specimen must be positioned 1 m (40 in) above the surface of the fuel source. For purposes of calculation, the surface absorptivity coefficient must be either that value which the package may be expected to possess if exposed to the fire specified or 0.8, whichever is greater; and the convective coefficient must be that value which may be demonstrated to exist if the package were exposed to the fire specified. Artificial cooling may not be applied after cessation of external heat input, and any combustion of materials of construction, must be allowed to proceed until it terminates naturally.

(5) Immersion--fissile material. For fissile material subject to § 71.55, in those cases where water inleakage has not been assumed for criticality analysis, immersion under a head of water of at least 0.9 m (3 ft) in the attitude for which maximum leakage is expected.

(6) Immersion--all packages. A separate, undamaged specimen must be subjected to water pressure equivalent to immersion under a head of water of at least 15 m (50 ft). For test purposes, an external pressure of water of 150 kPa (21.7 lbf/in2) gauge is considered to meet these conditions.

[69 FR 3795, Jan. 26, 2004]

§ 71.74 Accident conditions for air transport of plutonium.




(a) Test conditions--Sequence of tests. A package must be physically tested to the following conditions in the order indicated to determine their cumulative effect.

(1) Impact at a velocity of not less than 129 m/sec (422 ft/sec) at a right angle onto a flat, essentially unyielding, horizontal surface, in the orientation (e.g., side, end, corner) expected to result in maximum damage at the conclusion of the test sequence.

(2) A static compressive load of 31,800 kg (70,000 lbs) applied in the orientation expected to result in maximum damage at the conclusion of the test sequence. The force on the package must be developed between a flat steel surface and a 5 cm (2 in) wide, straight, solid, steel bar. The length of the bar must be at least as long as the diameter of the package, and the longitudinal axis of the bar must be parallel to the plane of the flat surface. The load must be applied to the bar in a manner that prevents any members or devices used to support the bar from contacting the package.

(3) Packages weighing less than 227 kg (500 lbs) must be placed on a flat, essentially unyielding, horizontal surface, and subjected to a weight of 227 kg (500 lbs) falling from a height of 3 m (10 ft) and striking in the position expected to result in maximum damage at the conclusion of the test sequence. The end of the weight contacting the package must be a solid probe made of mild steel. The probe must be the shape of the frustum of a right circular cone, 30 cm (12 in) long, 20 cm (8 in) in diameter at the base, and 2.5 cm (1 in) in diameter at the end. The longitudinal axis of the probe must be perpendicular to the horizontal surface. For packages weighing 227 kg (500 lbs) or more, the base of the probe must be placed on a flat, essentially unyielding horizontal surface, and the package dropped from a height of 3 m (10 ft) onto the probe, striking in the position expected to result in maximum damage at the conclusion of the test sequence.

(4) The package must be firmly restrained and supported such that its longitudinal axis is inclined approximately 45° to the horizontal. The area of the package that made first contact with the impact surface in paragraph (a)(1) of this section must be in the lowermost position. The package must be struck at approximately the center of its vertical projection by the end of a structural steel angle section falling from a height of at least 46 m (150 ft). The angle section must be at least 1.8 m (6 ft) in length with equal legs at least 13 cm (5 in) long and 1.3 cm (0.5 in) thick. The angle section must be guided in such a way as to fall end-on, without tumbling. The package must be rotated approximately 90° about its longitudinal axis and struck by the steel angle section falling as before.

(5) The package must be exposed to luminous flames from a pool fire of JP-4 or JP-5 aviation fuel for a period of at least 60 minutes. The luminous flames must extend an average of at least 0.9 m (3 ft) and no more than 3 m (10 ft) beyond the package in all horizontal directions. The position and orientation of the package in relation to the fuel must be that which is expected to result in maximum damage at the conclusion of the test sequence. An alternate method of thermal testing may be substituted for this fire test, provided that the alternate test is not of shorter duration and would not result in a lower heating rate to the package. At the conclusion of the thermal test, the package must be allowed to cool naturally or must be cooled by water sprinkling, whichever is expected to result in maximum damage at the conclusion of the test sequence.

(6) Immersion under at least 0.9 m (3 ft) of water.

(b) Individual free-fall impact test.

(1) An undamaged package must be physically subjected to an impact at a velocity not less than the calculated terminal free-fall velocity, at mean sea level, at a right angle onto a flat, essentially unyielding, horizontal surface, in the orientation (e.g., side, end, corner) expected to result in maximum damage.

(2) This test is not required if the calculated terminal free-fall velocity of the package is less than 129 m/sec (422 ft/sec), or if a velocity not less than either 129 m/sec (422 ft/sec) or the calculated terminal free-fall velocity of the package is used in the sequential test of paragraph (a)(1) of this section.

(c) Individual deep submersion test. An undamaged package must be physically submerged and physically subjected to an external water pressure of at least 4 MPa (600 lbs/in2).

§ 71.75 Qualification of special form radioactive material.




(a) Special form radioactive materials must meet the test requirements of paragraph (b) of this section. Each solid radioactive material or capsule specimen to be tested must be manufactured or fabricated so that it is representative of the actual solid material or capsule that will be transported, with the proposed radioactive content duplicated as closely as practicable. Any differences between the material to be transported and the test material, such as the use of non-radioactive contents, must be taken into account in determining whether the test requirements have been met. In addition:

(1) A different specimen may be used for each of the tests;

(2) The specimen may not break or shatter when subjected to the impact, percussion, or bending tests;

(3) The specimen may not melt or disperse when subjected to the heat test;

(4) After each test, leaktightness or indispersibility of the specimen must be determined by a method no less sensitive than the leaching assessment procedure prescribed in paragraph (c) of this section. For a capsule resistant to corrosion by water, and which has an internal void volume greater than 0.1 milliliter, an alternative to the leaching assessment is a demonstration of leaktightness of x10-4 torr-liter/s (1.3xx10-4 atm-cm3/s) based on air at 25°C (77°F) and one atmosphere differential pressure for solid radioactive content, or x10-6 torr-liter/s (1.30xx10-6 atm-cm3/s) for liquid or gaseous radioactive content; and

(5) A specimen that comprises or simulates radioactive material contained in a sealed capsule need not be subjected to the leaktightness procedure specified in this section, provided it is alternatively subjected to any of the tests prescribed in ISO/TR4826-1979(E), "Sealed radioactive sources leak test methods" which is available from the American National Standards Institute, 1430 Broadway, New York, N.Y. 10018.

(b) Test methods. (1) Impact Test. The specimen must fall onto the target from a height of 9 m (30 ft) or greater in the orientation expected to result in maximum damage. The target must be a flat, horizontal surface of such mass and rigidity that any increase in its resistance to displacement or deformation, on impact by the specimen, would not significantly increase the damage to the specimen.

(2) Percussion Test. (i) The specimen must be placed on a sheet of lead that is supported by a smooth solid surface, and struck by the flat face of a steel billet so as to produce an impact equivalent to that resulting from a free drop of 1.4 kg (3 lbs) through 1 m (40 in);

(ii) The flat face of the billet must be 25 millimeters (mm) (1 inch) in diameter with the edges rounded off to a radius of 3 mm±0.3 mm(.12 in±0.012 in);

(iii) The lead must be hardness number 3.5 to 4.5 on the Vickers scale and thickness 25 mm (1 in) or greater, and must cover an area greater than that covered by the specimen;

(iv) A fresh surface of lead must be used for each impact; and

(v) The billet must strike the specimen so as to cause maximum damage.

(3) Bending test. (i) This test applies only to long, slender sources with a length of 10 cm (4 inches) or greater and a length to width ratio of 10 or greater;

(ii) The specimen must be rigidly clamped in a horizontal position so that one half of its length protrudes from the face of the clamp;

(iii) The orientation of the specimen must be such that the specimen will suffer maximum damage when its free end is struck by the flat face of a steel billet;

(iv) The billet must strike the specimen so as to produce an impact equivalent to that resulting from a free vertical drop of 1.4 kg (3 lbs) through 1 m (40 in); and

(v) The flat face of the billet must be 25 mm (1 inch) in diameter with the edges rounded off to a radius of 3 mm±0.3 mm (.12 in±0.012 in).

(4) Heat test. The specimen must be heated in air to a temperature of not less than 800°C (1475°F), held at that temperature for a period of 10 minutes, and then allowed to cool.

(c) Leaching assessment methods. (1) For indispersible solid material —

(i) The specimen must be immersed for 7 days in water at ambient temperature. The water must have a pH of 6-8 and a maximum conductivity of 10 micromho per centimeter at 20° (68°F);

(ii) The water with specimen must then be heated to a temperature of 50°C±5°C (122°F±9°F) and maintained at this temperature for 4 hours.

(iii) The activity of the water must then be determined;

(iv) The specimen must then be stored for at least 7 days in still air of relative humidity not less than 90 percent at 30°C (86°F);

(v) The specimen must then be immersed in water under the same conditions as in paragraph (c)(1)(i) of this section, and the water with specimen must be heated to 50°C±5°C (122°F±9°F) and maintained at that temperature for 4 hours;

(vi) The activity of the water must then be determined. The sum of the activities determined here and in paragraph (c)(1)(iii) of this section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie (µCi)).

(2) For encapsulated material —

(i) The specimen must be immersed in water at ambient temperature. The water must have a pH of 6-8 and a maximum conductivity of 10 micromho per centimeter;

(ii) The water and specimen must be heated to a temperature of 50°C±5°C (122°F±9°F) and maintained at this temperature for 4 hours;

(iii) The activity of the water must then be determined;

(iv) The specimen must then be stored for at least 7 days in still air at a temperature of 30°C (86°F) or greater;

(v) The process in paragraph (c)(2)(i), (ii), and (iii) of this section must be repeated; and

(vi) The activity of the water must then be determined. The sum of the activities determined here and in paragraph (c)(2)(iii) of this section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie (Ci)).

(d) A specimen that comprises or simulates radioactive material contained in a sealed capsule need not be subjected to —

(1) The impact test and the percussion test of this section, provided that the specimen is alternatively subjected to the Class 4 impact test prescribed in ISO 2919-1980(e), "Sealed Radioactive Sources Classification" (see § 71.75(a)(5) for statement of availability); and

(2) The heat test of this section, provided the specimen is alternatively subjected to the Class 6 temperature test specified in the International Organization for Standardization document ISO 2919-1980(e), "Sealed Radioactive Sources Classification."

§ 71.77 Qualification of LSA-III Material




(a) LSA-III material must meet the test requirements of paragraph (b) of this section. Any differences between the specimen to be tested and the material to be transported must be taken into account in determining whether the test requirements have been met.

(b) Leaching Test. (1) The specimen, representing no less than the entire contents of the package, must be immersed for 7 days in water at ambient temperature;

(2) The volume of water to be used in the test must be sufficient to ensure that at the end of the test period the free volume of the unabsorbed and unreacted water remaining will be at least 10% of the volume of the specimen itself;

(3) The water must have an initial pH of 6-8 and a maximum conductivity 10 micromho/cm at 20°C (68°F); and

(4) The total activity of the free volume of water must be measured following the 7 day immersion test and must not exceed 0.1 A2.

Subpart G--Operating Controls and Procedures




§ 71.81 Applicability of operating controls and procedures.


A licensee subject to this part, who, under a general or specific license, transports licensed material or delivers licensed material to a carrier for transport, shall comply with the requirements of this subpart G, with the quality assurance requirements of subpart H of this part, and with the general provisions of subpart A of this part.

§ 71.83 Assumptions as to unknown properties.




When the isotopic abundance, mass, concentration, degree of irradiation, degree of moderation, or other pertinent property of fissile material in any package is not known, the licensee shall package the fissile material as if the unknown properties have credible values that will cause the maximum neutron multiplication.

§ 71.85 Preliminary determinations.




Before the first use of any packaging for the shipment of licensed material --

(a) The licensee shall ascertain that there are no cracks, pinholes, uncontrolled voids, or other defects that could significantly reduce the effectiveness of the packaging;

(b) Where the maximum normal operating pressure will exceed 35 kPa (5 lbf/in2) gauge, the licensee shall test the containment system at an internal pressure at least 50 percent higher than the maximum normal operating pressure, to verify the capability of that system to maintain its structural integrity at that pressure; and

(c) The licensee shall conspicuously and durably mark the packaging with its model number, serial number, gross weight, and a package identification number assigned by NRC. Before applying the model number, the licensee shall determine that the packaging has been fabricated in accordance with the design approved by the Commission.

§ 71.87 Routine determinations.




Before each shipment of licensed material, the licensee shall ensure that the package with its contents satisfies the applicable requirements of this part and of the license. The licensee shall determine that --

(a) The package is proper for the contents to be shipped;

(b) The package is in unimpaired physical condition except for superficial defects such as marks or dents;

(c) Each closure device of the packaging, including any required gasket, is properly installed and secured and free of defects;

(d) Any system for containing liquid is adequately sealed and has adequate space or other specified provision for expansion of the liquid;

(e) Any pressure relief device is operable and set in accordance with written procedures;

(f) The package has been loaded and closed in accordance with written procedures;

(g) For fissile material, any moderator or neutron absorber, if required, is present and in proper condition;

(h) Any structural part of the package that could be used to lift or tie down the package during transport is rendered inoperable for that purpose, unless it satisfies the design requirements of § 71.45;

(i) The level of non-fixed (removable) radioactive contamination on the external surfaces of each package offered for shipment is as low as reasonably achievable, and within the limits specified in DOT regulations in 49 CFR 173.443;

(j) External radiation levels around the package and around the vehicle, if applicable, will not exceed the limits specified in § 71.47 at any time during transportation; and

(k) Accessible package surface temperatures will not exceed the limits specified in § 71.43(g) at any time during transportation.

§ 71.88 Air transport of plutonium.




(a) Notwithstanding the provisions of any general licenses and notwithstanding any exemptions stated directly in this part or included indirectly by citation of 49 CFR chapter I, as may be applicable, the licensee shall assure that plutonium in any form, whether for import, export, or domestic shipment, is not transported by air or delivered to a carrier for air transport unless:

(1) The plutonium is contained in a medical device designed for individual human application; or

(2) The plutonium is contained in a material in which the specific activity is less than or equal to the activity concentration values for plutonium specified in Appendix A, Table A-2, of this part, and in which the radioactivity is essentially uniformly distributed; or

(3) The plutonium is shipped in a single package containing no more than an A2 quantity of plutonium in any isotope or form, and is shipped in accordance with § 71.5; or

(4) The plutonium is shipped in a package specifically authorized for the shipment of plutonium by air in the Certificate of Compliance for that package issued by the Commission.

(b) Nothing in paragraph (a) of this section is to be interpreted as removing or diminishing the requirements of § 73.24 of this chapter.

(c) For a shipment of plutonium by air which is subject to paragraph (a)(4) of this section, the licensee shall, through special arrangement with the carrier, require compliance with 49 CFR 175.704, U.S. Department of Transportation regulations applicable to the air transport of plutonium.

[69 FR 3795, Jan. 26, 2004]

§ 71.89 Opening instructions.




Before delivery of a package to a carrier for transport, the licensee shall ensure that any special instructions needed to safely open the package have been sent to, or otherwise made available to, the consignee for the consignee's use in accordance with 10 CFR 20.1906(e).

§ 71.91 Records.




(a) Each licensee shall maintain, for a period of 3 years after shipment, a record of each shipment of licensed material not exempt under § 71.10, showing where applicable --

(1) Identification of the packaging by model number and serial number;

(2) Verification that there are no significant defects in the packaging, as shipped;

(3) Volume and identification of coolant;

(4) Type and quantity of licensed material in each package, and the total quantity of each shipment;

(5) For each item of irradiated fissile material --

(i) Identification by model number and serial number;

(ii) Irradiation and decay history to the extent appropriate to demonstrate that its nuclear and thermal characteristics comply with license conditions; and

(iii) Any abnormal or unusual condition relevant to radiation safety;

(6) Date of the shipment;

(7) For fissile packages and for Type B packages, any special controls exercised;

(8) Name and address of the transferee;

(9) Address to which the shipment was made; and

(10) Results of the determinations required by § 71.87 and by the conditions of the package approval.

(b) Each certificate holder shall maintain, for a period of 3 years after the life of the packaging to which they apply, records identifying the packaging by model number, serial number, and date of manufacture.

(c) The licensee, certificate holder, and an applicant for a CoC, shall make available to the Commission for inspection, upon reasonable notice, all records required by this part. Records are only valid if stamped, initialed, or signed and dated by authorized personnel, or otherwise authenticated.

(d) The licensee, certificate holder, and an applicant for a CoC shall maintain sufficient written records to furnish evidence of the quality of packaging. The records to be maintained include results of the determinations required by § 71.85; design, fabrication, and assembly records; results of reviews, inspections, tests, and audits; results of monitoring work performance and materials analyses; and results of maintenance, modification, and repair activities. Inspection, test, and audit records must identify the inspector or data recorder, the type of observation, the results, the acceptability, and the action taken in connection with any deficiencies noted. These records must be retained for 3 years after the life of the packaging to which they apply.

[69 FR 3796, Jan. 26, 2004]

§ 71.93 Inspection and tests.




(a) The licensee, certificate holder, and applicant for a CoC shall permit the Commission, at all reasonable times, to inspect the licensed material, packaging, premises, and facilities in which the licensed material or packaging is used, provided, constructed, fabricated, tested, stored, or shipped.

(b) The licensee, certificate holder, and applicant for a CoC shall perform, and permit the Commission to perform, any tests the Commission deems necessary or appropriate for the administration of the regulations in this chapter.

(c) The certificate holder and applicant for a CoC shall notify the NRC, in accordance with § 71.1, 45 days in advance of starting fabrication of the first packaging under a CoC. This paragraph applies to any packaging used for the shipment of licensed material which has either--

(1) A decay heat load in excess of 5 kW; or

(2) A maximum normal operating pressure in excess of 103 kPa (15 lbf/in2) gauge.

[69 FR 3796, Jan. 26, 2004]

§ 71.95 Reports.




(a) The licensee, after requesting the certificate holder's input, shall submit a written report to the Commission of--

(1) Instances in which there is a significant reduction in the effectiveness of any NRC-approved Type B or Type AF packaging during use; or

(2) Details of any defects with safety significance in any NRC-approved Type B or fissile material packaging, after first use.

(3) Instances in which the conditions of approval in the Certificate of Compliance were not observed in making a shipment.

(b) The licensee shall submit a written report to the Commission of instances in which the conditions in the certificate of compliance were not followed during a shipment.

(c) Each licensee shall submit, in accordance with § 71.1, a written report required by paragraph (a) or (b) of this section within 60 days of the event or discovery of the event. The licensee shall also provide a copy of each report submitted to the NRC to the applicable certificate holder. Written reports prepared under other regulations may be submitted to fulfill this requirement if the reports contain all the necessary information, and the appropriate distribution is made. Using an appropriate method listed in § 71.1(a), the licensee shall report to: ATTN: Document Control Desk, Director, Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards. These written reports must include the following:

(1) A brief abstract describing the major occurrences during the event, including all component or system failures that contributed to the event and significant corrective action taken or planned to prevent recurrence.

(2) A clear, specific, narrative description of the event that occurred so that knowledgeable readers conversant with the requirements of part 71, but not familiar with the design of the packaging, can understand the complete event. The narrative description must include the following specific information as appropriate for the particular event.

(i) Status of components or systems that were inoperable at the start of the event and that contributed to the event;

(ii) Dates and approximate times of occurrences;

(iii) The cause of each component or system failure or personnel error, if known;

(iv) The failure mode, mechanism, and effect of each failed component, if known;

(v) A list of systems or secondary functions that were also affected for failures of components with multiple functions;

(vi) The method of discovery of each component or system failure or procedural error;

(vii) For each human performance-related root cause, a discussion of the cause(s) and circumstances;

(viii) The manufacturer and model number (or other identification) of each component that failed during the event; and

(ix) For events occurring during use of a packaging, the quantities and chemical and physical form(s) of the package contents.

(3) An assessment of the safety consequences and implications of the event. This assessment must include the availability of other systems or components that could have performed the same function as the components and systems that failed during the event.

(4) A description of any corrective actions planned as a result of the event, including the means employed to repair any defects, and actions taken to reduce the probability of similar events occurring in the future.

(5) Reference to any previous similar events involving the same packaging that are known to the licensee or certificate holder.

(6) The name and telephone number of a person within the licensee's organization who is knowledgeable about the event and can provide additional information.

(7) The extent of exposure of individuals to radiation or to radioactive materials without identification of individuals by name.

(d) Report legibility. The reports submitted by licensees and/or certificate holders under this section must be of sufficient quality to permit reproduction and micrographic processing.

[60 FR 50264, Sept. 28, 1995, as amended at 67 FR 3585, Jan. 25, 2002; 68 FR 58818, Oct. 10, 2003; 69 FR 3796, Jan. 26, 2004; 75 FR 73945, Nov. 30, 2010]

§ 71.97 Advance notification of shipment of irradiated reactor fuel and nuclear waste.




(a)(1) As specified in paragraphs (b), (c), and (d) of this section, each licensee shall provide advance notification to the governor of a State, or the governor’s designee, of the shipment of licensed material, within or across the boundary of the State, before the transport, or delivery to a carrier, for transport, of licensed material outside the confines of the licensee’s plant or other place of use or storage.

(2) As specified in paragraphs (b), (c), and (d) of this section, after June 11, 2013, each licensee shall provide advance notification to the Tribal official of participating Tribes referenced in paragraph (c)(3)(iii) of this section, or the official’s designee, of the shipment of licensed material, within or across the boundary of the Tribe’s reservation, before the transport, or delivery to a carrier, for transport, of licensed material outside the confines of the licensee’s plant or other place of use or storage.

(b) Advance notification is also required under this section for the shipment of licensed material, other than irradiated fuel, meeting the following three conditions:

(1) The licensed material is required by this part to be in Type B packaging for transportation;

(2) The licensed material is being transported to or across a State boundary en route to a disposal facility or to a collection point for transport to a disposal facility; and

(3) The quantity of licensed material in a single package exceeds the least of the following:

(i) 3000 times the A1 value of the radionuclides as specified in appendix A, Table A–1 for special form radioactive material;

(ii) 3000 times the A2 value of the radionuclides as specified in appendix A, Table A–1 for normal form radioactive material; or

(iii) 1000 TBq (27,000 Ci).

(c) Procedures for submitting advance notification. (1) The notification must be made in writing to:

(i) The office of each appropriate governor or governor’s designee;

(ii) The office of each appropriate Tribal official or Tribal official’s designee; and

(iii) The Director, Division of Security Policy, Office of Nuclear Security and Incident Response.

(2) A notification delivered by mail must be postmarked at least 7 days before the beginning of the 7-day period during which departure of the shipment is estimated to occur.

(3) A notification delivered by any other means than mail must reach the office of the governor or of the governor’s designee or the Tribal official or Tribal official’s designee at least 4 days before the beginning of the 7-day period during which departure of the shipment is estimated to occur.

(i) A list of the names and mailing addresses of the governors’ designees receiving advance notification of transportation of nuclear waste was published in the Federal Register on June 30, 1995 (60 FR 34306).

(ii) The list of governor’s designees and Tribal official’s designees of participating Tribes will be published annually in the Federal Register on or about June 30th to reflect any changes in information.

(iii) A list of the names and mailing addresses of the governors’ designees and Tribal officials’ designees of participating Tribes is available on request from the Director, Division of Intergovernmental Liaison and Rulemaking, Office of Federal and State Materials and Environmental Management Programs, U.S. Nuclear Regulatory Commission, Washington, DC 20555–0001.

(4) The licensee shall retain a copy of the notification as a record for 3 years.

(d) Information to be furnished in advance notification of shipment. Each advance notification of shipment of irradiated reactor fuel or nuclear waste must contain the following information:

(1) The name, address, and telephone number of the shipper, carrier, and receiver of the irradiated reactor fuel or nuclear waste shipment;

(2) A description of the irradiated reactor fuel or nuclear waste contained in the shipment, as specified in the regulations of DOT in 49 CFR 172.202 and 172.203(d);

(3) The point of origin of the shipment and the 7-day period during which departure of the shipment is estimated to occur;

(4) The 7-day period during which arrival of the shipment at State boundaries or Tribal reservation boundaries is estimated to occur;

(5) The destination of the shipment, and the 7-day period during which arrival of the shipment is estimated to occur; and

(6) A point of contact, with a telephone number, for current shipment information.

(e) Revision notice. A licensee who finds that schedule information previously furnished to a governor or governor’s designee or a Tribal official or Tribal official’s designee, in accordance with this section, will not be met, shall telephone a responsible individual in the office of the governor of the State or of the governor’s designee or the Tribal official or the Tribal official’s designee and inform that individual of the extent of the delay beyond the schedule originally reported. The licensee shall maintain a record of the name of the individual contacted for 3 years.

(f) Cancellation notice. (1) Each licensee who cancels an irradiated reactor fuel or nuclear waste shipment for which advance notification has been sent shall send a cancellation notice to the governor of each State or to the governor’s designee previously notified, each Tribal official or to the Tribal official’s designee previously notified, and to the Director, Division of Security Policy, Office of Nuclear Security and Incident Response.

(2) The licensee shall state in the notice that it is a cancellation and identify the advance notification that is being canceled. The licensee shall retain a copy of the notice as a record for 3 years.

[60 FR 50264, Sept. 28, 1995, as amended at 67 FR 3586, Jan. 25, 2002; 68 FR 58818, Oct. 10, 2003; 74 FR 62683, Dec. 1, 2009; 75 FR 73945, Nov. 30, 2010; 77 FR 34204, Jun. 11, 2012; 78 FR 17021, Mar. 19, 2013]

§ 71.99 Violations.




(a) The Commission may obtain an injunction or other court order to prevent a violation of the provisions of --

(1) The Atomic Energy Act of 1954, as amended;

(2) Title II of the Energy Reorganization Act of 1974, as amended; or (3) A regulation or order issued pursuant to those Acts.

(b) The Commission may obtain a court order for the payment of a civil penalty imposed under section 234 of the Atomic Energy Act:

(1) For violations of --

(i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of the Atomic Energy Act of 1954, as amended;

(ii) Section 206 of the Energy Reorganization Act;

(iii) Any rule, regulation, or order issued pursuant to the sections specified in paragraph (b)(1)(i) of this section; or

(iv) Any term , condition, or limitation of any license issued under the sections specified in paragraph (b)(1)(i) of this section.

(2) For any violation for which a license may be revoked under section 186 of the Atomic Energy Act of 1954, as amended.

§ 71.100 Criminal penalties.




(a) Section 223 of the Atomic Energy Act of 1954, as amended, provides for criminal sanctions for willful violation of, attempted violation of, or conspiracy to violate, any regulation issued under sections 161b, 161i, or 161o of the Act. For purposes of section 223, all the regulations in part 71 are issued under one or more of sections 161b, 161i, or 161o, except for the sections listed in paragraph (b) of this section.

(b) The regulations in part 71 that are not issued under sections 161b, 161i, or 161o for the purposes of section 223 are as follows: §§ 71.0, 71.2, 71.4, 71.6, 71.7, 71.10, 71.31, 71.33, 71.35, 71.37, 71.38, 71.39, 71.40, 71.41, 71.43, 71.45, 71.47, 71.51, 71.55, 71.59, 71.65, 71.71, 71.73, 71.74, 71.75, 71.77, 71.99, and 71.100.

[69 FR 3796, Jan. 26, 2004]

Subpart H--Quality Assurance




Source: 69 FR 3797, Jan. 26, 2004, unless otherwise noted.

§ 71.101 Quality assurance requirements.


(a) Purpose. This subpart describes quality assurance requirements applying to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and modification of components of packaging that are important to safety. As used in this subpart, "quality assurance" comprises all those planned and systematic actions necessary to provide adequate confidence that a system or component will perform satisfactorily in service. Quality assurance includes quality control, which comprises those quality assurance actions related to control of the physical characteristics and quality of the material or component to predetermined requirements. The licensee, certificate holder, and applicant for a CoC are responsible for the quality assurance requirements as they apply to design, fabrication, testing, and modification of packaging. Each licensee is responsible for the quality assurance provision which applies to its use of a packaging for the shipment of licensed material subject to this subpart.

(b) Establishment of program. Each licensee, certificate holder, and applicant for a CoC shall establish, maintain, and execute a quality assurance program satisfying each of the applicable criteria of §§ 71.101 through 71.137 and satisfying any specific provisions that are applicable to the licensee's activities including procurement of packaging. The licensee, certificate holder, and applicant for a CoC shall execute the applicable criteria in a graded approach to an extent that is commensurate with the quality assurance requirement's importance to safety.

(c) Approval of program. (1) Before the use of any package for the shipment of licensed material subject to this subpart, each licensee shall obtain Commission approval of its quality assurance program. Using an appropriate method listed in § 71.1(a), each licensee shall file a description of its quality assurance program, including a discussion of which requirements of this subpart are applicable and how they will be satisfied, by submitting the description to: ATTN: Document Control Desk, Director, Division of Spent Fuel Storage and Transportation, Office of Nuclear Material Safety and Safeguards.

(2) Before the fabrication, testing, or modification of any package for the shipment of licensed material subject to this subpart, each licensee, certificate holder, or applicant for a CoC shall obtain Commission approval of its quality assurance program. Each certificate holder or applicant for a CoC shall, in accordance with § 71.1, file a description of its quality assurance program, including a discussion of which requirements of this subpart are applicable and how they will be satisfied.

(d) Existing package designs. The provisions of this paragraph deal with packages that have been approved for use in accordance with this part before January 1, 1979, and which have been designed in accordance with the provisions of this part in effect at the time of application for package approval. Those packages will be accepted as having been designed in accordance with a quality assurance program that satisfies the provisions of paragraph (b) of this section.

(e) Existing packages. The provisions of this paragraph deal with packages that have been approved for use in accordance with this part before January 1, 1979, have been at least partially fabricated before that date, and for which the fabrication is in accordance with the provisions of this part in effect at the time of application for approval of package design. These packages will be accepted as having been fabricated and assembled in accordance with a quality assurance program that satisfies the provisions of paragraph (b) of this section.

(f) Previously approved programs. A Commission-approved quality assurance program that satisfies the applicable criteria of subpart H of this part, Appendix B of part 50 of this chapter, or subpart G of part 72 of this chapter, and that is established, maintained, and executed regarding transport packages, will be accepted as satisfying the requirements of paragraph (b) of this section. Before first use, the licensee, certificate holder, and applicant for a CoC shall notify the NRC, in accordance with § 71.1, of its intent to apply its previously approved subpart H, Appendix B, or subpart G quality assurance program to transportation activities. The licensee, certificate holder, and applicant for a CoC shall identify the program by date of submittal to the Commission, Docket Number, and date of Commission approval.

(g) Radiography containers. A program for transport container inspection and maintenance limited to radiographic exposure devices, source changers, or packages transporting these devices and meeting the requirements of § 34.31(b) of this chapter or equivalent Agreement State requirement, is deemed to satisfy the requirements of §§ 71.17(b) and 71.101(b).

[75 FR 73945, Nov. 30, 2010]

§ 71.103 Quality assurance organization.




(a) The licensee,2 certificate holder, and applicant for a CoC shall be responsible for the establishment and execution of the quality assurance program. The licensee, certificate holder, and applicant for a CoC may delegate to others, such as contractors, agents, or consultants, the work of establishing and executing the quality assurance program, or any part of the quality assurance program, but shall retain responsibility for the program. These activities include performing the functions associated with attaining quality objectives and the quality assurance functions.

(b) The quality assurance functions are--

(1) Assuring that an appropriate quality assurance program is established and effectively executed; and

(2) Verifying, by procedures such as checking, auditing, and inspection, that activities affecting the functions that are important to safety have been correctly performed.

(c) The persons and organizations performing quality assurance functions must have sufficient authority and organizational freedom to--

(1) Identify quality problems;

(2) Initiate, recommend, or provide solutions; and

(3) Verify implementation of solutions.

(d) The persons and organizations performing quality assurance functions shall report to a management level that assures that the required authority and organizational freedom, including sufficient independence from cost and schedule, when opposed to safety considerations, are provided.

(e) Because of the many variables involved, such as the number of personnel, the type of activity being performed, and the location or locations where activities are performed, the organizational structure for executing the quality assurance program may take various forms, provided that the persons and organizations assigned the quality assurance functions have the required authority and organizational freedom.

(f) Irrespective of the organizational structure, the individual(s) assigned the responsibility for assuring effective execution of any portion of the quality assurance program, at any location where activities subject to this section are being performed, must have direct access to the levels of management necessary to perform this function.

2 While the term "licensee" is used in these criteria, the requirements are applicable to whatever design, fabrication, assembly, and testing of the package is accomplished with respect to a package before the time a package approval is issued.

§ 71.105 Quality assurance program.




(a) The licensee, certificate holder, and applicant for a CoC shall establish, at the earliest practicable time consistent with the schedule for accomplishing the activities, a quality assurance program that complies with the requirements of §§ 71.101 through 71.137. The licensee, certificate holder, and applicant for a CoC shall document the quality assurance program by written procedures or instructions and shall carry out the program in accordance with those procedures throughout the period during which the packaging is used. The licensee, certificate holder, and applicant for a CoC shall identify the material and components to be covered by the quality assurance program, the major organizations participating in the program, and the designated functions of these organizations.

(b) The licensee, certificate holder, and applicant for a CoC, through its quality assurance program, shall provide control over activities affecting the quality of the identified materials and components to an extent consistent with their importance to safety, and as necessary to assure conformance to the approved design of each individual package used for the shipment of radioactive material. The licensee, certificate holder, and applicant for a CoC shall assure that activities affecting quality are accomplished under suitably controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanliness; and assurance that all prerequisites for the given activity have been satisfied. The licensee, certificate holder, and applicant for a CoC shall take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of quality by inspection and test.

(c) The licensee, certificate holder, and applicant for a CoC shall base the requirements and procedures of its quality assurance program on the following considerations concerning the complexity and proposed use of the package and its components:

(1) The impact of malfunction or failure of the item to safety;

(2) The design and fabrication complexity or uniqueness of the item;

(3) The need for special controls and surveillance over processes and equipment;

(4) The degree to which functional compliance can be demonstrated by inspection or test; and

(5) The quality history and degree of standardization of the item.

(d) The licensee, certificate holder, and applicant for a CoC shall provide for indoctrination and training of personnel performing activities affecting quality, as necessary to assure that suitable proficiency is achieved and maintained. The licensee, certificate holder, and applicant for a CoC shall review the status and adequacy of the quality assurance program at established intervals. Management of other organizations participating in the quality assurance program shall review regularly the status and adequacy of that part of the quality assurance program they are executing.

§ 71.107 Package design control.




(a) The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that applicable regulatory requirements and the package design, as specified in the license or CoC for those materials and components to which this section applies, are correctly translated into specifications, drawings, procedures, and instructions. These measures must include provisions to assure that appropriate quality standards are specified and included in design documents and that deviations from standards are controlled. Measures must be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the functions of the materials, parts, and components of the packaging that are important to safety.

(b) The licensee, certificate holder, and applicant for a CoC shall establish measures for the identification and control of design interfaces and for coordination among participating design organizations. These measures must include the establishment of written procedures, among participating design organizations, for the review, approval, release, distribution, and revision of documents involving design interfaces. The design control measures must provide for verifying or checking the adequacy of design, by methods such as design reviews, alternate or simplified calculational methods, or by a suitable testing program. For the verifying or checking process, the licensee shall designate individuals or groups other than those who were responsible for the original design, but who may be from the same organization. Where a test program is used to verify the adequacy of a specific design feature in lieu of other verifying or checking processes, the licensee, certificate holder, and applicant for a CoC shall include suitable qualification testing of a prototype or sample unit under the most adverse design conditions. The licensee, certificate holder, and applicant for a CoC shall apply design control measures to the following:

(1) Criticality physics, radiation shielding, stress, thermal, hydraulic, and accident analyses;

(2) Compatibility of materials;

(3) Accessibility for inservice inspection, maintenance, and repair;

(4) Features to facilitate decontamination; and

(5) Delineation of acceptance criteria for inspections and tests.

(c) The licensee, certificate holder, and applicant for a CoC shall subject design changes, including field changes, to design control measures commensurate with those applied to the original design. Changes in the conditions specified in the CoC require prior NRC approval.

§ 71.109 Procurement document control.




The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that adequate quality is required in the documents for procurement of material, equipment, and services, whether purchased by the licensee, certificate holder, and applicant for a CoC or by its contractors or subcontractors. To the extent necessary, the licensee, certificate holder, and applicant for a CoC shall require contractors or subcontractors to provide a quality assurance program consistent with the applicable provisions of this part.

§ 71.111 Instructions, procedures, and drawings.




The licensee, certificate holder, and applicant for a CoC shall prescribe activities affecting quality by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall require that these instructions, procedures, and drawings be followed. The instructions, procedures, and drawings must include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

§ 71.113 Document control.




The licensee, certificate holder, and applicant for a CoC shall establish measures to control the issuance of documents such as instructions, procedures, and drawings, including changes, that prescribe all activities affecting quality. These measures must assure that documents, including changes, are reviewed for adequacy, approved for release by authorized personnel, and distributed and used at the location where the prescribed activity is performed.

§ 71.115 Control of purchased material, equipment, and services.




(a) The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that purchased material, equipment, and services, whether purchased directly or through contractors and subcontractors, conform to the procurement documents. These measures must include provisions, as appropriate, for source evaluation and selection, objective evidence of quality furnished by the contractor or subcontractor, inspection at the contractor or subcontractor source, and examination of products on delivery.

(b) The licensee, certificate holder, and applicant for a CoC shall have available documentary evidence that material and equipment conform to the procurement specifications before installation or use of the material and equipment. The licensee, certificate holder, and applicant for a CoC shall retain, or have available, this documentary evidence for the life of the package to which it applies. The licensee, certificate holder, and applicant for a CoC shall assure that the evidence is sufficient to identify the specific requirements met by the purchased material and equipment.

(c) The licensee, certificate holder, and applicant for a CoC shall assess the effectiveness of the control of quality by contractors and subcontractors at intervals consistent with the importance, complexity, and quantity of the product or services.

§ 71.117 Identification and control of materials, parts, and components.




The licensee, certificate holder, and applicant for a CoC shall establish measures for the identification and control of materials, parts, and components. These measures must assure that identification of the item is maintained by heat number, part number, or other appropriate means, either on the item or on records traceable to the item, as required throughout fabrication, installation, and use of the item. These identification and control measures must be designed to prevent the use of incorrect or defective materials, parts, and components.

§ 71.119 Control of special processes.




The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that special processes, including welding, heat treating, and nondestructive testing are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.

§ 71.121 Internal inspection.




The licensee, certificate holder, and applicant for a CoC shall establish and execute a program for inspection of activities affecting quality by or for the organization performing the activity, to verify conformance with the documented instructions, procedures, and drawings for accomplishing the activity. The inspection must be performed by individuals other than those who performed the activity being inspected. Examination, measurements, or tests of material or products processed must be performed for each work operation where necessary to assure quality. If direct inspection of processed material or products is not carried out, indirect control by monitoring processing methods, equipment, and personnel must be provided. Both inspection and process monitoring must be provided when quality control is inadequate without both. If mandatory inspection hold points, which require witnessing or inspecting by the licensee's designated representative and beyond which work should not proceed without the consent of its designated representative, are required, the specific hold points must be indicated in appropriate documents.

§ 71.123 Test control.




The licensee, certificate holder, and applicant for a CoC shall establish a test program to assure that all testing required to demonstrate that the packaging components will perform satisfactorily in service is identified and performed in accordance with written test procedures that incorporate the requirements of this part and the requirements and acceptance limits contained in the package approval. The test procedures must include provisions for assuring that all prerequisites for the given test are met, that adequate test instrumentation is available and used, and that the test is performed under suitable environmental conditions. The licensee, certificate holder, and applicant for a CoC shall document and evaluate the test results to assure that test requirements have been satisfied.

§ 71.125 Control of measuring and test equipment.




The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that tools, gauges, instruments, and other measuring and testing devices used in activities affecting quality are properly controlled, calibrated, and adjusted at specified times to maintain accuracy within necessary limits.

§ 71.127 Handling, storage, and shipping control.




The licensee, certificate holder, and applicant for a CoC shall establish measures to control, in accordance with instructions, the handling, storage, shipping, cleaning, and preservation of materials and equipment to be used in packaging to prevent damage or deterioration. When necessary for particular products, special protective environments, such as inert gas atmosphere, and specific moisture content and temperature levels must be specified and provided.

§ 71.129 Inspection, test, and operating status.




(a) The licensee, certificate holder, and applicant for a CoC shall establish measures to indicate, by the use of markings such as stamps, tags, labels, routing cards, or other suitable means, the status of inspections and tests performed upon individual items of the packaging. These measures must provide for the identification of items that have satisfactorily passed required inspections and tests, where necessary to preclude inadvertent bypassing of the inspections and tests.

(b) The licensee shall establish measures to identify the operating status of components of the packaging, such as tagging valves and switches, to prevent inadvertent operation.

§ 71.131 Nonconforming materials, parts, or components.




The licensee, certificate holder, and applicant for a CoC shall establish measures to control materials, parts, or components that do not conform to the licensee's requirements to prevent their inadvertent use or installation. These measures must include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification to affected organizations. Nonconforming items must be reviewed and accepted, rejected, repaired, or reworked in accordance with documented procedures.

§ 71.133 Corrective action.




The licensee, certificate holder, and applicant for a CoC shall establish measures to assure that conditions adverse to quality, such as deficiencies, deviations, defective material and equipment, and nonconformances, are promptly identified and corrected. In the case of a significant condition adverse to quality, the measures must assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken must be documented and reported to appropriate levels of management.

§ 71.135 Quality assurance records.




The licensee, certificate holder, and applicant for a CoC shall maintain sufficient written records to describe the activities affecting quality. The records must include the instructions, procedures, and drawings required by § 71.111 to prescribe quality assurance activities and must include closely related specifications such as required qualifications of personnel, procedures, and equipment. The records must include the instructions or procedures which establish a records retention program that is consistent with applicable regulations and designates factors such as duration, location, and assigned responsibility. The licensee, certificate holder, and applicant for a CoC shall retain these records for 3 years beyond the date when the licensee, certificate holder, and applicant for a CoC last engage in the activity for which the quality assurance program was developed. If any portion of the written procedures or instructions is superseded, the licensee, certificate holder, and applicant for a CoC shall retain the superseded material for 3 years after it is superseded.

§ 71.137 Audits.




The licensee, certificate holder, and applicant for a CoC shall carry out a comprehensive system of planned and periodic audits to verify compliance with all aspects of the quality assurance program and to determine the effectiveness of the program. The audits must be performed in accordance with written procedures or checklists by appropriately trained personnel not having direct responsibilities in the areas being audited. Audited results must be documented and reviewed by management having responsibility in the area audited. Followup action, including reaudit of deficient areas, must be taken where indicated.

Appendix A to Part 71—Determination of A1 and A2




I. Values of A1 and A2 for individual radionuclides, which are the bases for many activity limits elsewhere in these regulations, are given in Table A-1. The curie (Ci) values specified are obtained by converting from the Terabecquerel (TBq) value. The Terabecquerel values are the regulatory standard. The curie values are for information only and are not intended to be the regulatory standard. Where values of A1 and A2 are unlimited, it is for radiation control purposes only. For nuclear criticality safety, some materials are subject to controls placed on fissile material.

II. a. For individual radionuclides whose identities are known, but which are not listed in Table A-1, the A1 and A2 values contained in Table A-3 may be used. Otherwise, the licensee shall obtain prior Commission approval of the A1 and A2 values for radionuclides not listed in Table A-1, before shipping the material.

b. For individual radionuclides whose identities are known, but which are not listed in Table A-2, the exempt material activity concentration and exempt consignment activity values contained in Table A-3 may be used. Otherwise, the licensee shall obtain prior Commission approval of the exempt material activity concentration and exempt consignment activity values for radionuclides not listed in Table A-2, before shipping the material.

c. The licensee shall submit requests for prior approval, described under paragraphs II(a) and II(b) of this Appendix, to the Commission, in accordance with § 71.1 of this part.

III. In the calculations of A1 and A2 for a radionuclide not in Table A-1, a single radioactive decay chain, in which radionuclides are present in their naturally occurring proportions, and in which no daughter radionuclide has a half-life either longer than 10 days, or longer than that of the parent radionuclide, shall be considered as a single radionuclide, and the activity to be taken into account, and the A1 or A2 value to be applied, shall be those corresponding to the parent radionuclide of that chain. In the case of radioactive decay chains in which any daughter radionuclide has a half-life either longer than 10 days, or greater than that of the parent radionuclide, the parent and those daughter radionuclides shall be considered as mixtures of different radionuclides.

IV. For mixtures of radionuclides whose identities and respective activities are known, the following conditions apply:

a. For special form radioactive material, the maximum quantity transported in a Type A package is as follows:


where B(i) is the activity of radionuclide i, and A1(i) is the A1 value for radionuclide I.

b. For normal form radioactive material, the maximum quantity transported in a Type A package is as follows:

∑B(i)/A2(i)≤1

where B(i) is the activity of radionuclide i, and A2(i) is the A2 value for radionuclide i.

c. Alternatively, the A1 value for mixtures of special form material may be determined as follows:


where f(i) is the fraction of activity for radionuclide I in the mixture, and A1(i) is the appropriate A1 value for radionuclide I.

d. Alternatively, the A2 value for mixtures of normal form material may be determined as follows:


where f(i) is the fraction of activity for radionuclide I in the mixture, and A2(i) is the appropriate A2 value for radionuclide I.

e. The exempt activity concentration for mixtures of nuclides may be determined as follows:


where f(i) is the fraction of activity concentration of radionuclide I in the mixture, and [A] is the activity concentration for exempt material containing radionuclide I.

f. The activity limit for an exempt consignment for mixtures of radionuclides may be determined as follows:


where f(i) is the fraction of activity of radionuclide I in the mixture, and A is the activity limit for exempt consignments for radionuclide I.

V. When the identity of each radionuclide is known, but the individual activities of some of the radionuclides are not known, the radionuclides may be grouped, and the lowest A1 or A2 value, as appropriate, for the radionuclides in each group may be used in applying the formulas in paragraph IV. Groups may be based on the total alpha activity and the total beta/gamma activity when these are known, using the lowest A1 or A2 values for the alpha emitters and beta/gamma emitters.

Table A-1—A1 and A2 VALUES FOR RADIONUCLIDES

Symbol of radionuclide Element and atomic number A1 (TBq) A1(Ci)b A2 (TBq) A2(Ci)b Specific activity
(TBq/g) (Ci/g)
Ac-225 (a) Actinium (89) 8.0X10-1 2.2X101 6.0X10-3 1.6X10-1 2.1X103 5.8X104
Ac-227 (a)
9.0X10-1 2.4X101 9.0X10-5 2.4X10-3 2.7 7.2X101
Ac-228
6.0X10-1 1.6X101 5.0X10-1 1.4X101 8.4X104 2.2X106
Ag-105 Silver (47) 2.0 5.4X101 2.0 5.4X101 1.1X103 3.0X104
Ag-108m (a)
7.0X10-1 1.9X101 7.0X10-1 1.9X101 9.7X10-1 2.6X101
Ag-110m (a)
4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.8X102 4.7X103
Ag-111
2.0 5.4X101 6.0X10-1 1.6X101 5.8X103 1.6X105
Al-26 Aluminum (13) 1.0X10-1 2.7 1.0X10-1 2.7 7.0X10-4 1.9X10-2
Am-241 Americium (95) 1.0X101 2.7X102 1.0X10-3 2.7X10-2 1.3X10-1 3.4
Am-242m (a)
1.0X101 2.7X102 1.0X10-3 2.7X10-2 3.6X10-1 1.0X101
Am-243 (a)
5.0 1.4X102 1.0X10-3 2.7X10-2 7.4X10-3 2.0X10-1
Ar-37 Argon (18) 4.0X101 1.1X103 4.0X101 1.1X103 3.7X103 9.9X104
Ar-39
4.0X101 1.1X103 2.0X101 5.4X102 1.3 3.4X101
Ar-41
3.0X10-1 8.1 3.0X10-1 8.1 1.5X106 4.2X107
As-72 Arsenic (33) 3.0X10-1 8.1 3.0X10-1 8.1 6.2X104 1.7X106
As-73
4.0X101 1.1X103 4.0X101 1.1X103 8.2X102 2.2X104
As-74
1.0 2.7X101 9.0X10-1 2.4X101 3.7X103 9.9X104
As-76
3.0X10-1 8.1 3.0X10-1 8.1 5.8X104 1.6X106
As-77
2.0X101 5.4X102 7.0X10-1 1.9X101 3.9X104 1.0X106
At-211 (a) Astatine (85) 2.0X101 5.4X102 5.0X10-1 1.4X101 7.6X104 2.1X106
Au-193 Gold (79) 7.0 1.9X102 2.0 5.4X101 3.4X104 9.2X105
Au-194
1.0 2.7X101 1.0 2.7X101 1.5X104 4.1X105
Au-195
1.0X101 2.7X102 6.0 1.6X102 1.4X102 3.7X103
Au-198
1.0 2.7X101 6.0X10-1 1.6X101 9.0X103 2.4X105
Au-199
1.0X101 2.7X102 6.0X10-1 1.6X101 7.7X103 2.1X105
Ba-131 (a) Barium (56) 2.0 5.4X101 2.0 5.4X101 3.1X103 8.4X104
Ba-133
3.0 8.1X101 3.0 8.1X101 9.4 2.6X102
Ba-133m
2.0X101 5.4X102 6.0X10-1 1.6X101 2.2X104 6.1X105
Ba-140 (a)
5.0X10-1 1.4X101 3.0X10-1 8.1 2.7X103 7.3X104
Be-7 Beryllium (4) 2.0X101 5.4X102 2.0X101 5.4X102 1.3X104 3.5X105
Be-10
4.0X101 1.1X103 6.0X10-1 1.6X101 8.3X10-4 2.2X10-2
Bi-205 Bismuth (83) 7.0X10-1 1.9X101 7.0X10-1 1.9X101 1.5X103 4.2X104
Bi-206
3.0X10-1 8.1 3.0X10-1 8.1 3.8X103 1.0X105
Bi-207
7.0X10-1 1.9X101 7.0X10-1 1.9X101 1.9 5.2X101
Bi-210
1.0 2.7X101 6.0X10-1 1.6X101 4.6X103 1.2X105
Bi-210m (a)
6.0X10-1 1.6X101 2.0X10-2 5.4X10-1 2.1X10-5 5.7X10-4
Bi-212 (a)
7.0X10-1 1.9X101 6.0X10-1 1.6X101 5.4X105 1.5X107
Bk-247 Berkelium (97) 8.0 2.2X102 8.0X10-4 2.2X10-2 3.8X10-2 1.0
Bk-249 (a)
4.0X101 1.1X103 3.0X10-1 8.1 6.1X101 1.6X103
Br-76 Bromine (35) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 9.4X104 2.5X106
Br-77
3.0 8.1X101 3.0 8.1X101 2.6X104 7.1X105
Br-82
4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.0X104 1.1X106
C-11 Carbon (6) 1.0 2.7X101 6.0X10-1 1.6X101 3.1X107 8.4X108
C-14
4.0X101 1.1X103 3.0 8.1X101 1.6X10-1 4.5
Ca-41 Calcium (20) Unlimited Unlimited Unlimited Unlimited 3.1X10-3 8.5X10-2
Ca-45
4.0X101 1.1X103 1.0 2.7X101 6.6X102 1.8X104
Ca-47 (a)
3.0 8.1X101 3.0X10-1 8.1 2.3X104 6.1X105
Cd-109 Cadmium (48) 3.0X101 8.1X102 2.0 5.4X101 9.6X101 2.6X103
Cd-113m
4.0X101 1.1X103 5.0X10-1 1.4X101 8.3 2.2X102
Cd-115 (a)
3.0 8.1X101 4.0X10-1 1.1X101 1.9X104 5.1X105
Cd-115m
5.0X10-1 1.4X101 5.0X10-1 1.4X101 9.4X102 2.5X104
Ce-139 Cerium (58) 7.0 1.9X102 2.0 5.4X101 2.5X102 6.8X103
Ce-141
2.0X101 5.4X102 6.0X10-1 1.6X101 1.1X103 2.8X104
Ce-143
9.0X10-1 2.4X101 6.0X10-1 1.6X101 2.5X104 6.6X105
Ce-144 (a)
2.0X10-1 5.4 2.0X10-1 5.4 1.2X102 3.2X103
Cf-248 Californium (98) 4.0X101 1.1X103 6.0X10-3 1.6X10-1 5.8X101 1.6X103
Cf-249
3.0 8.1X101 8.0X10-4 2.2X10-2 1.5X10-1 4.1
Cf-250
2.0X101 5.4X102 2.0X10-3 5.4X10-2 4.0 1.1X102
Cf-251
7.0 1.9X102 7.0X10-4 1.9X10-2 5.9X10-2 1.6
Cf-252 (h)
5.0X10-2 1.4 3.0X10-3 8.1X10-2 2.0X101 5.4X102
Cf-253 (a)
4.0X101 1.1X103 4.0X10-2 1.1 1.1X103 2.9X104
Cf-254
1.0X10-3 2.7X10-2 1.0X10-3 2.7X10-2 3.1X102 8.5X103
Cl-36 Chlorine (17) 1.0X101 2.7X102 6.0X10-1 1.6X101 1.2X10-3 3.3X10-2
Cl-38
2.0X10-1 5.4 2.0X10-1 5.4 4.9X106 1.3X108
Cm-240 Curium (96) 4.0X101 1.1X103 2.0X10-2 5.4X10-1 7.5X102 2.0X104
Cm-241
2.0 5.4X101 1.0 2.7X101 6.1X102 1.7X104
Cm-242
4.0X101 1.1X103 1.0X10-2 2.7X10-1 1.2X102 3.3X103
Cm-243
9.0 2.4X102 1.0X10-3 2.7X10-2 1.9X10-3 5.2X101
Cm-244
2.0X101 5.4X102 2.0X10-3 5.4X10-2 3.0 8.1X101
Cm-245
9.0 2.4X102 9.0X10-4 2.4X10-2 6.4X10-3 1.7X10-1
Cm-246
9.0 2.4X102 9.0X10-4 2.4X10-2 1.1X10-2 3.1X10-1
Cm-247 (a)
3.0 8.1X101 1.0X10-3 2.7X10-2 3.4X10-6 9.3X10-5
Cm-248
2.0X10-2 5.4X10-1 3.0X10-4 8.1X10-3 1.6X10-4 4.2X10-3
Co-55 Cobalt (27) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 1.1X105 3.1X106
Co-56
3.0X10-1 8.1 3.0X10-1 8.1 1.1X103 3.0X104
Co-57
1.0X101 2.7X102 1.0X101 2.7X102 3.1X102 8.4X103
Co-58
1.0 2.7X101 1.0 2.7X101 1.2X103 3.2X104
Co-58m
4.0X101 1.1X103 4.0X101 1.1X103 2.2X105 5.9X106
Co-60
4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.2X101 1.1X103
Cr-51 Chromium (24) 3.0X101 8.1X102 3.0X101 8.1X102 3.4X103 9.2X104
Cs-129 Cesium (55) 4.0 1.1X102 4.0 1.1X102 2.8X104 7.6X105
Cs-131
3.0X101 8.1X102 3.0X101 8.1X102 3.8X103 1.0X105
Cs-132
1.0 2.7X101 1.0 2.7X101 5.7X103 1.5X105
Cs-134
7.0X10-1 1.9X101 7.0X10-1 1.9X101 4.8X101 1.3X103
Cs-134m
4.0X101 1.1X103 6.0X10-1 1.6X101 3.0X105 8.0X106
Cs-135
4.0X101 1.1X103 1.0 2.7X101 4.3X10-5 1.2X10-3
Cs-136
5.0X10-1 1.4X101 5.0X10-1 1.4X101 2.7X103 7.3X104
Cs-137 (a)
2.0 5.4X101 6.0X10-1 1.6X101 3.2 8.7X101
Cu-64 Copper (29) 6.0 1.6X102 1.0 2.7X101 1.4X105 3.9X106
Cu-67
1.0X101 2.7X102 7.0X10-1 1.9X101 2.8X104 7.6X105
Dy-159 Dysprosium (66) 2.0X101 5.4X102 2.0X101 5.4X102 2.1X102 5.7X103
Dy-165
9.0X10-1 2.4X101 6.0X10-1 1.6X101 3.0X105 8.2X106
Dy-166 (a)
9.0X10-1 2.4X101 3.0X10-1 8.1 8.6X103 2.3X105
Er-169 Erbium (68) 4.0X101 1.1X103 1.0 2.7X101 3.1X103 8.3X104
Er-171
8.0X10-1 2.2X101 5.0X10-1 1.4X101 9.0X104 2.4X106
Eu-147 Europium (63) 2.0 5.4X101 2.0 5.4X101 1.4X103 3.7X104
Eu-148
5.0X10-1 1.4X101 5.0X10-1 1.4X101 6.0X102 1.6X104
Eu-149
2.0X101 5.4X102 2.0X101 5.4X102 3.5X102 9.4X103
Eu-150 (short lived)
2.0 5.4X101 7.0X10-1 1.9X101 6.1X104 1.6X106
Eu-150 (long lived)
7.0X10-1 1.9X101 7.0X10-1 1.9X101 6.1X104 1.6X106
Eu-152
1.0 2.7X101 1.0 2.7X101 6.5 1.8X102
Eu-152m
8.0X10-1 2.2X101 8.0X10-1 2.2X101 8.2X104 2.2X106
Eu-154
9.0X10-1 2.4X101 6.0X10-1 1.6X101 9.8 2.6X102
Eu-155
2.0X101 5.4X102 3.0 8.1X101 1.8X101 4.9X102
Eu-156
7.0X10-1 1.9X101 7.0X10-1 1.9X101 2.0X103 5.5X104
F-18 Fluorine (9) 1.0 2.7X101 6.0X10-1 1.6X101 3.5X106 9.5X107
Fe-52 (a) Iron (26) 3.0X10-1 8.1 3.0X10-1 8.1 2.7X105 7.3X106
Fe-55
4.0X101 1.1X103 4.0X101 1.1X103 8.8X101 2.4X103
Fe-59
9.0X10-1 2.4X101 9.0X10-1 2.4X101 1.8X103 5.0X104
Fe-60 (a)
4.0X101 1.1X103 2.0X10-1 5.4 7.4X10-4 2.0X10-2
Ga-67 Gallium (31) 7.0 1.9X102 3.0 8.1X101 2.2X104 6.0X105
Ga-68
5.0X10-1 1.4X101 5.0X10-1 1.4X101 1.5X106 4.1X107
Ga-72
4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.1X105 3.1X106
Gd-146 (a) Gadolinium (64) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 6.9X102 1.9X104
Gd-148
2.0X101 5.4X102 2.0X10-3 5.4X10-2 1.2 3.2X101
Gd-153
1.0X101 2.7X102 9.0 2.4X102 1.3X102 3.5X103
Gd-159
3.0 8.1X101 6.0X10-1 1.6X101 3.9X104 1.1X106
Ge-68 (a) Germanium (32) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 2.6X102 7.1X103
Ge-71
4.0X101 1.1X103 4.0X101 1.1X103 5.8X103 1.6X105
Ge-77
3.0X10-1 8.1 3.0X10-1 8.1 1.3X105 3.6X106
Hf-172 (a) Hafnium (72) 6.0X10-1 1.6X101 6.0X10-1 1.6X101 4.1X101 1.1X103
Hf-175
3.0 8.1X101 3.0 8.1X101 3.9X102 1.1X104
Hf-181
2.0 5.4X101 5.0X10-1 1.4X101 6.3X102 1.7X104
Hf-182
Unlimited Unlimited Unlimited Unlimited 8.1X10-6 2.2X10-4
Hg-194 (a) Mercury (80) 1.0 2.7X101 1.0 2.7X101 1.3X10-1 3.5
Hg-195m (a)
3.0 8.1X101 7.0X10-1 1.9X101 1.5X104 4.0X105
Hg-197
2.0X101 5.4X102 1.0X101 2.7X102 9.2X103 2.5X105
Hg-197m
1.0X101 2.7X102 4.0X10-1 1.1X101 2.5X104 6.7X105
Hg-203
5.0 1.4X102 1.0 2.7X101 5.1X102 1.4X104
Ho-166 Holmium (67) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 2.6X104 7.0X105
Ho-166m
6.0X10-1 1.6X101 5.0X10-1 1.4X101 6.6X10-2 1.8
I-123 Iodine (53) 6.0 1.6X102 3.0 8.1X101 7.1X104 1.9X106
I-124
1.0 2.7X101 1.0 2.7X101 9.3X103 2.5X105
I-125
2.0X101 5.4X102 3.0 8.1X101 6.4X102 1.7X104
I-126
2.0 5.4X101 1.0 2.7X101 2.9X103 8.0X104
I-129
Unlimited Unlimited Unlimited Unlimited 6.5X10-6 1.8X10-4
I-131
3.0 8.1X101 7.0X10-1 1.9X101 4.6X103 1.2X105
I-132
4.0X10-1 1.1X101 4.0X10-1 1.1X101 3.8X105 1.0X107
I-133
7.0X10-1 1.9X101 6.0X10-1 1.6X101 4.2X104 1.1X106
I-134
3.0X10-1 8.1 3.0X10-1 8.1 9.9X105 2.7X107
I-135 (a)
6.0X10-1 1.6X101 6.0X10-1 1.6X101 1.3X105 3.5X106
In-111 Indium (49) 3.0 8.1X101 3.0 8.1X101 1.5X104 4.2X105
In-113m
4.0 1.1X102 2.0 5.4X101 6.2X105 1.7X107
In-114m (a)
1.0X101 2.7X102 5.0X10-1 1.4X101 8.6X102 2.3X104
In-115m
7.0 1.9X102 1.0 2.7X101 2.2X105 6.1X106
Ir-189 (a) Iridium (77) 1.0X101 2.7X102 1.0X101 2.7X102 1.9X103 5.2X104
Ir-190
7.0X10-1 1.9X101 7.0X10-1 1.9X101 2.3X103 6.2X104
Ir-192 (c)
1.0 2.7X101 6.0X10-1 1.6X101 3.4X102 9.2X103
Ir-194
3.0X10-1 8.1 3.0X10-1 8.1 3.1X104 8.4X105
K-40 Potassium (19) 9.0X10-1 2.4X101 9.0X10-1 2.4X101 2.4X10-7 6.4X10-6
K-42
2.0X10-1 5.4 2.0X10-1 5.4 2.2X105 6.0X106
K-43
7.0X10-1 1.9X101 6.0X10-1 1.6X101 1.2X105 3.3X106
Kr-81 Krypton (36) 4.0X101 1.1X103 4.0X101 1.1X103 7.8X10-4 2.1X10-2
Kr-85
1.0X101 2.7X102 1.0X101 2.7X102 1.5X101 3.9X102
Kr-85m
8.0 2.2X102 3.0 8.1X101 3.0X105 8.2X106
Kr-87
2.0X10-1 5.4 2.0X10-1 5.4 1.0X106 2.8X107
La-137 Lanthanum (57) 3.0X101 8.1X102 6.0 1.6X102 1.6X10-3 4.4X10-2
La-140
4.0X10-1 1.1X101 4.0X10-1 1.1X101 2.1X104 5.6X105
Lu-172 Lutetium (71) 6.0X10-1 1.6X101 6.0X10-1 1.6X101 4.2X103 1.1X105
Lu-173
8.0 2.2X102 8.0 2.2X102 5.6X101 1.5X103
Lu-174
9.0 2.4X102 9.0 2.4X102 2.3X101 6.2X102
Lu-174m
2.0X101 5.4X102 1.0X101 2.7X102 2.0X102 5.3X103
Lu-177
3.0X101 8.1X102 7.0X10-1 1.9X101 4.1X103 1.1X105
Mg-28 (a) Magnesium (12) 3.0X10-1 8.1 3.0X10-1 8.1 2.0X105 5.4X106
Mn-52 Manganese (25) 3.0X10-1 8.1 3.0X10-1 8.1 1.6X104 4.4X105
Mn-53
Unlimited Unlimited Unlimited Unlimited 6.8X10-5 1.8X10-3
Mn-54
1.0 2.7X101 1.0 2.7X101 2.9X102 7.7X103
Mn-56
3.0X10-1 8.1 3.0X10-1 8.1 8.0X105 2.2X107
Mo-93 Molybdenum (42) 4.0X101 1.1X103 2.0X101 5.4X102 4.1X10-2 1.1
Mo-99 (a) (i)
1.0 2.7X101 6.0X10-1 1.6X101 1.8X104 4.8X105
N-13 Nitrogen (7) 9.0X10-1 2.4X101 6.0X10-1 1.6X101 5.4X107 1.5X109
Na-22 Sodium (11) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 2.3X102 6.3X103
Na-24
2.0X10-1 5.4 2.0X10-1 5.4 3.2X105 8.7X106
Nb-93m Niobium (41) 4.0X101 1.1X103 3.0X101 8.1X102 8.8 2.4X102
Nb-94
7.0X10-1 1.9X101 7.0X10-1 1.9X101 6.9X10-3 1.9X10-1
Nb-95
1.0 2.7X101 1.0 2.7X101 1.5X103 3.9X104
Nb-97
9.0X10-1 2.4X101 6.0X10-1 1.6X101 9.9X105 2.7X107
Nd-147 Neodymium (60) 6.0 1.6X102 6.0X10-1 1.6X101 3.0X103 8.1X104
Nd-149
6.0X10-1 1.6X101 5.0X10-1 1.4X101 4.5X105 1.2X107
Ni-59 Nickel (28) Unlimited Unlimited Unlimited Unlimited 3.0X10-3 8.0X10-2
Ni-63
4.0X101 1.1X103 3.0X101 8.1X102 2.1 5.7X101
Ni-65
4.0X10-1 1.1X101 4.0X10-1 1.1X101 7.1X105 1.9X107
Np-235 Neptunium (93) 4.0X101 1.1X103 4.0X101 1.1X103 5.2X101 1.4X103
Np-236 (short-lived)
2.0X101 5.4X102 2.0 5.4X101 4.7X10-4 1.3X10-2
Np-236 (long-lived)
9.0X100 2.4X102 2.0X10-2 5.4X10-1 4.7X10-4 1.3X10-2
Np-237
2.0X101 5.4X102 2.0X10-3 5.4X10-2 2.6X10-5 7.1X10-4
Np-239
7.0 1.9X102 4.0X10-1 1.1X101 8.6X103 2.3X105
Os-185 Osmium (76) 1.0 2.7X101 1.0 2.7X101 2.8X102 7.5X103
Os-191
1.0X101 2.7X102 2.0 5.4X101 1.6X103 4.4X104
Os-191m
4.0X101 1.1X103 3.0X101 8.1X102 4.6X104 1.3X106
Os-193
2.0 5.4X101 6.0X10-1 1.6X101 2.0X104 5.3X105
Os-194 (a)
3.0X10-1 8.1 3.0X10-1 8.1 1.1X101 3.1X102
P-32 Phosphorus (15) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 1.1X104 2.9X105
P-33
4.0X101 1.1X103 1.0 2.7X101 5.8X103 1.6X105
Pa-230 (a) Protactinium (91) 2.0 5.4X101 7.0X10-2 1.9 1.2X103 3.3X104
Pa-231
4.0 1.1X102 4.0X10-4 1.1X10-2 1.7X10-3 4.7X10-2
Pa-233
5.0 1.4X102 7.0X10-1 1.9X101 7.7X102 2.1X104
Pb-201 Lead (82) 1.0 2.7X101 1.0 2.7X101 6.2X104 1.7X106
Pb-202
4.0X101 1.1X103 2.0X101 5.4X102 1.2X10-4 3.4X10-3
Pb-203
4.0 1.1X102 3.0 8.1X101 1.1X104 3.0X105
Pb-205
Unlimited Unlimited Unlimited Unlimited 4.5X10-6 1.2X10-4
Pb-210 (a)
1.0 2.7X101 5.0X10-2 1.4 2.8 7.6X101
Pb-212 (a)
7.0X10-1 1.9X101 2.0X10-1 5.4 5.1X104 1.4X106
Pd-103 (a) Palladium (46) 4.0X101 1.1X103 4.0X101 1.1X103 2.8X103 7.5X104
Pd-107
Unlimited Unlimited Unlimited Unlimited 1.9X10-5 5.1X10-4
Pd-109
2.0 5.4X101 5.0X10-1 1.4X101 7.9X104 2.1X106
Pm-143 Promethium (61) 3.0 8.1X101 3.0 8.1X101 1.3X102 3.4X103
Pm-144
7.0X10-1 1.9X101 7.0X10-1 1.9X101 9.2X101 2.5X103
Pm-145
3.0X101 8.1X102 1.0X101 2.7X102 5.2 1.4X102
Pm-147
4.0X101 1.1X103 2.0 5.4X101 3.4X101 9.3X102
Pm-148m (a)
8.0X10-1 2.2X101 7.0X10-1 1.9X101 7.9X102 2.1X104
Pm-149
2.0 5.4X101 6.0X10-1 1.6X101 1.5X104 4.0X105
Pm-151
2.0 5.4X101 6.0X10-1 1.6X101 2.7X104 7.3X105
Po-210 Polonium (84) 4.0X101 1.1X103 2.0X10-2 5.4X10-1 1.7X102 4.5X103
Pr-142 Praseodymium (59) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.3X104 1.2X106
Pr-143
3.0 8.1X101 6.0X10-1 1.6X101 2.5X103 6.7X104
Pt-188 (a) Platinum (78) 1.0 2.7X101 8.0X10-1 2.2X101 2.5X103 6.8X104
Pt-191
4.0 1.1X102 3.0 8.1X101 8.7X103 2.4X105
Pt-193
4.0X101 1.1X103 4.0X101 1.1X103 1.4 3.7X101
Pt-193m
4.0X101 1.1X103 5.0X10-1 1.4X101 5.8X103 1.6X105
Pt-195m
1.0X101 2.7X102 5.0X10-1 1.4X101 6.2X103 1.7X105
Pt-197
2.0X101 5.4X102 6.0X10-1 1.6X101 3.2X104 8.7X105
Pt-197m
1.0X101 2.7X102 6.0X10-1 1.6X101 3.7X105 1.0X107
Pu-236 Plutonium (94) 3.0X101 8.1X102 3.0X10-3 8.1X10-2 2.0X101 5.3X102
Pu-237
2.0X101 5.4X102 2.0X101 5.4X102 4.5X102 1.2X104
Pu-238
1.0X101 2.7X102 1.0X10-3 2.7X10-2 6.3X10-1 1.7X101
Pu-239
1.0X101 2.7X102 1.0X10-3 2.7X10-2 2.3X10-3 6.2X10-2
Pu-240
1.0X101 2.7X102 1.0X10-3 2.7X10-2 8.4X10-3 2.3X10-1
Pu-241 (a)
4.0X101 1.1X103 6.0X10-2 1.6 3.8 1.0X102
Pu-242
1.0X101 2.7X102 1.0X10-3 2.7X10-2 1.5X10-4 3.9X10-3
Pu-244 (a)
4.0X10-1 1.1X101 1.0X10-3 2.7X10-2 6.7X10-7 1.8X10-5
Ra-223 (a) Radium (88) 4.0X10-1 1.1X101 7.0X10-3 1.9X10-1 1.9X103 5.1X104
Ra-224 (a)
4.0X10-1 1.1X101 2.0X10-2 5.4X10-1 5.9X103 1.6X105
Ra-225 (a)
2.0X10-1 5.4 4.0X10-3 1.1X10-1 1.5X103 3.9X104
Ra-226 (a)
2.0X10-1 5.4 3.0X10-3 8.1X10-2 3.7X10-2 1.0
Ra-228 (a)
6.0X10-1 1.6X101 2.0X10-2 5.4X10-1 1.0X101 2.7X102
Rb-81 Rubidium (37) 2.0 5.4X101 8.0X10-1 2.2X101 3.1X105 8.4X106
Rb-83 (a)
2.0 5.4X101 2.0 5.4X101 6.8X102 1.8X104
Rb-84
1.0 2.7X101 1.0 2.7X101 1.8X103 4.7X104
Rb-86
5.0X10-1 1.4X101 5.0X10-1 1.4X101 3.0X103 8.1X104
Rb-87
Unlimited Unlimited Unlimited Unlimited 3.2X10-9 8.6X10-8
Rb(nat)
Unlimited Unlimited Unlimited Unlimited 6.7X106 1.8X108
Re-184 Rhenium (75) 1.0 2.7X101 1.0 2.7X101 6.9X102 1.9X104
Re-184m
3.0 8.1X101 1.0 2.7X101 1.6X102 4.3X103
Re-186
2.0 5.4X101 6.0X10-1 1.6X101 6.9X103 1.9X105
Re-187
Unlimited Unlimited Unlimited Unlimited 1.4X10-9 3.8X10-8
Re-188
4.0X10-1 1.1X101 4.0X10-1 1.1X101 3.6X104 9.8X105
Re-189 (a)
3.0 8.1X101 6.0X10-1 1.6X101 2.5X104 6.8X105
Re(nat)
Unlimited Unlimited Unlimited Unlimited 0.0 2.4X10-8
Rh-99 Rhodium (45) 2.0 5.4X101 2.0 5.4X101 3.0X103 8.2X104
Rh-101
4.0 1.1X102 3.0 8.1X101 4.1X101 1.1X103
Rh-102
5.0X10-1 1.4X101 5.0X10-1 1.4X101 4.5X101 1.2X103
Rh-102m
2.0 5.4X101 2.0 5.4X101 2.3X102 6.2X103
Rh-103m
4.0X101 1.1X103 4.0X101 1.1X103 1.2X106 3.3X107
Rh-105
1.0X101 2.7X102 8.0X10-1 2.2X101 3.1X104 8.4X105
Rn-222 (a) Radon (86) 3.0X10-1 8.1 4.0X10-3 1.1X10-1 5.7X103 1.5X105
Ru-97 Ruthenium (44) 5.0 1.4X102 5.0 1.4X102 1.7X104 4.6X105
Ru-103 (a)
2.0 5.4X101 2.0 5.4X101 1.2X103 3.2X104
Ru-105
1.0 2.7X101 6.0X10-1 1.6X101 2.5X105 6.7X106
Ru-106 (a)
2.0X10-1 5.4 2.0X10-1 5.4 1.2X102 3.3X103
S-35 Sulphur (16) 4.0X101 1.1X103 3.0 8.1X101 1.6X103 4.3X104
Sb-122 Antimony (51) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.5X104 4.0X105
Sb-124
6.0X10-1 1.6X101 6.0X10-1 1.6X101 6.5X102 1.7X104
Sb-125
2.0 5.4X101 1.0 2.7X101 3.9X101 1.0X103
Sb-126
4.0X10-1 1.1X101 4.0X10-1 1.1X101 3.1X103 8.4X104
Sc-44 Scandium (21) 5.0X10-1 1.4X101 5.0X10-1 1.4X101 6.7X105 1.8X107
Sc-46
5.0X10-1 1.4X101 5.0X10-1 1.4X101 1.3X103 3.4X104
Sc-47
1.0X101 2.7X102 7.0X10-1 1.9X101 3.1X104 8.3X105
Sc-48
3.0X10-1 8.1 3.0X10-1 8.1 5.5X104 1.5X106
Se-75 Selenium (34) 3.0 8.1X101 3.0 8.1X101 5.4X102 1.5X104
Se-79
4.0X101 1.1X103 2.0 5.4X101 2.6X10-3 7.0X10-2
Si-31 Silicon (14) 6.0X10-1 1.6X101 6.0X10-1 1.6X101 1.4X106 3.9X107
Si-32
4.0X101 1.1X103 5.0X10-1 1.4X101 3.9 1.1X102
Sm-145 Samarium (62) 1.0X101 2.7X102 1.0X101 2.7X102 9.8X101 2.6X103
Sm-147
Unlimited Unlimited Unlimited Unlimited 8.5X10-1 2.3X10-8
Sm-151
4.0X101 1.1X103 1.0X101 2.7X102 9.7X10-1 2.6X101
Sm-153
9.0 2.4X102 6.0X10-1 1.6X101 1.6X104 4.4X105
Sn-113 (a) Tin (50) 4.0 1.1X102 2.0 5.4X101 3.7X102 1.0X104
Sn-117m
7.0 1.9X102 4.0X10-1 1.1X101 3.0X103 8.2X104
Sn-119m
4.0X101 1.1X103 3.0X101 8.1X102 1.4X102 3.7X103
Sn-121m (a)
4.0X101 1.1X103 9.0X10-1 2.4X101 2.0 5.4X101
Sn-123
8.0X10-1 2.2X101 6.0X10-1 1.6X101 3.0X102 8.2X103
Sn-125
4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.0X103 1.1X105
Sn-126 (a)
6.0X10-1 1.6X101 4.0X10-1 1.1X101 1.0X10-3 2.8X10-2
Sr-82 (a) Strontium (38) 2.0X10-1 5.4 2.0X10-1 5.4 2.3X103 6.2X104
Sr-85
2.0 5.4X101 2.0 5.4X101 8.8X102 2.4X104
Sr-85m
5.0 1.4X102 5.0 1.4X102 1.2X106 3.3X107
Sr-87m
3.0 8.1X101 3.0 8.1X101 4.8X105 1.3X107
Sr-89
6.0X10-1 1.6X101 6.0X10-1 1.6X101 1.1X103 2.9X104
Sr-90 (a)
3.0X10-1 8.1 3.0X10-1 8.1 5.1 1.4X102
Sr-91 (a)
3.0X10-1 8.1 3.0X10-1 8.1 1.3X105 3.6X106
Sr-92 (a)
1.0 2.7X101 3.0X10-1 8.1 4.7X105 1.3X107
T(H-3) Tritium (1) 4.0X101 1.1X103 4.0X101 1.1X103 3.6X102 9.7X103
Ta-178 (long-lived) Tantalum (73) 1.0 2.7X101 8.0X10-1 2.2X101 4.2X106 1.1X108
Ta-179
3.0X101 8.1X102 3.0X101 8.1X102 4.1X101 1.1X103
Ta-182
9.0X10-1 2.4X101 5.0X10-1 1.4X101 2.3X102 6.2X103
Tb-157 Terbium (65) 4.0X101 1.1X103 4.0X101 1.1X103 5.6X10-1 1.5X101
Tb-158
1.0 2.7X101 1.0 2.7X101 5.6X10-1 1.5X101
Tb-160
1.0 2.7X101 6.0X10-1 1.6X101 4.2X102 1.1X104
Tc-95m (a) Technetium (43) 2.0 5.4X101 2.0 5.4X101 8.3X102 2.2X104
Tc-96
4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.2X104 3.2X105
Tc-96m (a)
4.0X10-1 1.1X101 4.0X10-1 1.1X101 1.4X106 3.8X107
Tc-97
Unlimited Unlimited Unlimited Unlimited 5.2X10-5 1.4X10-3
Tc-97m
4.0X101 1.1X103 1.0 2.7X101 5.6X102 1.5X104
Tc-98
8.0X10-1 2.2X101 7.0X10-1 1.9X101 3.2X10-5 8.7X10-4
Tc-99
4.0X101 1.1X103 9.0X10-1 2.4X101 6.3X10-4 1.7X10-2
Tc-99m
1.0X101 2.7X102 4.0 1.1X102 1.9X105 5.3X106
Te-121 Tellurium (52) 2.0 5.4X101 2.0 5.4X101 2.4X103 6.4X104
Te-121m
5.0 1.4X102 3.0 8.1X101 2.6X102 7.0X103
Te-123m
8.0 2.2X102 1.0 2.7X101 3.3X102 8.9X103
Te-125m
2.0X101 5.4X102 9.0X10-1 2.4X101 6.7X102 1.8X104
Te-127
2.0X101 5.4X102 7.0X10-1 1.9X101 9.8X104 2.6X106
Te-127m (a)
2.0X101 5.4X102 5.0X10-1 1.4X101 3.5X102 9.4X103
Te-129
7.0X10-1 1.9X101 6.0X10-1 1.6X101 7.7X105 2.1X107
Te-129m (a)
8.0X10-1 2.2X101 4.0X10-1 1.1X101 1.1X103 3.0X104
Te-131m (a)
7.0X10-1 1.9X101 5.0X10-1 1.4X101 3.0X104 8.0X105
Te-132 (a)
5.0X10-1 1.4X101 4.0X10-1 1.1X101 1.1X104 3.0X105
Th-227 Thorium (90) 1.0X101 2.7X102 5.0X10-3 1.4X10-1 1.1X103 3.1X104
Th-228 (a)
5.0X10-1 1.4X101 1.0X10-3 2.7X10-2 3.0X101 8.2X102
Th-229
5.0 1.4X102 5.0X10-4 1.4X10-2 7.9X10-3 2.1X10-1
Th-230
1.0X101 2.7X102 1.0X10-3 2.7X10-2 7.6X10-4 2.1X10-2
Th-231
4.0X101 1.1X103 2.0X10-2 5.4X10-1 2.0X104 5.3X105
Th-232
Unlimited Unlimited Unlimited Unlimited 4.0X10-9 1.1X10-7
Th-234 (a)
3.0X10-1 8.1 3.0X10-1 8.1 8.6X102 2.3X104
Th(nat)
Unlimited Unlimited Unlimited Unlimited 8.1X10-9 2.2X10-7
Ti-44 (a) Titanium (22) 5.0X10-1 1.4X101 4.0X10-1 1.1X101 6.4 1.7X102
Tl-200 Thallium (81) 9.0X10-1 2.4X101 9.0X10-1 2.4X101 2.2X104 6.0X105
Tl-201
1.0X101 2.7X102 4.0 1.1X102 7.9X103 2.1X105
Tl-202
2.0 5.4X101 2.0 5.4X101 2.0X103 5.3X104
Tl-204
1.0X101 2.7X102 7.0X10-1 1.9X101 1.7X101 4.6X102
Tm-167 Thulium (69) 7.0 1.9X102 8.0X10-1 2.2X101 3.1X103 8.5X104
Tm-170
3.0 8.1X101 6.0X10-1 1.6X101 2.2X102 6.0X103
Tm-171
4.0X101 1.1X103 4.0X101 1.1X103 4.0X101 1.1X103
U-230 (fast lung absorption) (a)(d) Uranium (92) 4.0X101 1.1X103 1.0X10-1 2.7 1.0X103 2.7X104
U-230 (medium lung absorption) (a)(e)
4.0X101 1.1X103 4.0X10-3 1.1X10-1 1.0X103 2.7X104
U-230 (slow lung absorption) (a)(f)
3.0X101 8.1X102 3.0X10-3 8.1X10-2 1.0X103 2.7X104
U-232 (fast lung absorption) (d)
4.0X101 1.1X103 1.0X10-2 2.7X10-1 8.3X10-1 2.2X101
U-232 (medium lung absorption) (e)
4.0X101 1.1X103 7.0X10-3 1.9X10-1 8.3X10-1 2.2X101
U-232 (slow lung absorption) (f)
1.0X101 2.7X102 1.0X10-3 2.7X10-2 8.3X10-1 2.2X101
U-233 (fast lung absorption) (d)
4.0X101 1.1X103 9.0X10-2 2.4 3.6X10-4 9.7X10-3
U-233 (medium lung absorption) (e)
4.0X101 1.1X103 2.0X10-2 5.4X10-1 3.6X10-4 9.7X10-3
U-233 (slow lung absorption) (f)
4.0X101 1.1X103 6.0X10-3 1.6X10-1 3.6X10-4 9.7X10-3
U-234 (fast lung absorption) (d)
4.0X101 1.1X103 9.0X10-2 2.4 2.3X10-4 6.2X10-3
U-234 (medium lung absorption) (e)
4.0X101 1.1X103 2.0X10-2 5.4X10-1 2.3X10-4 6.2X10-3
U-234 (slow lung absorption) (f)
4.0X101 1.1X103 6.0X10-3 1.6X10-1 2.3X10-4 6.2X10-3
U-235 (all lung absorption types) (a),(d),(e),(f)
Unlimited Unlimited Unlimited Unlimited 8.0X10-8 2.2X10-6
U-236 (fast lung absorption) (d)
Unlimited Unlimited Unlimited Unlimited 2.4X10-6 6.5X10-5
U-236 (medium lung absorption) (e)
4.0X101 1.1X103 2.0X10-2 5.4X10-1 2.4X10-6 6.5X10-5
U-236 (slow lung absorption) (f)
4.0X101 1.1X103 6.0X10-3 1.6X10-1 2.4X10-6 6.5X10-5
U-238 (all lung absorption types) (d),(e),(f)
Unlimited Unlimited Unlimited Unlimited 1.2X10-8 3.4X10-7
U (nat)
Unlimited Unlimited Unlimited Unlimited 2.6X10-8 7.1X10-7
U (enriched to 20% or less) (g)
Unlimited Unlimited Unlimited Unlimited See Table A-4 See Table A-4
U (dep)
Unlimited Unlimited Unlimited Unlimited See Table A-4 (See Table A-3)
V-48 Vanadium (23) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 6.3X103 1.7X105
V-49
4.0X101 1.1X103 4.0X101 1.1X103 3.0X102 8.1X103
W-178 (a) Tungsten (74) 9.0 2.4X102 5.0 1.4X102 1.3X103 3.4X104
W-181
3.0X101 8.1X102 3.0X101 8.1X102 2.2X102 6.0X103
W-185
4.0X101 1.1X103 8.0X10-1 2.2X101 3.5X102 9.4X103
W-187
2.0 5.4X101 6.0X10-1 1.6X101 2.6X104 7.0X105
W-188 (a)
4.0X10-1 1.1X101 3.0X10-1 8.1 3.7X102 1.0X104
Xe-122 (a) Xenon (54) 4.0X10-1 1.1X101 4.0X10-1 1.1X101 4.8X104 1.3X106
Xe-123
2.0 5.4X101 7.0X10-1 1.9X101 4.4X105 1.2X107
Xe-127
4.0 1.1X102 2.0 5.4X101 1.0X103 2.8X104
Xe-131m
4.0X101 1.1X103 4.0X101 1.1X103 3.1X103 8.4X104
Xe-133
2.0X101 5.4X102 1.0X101 2.7X102 6.9X103 1.9X105
Xe-135
3.0 8.1X101 2.0 5.4X101 9.5X104 2.6X106
Y-87 (a) Yttrium (39) 1.0 2.7X101 1.0 2.7X101 1.7X104 4.5X105
Y-88
4.0X10-1 1.1X101 4.0X10-1 1.1X101 5.2X102 1.4X104
Y-90
3.0X10-1 8.1 3.0X10-1 8.1 2.0X104 5.4X105
Y-91
6.0X10-1 1.6X101 6.0X10-1 1.6X101 9.1X102 2.5X104
Y-91m
2.0 5.4X101 2.0 5.4X101 1.5X106 4.2X107
Y-92
2.0X10-1 5.4 2.0X10-1 5.4 3.6X105 9.6X106
Y-93
3.0X10-1 8.1 3.0X10-1 8.1 1.2X105 3.3X106
Yb-169 Ytterbium (70) 4.0 1.1X102 1.0 2.7X101 8.9X102 2.4X104
Yb-175
3.0X101 8.1X102 9.0X10-1 2.4X101 6.6X103 1.8X105
Zn-65 Zinc (30) 2.0 5.4X101 2.0 5.4X101 3.0X102 8.2X103
Zn-69
3.0 8.1X101 6.0X10-1 1.6X101 1.8X106 4.9X107
Zn-69m (a)
3.0 8.1X101 6.0X10-1 1.6X101 1.2X105 3.3X106
Zr-88 Zirconium (40) 3.0 8.1X101 3.0 8.1X101 6.6X102 1.8X104
Zr-93
Unlimited Unlimited Unlimited Unlimited 9.3X10-5 2.5X10-3
Zr-95 (a)
2.0 5.4X101 8.0X10-1 2.2X101 7.9X102 2.1X104
Zr-97 (a)
4.0X10-1 1.1X101 4.0X10-1 1.1X101 7.1X104 1.9X106

a A1 and/or A2 values include contributions from daughter nuclides with half-lives less than 10 days.
b The values of A1 and A2 in Curies (Ci) are approximate and for information only; the regulatory standard units are Terabecquerels (TBq) (see Appendix A to Part 71—Determination of A1 and A2, Section I).
c The quantity may be determined from a measurement of the rate of decay or a measurement of the radiation level at a prescribed distance from the source.
d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of transport.
e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident conditions of transport.
f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.
g These values apply to unirradiated uranium only.
h A1 = 0.1 TBq (2.7 Ci) and A2 = 0.001 TBq (0.027 Ci) for Cf-252 for domestic use.
i A2 = 0.74 TBq (20 Ci) for Mo-99 for domestic use.

Table A-2—EXEMPT MATERIAL ACTIVITY CONCENTRATIONS AND EXEMPT CONSIGNMENT ACTIVITY LIMITS FOR RADIONUCLIDES

Symbol of radionuclide Element and atomic number Activity concentration for exempt material (Bq/g) Activity concentration for exempt material (Ci/g) Activity limit for exempt consignment (Bq) Activity limit for exempt consignment (Ci)
Ac-225 Actinium (89) 1.0X101 2.7X10-10 1.0X104 2.7X10-7
Ac-227
1.0X10-1 2.7X10-12 1.0X103 2.7X10-8
Ac-228
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Ag-105 Silver (47) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ag-108m (b)
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Ag-110m
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Ag-111
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Al-26 Aluminum (13) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
Am-241 Americium (95) 1.0 2.7X10-11 1.0X104 2.7X10-7
Am-242m (b)
1.0 2.7X10-11 1.0X104 2.7X10-7
Am-243 (b)
1.0 2.7X10-11 1.0X103 2.7X10-8
Ar-37 Argon (18) 1.0X106 2.7X10-5 1.0X108 2.7X10-3
Ar-39
1.0X107 2.7X10-4 1.0X104 2.7X10-7
Ar-41
1.0X102 2.7X10-9 1.0X109 2.7X10-2
As-72 Arsenic (33) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
As-73
1.0X103 2.7X10-8 1.0X107 2.7X10-4
As-74
1.0X101 2.7X10-10 1.0X106 2.7X10-5
As-76
1.0X102 2.7X10-9 1.0X105 2.7X10-6
As-77
1.0X103 2.7X10-8 1.0X106 2.7X10-5
At-211 Astatine (85) 1.0X103 2.7X10-8 1.0X107 2.7X10-4
Au-193 Gold (79) 1.0X102 2.7X10-9 1.0X107 2.7X10-4
Au-194
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Au-195
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Au-198
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Au-199
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ba-131 Barium (56) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ba-133
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ba-133m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ba-140 (b)
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Be-7 Beryllium (4) 1.0X103 2.7X10-8 1.0X107 2.7X10-4
Be-10
1.0X104 2.7X10-7 1.0X106 2.7X10-5
Bi-205 Bismuth (83) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Bi-206
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Bi-207
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Bi-210
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Bi-210m
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Bi-212 (b)
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Bk-247 Berkelium (97) 1.0 2.7X10-11 1.0X104 2.7X10-7
Bk-249
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Br-76 Bromine (35) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
Br-77
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Br-82
1.0X101 2.7X10-10 1.0X106 2.7X10-5
C-11 Carbon (6) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
C-14
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Ca-41 Calcium (20) 1.0X105 2.7X10-6 1.0X107 2.7X10-4
Ca-45
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Ca-47
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Cd-109 Cadmium (48) 1.0X104 2.7X10-7 1.0X106 2.7X10-5
Cd-113m
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Cd-115
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Cd-115m
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Ce-139 Cerium (58) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ce-141
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Ce-143
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ce-144 (b)
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Cf-248 Californium (98) 1.0X101 2.7X10-10 1.0X104 2.7X10-7
Cf-249
1.0 2.7X10-11 1.0X103 2.7X10-8
Cf-250
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Cf-251
1.0 2.7X10-11 1.0X103 2.7X10-8
Cf-252
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Cf-253
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Cf-254
1.0 2.7X10-11 1.0X103 2.7X10-8
Cl-36 Chlorine (17) 1.0X104 2.7X10-7 1.0X106 2.7X10-5
Cl-38
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Cm-240 Curium (96) 1.0X102 2.7X10-9 1.0X105 2.7X10-6
Cm-241
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Cm-242
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Cm-243
1.0 2.7X10-11 1.0X104 2.7X10-7
Cm-244
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Cm-245
1.0 2.7X10-11 1.0X103 2.7X10-8
Cm-246
1.0 2.7X10-11 1.0X103 2.7X10-8
Cm-247
1.0 2.7X10-11 1.0X104 2.7X10-7
Cm-248
1.0 2.7X10-11 1.0X103 2.7X10-8
Co-55 Cobalt (27) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Co-56
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Co-57
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Co-58
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Co-58m
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Co-60
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Cr-51 Chromium (24) 1.0X103 2.7X10-8 1.0X107 2.7X10-4
Cs-129 Cesium (55) 1.0X102 2.7X10-9 1.0X105 2.7X10-6
Cs-131
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Cs-132
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Cs-134
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Cs-134m
1.0X103 2.7X10-8 1.0X105 2.7X10-6
Cs-135
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Cs-136
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Cs-137 (b)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Cu-64 Copper (29) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Cu-67
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Dy-159 Dysprosium (66) 1.0X103 2.7X10-8 1.0X107 2.7X10-4
Dy-165
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Dy-166
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Er-169 Erbium (68) 1.0X104 2.7X10-7 1.0X107 2.7X10-4
Er-171
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Eu-147 Europium (63) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Eu-148
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Eu-149
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Eu-150 (short lived)
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Eu-150 (long lived)
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Eu-152
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Eu-152m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Eu-154
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Eu-155
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Eu-156
1.0X101 2.7X10-10 1.0X106 2.7X10-5
F-18 Fluorine (9) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Fe-52 Iron (26) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Fe-55
1.0X104 2.7X10-7 1.0X106 2.7X10-5
Fe-59
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Fe-60
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Ga-67 Gallium (31) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ga-68
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Ga-72
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Gd-146 Gadolinium (64) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Gd-148
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Gd-153
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Gd-159
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Ge-68 Germanium (32) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
Ge-71
1.0X104 2.7X10-7 1.0X108 2.7X10-3
Ge-77
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Hf-172 Hafnium (72) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Hf-175
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Hf-181
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Hf-182
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Hg-194 Mercury (80) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Hg-195m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Hg-197
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Hg-197m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Hg-203
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Ho-166 Holmium (67) 1.0X103 2.7X10-8 1.0X105 2.7X10-6
Ho-166m
1.0X101 2.7X10-10 1.0X106 2.7X10-5
I-123 Iodine (53) 1.0X102 2.7X10-9 1.0X107 2.7X10-4
I-124
1.0X101 2.7X10-10 1.0X106 2.7X10-5
I-125
1.0X103 2.7X10-8 1.0X106 2.7X10-5
I-126
1.0X102 2.7X10-9 1.0X106 2.7X10-5
I-129
1.0X102 2.7X10-9 1.0X105 2.7X10-6
I-131
1.0X102 2.7X10-9 1.0X106 2.7X10-5
I-132
1.0X101 2.7X10-10 1.0X105 2.7X10-6
I-133
1.0X101 2.7X10-10 1.0X106 2.7X10-5
I-134
1.0X101 2.7X10-10 1.0X105 2.7X10-6
I-135
1.0X101 2.7X10-10 1.0X106 2.7X10-5
In-111 Indium (49) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
In-113m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
In-114m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
In-115m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ir-189 Iridium (77) 1.0X102 2.7X10-9 1.0X107 2.7X10-4
Ir-190
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Ir-192
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Ir-194
1.0X102 2.7X10-9 1.0X105 2.7X10-6
K-40 Potassium (19) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
K-42
1.0X102 2.7X10-9 1.0X106 2.7X10-5
K-43
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Kr-81 Krypton (36) 1.0X104 2.7X10-7 1.0X107 2.7X10-4
Kr-85
1.0X105 2.7X10-6 1.0X104 2.7X10-7
Kr-85m
1.0X103 2.7X10-8 1.0X1010 2.7X10-1
Kr-87
1.0X102 2.7X10-9 1.0X109 2.7X10-2
La-137 Lanthanum (57) 1.0X103 2.7X10-8 1.0X107 2.7X10-4
La-140
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Lu-172 Lutetium (71) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Lu-173
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Lu-174
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Lu-174m
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Lu-177
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Mg-28 Magnesium (12) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
Mn-52 Manganese (25) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
Mn-53
1.0X104 2.7X10-7 1.0X109 2.7X10-2
Mn-54
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Mn-56
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Mo-93 Molybdenum (42) 1.0X103 2.7X10-8 1.0X108 2.7X10-3
Mo-99
1.0X102 2.7X10-9 1.0X106 2.7X10-5
N-13 Nitrogen (7) 1.0X102 2.7X10-9 1.0X109 2.7X10-2
Na-22 Sodium (11) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Na-24
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Nb-93m Niobium (41) 1.0X104 2.7X10-7 1.0X107 2.7X10-4
Nb-94
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Nb-95
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Nb-97
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Nd-147 Neodymium (60) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Nd-149
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ni-59 Nickel (28) 1.0X104 2.7X10-7 1.0X108 2.7X10-3
Ni-63
1.0X105 2.7X10-6 1.0X108 2.7X10-3
Ni-65
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Np-235 Neptunium (93) 1.0X103 2.7X10-8 1.0X107 2.7X10-4
Np-236 (short-lived)
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Np-236 (long-lived)
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Np-237 (b)
1.0 2.7X10-11 1.0X103 2.7X10-8
Np-239
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Os-185 Osmium (76) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Os-191
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Os-191m
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Os-193
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Os-194
1.0X102 2.7X10-9 1.0X105 2.7X10-6
P-32 Phosphorus (15) 1.0X103 2.7X10-8 1.0X105 2.7X10-6
P-33
1.0X105 2.7X10-6 1.0X108 2.7X10-3
Pa-230 Protactinium (91) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Pa-231
1.0 2.7X10-11 1.0X103 2.7X10-8
Pa-233
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Pb-201 Lead (82) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Pb-202
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Pb-203
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Pb-205
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Pb-210 (b)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Pb-212 (b)
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Pd-103 Palladium (46) 1.0X103 2.7X10-8 1.0X108 2.7X10-3
Pd-107
1.0X105 2.7X10-6 1.0X108 2.7X10-3
Pd-109
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Pm-143 Promethium (61) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Pm-144
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Pm-145
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Pm-147
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Pm-148m
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Pm-149
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Pm-151
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Po-210 Polonium (84) 1.0X101 2.7X10-10 1.0X104 2.7X10-7
Pr-142 Praseodymium (59) 1.0X102 2.7X10-9 1.0X105 2.7X10-6
Pr-143
1.0X104 2.7X10-7 1.0X106 2.7X10-5
Pt-188 Platinum (78) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Pt-191
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Pt-193
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Pt-193m
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Pt-195m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Pt-197
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Pt-197m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Pu-236 Plutonium (94) 1.0X101 2.7X10-10 1.0X104 2.7X10-7
Pu-237
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Pu-238
1.0 2.7X10-11 1.0X104 2.7X10-7
Pu-239
1.0 2.7X10-11 1.0X104 2.7X10-7
Pu-240
1.0 2.7X10-11 1.0X103 2.7X10-8
Pu-241
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Pu-242
1.0 2.7X10-11 1.0X104 2.7X10-7
Pu-244
1.0 2.7X10-11 1.0X104 2.7X10-7
Ra-223 (b) Radium (88) 1.0X102 2.7X10-9 1.0X105 2.7X10-6
Ra-224 (b)
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Ra-225
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Ra-226 (b)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Ra-228 (b)
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Rb-81 Rubidium (37) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Rb-83
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Rb-84
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Rb-86
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Rb-87
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Rb(nat)
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Re-184 Rhenium (75) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Re-184m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Re-186
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Re-187
1.0X106 2.7X10-5 1.0X109 2.7X10-2
Re-188
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Re-189
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Re(nat)
1.0X106 2.7X10-5 1.0X109 2.7X10-2
Rh-99 Rhodium (45) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Rh-101
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Rh-102
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Rh-102m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Rh-103m
1.0X104 2.7X10-7 1.0X108 2.7X10-3
Rh-105
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Rn-222 (b) Radon (86) 1.0X101 2.7X10-10 1.0X108 2.7X10-3
Ru-97 Ruthenium (44) 1.0X102 2.7X10-9 1.0X107 2.7X10-4
Ru-103
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Ru-105
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Ru-106 (b)
1.0X102 2.7X10-9 1.0X105 2.7X10-6
S-35 Sulphur (16) 1.0X105 2.7X10-6 1.0X108 2.7X10-3
Sb-122 Antimony (51) 1.0X102 2.7X10-9 1.0X104 2.7X10-7
Sb-124
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Sb-125
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Sb-126
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Sc-44 Scandium (21) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
Sc-46
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Sc-47
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Sc-48
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Se-75 Selenium (34) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Se-79
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Si-31 Silicon (14) 1.0X103 2.7X10-8 1.0X106 2.7X10-5
Si-32
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Sm-145 Samarium (62) 1.0X102 2.7X10-9 1.0X107 2.7X10-4
Sm-147
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Sm-151
1.0X104 2.7X10-7 1.0X108 2.7X10-3
Sm-153
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Sn-113 Tin (50) 1.0X103 2.7X10-8 1.0X107 2.7X10-4
Sn-117m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Sn-119m
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Sn-121m
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Sn-123
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Sn-125
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Sn-126
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Sr-82 Strontium (38) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
Sr-85
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Sr-85m
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Sr-87m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Sr-89
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Sr-90 (b)
1.0X102 2.7X10-9 1.0X104 2.7X10-7
Sr-91
1.0X101 2.7X10-10 1.0X105 2.7X10-6
Sr-92
1.0X101 2.7X10-10 1.0X106 2.7X10-5
T(H-3) Tritium (1) 1.0X106 2.7X10-5 1.0X109 2.7X10-2
Ta-178 (long-lived) Tantalum (73) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Ta-179
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Ta-182
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Tb-157 Terbium (65) 1.0X104 2.7X10-7 1.0X107 2.7X10-4
Tb-158
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Tb-160
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Tc-95m Technetium (43) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Tc-96
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Tc-96m
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Tc-97
1.0X103 2.7X10-8 1.0X108 2.7X10-3
Tc-97m
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Tc-98
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Tc-99
1.0X104 2.7X10-7 1.0X107 2.7X10-4
Tc-99m
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Te-121 Tellurium (52) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Te-121m
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Te-123m
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Te-125m
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Te-127
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Te-127m
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Te-129
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Te-129m
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Te-131m
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Te-132
1.0X102 2.7X10-9 1.0X107 2.7X10-4
Th-227 Thorium (90) 1.0X101 2.7X10-10 1.0X104 2.7X10-7
Th-228 (b)
1.0 2.7X10-11 1.0X104 2.7X10-7
Th-229 (b)
1.0 2.7X10-11 1.0X103 2.7X10-8
Th-230
1.0 2.7X10-11 1.0X104 2.7X10-7
Th-231
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Th-232
1.0X101 2.7X10-10 1.0X104 2.7X10-7
Th-234 (b)
1.0X103 2.7X10-8 1.0X105 2.7X10-6
Th (nat) (b)
1.0 2.7X10-11 1.0X103 2.7X10-8
Ti-44 Titanium (22) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
Tl-200 Thallium (81) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Tl-201
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Tl-202
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Tl-204
1.0X104 2.7X10-7 1.0X104 2.7X10-7
Tm-167 Thulium (69) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Tm-170
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Tm-171
1.0X104 2.7X10-7 1.0X108 2.7X10-3
U-230 (fast lung absorption) (b),(d) Uranium (92) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
U-230 (medium lung absorption) (e)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U-230 (slow lung absorption) (f)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U-232 (fast lung absorption) (b),(d)
1.0 2.7X10-11 1.0X103 2.7X10-8
U-232 (medium lung absorption) (e)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U-232 (slow lung absorption) (f)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U-233 (fast lung absorption) (d)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U-233 (medium lung absorption) (e)
1.0X102 2.7X10-9 1.0X105 2.7X10-6
U-233 (slow lung absorption) (f)
1.0X101 2.7X10-10 1.0X105 2.7X10-6
U-234 (fast lung absorption) (d)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U-234 (medium lung absorption) (e)
1.0X102 2.7X10-9 1.0X105 2.7X10-6
U-234 (slow lung absorption) (f)
1.0X101 2.7X10-10 1.0X105 2.7X10-6
U-235 (all lung absorption types) (b),(d),(e),(f )
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U-236 (fast lung absorption) (d)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U-236 (medium lung absorption) (e)
1.0X102 2.7X10-9 1.0X105 2.7X10-6
U-236 (slow lung absorption) (f)
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U-238 (all lung absorption types) (b),(d),(e),(f )
1.0X101 2.7X10-10 1.0X104 2.7X10-7
U (nat) (b)
1.0 2.7X10-11 1.0X103 2.7X10-8
U (enriched to 20% or less) (g)
1.0 2.7X10-11 1.0X103 2.7X10-8
U (dep)
1.0 2.7X10-11 1.0X103 2.7X10-8
V-48 Vanadium (23) 1.0X101 2.7X10-10 1.0X105 2.7X10-6
V-49
1.0X104 2.7X10-7 1.0X107 2.7X10-4
W-178 Tungsten (74) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
W-181
1.0X103 2.7X10-8 1.0X107 2.7X10-4
W-185
1.0X104 2.7X10-7 1.0X107 2.7X10-4
W-187
1.0X102 2.7X10-9 1.0X106 2.7X10-5
W-188
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Xe-122 Xenon (54) 1.0X102 2.7X10-9 1.0X109 2.7X10-2
Xe-123
1.0X102 2.7X10-9 1.0X109 2.7X10-2
Xe-127
1.0X103 2.7X10-8 1.0X105 2.7X10-6
Xe-131m
1.0X104 2.7X10-7 1.0X104 2.7X10-7
Xe-133
1.0X103 2.7X10-8 1.0X104 2.7X10-7
Xe-135
1.0X103 2.7X10-8 1.0X1010 2.7X10-1
Y-87 Yttrium (39) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Y-88
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Y-90
1.0X103 2.7X10-8 1.0X105 2.7X10-6
Y-91
1.0X103 2.7X10-8 1.0X106 2.7X10-5
Y-91m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Y-92
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Y-93
1.0X102 2.7X10-9 1.0X105 2.7X10-6
Yb-169 Ytterbium (70) 1.0X102 2.7X10-9 1.0X107 2.7X10-4
Yb-175
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Zn-65 Zinc (30) 1.0X101 2.7X10-10 1.0X106 2.7X10-5
Zn-69
1.0X104 2.7X10-7 1.0X106 2.7X10-5
Zn-69m
1.0X102 2.7X10-9 1.0X106 2.7X10-5
Zr-88 Zirconium (40) 1.0X102 2.7X10-9 1.0X106 2.7X10-5
Zr-93 (b)
1.0X103 2.7X10-8 1.0X107 2.7X10-4
Zr-95
1.0X101 2.7X10-10 1.0X106 2.7X10-5
Zr-97 (b)
1.0X101 2.7X10-10 1.0X105 2.7X10-6

a [Reserved]
b Parent nuclides and their progeny included in secular equilibrium are listed in the following:

Sr-90 Y-90
Zr-93 Nb-93m
Zr-97 Nb-97
Ru-106 Rh-106
Cs-137 Ba-137m
Ce-134 La-134
Ce-144 Pr-144
Ba-140 La-140
Bi-212 Tl-208 (0.36), Po-212 (0.64)
Pb-210 Bi-210, Po-210
Pb-212 Bi-212, Tl-208 (0.36), Po-212 (0.64)
Rn-220 Po-216
Rn-222 Po-218, Pb-214, Bi-214, Po-214
Ra-223 Rn-219, Po-215, Pb-211, Bi-211, Tl-207
Ra-224 Rn-220, Po-216, Pb-212, Bi-212, Tl-208(0.36), Po-212 (0.64)
Ra-226 Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210
Ra-228 Ac-228
Th-226 Ra-222, Rn-218, Po-214
Th-228 Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Th-229 Ra-225, Ac-225, Fr-221, At-217, Bi-213, Po-213, Pb-209
Th-nat Ra-228, Ac-228, Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Th-234 Pa-234m
U-230 Th-226, Ra-222, Rn-218, Po-214
U-232 Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
U-235 Th-231
U-238 Th-234, Pa-234m
U-nat Th-234, Pa-234m, U-234, Th-230, Ra-226, Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210
U-240 Np-240m
Np-237 Pa-233
Am-242m Am-242
Am-243 Np-239

c [Reserved]
d These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of transport.
e These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident conditions of transport.
f These values apply to all compounds of uranium other than those specified in notes (d) and (e) of this table.
g These values apply to unirradiated uranium only.

TABLE A-3—GENERAL VALUES FOR A1 AND A2

Contents A1 A2 Activity concentration for exempt material
(Bq/g)
Activity concentration for exempt material (Ci/g) Activity limits for exempt consignments (Bq) Activity limits for exempt consignments (Ci)
(TBq) (Ci) (TBq) (Ci)
Only beta or gamma emitting radionuclides are known to be present 1 x 10-1 2.7 x 100 2 x 10 -2 5.4 x 10-1 1 x 101 2.7 x10-10 1 x 104 2.7 x10-7
Only alpha emitting
radionuclides are known
to be present
2 x 10-1 5.4 x 100 9 x 10-5 2.4 x 10-3 1 x 10-1 2.7 x10-12 1 x 103 2.7 x10-8
No relevant data are available 1 x 10-3 2.7 x 10-2 9 x 10-5 2.4 x 10-3 1 x 10-1 2.7 x 10-12 1 x 103 2.7 x 10-8

TABLE A-4—ACTIVITY-MASS RELATIONSHIPS FOR URANIUM

Uranium Enrichment1
wt % U-235 present
Specific Activity
TBq/g Ci/g
0.45 1.8 x 10-8 5.0 x 10-7
0.72 2.6 x 10-8 7.1 x 10-7
1 2.8 x 10-8 7.6 x 10-7
1.5 3.7 x 10-8 1.0 x 10-6
5 1.0 x 10-7 2.7 x 10-6
10 1.8 x 10-7 4.8 x 10-6
20 3.7 x 10-7 1.0 x 10-5
35 7.4 x 10-7 2.0 x 10-5
50 9.3 x 10-7 2.5 x 10-5
90 2.2 x 10-6 5.8 x 10-5
93 2.6 x 10-6 7.0 x 10-5
95 3.4 x 10-6 9.1 x 10-5

1 The figures for uranium include representative values for the activity of the uranium-234 that is concentrated during the enrichment process.

[60 FR 50264, Sept. 28, 1995 as amended at 61 FR 28724, June 6, 1996; 69 FR 3800, Jan. 26, 2004; 77 FR 39908, Jul. 6, 2012]

Page Last Reviewed/Updated Wednesday, July 31, 2013