ML23156A052
| ML23156A052 | |
| Person / Time | |
|---|---|
| Issue date: | 02/19/1998 |
| From: | NRC/SECY |
| To: | |
| References | |
| PRM-071-012, 63FR08362 | |
| Download: ML23156A052 (1) | |
Text
ADAMS Template: SECY-067 DOCUMENT DATE: 02/19/1998 TITLE: PRM-071-012 - 63FR08362 - INTERNATIONAL ENERGY CONSULTANTS, INC.; RECEIPT OF PETITION FOR RULEMAKING CASE
REFERENCE:
PRM-071-012 63FR08362 KEYWORD: RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete
DOCKET NO. PRM-71-12 (63FR08362)
In the Matter of INTERNATIONAL ENERGY CONSULTANTS, INC.; RECEIPT OF PETITION FOR RULEMAKING DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT
~ 10/22/'>7 ** otif'l.5/97 LTR FM FRANK P. FALCI, PRESIDENT, INT'L ENERGY CONSULTANTS, INC. FILING PETITION FOR RULEMAKING RE SPECIAL REQUIREMENTS FOR PLUTONIUM SHIPMENTS 02/12/98 02/11/98 FEDERAL REGISTER NOTICE - RECEIPT OF PETITION FOR RULEMAKING 03/24/98 03/20/98 COMMENT OF FRANK SMITH ( 1) 03/25/98 03/20/98 COMMENT OF ANN R. SMITH ( 2) 04/27/98 04/22/98 COMMENT OF ENVIRONMENTAL EVALUATION GROUP (ROBERT H. NEILL, DIRECTOR) ( 3) 04/30/98 04/29/98 LTR FM ROBERT H. NEILL, DIRECTOR, EEG, PROVIDING SUPPLEMENT TO COMMENT NO. 3. (PREVIOUSLY ASSIGNED AS COMMENT NO. 4) 5/05/98 05/05/98 COMMENT OF DEPARTMENT OF ENERGY (KELVIN J. KELKENBERG) ( 5) 05/06/98 05/04/98 COMMENT OF WESTERN GOVERNORS' ASSOCIATION (JAMES M. SOUBY, EXECUTIVE DIRECTOR) ( 6) 05/07/98 05/05/98 COMMENT OF TRANSNUCLEAR, INC.
(JOHN MANGUS!, GEN. MGR. - OPS.) ( 7) 05/18/98 05/15/98 COMMENT OF JOHN C. RODGERS, CHP ( 8) 05/26/98 04/13/98 COMMENT OF DOE - OFC. OF CIVILIAN RADIOACTIVE WASTE MGMT.
(LAKE H. BARRETT, ACTING DIRECTOR) ( 9) 05/26/98 05/20/98 LTR FM FRANK P. FALCI, PRESIDENT, INT'L ENERGY CONSULTANTS, INC., TRANSMITTING LAKE H. BARRETT LETTER (COMMENT NO. 9) 06/02/98 05/26/98 COMMENT OF GENERAL ATOMICS (AL ZIMMER) ( 10) 06/18/98 06/17/98 FEDERAL REGISTER NOTICE - PETITON FOR RULEMAKING:
EXTENSION OF COMMENT PERIOD
I .
IE C ,,..- International Energy Consultants, Inc.
8905 Copenhaver Drive D0CKlr Potomac, Maryland 20854 H!IIIDJFU.1 - 7£--/2, Tel: (301) 340-1047
('1>3FA. 83~) Fax: (301) 340-2229 DOCKETED February 6, 2004 USNRC Ms. Annette L. Vietti-Cook March 15, 2004 (11 :31 AM)
Secretary of the Commission United States Nuclear Regulatory Commission OFFICE OF THE SECRETARY Washington, D.C. 20555-0001 RULEMAKINGS AND ADJUDICATIONS STAFF
Dear Ms. Vietti-Cook:
Thank you for your letter dated December 29, 2003 in which you announce the Nuclear Regulatory Commission's (NRC's) decision on my petition for rulemaking, docketed as PRM-71-12, which I submitted to the NRC on September 25, 1997. In your letter you inform me that NRC has reached agreement with me about removal of the double containment requirement of 10 CFR 71.63(b) and that accordingly NRC will "remove the double containment requirement of
§ 71.63(b)" in keeping with my request.
Your letter continues beyond this announcement to assert that my original petition also requested elimination of 10 CFR 71.63(a). This assertion is not correct.
My original petition letter is explicit in the expression of the petition. My petition states, "I request that 10 CFR 71 .63(b) be deleted in its entirety." There is no request for the elimination of 10 CFR 71.63(a). (For your convenience, I have enclosed a copy of my original letter dated September 25, 1997 which conveys my petition.)
- My comment regarding 10 CFR 71.63(a) at the close of my original petition letter is simply an acknowledgment that 10 CFR 71 .63(a) also relates only to a limited group of radionuclides as does 10 CFR 71.63(b), and therefore, NRC should consider the possible inference ("a corollary") that could be reached by an interested party regarding 10 CFR 71.63(a) if the requested elimination of 10 CFR 71.63(b) were to be accepted by NRC. This was not a request by me to eliminate 10 CFR 71.63(a).
I request that this clarifying letter be included in the record for the NRC rulemaking and decision for PRM-71-12.
Sincerely, dFrank P. Falci President Enclosure
r .
INTERNATIONAL ENERGY CONSULTANTS. INC.
8905 COPENHAVER DRIVE POTOMAC, MARYLAND 20854 (301) 340-1047 (301) 340-2229 FAX September 25, 1997 Secretary U.S. Nuclear Regulatory Commission Attn: Chief, Docketing and Service Branch Washington, DC 20555-0001 Re: Petition for Ru.lemaking Pursuant to Provisions of 10 CFR 2.802 Regarding the Special Requirements for Plutonium Shipments", 10 CPR 71.63_
Dear Mr. Secretary:
I have previously communicated with you about the NRC Rulemaking on Requirements for Shipping Packages Used to Transport Vitrified High Level Waste, Docket Number 71-11, by my letter to you dated July 22, 1997. Receipt of my letter was acknowledged by your office in a notice postmarked July 25, 1997.
One of my earlier comments was:
"The Notice of Proposed Rulemaking referred to a NRC Staff Requirements Memorandum (SRM) dated October 31, 1996. This SRM directed the staff to initiate and expedite the Notice of Proposed Rulemaking and this has been accomplished. The SRM also stated, 'In the longer term, the staff should also address wheth('r the technical basis for 10 CFR 71.63 remain valid, or whether a revision or elimination of portions of 10 CFR 71.63 is needed to provide flexibility for current and future technologies.* This broader look at 10 CFR 71.63 is to be completed by September 26, 1997. I recommend that pursuit of the total elimination of 10 CFR 71.63 be undertaken forthwith."
In keeping with the intent of my earlier recommendation, I hereby petition the U. S. Nuclear Regulatory Commission (NRC) for a rulemaking under the provisions of 10 CFR 2.802. I request that 10 CFR 71.63 (b) be deleted in its entirety.
This action is needed to remove from Nuclear Regulatory Commission regulations, provisions which cannot be supported technically or logically. Based on the "Q-System for the Calculation of At and Ai Values", an Ai quantity of any radionuclide has the same potential for damaging the environment and the human species as an A,, quantity of any other radionuclide. Therefore, the requirement that a Type B package must be used whenever package content exceeds an A,, quantity should be applied consistently for any radionuclide. That is, if a Type B package is sufficient for a quantity of radionuclide X which exceeds ~. then a Type B package should be sufficient for a quantity of radionuclide Y which exceeds A,,. And this should be similarly so for every other radionuclide.
- I U.S. Nuclear Regulatory Commission September 2S, 1997 While the regulations do embrace this simple logical congruence for the most part, the congruence fails under 10 CFR 71.63 (b) wherein packages containing plutoQium must include a separate inner container for quantities of plutonium having a radioactivity exceeding 20 curies (with certain exceptions).
If the NRC allows ~is failure of congruence to persist, these regulations wilJ be vulnerable to the following challenges:
( 1) The logical foundation of the adequacy of A2 values as a proper measure of the potential for damaging the environment and the human species, as set forth under the Q-System, is compromised.
(2) The absence of a radioactivity limit for every radionuclide which, if exceeded, would require a separate inner container, is an inherently inconsistent safety practice.
(3) The performance requirements for Type B packages as called for by 10 CFR 71 establish containment conditions under different levels of package trauma. The satisfaction of these requirements should be a matter of proper design work by the package designer and proper evaluation of the design through regulatory review. The imposition of any specific package design feature such as that contained in 10 CFR 71.63 (b) is gratuitous. The regulations are not formulated as package design specifications, nor should they be.
In addition, the continuing presence of 10 CFR 71.63 (b) engenders excessively high costs in the transport of some radioactive materials with no clearly measurable net safety benefit This is so in part because the ultimate release limits allowed under 10 CFR 71 package performance requirements are identical with or without a "separate inner container", and because the presence of a "separate inner container promotes additional exposures to radiation through the additional handling required for the "separate inner container".
Further, excessively high costs occur in some transport campaigns. One instance of such damage to our national budget is in the transport of transuranic wastes. Large numbers of transuranic waste drums must*
be shipped in packages which have a "separate inner container' to comply with the existing rule. Large savings would accrue without this rule.
- The elimination of 10 CFR 71.63 (b) in its entirety would resolve these regulatory defects.
I recognize that the arguments presented here also reflect upon 10 CFR 71.63 (a) bee~ 10 CFR 71.63 (a) also relates only to a limited group of radionuclides. I request that NRC consider this matter as a corollary to the basic petition.
- Sincerely, President FPF/aw
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 29, 2003 Mr. Frank P. Falci, President International Energy Consultants, Inc.
8905 Copenhaver Drive Potomac, Maryland 20854
Dear Mr. Falci:
The purpose of this letter is to provide you with the Nuclear Regulatory Commission's (NRC's) decision on your petition for rulemaking, docketed as PRM-71-12, which you submitted to the NRC on September 25, 1997. The petition requested that the NRC delete 10 CFR 71.63 -- the special requirements for the transportation of plutonium.
The N RC has considered the petition and the related public comments and has decided to grant your request to remove the double containment requirement of§ 71.63(b). However, the requirement of§ 71.63(a), which specifies that shipments whose contents contain greater than 0.74 Tbq (20 Ci) of plutonium be made with the contents in solid form, is retained. Thus, this portion of your petition is denied. You can find a detailed description of the NRC decision on your petition in the Part 71 final rule, under Issue 17. This completes action on PRM-71-12.
If you have any questions on this matter, please contact NaiemTanious at 301-415-6103.
Sincerely, C?rY~~ V,,:~- f~r:0"-..-/
Annette L. Vietti-Cook Secretary of the Commission
Enclosure:
Federal Register Notice for Final Rule
r
...... ,., ,,,_,~ -
t DOCKETtU&P"
( '13 Fil g.3,~)
May 14, 2004 DOCKETED USNRC June 15, 2004 (11 :31AM)
Mr. Frank P. Falci, President International Energy Consultants, Inc. OFFICE OF SECRETARY 8905 Copenhaver Drive RULEMAKINGS AND Potomac, MD 20854 ADJUDICATIONS STAFF
Dear Mr. Falci:
I am responding on behalf of the U.S. Nuclear Regulatory Commission (NRC), to your February 6, 2004, letter in which you requested that NRC include it in the record for the decision for PRM-71-12.
Your letter has been docketed on March 15, 2004, by the Office of the Secretary, Rulemaking and Adjudications Staff, and is now posted on the NRC ADAMS system, Accession Number ML040770244. We have also posted your letter, as a comment letter, on the NRC rulemaking website for the Part 71 final rule. This action satisfies your request to include your letter in the record for the decision for PRM-71-12.
Sincerely, orig. signed by AP for Scott W. Moore, Chief Rulemaking and Guidance Branch Division of Industrial and Medical Nuclear Safety Office of Nuclear Material Safety and Safeguards Distribution:
IMNS R/F CPoland C:\MyFiles\Checkout\Part71-Ltr-to-petitioner-4-29-04.WPD (Packa e Number: ML041340147) *See Previous concurrence OFFICE: RGB:IMNS RGB:IMNS TSSl:SFPO NAME: NTanious* APersinko* RLewis*
DATE: 05 /07 / 04 05 /10 /04 05/11 /04 5 11'-f /04 OFFICIAL RECORD COPY
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 14, 2004 Mr. Frank P. Falci, President International Energy Consultants, Inc.
8905 Copenhaver Drive Potomac, MD 20854
Dear Mr. Falci:
I am responding to your February 6, 2004, letter in which you clarified your intentions with respect to your September 25, 1997, petition for rulemaking on 10 CFR 71.63. You also requested that NRC include your February 6, 2004, clarification letter in the record for the decision for PRM-71-12.
Your letter has been docketed on March 15, 2004, by the Office of the Secretary, Rulemaking and Adjudications Staff, and is now posted on the NRC ADAMS system, Accession Number ML040770244. We have also posted your letter, as a comment letter, on the NRC rulemaking website for the Part 71 final rule. Thank you for bringing this clarification to our attention.
Sincerely, AAJW-<JWJ~ £r Scott W. Moore, Chief Rulemaking and Guidance Branch Division of Industrial and Medical Nuclear Safety Office of Nuclear Material Safety and Safeguards
NRC FORM 665P U.S. NUCLEAR REGULATORY COMMISSION
- (4*2002)
PACKAGE ADAMS DOCUMENT SUBMISSION (Multiple Documents)
Originated By Naiem Tanious I
If you have more than two documents in vour packaqe, copy this paqe as needed.
Telephone 301-415-6103 IMail Stop T9C24 ILAN ID NST I Date 05/13/04 List below the Document Titles or Accession Numbers in the exact order they should be in the ADAMS Package 1 Document No.
Document Title or Accession No. ML040770244 - Letter from the petitioner (Frank Falci)
~ Is this a brief title that can be changed by DPC according to template instructions?
Is this an exact title formatted according to template instructions that should not be changed by DPC?
Document AVAILABILITY (select one)
[fil Publicly Available B Document SENSITIVITY (select one)
Sensitive ~ Non-Sensitive
§ (Indicate release date)
Immediate Release Normal Release § Sensitive-Copyright Document SECURITY ACCESS LEVEL (select one)
Document Processing Center Non-Sensitive Copyright
=Owner Delay Release Until NRG Users =Viewer Limited Document Security (Defined by User)
=Viewer Date =Viewer Non-Publicly Available =Viewer Package Accession No. ADAMS Template No. RIDS Code (if applicable) Other Identifiers ML041340147 NMSS-010
? Document No.
Document Title or Accession No. ML041330189 - Letter to petitioner (Frank Falci)
- ~ Is this a brief title that can be changed by DPC according to template instructions?
Is this an exact title formatted according to template instructions that should not be changed by DPC?
Document AVAILABILITY (select one)
[fil Publicly Available (Indicate release date)
BDocument SENSITIVITY (select one)
Sensitive Sensitive-Copyright
~ Non-Sensitive Non-Sensitive Copyright
~ Immediate Release Document SECURITY ACCESS LEVEL (select one)
Normal Release ~ Document Processing Center = Owner Delay Release Until NRG Users =Viewer Limited Document Security (Defined by User)
= Viewer Date = Viewer 0 Non-Publicly Available =Viewer ADAMS Template No. RIDS Code (if applicable) Other Identifiers NMSS-010 Submitted By Roberta Gordon I Telephone 301-415-7555 Mail Stop T9C24 ILAN ID REG I Date submitted to DPC 05/17/04 Accession No. ML020170279
'EC ., International Energy Consultants, Inc .
8905 Copenhaver Drive OOCKiif - ; Potomac, Maryland 20854-PEAD FIULI fl! 7/-1 ~ Tel: {301) 340-1047 Fax: (301) 340-2229
( lo:; FR. 3 3 t,J.,)
DOCKETED February 6, 2004 USNRC Ms. Annette L. Vietti-Cook March 15, 2004 (11 :31 AM)
Secretary of the Commission United States Nuclear Regulatory Commission OFFICE OF THE SECRETARY Washington, D.C. 20555-0001 RULEMAKINGS AND ADJUDICATIONS STAFF
Dear Ms. Vietti-Cook:
Thank you for your letter dated -December 29, 2003 in which you announce the Nuclear Regulatory Commission's (NRC's) decision on my petition for rulemaking, docketed as PRM-71-12, which I submitted to the NRC on September 25, 1997. In your letter you inform me that NRC has reached agFe~ent W!th me about removal of the double containment requirement of 10 CFR 71.63(b) and that accordingly NRC will "remove the double containment requirement of
§ 71.63(b)" in keeping with my request.
Your letter continues beyond this announcement to assert that my original petition also requested elimination of 10 CFR 71.63(a). This assertion is not correct.
My original petition letter is explicit in the expression of the petition. My petition states, "I request that 10 CFR 71.63(b) be deleted in its entirety." There is no request for the elimination of 10 CFR 71.63(a). (For your convenience, I have enclosed a copy of my original letter dated September 25, 1997 which conveys my petition.)
- My comment regarding 10 CFR 71.63(a) at the close of my original petition letter is simply an acknowledgment that 10 CFR 71.63(a) also relates only to a limited group of radionuclides as does 10 CFR 71.63(b), and therefore, NRG should consider the pcissible inference ("a corollary") that could be reached by an interestod party regarding 10 CFR 71.63(a) if the requested elimination of 10 CFR 71.63(b) were to be accepted by NRG. This was not a request by me to eliminate 10 CFR 71.63(a).
I request that this clarifying letter be included in the record for the NRC rulemaking and decision for PRM-71-12.
Sincerely, dFrank P. Falci President Enclosure
INTERNATIONAL ENERGY *CONSULT.AN'l:'S. INC.
8905 COPENHAVER DRIVE POTOMAC, MARYLAND 20854 (301) 340-1047 (301) 340-2229 FAX September 25, 1997 Secretary U.S. Nuclear Regulatory Commission Attn: Chief, Docketing and Service Branch Washington, DC 20555-0001 Re: Petition for Rulemaking Pursuant to Provisions of 10 CFR 2.802 Regarding the ..Special Requirements for Plutonium Shipments", 10 CFR 71.63_
Dear Mr. Secretary:
I have previously communicated with you about the NRC Rulemaking on Requirements for Shipping Packages Used to Transport Vitrified High Leve] Waste, Docket Number 71-11, by my Jetter to you dated July 22, 1997. Receipt ofmy letter was acknowledged by your office in a notice postmarked July 25,1997.
One of my earlier comments was:
"The Notice of Proposed Rulemaking referred to a NRC Staff Requirements Memorandum (SRM) dated October 31, 1996. This SRM directed the staff to initiate and expedite the Notice of Proposed Rulemaking and this has been accomplished. The SRM also stated, 'In the tonger tenn, the staff should also address whetlu~r the technical basis for 10 CFR 71.63 remain vaJid, or whether a revision or elimination of portions of 10 CFR 71.63 is needed to provide flexibility for current and future technologies.' This broader look at 10 CFR 71.63 is to be completed by September 26, 1997. I recommend that pursuit of the total elimination of 10 CFR 71.63 be undertaken forthwith."
In keeping with the i~tent of my earlier recommendation, I hereby petition the U. S. Nuclear Regulatory Commission (NRC) for a rulemaking under the provisions of l 0 CFR 2.802. I request that 10 CFR 71.63 (b) be deleted in its entirety.
This action is needed to remove from Nuclear Regulatory Commission regulations, provisions which cannot be supported technically or logically. Based on the Q-System for the Calculation of A1 and Az Values", an A2 quantity of any radionuclide has the same potentia! for damaging the environment and the human species as an A2 quantity of any other radionuclide. Therefore, the requirement that a Type B package must be used whenever package content exceeds an Ai. quantity should be applied consistently for any radionuclide. That is, if a Type B package is sufficient for a quantity of radionuclide X which exceeds A2, then a Type B package should be sufficient for a quantity of radionuclide Y which exceeds Az., And this should be similarly so for every other ra~ionuclide.
U.S. Nuclear Regulatory Commission . September 25, 1997 While the regulations do embrace this simple logical congruence for the most part, the congruence fails under IO CFR 71.63 (b) wherein packages containing plutoqium must include a separate inner container for quantities of plutonium having a radioactivity exceeding 20 curies (with certain exceptions).
If the NRC allows *is failure of congruence to persist, these regulations will be vulnerable to the following challenges: *
(I) The logical foundation of the adequacy of A2 values as a proper measure of the potential for damaging the environment and the human species, as set forth under the Q-System, is compromised.
(2) The absence of a radioactivity limit far every radionuclide which, if exceeded, would require a separate inner container, is an inherently inconsistent safety practice.
(3) The performance requirements for Type B packages as called for by 10 CFR 71 establish containment conditions under different levels of package trauma. The satisfaction of these requirements should be a matter of proper design work by the package designer and proper evaluation of the design through regulatory review. The imposition of any specific package design feature such as that contained in 10 CFR 71.63 (b) is gratuitous: The regulations are not formulated as package design specifications, nor should they be.
In addition, the continuing presence of IO CFR 71.63 (b) engenders excessively high costs in the transport of some radioactive materials with no clearly measurable net safety benefit This is so in part because the ultimate release limits allowed under 10 CPR 71 package perfonnance requirements are identical with or without a "separate inner container", and because the presence of a "separate inner container promotes additional exposures to radiation through the additional handling required for the 11separate inner container".
Further, excessively high costs occur in some transport campaigns. One instance of such damage to our national budget is in the transport of transuranic wastes. Large numbers of transuranic waste drums must.
be shipped in packages which have a "separate inner container to comply with the existing rule. Large savings would accrue without this rule. *
- The elimination of 10 CFR 71.63 (b) in its entirety would resolve these regulatory defects.
I recognize that the arguments presented here also reflect upon 10 CPR 71.63 (a) bec~use 10 CFR 71.63 (a) also relates only to a limited group *of radionuclides. I request that NRC consider this matter as a corollary to the basic petition. *
- President FPF/aw
r aiem Tanious - Re: Letter from Petitoner
.- ~- ~- **. " ~*-~ ......,--.-.~* . .r *-*-**-*~-~- . -*~~-*. - . ........- ....... ?\_..,, ____ ,.,~,.,,,. .~ _..-.,_. ..... --****n** ' .. -***-*~ *- r " "'""'"' .. ,, . .... ~- r **,*- **" ~., ..,. _, ............. -, ..,. _ _..,,....,..
From: E. Neil Jensen To: Naiem Tanious; Robert Lewis; Sandra Wastler Date: 3/18/04 4:35PM
Subject:
Re: Letter from Petitoner I think Rob's proposed response is exactly what's called for.
>>> Robert Lewis 03/18/04 04:28PM >>>
Naiem, Maybe we dont need a meeting -
I looked at the letter. I agree with your interpretation that the reasonable way to read his letter is that he DID request elimination of 71.63(a) [his last paragraph specifically asks for its elimination as a 'corrollary' of the basic PRM].
Falci only asks that his 2/6 clarifying letter be "included in the record" for the PRM. The docket stamp indicates this was done. Perhaps a short letter simply thanking him for the clarification an noting that we included his letter in the PRM 71-12 docket is all that is needed.
-Rob
>>> Naiem Tanious 03/18/04 03:S0PM >>>
Rob: O.k.
Neil and Sandi:
Would you be available for a short meeting next Tuesay or Wed?
>>> Robert Lewis 03/18/04 03:47PM >>>
I am free Friday, next tues and wed, then on travel for 2 weeks.
>>> Naiem Tanious 03/18/04 03:39PM >>>
We recevied a letter from Mr. frank Falci, the petitoner. He is unhappy with the way we worded the NRC decision.
I personnally don't agree with his objections, but I will defer to our collective judgment/decision, therefore, we need to meet to discuss our response.
I left copies with Rob Temps, and talked to Sandi and Neil about having a meeing. Neil said he can't meet this coming week. Please indicate your availbility. I will bring copies of previous material/memos on this petition for our discussions. thanks NaiemT
I LJ liKET PETIT~ R MBER 11-1~
( ftJ3Ff<. g 3&,~)
NUCLEAR REGULATORY COMMISSION "98 JUN 18 AlO :56 10 CFR Part 71
[Docket No. PRM-71-12)
QFF-1', *- , )F *0C,"'rJ K
- r*, 1Y RULr.:rv .: *1~ " ,.. i:J AOJUDIC1 :1 1- ,, ; ._, l AFF Petition from International Energy Consultants, Inc.; ..,
Extension of Comment Period AGENCY: Nuclear Regulatory Commission .
- ACTION: Petition for rulemaking: Extension of comment period.
SUMMARY
- On February 19, 1998 (63 FR 8362), the Nuclear Regulatory Commission (NRG) published for public comment a petition for rulemaking filed by the International Energy Consultants, Inc. The comment period was to have expired on May 5, 1998. General Atomics submitted a comment on May 26, 1998, and requested that the comment period be extended so that their comment, and comments by other industry people, be considered. In view of this request, the NRG believes it is appropriate to extend the comment period, which now expires on July 31, 1998.
DATES: The comment period has been extended and now expires July 31 , 1998. Comments received after this date will be considered if it is practical to do so but the Commission is able to ensure consideration only for comments received on or before this date.
ADDRESSES: Send comments by mail addressed to the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 . Attention: Rulemakings and Adjudications Staff.
f,.J,. ~ to/:J'I/C/8' of fd3F~3'/335
, If t
I Hand-deliver comments to: 11555 Rockville Pike, Rockville, Maryland, between 7:30 am and 4:15 pm on Federal workdays.
You may also provide comments via the NRC's interactive rulemaking web site through the NRC home page (http://www.nrc.gov). From the NRC home page, select "Rulemaking" from the tool bar. The interactive rulemaking website can then be accessed by selecting "Rulemaking Forum." This site provides the availability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking site, contact Ms. Carol Gallagher, (301) 415-5905; e-mail CAG@nrc.gov.
Certain documents related to this rulemaking, including comments received and the environmental assessment and finding of no significant impact, may be examined at the NRC Public Document Room, 2120 L Street NW., (Lower Level), Washington, DC. These same documents also may be viewed and downloaded electronically via the interactive rulemaking website established by NRC for this rulemaking.
FOR FURTHER INFORMATION CONTACT: Mark Haisfield [telephone (301) 415-6196, e-mail MFH@nrc.gov] of the Office of Nuclear Material Safety and Safeguards, U.S. Nuclear
- Regulatory Commission, Washington, DC 20555-0001 .
1**/
i ') -,~-
Dated at Rockville, Maryland, this / / *~ day of June, 1998.
For the Nuclear Regulatory Commission.
John C. Hoyle, Secretary of the Commission .
2
DOCKET NUMBER
+ GENERAL ATOMICS PETITION RULE PRM 7/-/ ~
( fo3FI< 83b:J.)
DOCKETED File 6.11.31.5.1 us rnc May 26, 1998 "98 JUN -2 P4 :50 Secretary, U.S. Nuclear Regulatory Commission Rulemakings and Adjudications Staff Washington, DC 20555-0001
Subject:
Petition for Rulemaking Pursuant to Provisions of 10 CFR 2.802 Regarding the "Special Requirements for Plutonium Shipments", 10 CFR 71.63, (Docket No. PRM 12)
Dear Mr. Secretary:
General Atomics missed the May 5, 1998 comment period deadline for commenting on Federal Register Notice of February 19, 1998 (Vol. 63. No. 33. PP 8362-8363),
International Energy Consultants, Inc. (IEC): Receipt of Petition for Rulemaking [Docket No. PRM-71-12]. The petitioner requested that the special requirements for plutonium shipments contained in 10 CFR 71.63 be eliminated. We offer the following comments on the petition. We request that the comment period be extended so that the enclosed comment and comments by other industry people be considered.
General Atomics agrees that this action is needed to remove from Nuclear Regulatory Commission regulations, provisions that cannot be supported technically or logically.
The basis of the petition is valid. 10 CFR 71.63 is inconsistent with the use of the Q-system for the calculation of Al and A2 values since an A2 quantity of any radionuclide has the same potential for damaging the environment and the human species as an A2 quantity of any other radionuclide. Therefore, the requirement that a Type B package must be used whenever package content exceeds an A2 quantity should be applied consistently for any radionuclide, including plutonium. Removal of the special requirements for plutonium will also bring U.S. regulations closer to the International Atomic Energy Agency (IAEA) regulations. The IAEA accepts that the Q-System effectively provides radionuclide quantity limits for identifying the threshold for specific requirements in packaging design.
General Atomics also agrees that the elimination of the 10 CFR 71.63 requirements will reduce personnel exposure through reduced handling by not having to handle the separate inner container.
Sincerely, cc: L. Johnson K. Asmussen JUN - 4 1998 Acknowledged by card .....,, , ,, ,
3550 GENERAL ATOMICS COURT, SAN DIEGO, CA 92121-1194 P.O. BOX 85608, SAN DIEGO, CA 92186-9784 (619) 455-3000
U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS &ADJUDICATIONS STAFF OFRCE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Data 5 :). q CJ K Cq>les Fkloatled / -I-Add'! Copies Reproduced ~
IJIS1llDIJIUO!I' i'n~
1:>vff.Pt os
INTERNATIONAL ENERGY CONSULTANTS, INC.
8905 COPENHAVER DRIVE DOCK ETED POTOMAC, MARYLAND 20854 USNRC (301) 340-1047 (301) 340-2229 FAX *95 MAY 26 P 1 :25 OF-rr,,,.,..11_,.* ,,. . .
RU1-Ei\l1 ADJUDIC, . AFF May 20, 1998 DOCKET NUMBER PETITION RULE PRM 7 /,. /p
( ~3 Fl<. 83G:-~)
Secretary U.S. Nuclear Regulatory Commission Attn: Rulemaking and Adjudications Staff Washington, DC 20555-0001 Re: Docket No. PRM-71-12
Dear Mr. Secretary:
In response to the Federal Register Notice, February 19, 1998 (Vol. 63. No. 33, pp 8362-8363), International Energy Consultants, Inc. (/EC): Receipt of Petition for Rulemaking [Docket No. PRM 71-12], I have enclosed herein a copy of a letter dated April 13, 1998, which was sent to me by Mr. Lake H. Barrett, Acting Director, Office of Civilian Radioactive Waste Management (RW), U.S. Department of Energy .
- In his letter to me, Mr. Barrett expressed support for this petition on behalf of RW. I request that Mr. Barrett's letter to me be accepted by NRC as further comments, from an expert organization in the field to which the petition relates, in support of this petition.
Sincerely, Frank P. Falci President FPF/aw Enclosure
U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION I/
Postmark Oat" 5 ;; '18____
Copies Receiver'_ _ _/ ~ - - -
Add'I Copies RC" sn ~ ~~X l'4S-
Department of Energj Washington, DC 20585 April 13, 1998 DOCKET NU!\,JBER Mr. Frank P. Falci President PETITION hvLE ?Fii,111-/~
International Energy Consultants, Inc. ( (;3Flf ~ g*~ :i) 8905 Copenhaver Drive .*
Potomac, Matyland 20854
Dear Mr. Falci:
The Office of Civilian Radioactive Waste Manag~ent (OCRWM) bas reviewed your letter dated March 9, 1998, regarding the ~tion for rulemaldng which you tiled with the Nuclear Regulatory Commission (NRC) on September 25, i 997 (PRM-71-12). We agree that the content ofNRC regulations pertaining to special requirements fur piutoniuin shipments could have a
- significant cost impact on the Department ofEnergy (DOE) without a clear safety benefit We also agree that the rule pertaining to double containment appears to be inconsistent and excessive when compared to containment requirements for~~ radionucli~.
OCRWM anticipates a final rule from the NRC that will exempt vitrified high-level waste from the double containment requirements of 10 CFR Part 71.63. The NRC's ntlemaking was prompted by DOE' s 1993 petition for rulemaking (PRM-71-11 ), which argued that vitrified high-level waste is botmded by spent fuel elements, which are already exempt In OCR.WM's July 17, 1997, letter, which responded to NRC's May 1997.~otice ofProposcd Rulemaking (PRM 11 ), Requirements for Shipping Packagu Used to D-ansport Vitrified High-Lever Waste, we recommended a less prescriptive approach than was *contained in the pl'()JJOScdrule for the longer .
term consideration of disposition of plutoniwn bearing materials that exist in a variety of forms.
- Your petition (PRM-71-12) is different in that it questions the need for double containment at all.
The manner in which NRC responds to your petition for-rulemaking ( PRM-71-12) could affect several DO&managed prqgrams. OOE's Office of Environmental Management (EM) will prepare integrated comments to.the NRC's request for comment on yom petition. OCRWM will coordinate with EM in preparing such comments.
Should you have any questions or require additional information, please feel free to contact Mr. William Lake of my staff at (202) 586-2840.
Sincerely,
. 'i~a-~~ Lake H. Barrett, Actfu;abirector f- Office of Civilian Radioactive
- Waste Management l'Mlldwllleci,Hlan-,cllld,-s-
Department of Energ}'
Washington, DC 20585 April 13, 1998 Mr. FrankP. Falci DOCKET NUMBER President PETITION fiL ** P, 11 1 /-/Ji
( &3FI<~ 3~ :i)
International Energy Consultants, Inc.
- 8905 Copenhaver Drive .*
Potomac, Maryland 20854
. 0)
Dear Mr. Falci:
The Office of Civilian Radioactive Waste Management (OCRWM) has reviewed your letter dated March 9, 1998, regarding the petition for rulemaking which you filed with the Nuclear Regulatory Commission (NR.C) on September 25, 1997 (PRM-71-12). We agree that the content ofNRC regulations pertaining to special requirements for piutoniuin shipments could have a significant cost impact on the Department of Energy (DOE) without a clear safety benefit We also agree that the rule pertaining to double containment appears to be inconsistent and excessive when compared to containment requirements for o~~r radionucli~s.
OCRWM anticipates a final rule from the NRC that will exempt vitrified high-level waste from the double containment requirements of 10 CFR Part 71.63. The NRC's rulemalcjng was prompted by DOE's 1993 petition for rulemaldng (PRM-71-11 ), which argued that vitrified high-level waste is bounded by spent fuel elements, which are already exempt. In OCRWM's July 17, 1997, letter, which responded to NRC's May 1997."Notice ofProposedRulemmng (PRM 11 ), Requirements for Shipping Packages Used to Transport Vitrified High-Lever Waste, we recommended a less prescriptive approach than was *contained in the proposed rule for the longer .
term consideration of disposition of plutonium bearing materials that exist in a variety of forms.
- Your petition (PRM-71-12) is different in tliat it questions the need for double containment at all.
The manner in which NRC responds to your petition for-rulemalcjng ( PRM-71-12) could affect several DOE-managed prQgrams. DOE's Office of Environmental Management (EM) will prepare integrated comments to.the NRC's request for comment on your petition. OCRWM will coordinate with EM in preparing such comments.
Should you have any questions or require additional information, please feel free to contact Mr. William Lake of my staff at (202) 586-2840.
Sincerely,
. r?~a-~~ -Lake H. Barrett, Actiii;£irector Office of Civilian Radioactive
- Waste Management Pmtad wllh fKTf Hl on nlCydlld paper MAY 2 8 1998 Acknowledge~ by card ...............................,...
U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS &ADJUDICATIONS STAFF OFFICE OFTtfE SECRETARY OF THE COMMISSION Document Statistics Postmark Date ---'-+>-o...;..+--'-=--- - -
Copies Received _ _ _ _ _ _ __
Add'I Copies Repro Speci istribution
- , 7(ID.s
DOCKET f lUM ER PETITION U E PRM 11::J.?
Secrretary ( (;3 Fr< 8 31P:l) oouct~fRTJD U.S. Nuclear Regulatory Commission Attn. Chief, Docketing and Service Branch Washington, DC
'98 MAY 18 A11 :53 Re: Petition for Rulemaking regarding "Special Requirements for Plutonium Shipments, IOCFR 71.63 OFFl ~ - *- _I 1**. nr
\.. ' - 1 I. *i I ;*f D M s AUL,_ ~ ' . , ' \
ear r. ecretary: ADJUDi6i', ;; ;,!\ ,AFF The petition by Mr. Frank Falci on behalf of the International Energy Consultants, Inc., appears to me to be without merit. He argues that the provision for a separate inner container for shipment of large quantities of plutonium is unacceptable due to an alleged failure of "logical congruence" with the A2 quantity concept, and considerations of worker exposure and cost. In fact there has always been a special treatment of the category of transuranic radionuclides in the regulatory process. There are good reasons for this in light of the high radiological toxicity oftransuranics, the potential for the generation ofradiolytic gases, the formation of readily suspended particulates by alpha radiation interactions with package contents, etc. Moreover, the public has demanded extra precautions be taken in the transport of transuranic wastes on public highways. There is no good reason to believe that the special treatment accorded to transuranic waste will somehow precipitate a regulatory avalanche of requirements for special containers for every radionuclide shipped in large quantities; this does not follow simply as a matter of 'logical congruence as he asserts. The claim that the requirement for a separate inner container for transuranics will somehow cause added radiation exposure of workers through supposed additional handling requirement ,
apparently ignores the design characteristics of the TRUPACT shipping package, since the so-called additional handling would amount to only handling a second inner closure lid, and not somehow handling the waste itself twice. In addition, the category ofTRU waste does not include wastes having high external beta/gamma radiation fields. Finally, the cost issue, while real, is not compelling, since the added cost of the double container is only a very small fraction of the overall costs of shipment of transuranic wastes.
My recommendation to the Commission would be that this petition be denied.
Sincerely, John C. Rodgers , CHP 2795 Via Caballero del Sur Santa Fe, NM 87505
- May 15, 1998 MAY 2 l 1998 Acknowledged by card...... ** .. .* ,.. ..,...
U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS & ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Dcr:lli .e:-t Statistics Postman, oat _ ..S !fl h 8 - ~ /um, ~ ~
Copies RecGiv._ _ ____,,..,.;/'--'.__...,.._-
Add'I Cooies RAprocluccd _...:-L.,r -= -- -
s~
..::EJ>I! ~ ~
p~/4-
May 15, 1998 NOTE TO: Emile Julian Chief, Docketing and Services Branch FROM: Carol Gallagher ADM,DAS
/) _
{jOA1
~ ~
SUBJECT:
DOCKETING OF COMMENT ON PRM-71-12 Attached for docketing is a comment related to the Petition for Rulemaking on Special Requirements for Plutonium Shipments (PRM-71-12). This comment was received via the rulemaking website on May 15, 1998. The submitter's name is John Rodgers, CHP, 2795 Via Caballero del Sur, Santa Fe, NM 87505. Please send a copy of the docketed comment to Mark Haisfield (mail stop T9-F-31) for his records.
Attachment:
As stated cc w/o attachment:
M. Haisfield
DOCKETED USNRC
-3§- TRANSNUCLEAR, INC.
'98 MAY - 7 All :11 DOCKET NUMBER May 5, 1998 PETI I N UL PRM '11-- IJ.,.
(p3 Fl<ff 3t.:J..) (j)
Secretary US Nuclear Regulatory Commission Attn: Chief, Docketing and Service Branch Washington, DC 20555
- Re: Petition for Rulemaking Pursuant to Provisions of 10 CFR 2.802 Regarding the "Special Requirements for Plutonium Shipments" in 10 CFR 71.63. [Docket No. PRM-71-12]
Dear Mr. Secretary:
This letter is to formally state that Transnuclear, Inc. supports the petition for rulemaking filed by International Energy Consultants, Inc.
Transnuclear, Inc. is a designer of packagings for the transport of various types of radioactive materials and is also a transporter of such packages on a domestic and international basis.
Transnuclear believes that, given the validity of the "Q-System for the calculation of A 1 and A 2 Values" and the definition of Type B packages, there should not be a special distinction or additional requirements placed on the packaging of one isotope over another if the potential for damaging the environment and the human species is the same. Therefore, we support the petitioner's request that the NRC revise its regulations to eliminate paragraph 71.63(b).
Sincerely,
\ ran: nu: l,_,e,a....r...___..,,,,.,
a : :angusi General Manager - Operations MAY - 7 1998 Acknowledged by card ......................,,u,UJ,v.m FOUR SKYLINE DRIVE* HAWTHORNE, NEW YORK 10532-2176 TELEPHONE: 914-347-2345
- FAX: 914-347-2346
U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS &ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Doc* 1 e* , ,..+~ i .cs Postmark Data S 15j_"'--:=- g, _ __
Copies Reoollroo _____,_/-.--_ __
Add'I Copl99 Raproduc rj -,,--.....__ __
s~ ~~~i>1 ns:
MAY. 5 . 1998 2:55AM WESTERN GOVERNORS' N0.396 P.2 DOCKET NUMBER PETITION RULE PRM 71 ~ I;l DOCKETED US RC
( ~3F~ 83"~) @
'98 t1A'1' -6 p 3 :15 May 4, 1998 Secretary U.S. Nuclear Regulatory Commission WESTERN Attn: Rulemakings and Adjudications Staff Washington, DC 20555 GOVERNORS' RE; Petition for Rulemaking Pursuant to Provisions of 10 CFR ASSOCIATION 2.802 Regarding the "Special Requirements for Plutonium Shipments", 10 CFR 71.63, NRC Docket No. PRM-71-12
.~ Tony Knowles Jim Geringer Governor ofWyoming
Dear Mr. Secretary,
The Western Governors' Association and its Technical Advisory Group for WIPP Transport strongly oppose the petition by Mr. Frank Falci, International Energy Consultants, Inc., to weaken the Vice Chairman regulations governing the packaging and transportation of radioactive material contained in 10 CFR 71.63. The following are comments James M. Souby Exccucive Director from the Advisory Group:
- 1. Having worked for more than a decade to implement a safe system for transporting plutonium-contaminated transuranic (TRU) waste to the Waste Isolation Pilot Plant, we find the proposed weakening of packaging standards particularly distressing and damaging towards efforts to safely ship TRU waste to the WIPP facility. More than a decade ago, western states objected to the U.S. Department of Energy's proposal to ship waste to WIPP in uncertified single-wall, vented packages.
As a result, DOE redesigned the package. The new design not only met the double containment requirements the petitioner wants NRC to abandon, but one which also increased payload Headquamrs: capacity and lowered operating costs. We find the petitioner's 600 17th Street claim of excessive costs in <:9mplying with the existing Suite 1705 South Tower Denver, Colorado 80202-5452 regulation that mandates the double containment unsupportable, as evidenced by the experience with the (303) 623-9378 development of the Transuranic Packaging Transporter Fax (303) 534-7309 (TRUPACT-11).
Wawnpn, D.C. Office-:
400 N. C-apitol Street, N.W.
Suite 388 Washington, D.C. 20001 (202) 624-5402 Fax (202) 624-7707 MAY - 7 1998 www.wes1gov.org Acknowledged by card ........................., ..-
U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS&ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics PostmarkDate 6/lt?/1? ,_~~~ ~ ~
Copies Received I Add'I Copies Reproduced _ __,___ _
S~ llS
MAY. 5.1998 2:56AM WESTERN GOVERNORS' N0.396 P.3 U.S. NRC Letter Page 2 May 4, 1998
- 2. The Advisory Group has developed a comprehensive set of procedures and protocols contained in the 'Western Governors' Association Waste Isolation Pilot Plan Transportation Safety Program Implementation Guide" that are based on the regulations in 10 CFR 71.63 to assure the public and local emergency responders that all safety measures are being taken to make this a safe shipping campaign. Along with state representative educating the public, road shows carrying the TRU'PACT-11 containers are traveling across the West and Nation showing the public what the inside of the containers look like and the safety efforts being put forth to ensure protection of public health and the environment.
To remove the double containment standard would only cause public concern that is unwarranted and would be detrimental to the entire transportation system for WIPP shipments, planned to begin in June, 1998.
- 3. We find that the petitioner's proposal to allow the significant quantity of shipments containing plutonium liquids a threat to the public health and safety.
Based on the recent events related to the leaking of liquids from a shipment of DOE low level radioactive waste from Fernald, Ohio to the Nevada Test Site, the perception of lessening the Federal Government's standard would further erode public confidence in the safety of any DOE radioactive materials shipment.
Unless the petitioner can prove that there will be no significant impacts to public health and safety or reduction in transport efficiency or cost as a result of the proposed relaxation of NRC plutonium packaging requirements, the petition should be denied.
For these reasons, we urge the NRC to act expeditiously to deny the petition by International Energy Consultants, Inc.
James Sou Executive Director Western Governors' Association cc: WGA Technical Advisory Group on WIPP Transport
DOCKET NUMBER PETITION RULE PRM 7 / ~ / :Z..
( t, 3 F,q, 8 3 '-':2 Department of Energy DOCKETED Germantown, MD 20874-1290 SNRC MAY O5 1998 Secretary °98 MAY -5 P4 :OS U.S. Nuclear Regulatory Commission Attn. Rulemakings and Adjudications Staff OFFlc- sp-**
Washington . DC 20555-0001 RULE. '"I ADJU I
Dear Mr . Secretary:
This is in response to the Federal Register Notice, February 19, 1998 (Vol.
- 63. No . 33. pp 8362-8363). International Energy Consultants, Inc. (!EC);
Receipt of Petition for Rulemaking [Docket No. PRM-71-12]. The petitioner.
IEC. requests that the special requirements for plutonium shipments contained in Title 10. Code of Federal Regulations. Part 71. paragraph 63 (10CFR71.63) be eliminated. The primary basis for IEC's petition is the regulations in 10 CFR 71 are internally inconsistent regarding the adequacy of a Type B package.
Based on the "Q-System for the Calculation of A1 and A2 values." which is embraced by NRC regulations. a Type B package is sufficient for all radionuclides whose quantity exceeds A2 . The additional regulatory requirement of a "separate inner container" for packages containing plutonium is not congruent with the requirements for all other radionuclides. This compromises the logical foundation of the adequacy of A2 values set forth by the Q-System for all radionuclides. The petitioner also notes that similar arguments can be made for the prohibition of liquid forms of plutonium . The Department agrees with the petitioner's arguments. and believes that the Commission should change the special requirements for plutonium shipments accordingly.
Since the advent of the Q-System in the RAM transport regulations in the mid 70's, there has been no remaining justification for the double containment requirement for plutonium . The Q-System provides a consistent method for setting the quantity of a radionuclide that is considered to provide the
- potential for exceeding a dose limit by the most limiting dose pathway. This was not always the case. In the previous hazard categorization scheme nuclides were considered in classes, but not every nuclide in a class was of an equal hazard. This situation lead to special treatment of some nuclides/physical forms that were considered particularly hazardous. Thus .
the requirement that packagings used for Pu shipments to be doubly contained (resulting from concern about a specific material) is not justified. NRC has adopted the Q-System methodology as part of an effort to harmonize with the international regulatory system. This special treatment (10CFR71.63) is internally inconsistent with the changes that NRC has already made.
The Department believes that if the special requirements are eliminated personnel exposures from routine handing will decrease through reduced process time and that costs will be reduced substantially through more efficient handling and packaging. We also note that the elimination of the prescri ptive special requirements is consistent with the trend towards "performance based" regulations and will help move our domestic regulations closer to those accepted by the international community.
MAY - 7 1998 Acknowledged by card .............................._
@ Printed with soy ink on recycled paper
U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS & ADJUDICATIONS STAFF OFFICE FTHE SECRETARY OFT, .E COMMISSION Postmark Date .
Copies Rec6iv 1_
..5 Ls /qg
])
Add'I Copies Repro J c '"' _ 4-S~ J;;;_,~ 7? 1[).s
2 International harmonization of regulations is another benefit of the proposed change. The International Atomic Energy Agency (IAEA) and national regulations of member countries do not generally make nuclide specific regulations such as the requirement being considered in this petition. They accept the fact that the Q-System does effectively provide quantity limits that delineate the threshold for specific additional requirements in packaging capability to withstand the rigors of transport. Removal of this element-specific requirement will further the harmonization of the US regulations with IAEA and many nations of the world and could be a factor in enhancing US participation in the global nuclear energy industry.
Significant amounts of plutonium will be transported from Department of Energy facilities to storage and disposal sites under the proposed strategy of Accelerating Cleanup : Paths to Closure. As a result. the Department is encouraged by the prospect of the Commission's reassessment and elimination of the special requirements for plutonium shipments . The Department urges the Commission to evaluate the need for continuance of these special requirements in light of the basic containment regulations of 10CFR71.51. current practices for compliance with the rules for containment. and expected quantities and types of plutonium bearing materials to be shipped.
The opportunity to provide comment on the petition for rulemaking to modify 10CFR71.63 is appreciated. If you would like more information I can be reached on (301) 903-1969.
. Ke l kenb f9~--
Act i g Director Office of Transportation and Emergency Management Office of Site Operations Office of Environmental Management
cc:
G. Ives. DP-20 L. Lee. DP-22
- 0. Pearson. EH-3 D. Berkovitz. EM-20 M. Frei. EM-30 J. Fiore. EM-40 G. Boyd, EM-50 D. Huizenga, EM-60 E. Schmitt. EM-70 C. Guidice. EM-70 M. Wangler. EM-70 J. Thompson. EM-75 P. Karcz. EM-75 A. Kapoor. EM-76 M. Keane. EM-76 R. Lange, NE-40 S. Franks. NE-50
- 0. Lowe. NE-70 E. McCallum. NN-51 J. Brodman. PO-71 W. Lake. RW-40 R. Milner. RW-40 J. Arthur. AL M. Wi 11 i ams . AL G. Binns. AL R. Senna. AL F. Holmes. ID
ENVIRONMENTAL EVALUATION GROUP DOCKETED US RC
- - - - - - - - - - - - - - - - - - - - - A N EQUAL OPPORTUNITY/ AFRRMATIVEACTIONEMPLOYER --
7007 WYOMING BOULEVARD, N.E.
SUITE F-2 "98 APR 30 Pl2 :27 ALBUQUERQUE, NEW MEXICO 87109 (505) 828-1003 FAX (505) 828-1062 tLJ AFF April 29, 1998 OCr<ET NUMBER Secretary TITIO RULE PAM --:-----'lb.*
U.S . Nuclear Regulatory Commission ( C,3FI< 8'3'-~)
Attn: Chief, Docketing and Service Branch Washington, DC 20555-0001 Re: Petition for Rulemaking Pursuant to Provisions of 10 CFR 2.802 Regarding the "Special Requirements for Plutonium Shipments", 10 CFR 71.63, NRC docket No . PRM-71-12.
Dear Mr. Secretary:
The following comments are provided by EEG on the subject petition by Mr. Frank Falci, former DOE employee who is now President, International Energy Consultants, Inc., to relax the requirements for transportation of plutonium by deleting 10 CFR 71 .63 in its entirety.
Enclosed are two EEG reports :
EEG-24 Potential Problems from Shipment of High-Curie Content Contact-Handled Transuranic (CH-TRU) Waste to WIPP, August 1983, by Robert H. Neill and James K. Channell.
EEG-33 Adequacy of TR UPACT-I Design for Transporting Contact-Handled Transuranic Wastes to WIPP, June 1986, by James K. Channell, John C. Rodgers and Robert H . Neill.
Robert H. Neill Director RHN:js Enclosures(2) cc: w/enclosures: C.R. Chappell, SFPO/NMSS APR 3 0 1998 Aclt11o*ul dged by card ...........,....n .....,,,._.,
Providing an independent technical analysis of the Waste Isolation Pilot Plant (WIPP),
a federal transuranic nuclear waste repository.
U.S. NUCLEAR REGULATORY COMMISSION RULEMAKINGS &ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION
EEG-24 POTENTIAL PROBLEMS FROM SHIPMENT OF HIGH-CURIE CONTENT CONTACT-HANDLED TRANSURANIC (CH-TRU)
WASTE TO WIPP Robert H. Neill and James K. Channell Environmental Evaluation Group New Mexico August 1983
Emr.innnental. Evaluaticn GraJp Reports EEX;-1 Goad, Ik>nna, A Conpilation of Site Selection Criteria Considerations am Concerns Appearing in the Literature on the Deep Disposal of Radioactive Wastes, June 1979 EEX;-2 Review Connnents on Geolcx:dcal <l'laracterization Report. Waste Isolation Pilot Plant (WIPP) Site. Southeastern New Mexico SAND 78-1596. Volumes I am II, December 1978.
EEX;-3 Neill, Robert H., et al, (eds.) Radiological Health Review of the Draft Envirornnental Impact statement (OOE/EIS-0026-D)
Waste Isolation Pilot Plant. U. S. Deparbnent of Enei;gy.
August 1979.
EEX;-4 Little, Marshall s. , Review Comments on the Report of the Steering Committee on Waste Acceptance Criteria for the Waste Isolation Pilot Plant, Februai:y 1980.
EEX;-5 Cl'lannell, James K. , calculated Radiation Ibses From Deposition of Material Released in Hypothetical Transportation Accidents Involving WIPP-Related Radioactive Wastes, November 1980.
EEX;-6 Geotechnical Considerations for Radiological Hazard Assessment of WIPP. A Report of a Meeting Held on January 17-18. 1980, April 1980.
EEX;-7 Olatw:vedi, Lokesh, WIPP Site am Vicinity Geological Field Trip. A Report of a Field Trip to the Proposed Waste Isolation Pilot Plant Project in Southeastern New Mexico.
June 16 to 18. 1980, November 1980.
EEX;-8 EEX;-9 Wofsy, carla, '!he Significance of Certain Rustler Aquifer Parameters for Predicting Long-Term Radiation Ibses from WIPP, September 1980.
Spiegler, Peter, An Approach to calculating Upper Bounds on Maximum Individual Ibses From the Use of Contaminated Well Water Following a WIPP Repository Breach, September 1981.
EEX;-10 Radiological Health Review of the Final Envirornnental Impact Statement (OOE/EIS-0026) Waste Isolation Pilot Plant. U.S.
Department of Energy. January 1981.
EEX;-11 Cl'lannell, James K., calculated Radiation Ibses From Radionuclides Brought to the SUrface if Future Drilling Intercepts the WIPP Repository am Pressurized Brine, January 1982.
(Continued on Back Cover)
rorENTIAL PROBifflS FRCM SHilMENT OF HIGH-aJRIE CX>Nl'ENT CX>NTACI'-HANDIED TRANSURANIC (CH-TRU) WASTE 'ID WIPP Robert H. Neill James K. Olannell Environmental Evaluation Group 7007 Wyoming Boulevard NE, SUite F-2 Albuquerque, New Mexico 87109 August 1983
- Reprinted November 1990
'Ihis report was first published in August 1983 when the Environmental Evaluation Group was part of the New Mexico Health and Enviromnent Department, Environmental Improvement Division
- TABIE OF CONTENTS Page FOREWORD ********e****************e***o**********************************e i STAFF AND CONSUIIT'ANTS ii iii
SUMMARY
- ct**********C11******************************0*
CONCIIJSIONS AND ~ O N S *********1>eooooe ***eeeeeeo eoeoeeoo ***** V I. ~OOcrION *********** o ., * * * *
- o * * * * * * * * * * * * * * * *
- o * * * * * *
- o * * * * * * * * * *
- 1 II. rossIBIE PROBUM AREAS eo**********ei;i eeeso oeoQceoQGD********e****** 4 III HIGH-aJRIE rnNI'ENT CH-TRU CONTAINERS o 6 IVo MJFQJAC'l OF TRUPACT CONTAINERS FOR SHIIMENT **** o o ******** ~ ********** 16 GI.DS.S.AR..Y Q ** 0 0 ***** I) ************* 0 ****** 0 **** 0 0
- C
- Cl O
- Cl O O O
- 0 E) g
- Cl *** Q O ** "
- 19 REFERENCES oeeee ***e******* .. * * * * * * * * * * * *
- e e e o o e e o c i o o o e o o c o e e e e e o o e 1 1 e o o *
- e 20
- .APPfl\lOIX A ID G ., e e
- o * * * * * *
- e o * *
- e * *
- a
- e o *
- o
- o o m o e o
- o o o o *
- 10 o o o o *
- o o o *
- o o *
- 23
FOREWORD The purpose of the Environmental Evaluation Group (EEG) is to conduct an independent technical evaluation of the potential radiation exposure to people from the proposed Federal radioactive Waste Isolation Pilot Plant (WIPP) near Carlsbad, in order to protect the public health and safety and ensure that there is minimal environmental degradation. The EEG is part of the Environmental Improvement Division, a component of the New Mexico Health and Environment Department -- the agency charged with the primary responsibility for protecting the health of the citizens of New Mexico.
The Group is neither a proponent nor an opponent of WIPP.
Analyses are conducted of available data concerning the proposed site, the design of the repository, its planned operation, and its long-term stability.
These analyses include assessments of reports issued by the U.S. Department of Energy (DOE) and its contractors, other Federal agencies and organizations, as they relate to the potential health, safety and environmental impacts from WIPP.
The project is funded entirely by the U.S. Department of Energy through Contract DE-ACO4-79AL1O752 with the New Mexico Health and Environment Department *
- [\
t~fti'.Jdl Robert H. Nei 1 l
\
Di rector i
EEx; STAFF AND CONSIJilI'ANTS (1983)
James Kc Channell, Ph.D., P.E., Enviromnental Engineer I.okesh Cllaturvedi 8 Ph.D., Engineering Geologist Inz Elena Garcia, B.B.E., Administrative Secretary Marshall S. Little ( 1 ), M.S., Health Physicist Mary Ann Lynch, Secretary
- Jack M. Mobley, B.A. , Scientific Liaison Officer Robert H. Neill, M.S., Director Kenneth Rehfeldt, M. S. , Hydrologist Nonna I. Silva, Administrative Officer Peter Spiegler (1 ) ( 2 ), Ph.D., Radiological Health Analyst Rabbe Tucker, M.L.S., Librarian (1 ) certified, American Board of Health Physicis (2 ) Certified, American College of Radiology ii
SUMMARY
There are approximately 1000 drums of contact-handled transuranic (CH-TRU) wastes containing more than 100 Ci/drum of Pl utoni um-238 that are stored at the Savannah River Plant (SRP) and at the Los Alamos National Laboratory (LANL). As much as one-half of these high-curie content containers may contain waste in a combustible form. To date, no plans have been announced to process these wastes prior to shipment to the WIPP.
Studies performed at DOE laboratories have shown that large quantities of gases are generated in stored drums containing greater than 100 curies of Plutonium-238. Concentrations of hydrogen gas in the void space of the drums are often found to be high enough to be explosive. The calculated explosive energy for a truck shipment consisting of 36 drums with 6% hydrogen (lower explosive limit)
- would be equivalent to about 0.7 pounds of trinitroglycerol. None of the analy-ses in the DOE WIPP Final Environmental Impact Statement (-FEIS), Safety Analysis Report (SAR), and Preliminary Transportation Analysis (PTA) have considered the possibility that the generation of hydrogen gas by radiolysis may create an ex-plosive or flammable hazard that could increase the frequency and severity of accidental releases of radionuclides during transportation or handling.
These high Plutonium-238 concentration containers would also increase the esti-mated doses received by individuals and populations from transportation, WIPP site operations, and human intrusion scenarios even if the possibility of gas-enhanced releases is ignored. The WIPP Project Office has evaluated this effect on WIPP site operations and is suggesting a maximum limit of 140 Plutonium-239 equivalent curies (P-Ci) per drum so that postulated accidental off-site doses will not be larger than those listed in the FEIS. No actions have been sug-gested by DOE to maintain the transportation, occupational, and human intrusion doses that would result from accidents involving high-curie content drums to those listed in the FEIS.
I Additionally~ the TRUPACT container, which is being designed for the transporta-tion of Contact-Handled Transuranic (CH-TRU) wastes to WIPP, does not appear to meet the Nuclear Regulatory Commission (NRC) regulations requiring double iii
containment for the transportation of plutonium in quantities greater than 20 Ci. A 20 alpha Ci/shipment limit would require approximately 200,000 shipments for the 4 million curies of alpha emitters slated for WIPP.
The WIPP Project Office has not brought the gas generation problem or the adequacy of the TRUPACT shipping container to the attention of the Environmental Evaluation Group (EEG) either in reports, letters, or verbal communications
- iv
I. INTROIXJCrION A. Final Environmental Impact statement Analysis The April 1979 Draft Environmental Impact statement (DEIS) (Ref.1) for WIPP provided estimates of doses to people from the transportation of radioactive waste to WIPP for the expected radiation exposure for each of the waste gen-erating facilities scheduled to ship wastes to WIPP. When the Final Envirornnental Impact statenv:mt (FEIS) on WIPP (Ref. 2) was published in October 1980 the only estimates included were based on waste coming from Idaho National Engineering Iaborato:cy (INEL) and the Rocky Flats Plant (RFP) and all the calculated doses from other facilities were deleted. Transpor-tation accident calculations were included in both reports using typical RFP waste concentrations.
B. Post FEIS Analysis An October 1982 OOE report (Ref. 3) on radioactive wastes indicated that the radionuclide composition and concentrations of 'lHJ wastes va:cy considerably between the several generating facilities and are greatly different than the "typical" waste used in the October 1980 Final EIS and the various revisions to the Safety Analysis Report (SAR) (Ref. 4). Em used the 1982 data to cal-culate the average composition am concentrations for waste coming from each of the facilities including the savannah River Plant (SRP). 'Ihese values
- were transmitted on November 10, 1982 to OOE (Ref. 5) with the observation that the revised invento:cy could significantly change the calculated quanti-ties of nuclides released and doses incurred in IOOSt transportation, opera-tion, and human intrusion scenarios. '1he following statement was also made by Em in this letter:
"Em _is hereby requesting that OOE: (1) mcxlify the SAR values to be consistent with present data and plans ; and (2) begin to 1
keep us fully and a.irrently infontei about all changes and con-templated changes involving wastes that may be brought to WIPP.
Infonnation needed includes shipment schedules from each lab, total radionuclide concentrations, and percent composition of the nr,re .important radionuclides.
Current and detailed infonnation concerning waste fonns, plans for processing, and waste acceptance criteria certification and compliance procedures are all part of the same problem. We must r know these details in order to provide an adequate independent technical review of the project. 11 DOE responded to the November 10, 1982 letter by developing Pu-239 Equivalent TRU Activity Limits for the WIPP Containers and presented a pro-posed methodology and rationale for detennining the limit at a February 15, 1983 meeting with EEG. The limit chosen by DOE at that time was 169 P-Ci (Pu-239 Equivalent Curies) per 55-gallon drum on the basis that the dose to an off-site individual fran the limiting accident would be no greater than calculated in the FE IS.
releases from the Transportation accident scenario doses, long-term repository, and possible problems with radiolytic *gas genera ti on at higher container 1oadi ngs were not considered by DOE in the analysis. EEG has not received any updated or more detailed data on radio-nuclide content, waste fonns, etc.
The EEG agreed with the method and limit of the P-Ci activity for opera-tional accidents at the WIPP facility but made the following additional 11 comment on March 17, 1983 (Ref. 6): it also seems consistent to set equi-valent curie limits on specific groups of containers (e.ge on the hoist and in transportation shipments) and to consider all types of radiation exposure (operational, transportation, and long-term).11 Although, DOE has not specifically answered this March 18, 1983 letter, their April 13, 1983 Preliminary Transportation Analysis (PTA)(Ref. 7) addressed some of the concerns.
The Preliminary Transportation Analysis report used an average of the new inventory rather than facility specific inventories. This average was 303 P-Ci per TRUPACTs a value 10.7 times that used in the FEIS (42 average loaded drums with 0.68 P-Ci per drum)o The Preliminary Transportation Analysis report did not specifically discuss radiation doses to the maximum i ndi vi dual from transportation accidents but the doses vs. square meters curves in Appendices J and L indicate that severity category 6 or greater 2
accidents (whi°ch have an estimated probability of occurence during the re-pository lifetime of about 0.006) would lead to doses10-100 times those presented in the FEIS.
After an exchange of comments on the PTA (Ref. 8, 9, 10) between DOE and EEG the following issues remain on those topics addressed in the PTA:
(1) The PTA provided no nB-J data or analysis relative to the shipment of RH-TRU or Experimental HLW to WIPP since the October 1980 calculations in the FEIS, although the origin is known as well as the routes to be taken.
(2) The analysis does not use the latest inventory for each Plant and fails to address the fact that the average Savannah River Pl ant waste con-tainers contain 4 times the number of P-Ci assumed in the analysiso (3) The increase in maximum individual accident doses by an order-of-magnitude or more compared to the FEIS has not been fully addressed
- 3
II. POSSIBLE PROBLEM AREAS The above discussion mentions some of the outstanding concerns the EEG has about the overall transportation issue. There are also other concerns that have not been raised by DOE or EEG in our previous transportation evalua-tionsc Because a number*of issues are involved, EEG decided to address them in this single document rather than continue to attempt to deal with them by commenting on various reports. The main topics are listed below. and will be dealt with in more depth in subsequent chapters.
A. High-curie content of CH-TRU containers. There are two issues in this category:
(a) The possibility that radiolytic hydrogen gas ge~eration in high-curie content containers may present an explosive hazard in transporting and handling these containers. The WI PP Project Office has never informed us that there is a potential gas generation problem in any WIPP Project Office correspondence or reports. However, review of several DOE c_ontractor reports (Ref. 11, 12, 13, 14, 17) indicates that some investigators at DOE laboratories have recognized this potential problem
- for four years and suggests that the explosive nature of the gas in the wastes deserves a comprehensive evaluation; (b) The effect of higher curie quantities in drums on operation, trans-portation, and human intrusion dose calculations.
B. Updated RH-TRU and Experimental HLW Data and Analyses. No new data or evaluations have been provided to the State since the FEIS was issued in October 1980. DOE stated in the PTA (Ref. 7) that there is not suffi-cient new information to warrant a reevaluat 1ion at this time. EEG be-lieves a great deal of information currently exists (models, routes, waste generator source) and merits analysis now. This report will only address CH-TRU waste problems and will not address the remote-handled transuranic (RH-TRU) or experimental high-level waste (HLW) issues.
4
C. Adequacy of TRUPACT Containers for Shipment. There is some question whether the current plans for transporting 55-gallon drums or boxes in the TRUPACT meets the requirements of a Type B package that woul ct be permitted by the Ue S. Nuclear Regulatory Commission and the Department of Transportation to transport more than 20 curies of plutonium per TRUPACT (Ref. 15, 16).
5
III. HIGH-CURIE CONTENT CH-TRU CONTAINERS A. Radiolytic Gas Generation The CH-TRU defense waste in retrievable storage at SRP that is scheduled to be disposed at WIPP (Ref. 3) includes substantial quantities of Pu-238, an alpha emitter with an 86.4 year half-life. Pu-238 has a specific acti-vity of 17.5 Ci/g and is 285 times more radioactive than Pu-239 with 6.13 X 10- 2 Ci/g.
Radiolysis by alpha anitters of organic wastes breaks down the chemical bonds and produces H2 , 0 2 , and CH 4
- Studies have indicated that hydrogen gas concentrations can increase to potentially explosive levels within a few weeks if 100 curies or more of alpha emitters are stored in a 55 gal-lon drum (Ref. 17).
The lower explosive limit has been defined as either 5 mol% H2 or 4 mol%
H2 in air (Ref. 17) or 6 mo1% H2 (Ref. 12). Even though waste storage drums usually do not contain gas-tight seals and pressurize little or not at all, their void space will still contain elevated concentrations of H2
- Although, this leakage is probably the main reason why H2 gas prob-lems have not occurred* during storage {Ref. 17), it would increase the H2 concentration inside the perfectly gas tight TRUPACT transporting the drums to WIPP. The hydrogen gas buildup in drums or TRUPACTS could poten-tially cause accidents or could increase the quantity of released during handling and transportation accidents.
lysis process radionuclides Also, the radio-has been observed to generate considerable quantities of powder when it occurs in a cellulosic matrix. One study indicated that 10% of the cellulosics, containing 50% of the radioactivity, was reduced to powder in a drum {Ref. 12). The FEIS transportation accident scenarios ass!,Jmed 10% powder with 10% of the radioactivity in the average drum1 after all combustibles had burnede Quantity of waste. Approximately one thousand drums presently stored at SRP and LANL contain over 100 Ci of Plutonium-238. The following table shows a break-down of wastes stored at SRP. Note that a considerable 6
fraction of these wastes originated at either Mound Laboratory or LANL. At the end of 1979 there were 560 30-gallon drums stored in concrete casks in covered trenches at LANL containing over 100,000 Ci of Pu-238 (Ref. 19).
Large amounts of Pu-238 are expected to be generated in the future and shipped to WIPP. Current projections are that over 3.4 mi 11 ion curies out of the 4.0 million curies of TRU radionuclides expected to come to WIPP wi 11 be Pu-238 (Ref. 3). Most of this Pu-238 wi 11 be f ram heat source wastes and unless special *precautions are taken there will be additional thousands of drums filled with over 100 Ci of Pu-238.
Pu-238 Wastes Stored at Savannah River Plant (a)
- Curi es of Pu-238 0-86 86-172 SRP Wastes 657 36 Number of Drums LANL Wastes(bl 36 15 Mound Wastes (C) 172-258 24 9 258-344 14 7 344-430 9 12 430-516 4 13 516-602 3 14 602-688 12
- Total Drums Total Curies 747 (a) Source is Appendix B of Reference 22.
118 49,600 440 128,000 (b) 30-gallon drums (C) Distribution not given. Average for the 55-gallon drums is 290 Ci.
There are also 125,000 Ci of Mound Wastes stored in boxes, cans, and tanks at SRP in some cases at >0.5 Ci/t concentrations.
Length of time for a drum to reach explosive concentrations. Using a G value of 1.9 molecule of gas/100 ev ionizing alpha energy absorbed {Ref.
- 17) and assuming H2 is 50% of the total gas evolved (Ref. 17), one curie of Pu-238 with a 5.5 Mev/disintegration will produce 0.044 liters H2 /week.
7
For a 208.k. drum with a 50% void space, the percentage of the void space filled with hydrogen produced by the radiolysis of 85 Ci Pu-238 each week will be (0.044)(85)(100)/100 = 3.7% H2 /week.
As the following graph shows, a concentration of 6% H2 could be realized within two weeks if there is no leakage fran the drum sealse Data from SRP and LANL indicate that leakage would be expected to occur. According to Zerwekh (Ref. 12), mixtures of 6% or more are explosive. The _following calculation indicates the extent. If all 36 drums in a TRUPACT truck ship-ment had 6% H2 , it would be equivalent to approximately 0.7 pounds of tri-nitroglycerol. Blasting operations of rock use approximately one pound of this explosive per cubic yard of rock.
Explosive equivalent of hydrogen in one TRUPACT from SRP.
Assumptions.
6 1 iters of hydrogen gas at STP per drum 36 drums per shipment Heat of combustion of Trinitroglycerol is 368.4 kcal/gm mole Heat of formation of hydrogen is 52.09 Kcal/gm mole Molecular weight of Trinitroglycerol is 227 Calculation.
Number of moles of hydrogen in one TRUPACT from SRP (6 x 36)/22.4 = 9.7 moles of hydrogen Energy relased in combustion of 9.7 moles of hydrogen 9.7 x 52.09 = 505 Kcal = 505,000 cal Combustion energy per gram of Trinitroglycerol 368,400/227 = 1621 cal/gm Equivalence of hydrogen in one TRUPACT fran SRP 505~000/1621 = 311 gm= 0e7 lbs.
Energy released in combustion of 9.7 moles of hydrogen equals combustion of 0.7 lbs. of Trinitroglycerol.
8
12 H2 GAS GENERATED IN 10 :,STD CH-TRU DRUM WITH
~ 85 Ci Pu-238
- 0:
0 z
- c C\I 8
6
- -EXPLOSIVE' LIMIT
-~
I I
I 4
4I I
2 I I
I I
1 2 3 4 5 TIME (WEEKS) 9
Length of time for a TRUPACT to reach explosive concentrations. Another estimate that could be made is the time it would take the entire void space of a TRUPACT to develop a 5 % by volume concentration of H2
- The time required depends on the number of curies in the TRUPACT, the G{gas) factor for H2 and total gas, the void space in drums and in the TRUPACT volume, and the initial H2 concerntration in the drums. The following tabulation assumes that the drums are vented, sealed and stored for 1-4 weeks before loading in the TRUPACT. The drums are assumed to diffuse gas through their seals so as to maintain one atmsophere of internal pressure. Other assump-tions are a G{gas) factor of 2.0 total, and 1.0 for H2 ; 70% void space in drums; 13.5 m3 of TRUPACT void space outside of drums; 140 Ci of alpha ra-diation per drum; and 36 drums per TRUPACT.
Time in TRUPACT before Drum Storage H2 Cone. in Final H2 Co_nc. becoming potentially Time-days Dru.ms, vo~ % in TRUPACT, Vol % explosive (days) 7 4.2 5.0 25.0 14 8.0 5.0 18.9 21 11.5 5.0 13.3 28 14.8 5.0 7.9 These assumptions for time of storage after venting and time in the TRUPACT are not unreasonable. Without a special effort it is reasonable to believe that drums would be stored for two to three weeks after venting and before shipping. Time in the TRUPACT could easily run 12-15 days when allowance is made for weekends and several days of outside storage at the WIPP site.
B. Effect on Radionuclide Release in Accidents A higher curie content in waste containers than assumed in the FEIS has the potential to significantly increase the calculated amounts of radionuclides rel eased in the various transportation, operation, and human intrusion re-l ease scenarios presented in the FEIS.
DOE recognized that high-curie content drums would increase the maximum dose that could occur to an off-site individual in the limiting operational 10
operational acci_dent at the WIPP facility itself. They responded to this by the following actions:
(a) setting a maximum Pu-239 equivalent curie (P-Ci) limit for 55-gallon drums and other containers.
(b) indicating an intent to require that underground waste hauling vehicles have a maximum possible velocity of 20 mph and puncture re-sistant fuel tanks so that the underground fire accident (which was limiting *in the FEIS analysi5) could be considered incredible.
The hoist drop accident becomes limiting if the underground fire can be considered as incredible. EEG accepted this change in the limiting acci-dent with the stipulation that speed limited and puncture resistant vehicles would be required underground. Also, on March 18, 1983 EEG agreed with the (169 P-Ci) limit (Ref. 6).
In the same response (Ref. 6), EEG also expressed the opinion that it was equally appropriate to set equivalent curie limits for the hoist, for transportation shipments, and for human intrusion situations. Since DOE has not responded to any of these other considerations they are discussed in more detail below.
Equivalent Curie Limitation on Hoist. DOE stated at a February 15, 1983 meeting with EEG that an equivalent curie limit on the hoist would be
- needed to insure that the hoist drop accident would not lead to greater doses than the limiting FEIS accident.
that they have set such a limit.
However, there is no indication The need for a limit is more than theo-retical, as indicated by the following discussion.
A maximum limit of 468 P-Ci on the hoist will insure that the hoist drop accident release will not exceed .the release assumed in the limiting accident in the FEIS (Ref. 6). This limit is unlikely to be met by random probability, especially for wastes from the Savannah River Pl ant and Los Alamos National Laboratory.
The average of all CH-TRU at SRP is now estimated to be about 31.7 curies per 55-ga 11 on drum (Ref. 5). The average of the Pu-238 wastes, which comprise. about 37% the total at SRP, is reported to be 60 P-Ci/drum (Ref.
11
11). Present plans are to load 48 drums on the hoist during regular operation. Thus, an average hoist load of SRP wastes would be 1530 Ci (overall) and 2880 Ci (Pu-238 waste only) which is 3.3 and 6.2 times the 468 Ci limit. Maximum loads could be as much as 6720 P-Ci, if all drums were at the maximum limit of 140 P-Cia Wastes from LANL average about 8.0 P-Ci per 55-gallon drum (Ref. 3). Thus*
a 48 drum load on the hoist would contain 385 P-Ci. If the hoist had only one maximally 1oaded drum and 47 average ones it would exceed the 1imi t by 11%.
Wastes from the other laboratories have average concentrations much lower than SRP and LANL and would be expected to be under the 468 P-Ci hoist limit most of the time. However, without more information on the distri-bution of waste concentrations in individual packages it's not possible to estimate the frequency with which this equivalent curie limit might be exceeded.
Since SRP. and LANL wastes are estimated to be 11.9% and 7.7% of wastes by volume coming to WIPP (Ref. 3), it appears that a 468 P-Ci limit on the hoist would be violated 5 to 15% of the time by random probability unless positive management controls are taken.
It is concluded that the FEIS dose estimates for the limiting operational accident will be exceeded unless a limit is enforced on the equivalent curies that can be placed on the hoist.
Transportation Accidents. If the transportation accident scenario assump-tions used in the FEIS are applied to transportation shipments in which the equivalent curie contents are one or two orders of magnitude greater than assumed in the FEIS, then it is apparent that the individual and po-pulation doses presented in the FEIS will be increased by a like amounta The Preliminary Transportation Analysis (Ref. 7), although it used slightly different rel ease rates and atmospheric diffusion methodology, did report individual doses10-100 times greater than in the FEIS (see discussion in Chapter I).
12
Scenario dose calculations could give even higher numbers than in the PTA, because the Preliminary Transportation Analysis assumed a TRUPACT 1oad of 303 P-Ci. A load of average SRP Pu-238 drums (2880 P-Ci) would contain 9.5 times this amount.
The probabilities of accidents and their severities developed in the PTA are based on route specific (when ava i 1able) and national transportation accident rates. The assumptions are also made that the wastes are in a relatively stable form and transported in a container system that meets Type B requirements. However, items discussed elsewhere in this report concerning possibly explosive gas generation and subsequent formation of powders in cellulosic waste (Ref. 12) and adequacy of the TRUPACT for 1arge quantity shipments cast doubt upon whether these assumpt i ans are valid. These deficiencies, if they actually exist, would be expected to increase the probability and, perhaps, the quantity of releases from transporation accidents.
The one to two order-of-magnitude increase in calculated radiation doses, compared to the FE IS, raises the question of whether a supplement to the FEIS would be appropriate.
Human Intrusion Scenarios. In EEG's March 18, 1983 letter to DOE (Ref. 6) the following statement was made about human intrusion:
"The scenarios that estimate doses from dri 11 i ng through a stack of
~rums or boxes (FEIS, EEG-15, and Scenario I in TME 3151) all assume a m~ximal loaded container is encountered~ Thus, a raising of the limit would increase the dose. However, decay before intrusion would signi-ficantly offset this 10-fold increase since the increases in loading (compared *to the FEIS) are primarily due to Pu-238 and Am-241. The TME 3151 Scenario I assumptions applied to a maximal-loaded and 2 average-loaded LANL drums (with typical radionuclide distribution) would bring to the surf ace about one-fifth of the nuc 1ides permitted by the proposed EPA (40_CFR 191) standard for a reasonably foreseeable release. However; this calculation is sensitive to the assumed radio-nuclide distribution; a drum containing 200 gm of Pu-239 and 82 Ci of 13
Am-241 would exceed the proposed standard at 100 years after closing.
We believe the maximum Am-241activity (including ingrowth from Pu-241) should be limited to 80 Ci per drum to insure that a borehole through the repository would not result in excess radionuclide quantities being brought to the surface~ Also, a loading plan would be necessary in the repository for separating high concentration containers to insure that a single borehole could not strike 2 or 3 high concentra-tion containers. 11 High Am-241 content containers apparently exist. Appendix M of the PfA (Ref. 7) mentions that at Hanford the waste product resulting from reworking 8-10 year old weapons grade plutonium could have an americium to plutonium weight ratio as high as 10 to le This is a radioactivity ratio of about 425 to 1. Also, the equivalent curie calculation considers Am=241 to be only one-third as toxic as the plutonium radioisotopes and a 140 P-Ci limit would allow 420 Ci of Am-241 in a 55 gallon dru*m.
A recalculation slightly changes the amount of Am-241 that could be in a stack of 3 drums and not exceed the permissible quantity that can be brought to the surface under the proposed EPA High Level and Transuranic Waste Standard (40 CFR 191). The permissible quantity (based on an
.assumed inventory at time of closing of 3.6 x 10 6 Ci of alpha emitting TRU) is 77 Ci Am-241 at the time of drilling. Since Am-241 will ingrow from Pu-241 decay the limitation on curies of Am-241 that wi 11 be present at the earliest feas*ible time for inadvertent human intrusion is the.
appropriate limitation. EEG believes the appropriate minimum time is 100 years after repository closing, hence our request in the Executive Summary of EEG-23 (Ref. 20) for 100 years of post-closure administrative control.
If there are no controls over placing of high-curie content drums in the repository, the Am-241 activity should be limited at 100 years after closing to 25 Ci/55-gallon drum.
Gas Generation in the Repositoryo The DOE has recognized that gas genera-tion. from CH-TRU wastes disposed of in WIPP could be a potential post-emplacement problem in the repository. DOE has concluded that an annual gas generation rate of <10 moles/m 3 in the.repository disposal rooms would 14
be acceptab 1e and pl aced a requirement in the Waste Acceptance Criteria (Ref. 21) that limits the concentration of organic material to less than 220 kg/m 3 in 55-gallcin drums (46 kg/55-gallon drum) and 100 kg/m 3 in boxes. This analysis, which assumed an average alpha-emitting radioacti-vity content of only 0.62 Ci/drum, concluded that gas generation from ra-diolysis was negligible compared to that from organics decomposition.
For high-curie content containers the radiolytic gas generation rate can be substantial. For example, an average concentration of 49 alpha curies per drum would generate 10 moles/m 3 per year if the G{gas) factor (mole-cules gas/100 ev of deposited energy) was 1.0 *
- The same drums could be substantial gas generators from both mechanisms since many organic matrices generate much gas by radiolysis. Therefore, it is appropriate to add the gas generation rate that results from organic decomposition to that from radiolysis in a waste storge room.
The appropriate expression for the limiting average concentration of alpha curies and kilograms of organics in all the 55-gallon drums in a reposi-tory disposal roan would be:
_ Ci/drum x G{gas) kg/drum organics Caverage - ( 49 ) + < 1.0.
46
- Gas generation in the repository is yet another situation where problems could arise if high-curi~ content containers are brought to WIPP without a positi-ve mechanism for mfxing them with containers of much lower radionu-clide concentration.
15
IV. ADEQUACY OF TRUPACT CONTAINERS FOR SHIPMENT OF CH-TRU WASTE The WIPP FEIS indicates that about 6 million cubic feet of CH-TRU waste
- will be shipped to WIPP in TRUPACT containers holding 42 drums each. A later DOE publication limits the TRUPACT to 36 drums (Ref. 7).
The authorizing legislation for WIPP (PL 96-164) stated that the defense transuranic waste scheduled for emplacement is exempted from regulation by the NRC. Nonetheless, the DOE FEIS stated 11 The transportation of radio-active waste to the WIPP will comply with the regulations of the U. S.
Department of Transportation (DOT) and the corresponding regulations of the
- u. s. Nuclear Regulatory Commission (NRC) 11 (Chapter 6, Ref. 2). Addition-ally, the DOE's internal order for the packaging of fissile material (Ref *
- 18) states that *when offered to the carrier, each shipment of radioactive material shall be in compliance with applicable DOT and NRC regulations, specifically Code of Federal Regul ati ans, Title 10, Part 71. 31 through 71.42.
Although DOE has voluntarily agreed to meet the NRC and DOT transportation requirements for these shipments, the design of the TRUPACT may preclude compliance with the double containment provisions of the NRC. If the TRUPACT does not meet the double containment provisions, shipments of Transport Group I Type B materials would be limited to 20 curies of pluto-nium per shipment~ a value considerably less than the average of 143 curies/shipment of fissionable material cited in the FEIS (3.4 curies/drum and 42 drums). The number of shipments to WIPP could increase substan-tially if the 4 million curies of alpha nuclides were limited to 20 Ci/shipment. To transport the 6 million cubic feet may require more than 200,000 shipments.
The DOE Waste Acceptance Criteria for CH-TRU waste to be shipped to WI PP (Ref. 21) has the following restriction:
11 Explosives and compressed gases. TRU waste shall contain no explo-sives or compressed gases as defined by 49 CFR 173 Subpart C and G. 11 16
Subpart C applies only to materials that are intended to be explosive. The TRU wastes are not intended as explosives. Subpart G applies only to com-pressed gases whose pressure exceeds 40 psia. The pressure in the drums is not likely to be greater than 2.6 times the atmospheric pressure.
Explosive mixtures of hydrogen, oxygen and methane can occur in these drums at pressures considerably less than 40 psia. As a result, it is question-able whether subparts C and G of 49 CFR 173 provide assurance that explosive gases will not be sent to WIPP. The DOE certification compliance requirements (Ref. 21) also do not satisfactorily address this problem of gas formation from high-curie content drums.
The Code of Federal. Regulations Title 10 Part 71.42 contains the following NRC regulation:
- 11 Plutonium in excess of twenty (20) curies per package packaged in a separate inner container placed within outer packaging that meets the requirement of Subpart C for packaging of material in shali be normal form. The separate inner container shall not release plutonium v,hen the. entire package is subjected to the normal and accident test conditions specified in Appendices A and B. Solid plutonium in the following forms is exempt.from the requirements of this paragraph:
(1) Reactor fuel elements (2) Metal or metal alloy; or (3) Other plutonium bearing solids that the Commission determines should be exempt from the requirements of this section. 11 While the application of this provision to CH-TRU waste is not clear, the preamble to NRC's recent revision to 10 CFR Part 71 (48 FR 356.00 et.seq.)
includes the following: "The Commission considers it most important that solid form plutonium be* doubly contained and that both barriers in the packaging maintain their integrity under normal and accident test condi -
11 tion. Thus, the above regulation suggests that a Type B container within a Type B container would be required for plutonium shipments greater than 20 curies. While the actual testing of the TRUPACT has yet to be per-formed, the ability of the waste containing drums or boxes (Type -A con-tainers) not to open inside the TRUPACT following Type B testing may be difficult to demonstrate. With reference to the three exemptions, NRC 17
11 stated: Since the double containment provision compensates for the fact 11 that the plutonium may not be in a nonrespirable form/ 1 solid forms of plutonium that are essentially nonrespirable should be exempted from the double containment r-equirements. 11 Fuel elements and metals or metal alloys are considered in a nonrespirable form and hence they are exempted.
However, exemption (3) may not apply for the CH-TRU waste since one percent by weight of the Y1aste is allowed to be in respirable form (particle size less than 10 microns).
There is also some question on the compliance of the WIPP waste shipments with the NRC and DOT thermal limits for fissile material. The DOE Waste Acceptance Criteria, the DOE FEIS or internal order for the packaging of fissile materials does not establish a thermal limit. The DOT in 49 CFR Part 173.396 (i)(2), Fissile Radioactive Material~ provides that for Spece 6M metal packaging (a container used at INEL according to the FE IS} the radioactive thermal decay energy output shall not exceed 10 watts. Because of this limitation, the Spec. 6M metal packaging is limited to a content of 0.020 kilograms of Pu-238 (348 curies).
Although DOE has not addressed this important problem with EEG, it appears that there are five possible solutions to comply with the NRC and DOT regu-lations: a) redesign the TRUPACT to assure double containment; b} limit shipments to 20 Ci plutonium; c) NRC could issue new regulations for CH-TRU waste which do not require double containment; or d) DOE could process the waste to remove the high concentrations of Plutonium 238 in -many of the packages or to insure that the material is in non-respirable form; e) grant themselves a variance when they self-certify compliance with the NRC and DOT regulations.
18
GLOSSARY OF ABBREVIATIONS AND ACRONYMS CH-TRU Contact-handled Transuranic DOE U. S. Department of Energy FEIS Final Environmental Impact Statement on WIPP INEL Idaho National Engineering Laboratory, Idaho Fa 11 s, ID LANL Los Alamos National Laboratory, Los Alamos, NM NRC U. S. Nulcear Regulatory Commission P-Ci Plutonium-239 Equivalent curies
- PTA RFP RH SAND DOE Preliminary Transportation Analyses Report Rocky Flats Plant, Golden, CO Remote-handled Sandia National Laboratory, Albuquerque, NM SAR Safety Assessment Report SRP Savannah River Plant, Aiken, SC TRU Transuranic waste WIPP Waste Isolation Pilot Plant near Carlsbad, NM 19
REFERENCES
- 1. U. S. Department of Energy, Draft Environmental Impact Statement Waste Isolation Pilot Plant, DOE/EIS-0026-D, 2 vols., April 1979.
- 2. U. s. Department of Energy, Final Environmental Impact Statement Waste Isolation Pilot Plant, DOE/EIS-0026, 2 vols., October 1980.
- 3. U. S. Department of Energy, Spent Fuel and Radioactive Waste Inventories, Projections, and Characteristics, DOE/NE-0017-1, October 1982.
- 4. U. S. Department of Energy, Waste Isolation Pilot Plant Safety Analysis 5.
Report, 5 vols.
Neil 1, Robert H., Environmental Evaluation Group Di rector, Correspondence to Joseph M. McGough, U. S. Department of Energy WIPP Project Manager,-
November 10, 1982.
- 6. Nei 11, Robert H., Environmental Evaluation Group Di rector, Correspondence to Joseph M. McGough, U. s. Department of Energy WIPP Project Manager, March 18, 1983.
- 7. u. S. Department of Energy, Preliminary WIPP Transportation Analysis, WTSD-TME-002, April 1983.
- 8. Environmental Evaluation Group, Partial Comments on U.S. Department of Energy, Preliminary WIPP Transportation Analysis, WTSD-TME-002, April 1983, May 1983.
- 9. U. s. Department of Energy, DOE Response to the State of New Mexico'sCom-ments on "Summary of the Results of the Evaluation of the WIPP Site and Preliminary Design Validation Program 11 WIPP-DOE=l61, WIPP-DOE-174, June 1983.
- 10. Envi ronmen_tal Evaluation Group, Response to DOE response to EEG comments on U. s. Department of Energy, Preliminary WIPP Transportation Analysi*s,
WTSD-TME-002. Draft memorandum for use in meeting with U. S. Department of Energy, July 22, 1983.
- 11. Roberts, F. P., An Assessment of Radiation Effects in Defense Transuranic Waste Forms, PNL 3913, July 1981.
- 12. Zerwekh, Al, Gas Generation From Radiolytic Attack of TRU-Contaminated Hydrogenous Waste, LA-7674-MS, June 1979.
- 13. Molecke, Martin A., Gas Generation From Transuranic Waste Degradation, SAND 79-09llC, November 1979.
- 14. Molecke, Martin A., Gas Generation From Transuranic Waste Degradation: An Interim Assessment, SAND 79-0117, 1979.
- 15. 10 Code of Federal Regulations," Part 71.42(b). (Also 49 Code of Federal Regulations, Part 143.)
I I
I
- I 16. Jefferson, Robert M., ed., Program Strategy Document for the Nuclear Materials Transportation Technology Center {FY 80), SAND 80-0784, Ap_ril 1980.
- 17. Ryan, John P., Radiogenic Gas Accumulation in TRU Waste Storage Drums, DP-1604, January 1982.
- 18. U.S. Department of Energy, Orders, DOE 5480.l chg 3, Chapter III, "Safety Requirements for the Packaging of Fissile and Other Radioactive Mater:-ials,"
Section 7, Requirements, May 1, 1981.
- 19. Los Alamos National Laboratories, Alternative Transuranic Waste Management Strategies at Los Alamos National Laboratory, LA-8982-MS, September 1981.
- 20. Environmental Evaluation Group, Evaluation of the Suitability of the WIPP Site, EEG-23, May 1983.
- 21. U. S. Department of Energy, TRU Waste Acceptance Criteria for the Waste
- Isolation Pilot Plant, Revision 1, September 1981.
21
~-------------------------~- --
- 22. U. s. Department of Energy, Alternatives for Long-Term Management of Defense Transuranic Wastes at the Savannah River Pl ant , Ai ken, South
.__ Carolina, D0E/SR-WM-79-1, July 1979.
)
\
22
The following page, explaining the method of calculating the Pu-239 equ~valent TRU activity, was reproduced from the Preliminary Transportation Analysis.
(Reference 7).
APPENDIX A Pu-239 EQUIVALENT TRU ACTIVITY The maximum Pu-2_39 equivalent TRU activity limit (AM) for 55-gallon drums is 140 P-Ci.
The Pu-239 equivalent correction factor is ba~ d on the maximum permissible concentration {mpc) from 10CFR20, 5 Appendix B, Table 1, Column 1 for limiting form.
The activity (AM) can be characterized bY:-
K A.1 AM = I where there are K TRU isotopes, Ai is the maximum activity of isotope i, and CF; is an mpc correction factor for isotope i obtained by multiplying the mpc specifie:d above by 5 x 1011 ml/µCj to normalize the factor relative to Pu-239.
The correction factors used in these analyses are:
Isotope Correction Factor (CF;)
U-233 250
. Pu-238 1 Pu-239 .1 Pu-240 1 Pu-241 45 Pu-242 1 Am-241 3 Am-243 3 Cm-244 4 .. 5
- Cf-252 3 Values of the maximum permissible concentration (mpc) are currently under review~~ the Nuclear Regulatory Commission {NRC) in their revision of 10CFR20 °. It is recognized that the correction factors (CF) listed are not consistent with the dose conversion factors used in the analysis. It is expected that the CF values for U-233, Ara-241, Am-243, Cr.1-244, and Cf-252 may be reduced when 10CFR20 is revised, and the value for Pu-241 may focrease .. However, the impact of the activity assumed to be present in this analysis *will remain unchanged by the revision, but the allowed contents of a container may increase or decrease for various isotopes. To allow a preliminary assessment of these potential changes, Table A-1 provides a summary of CF values based on other data sets. The data sets include the dose conversion factors used in this analysis26, which it is considered will produce correction factors similar to the mpc's expected in the NRC's revision of JOCFR20.
23
Envimnental. Evaluat:i.al Gr:aJp Reports (Continued from Front Cover)
EEX3-12 Little, Marshall s., Potential Release Scenario and Radiological Consequence Evaluation of Mineral Resources at WIPP, May 1982.
E00-13 Spiegler, Peter., Analysis of the Potential Fonnation of a Breccia Chimney Beneath the WIPP :Repositocy, May, 1982.
E00-14 Not published.
E00-15 Bard, stephen T., Fstimated Radiation Doses Resulting if an Explorato:cy Borehole Penetrates a Pressurized Brine Reservoir Assumed to EKist Below the WIPP Repositocy Horizon, March 1982.
E00-16. Radionuclide Release. Transj;)ort and Consequence Modelim for WIPP. A Reoort of a Workshop Held on September 16-17. 1981, February 1982.
E00-17 Spiegler, Peter, Hydrologic Analyses of 'lwo Brine Encounters in the Vicinity of the Waste Isolation Pilot Plant (WIPP)
Site, December 1982.
E00-18 Spiegler, Peter, Origin of the Brines Near WIPP from the Drill Holes ERDA-6 and WIPP-12 Based on stable Isotope Concentration of Hydrogen and Oxygen, March 1983.
- E00-19 Clanne;l.l, James K. , Review Comments* on Erwirornnental Analysis Cost Reduction Proposals (WIPP/OOE-136) July 1982, November 1982 *
- E00-20 E00-21 Baca, 'Ihamas E. , An Evaluation of the Non-radiological Erwirornnental Problems Relating to the WIPP, Februazy 1983.
Faith, stuart, et al., 'Ihe Geochemistry *of 'lwo Pressurized Brines From the castile Fonnation in the Vicinity of the Waste Isolation Pilot Plant <WIPP) Site, April 1983. -
E00-22 Em Review Comments on the Geotechnical Reports Provided by OOE to Em Under the stipulated Agreement 'lhrough March 1.
1983, April 1983.
E00-23 Neill, Robert H., et al., Evaluation of the SUitability of the WIPP Site, May 1983.
E00-24 Neill, Robert H. and James K. Qiannell Potential Problems From Shipment of High-curie Content Contact-Handled Transuranic ( CH..JI'RCJ) Waste to WIPP, August 1983
- EEX;-25 ara.turvedi, Iokesh, Occurrence of Gases in the Salado Fonnation, March 1984.
EEX;-26 Spiegler, Peter, Erwi.ronmental Evaluation Group's Environmental Moni:toring Program for WIPP, October 1984.
EEX;-27 Rehfeldt, Kenneth, Sensitivity Analysis of Solute Transport in Fractures and Detenni.na.tion of .Anisotropy Within the CUlebra Iblamite, September 1984.
EEX;-28 Knowles, H. B., Radiation Shielding in the Hot Cell Facility at the Waste Isolation Pilot Plant: A Review, November 1984.
EEX;-29 Little, Marshall s., Evaluation of the Safety Analysis Report for the Waste Isolation Pilot Plant Project, May 1985.
EEX;-30 Dougherty, Frank, Tenera. Corporation, Evaluation of the Waste Isolation Pilot Plant Classification of Systems. structures and Components, July 1985.
EEX;-31 Ramey, Dan, Olemist;cy of the Rustler Fluids, July 1985.
EEX;-32 Cbaturvecli, Iokesh an::l James K. Cl'rannell, 'Ihe Rustler Fonnation as a Transport Medium for Contaminated Groundwater, December 1985.
EEX;-33 Olannel.l, James K., John C. Rodgers and Robert H. Neill; Adequacy of TRUPACl'-I Design for Transportim eontact-Handled Transuranic Wastes to WIPP, Jt.me 1986.
EEX;-34 Olaturvedi, Iokesh, (ed), '!he Rustler Fonnation at the WIPP
. Site, Januacy 1987.
EEX;-35 Cl1apman, Jenny B. , stable Isotopes in Southeastern New Mexico Groundwater: Implications for rating Recharge in the WIPP Area, October 1986.
EEX;-36 Iowenstein, Tim K. , Post Burial Alteration of the Pennian Rustler Fonnation Evaoorites. WIPP Site. New Mexico, April 1987.
EEX;-37 Roogers, John C., E>maust stack Monitoring Issues at the Waste Isolation Pilot Plant, November 1987.
EEX;-38 Rodgers; John C., Kenney, Jim w., A Critical Assessment of Continuous Air Monitoring Systems At '!he Waste Isolation Pilot Plant, March 1988.
EEX;-39 01.apman, Jenny B., Chemical and Radiochemical Cll.aracteristics of Groundwater in the CUlebra Dolomite. Southeastern New Mexico, March 1988.
EEX:;-40 Review of the Final Safety Analysis Report (Draft). OOE Waste Isolation Pilot Plant, May 1989.
EEX:;-41 Review of the Draft SUpplement Erwirornoontal Impact statement, OOE Waste Isolation Pilot Plant, July 1989.
EEX:;-42 <llatw:vedi, Iokesh, Evaluation of the OOE Plans for Radioactive Experiments arrl Operational Dem:>nstration at WIPP, September, 1989.
EEX:;-43 Kenney, Jim W., Jahn C. Rodgers, Jenny B. Cllapnan, arrl Kevin J. Shenk, Preoperational Radiation SW:Veillance of the WIPP Project by EEX:i. 1985-1988, Januacy 1990.
EEX:;-44 Greenfield, Moses A., Probabilities of a catastrophic Waste Hoist Accident at the Waste Isolation Pilot Plant, Januacy 1990.
EEX:;-45 Silva, Matthew K., Preliminary Investigation into the Explosion Potential of Volatile Organic conp:,unds in WIPP OI-TRU Waste, June 1990 *
- f
EEG-33 DOE/ AL/ 10752-33
- NEWNDJCO HEALTH AHO ENVIRONMENT OEPAI\TMENT ADEQUACY OF TRUPACT-I DESIGN FOR TRANSPORTING CONTACT-HANDLED TRANSURANIC WASTES TO WIPP James K. Channell, John C. Rodgers, and Robert H. Neill Environmental Evaluation Group Environmental Improvement Division Health and Environment Department J State of New Mexico P.O. Box 968 Santa Fe, NM 87503 June 1986
Environmental Evaluation Group Reports EEG-1 Goad. Donna. A Compilation of Site Selection Criteria Considerations and Concerns Appearing in the Literature on the Deep Disposal of Radioactive Wastes. June 1979.
EEG-2 Review Comments on Geological Characterization Report.
Waste Isolation Pilot Plant (WIPP) Site. Southeastern New Mexico SAND 78-1596. Volumes I and II. December 1978.
EEG-3 Neill. Robert H .* et al. eds .* Radiological Hea1th Review of the Draft Environmental Impact Statement (DOE/EIS-0026-D)
Waste Isolation Pilot Plant. U.S. Department of Energy.
August 1979.
EEG- 4 EEG-5 Little. Marshall s .. Review Comments on the Report of the Steering Committee on Waste Acceptance Criteria for the Waste Isolation Pilot Plant. February 1980.
Channell. James K ** Calculated Radiation Doses From Deposition of Material Released in Hypothetical Transportation Accidents Involving WIPP-Related Radioactive Wastes. November 1980.
EEG-6 Geotechnical Considerations for Radiological Hazard Assessment of WIPP. A Report of a Meeting Held on January 17 - 18. 1980. April 1980.
EEG- 7 Chaturvedi. Lokesh. WIPP Site and Vicinity Geological Field Trip. A Report of a Field Trip to the Proposed Waste Isolation Pilot Plant Project in Southeastern New Mexico.
June 16 to 18. 1980. November 1980.
Wofsy. Carla. The Significance of Certain Rustler Aquifer Parameters for Predicting Long-Term Radiation Doses from WIPP. September 1980.
t EEG-9 Spiegler. Peter. An Approach to Calculating Upper Bounds on Maximum Individual Doses From the Use of Contaminated Well Water Following a WIPP Repository Breach. September 1981.
EEG-10 Radiological Health Review of the Final Environmental Impact Statement (DOE/EIS-0026) Waste Isolation Pilot Plant. U. s. Department of Energy. January 1981.
(continued back page)
FOREWORD The purpose of the Environmental Evaluation Group (EEG) is to conduct an independent technical evaluation of the potential radiation exposure to people from the proposed Federal Radioactive Waste Isolation Pilot Plant (WIPP) near Carlsbad. in order to protect the public health and safety and ensure that there is minimal environmental degradation. The 'EEG is part of the Environmental Improvement Division. a component of the New Mexico Health and Environment Department -- the agency charged with the primary responsibility for protecting the health of the citizens of New Mexico .
The Group is neither a proponent nor an opponent of WIPP.
Analyses are conducted of available data concerning the proposed site. the design of the repository. its planned operation. and its long-term stability. These analyses include assessments of reports issued by the U.S. Department of Energy (DOE) and its contractors. other Federal agencies and organizations. as they relate to the potential health. safety and environmental impacts from WIPP.
The project is funded entirely by the U.S. Department of Energy through Contract DE-AC04-79AL10752 with the New Mexico health and Environment Department.
Robert H. Neill Director
-i-
EEG STAFF (1) .
James K. Channell
- Ph.D ** P.E ** Environmental Engineer i
Jenny B. Chapman. M.s ** Hydrogeologist Lokesh Chaturvedi. Ph.D ** Engineering Geologist William c. Foege. M~L.S./M.s ** Librarian 1.
Jim Kenney. M.s ** Environmental Scientist Miriam L. Kramer. B.A ** Secretary III
- c. Robert McFarland. B.s ** Quality Assurance Engineer Jack M. Mobley. B.A ** Scientific Liaison Officer Robert H. Neill. M.S *.* Director.
Teresa Ortiz. Administrative Secretary John c. Rodgers( 1 ). M.S .* ~ealth Physicist Norma Silva. Administrative Officer (1) Cert~fied. American Board of Health Physics.
-ii-
EXECUTIVE
SUMMARY
TRUPACT I is the shipping container designed by the U. s.
Department of Energy (DOE) to transport contact-handled transuranic (CH-TRU) radioactive waste to the Waste Isolation Pilot Plant near Carlsbad. New Mexico. Approximately 24.000 shipments will be required to transport the 6 million cubic feet of waste to WIPP over a 20 year period.
Transportation regulations that have been issued by the U. s.
Department of Transportation permit the DOE to evaluate. approve and certify their own packages provided the regulations are equivalent in safety to those specified by the U.S. Nuclear Regulatory Commission.
TRUPACT I was designed with two features that do not meet the NRC and DOT transportation regulations:
(i) i t has only single containment. which is not permitted for most forms of radioactive material if the shipment contains greater than 20 Curies of plutonium: and (2) the waste storage cavity is continuously vented through filters to the atmosphere.
The evaluation addressed these two design features as well as the problem of hydrogen gas generation in the wastes and the limits of radioactive materials proposed by DOE for a TRUPACT shipment.
A review of the history of regulations pertaining to the double I
containment requirement indicated that they clearly apply to transuranic waste shipments unless i t can be shown that the waste forms are "sufficiently nonrespirable". However. waste
-iii-
forms which are permitted by WIPP waste certification criteria to contain 1% respirable fines. average 25% combustible material. and can generate potentially flammable or explosive 9oncentrations of hydrogen gas should not be considered either nonrespirable or stable.
A principal advantage of a TRUPACT with double containment is
~he estimated decrease from 12 to 0.02 in the number of accidents involving radionuclide releases during the WIPP Project. Even minor accidents involving little public radiation exposure are costly to monitor and clean up and can decrease public confidence in the safety of radioactive material shipments. An additional advantage of double containment is the extra protection i t is expected to provide in the event of a low probability (0.1-1%)/high consequence accident.
fatalities with the present design.
These very severe accidents could result in up to 10-30 latent cancer Double containment is estimated to reduce this by at least 60% to BO%.
NRC regulations prohibit all forms of venting and do not permit reliance on filters to meet permissible radionuclide releases.
The TRUPACT I design has incorporated continuous venting through filters. The purpose of TRUPACT venting is to reduce the probability of failure from fatigue in the package due to pressure changes caused by altitude and temperature variation.
There is also concern whether hydrogen buildup through alpha induced radiolysis of organic material in a sealed TRUPACT would be a problem. EEG is opposed to continuous venting of the TRUPACT on the grounds that i t compromises the integrity of the package by providing a pathway for release in case of filter malfunction and the possibility that the vent area is more 1
susceptible to failure during a severe accident and because viable alternatives exist for hydrogen control.
-iv-
The report evaluates in detail hydroge~. generation in TRU wastes because of its relation to the venting issue. While venting of both drums and the TRUPACT might be able to maintain hydrogen concentrations below the minimum flammable concentration of 4%
for low-curie loads. i t is questionable if control would be adequate for some high-curie ioads.
Although DOE has concentrated on venting mechanisms for controlling hydrogen concentrations. promising. alternate methods exist and should be investigated. These include the use of hydrogen-getters or hydrogen-oxygen recombiners along with the use of administrative controls. One or more of these alternate methods hold the promise of being more reliable gas control mechanisms than ventlhg and their use would remove the need for venting to control hydrogen concentrations.
DOE has established an tipper limit of 12.000 curies of TRU waste in a TRUPACT-I load. This load wouid contain a more toxic inventory than a spent fuel shipment. Also. because of differences in waste form and package design i t is expected that a somewhat higher fraction of the wastes would be released from the TRUPACT than from a spent fue.l cask following a severe accident. Since no waste generating site has average waste concentrations as high as 2.000 curies i t is not necessary to establish such a high upper limit in order to transport defense wastes to WIPP.
EEG recommends that TRUPACT-I not be certified for transporting any waste to WIPP unless the vents are sealed and the package is limited to 20 curies of plutonium per load. We further recommend that: (1) the TRUPACT be redesigned to include double containment and eliminate continuous venting: (2) the use of methods other than venting for hydrogen gas control be seriously
-v-
CONTENTS Title Foreword. . . . * . . . * * . . . . . . . * . . * . . * * . . . . . . . * . . . . * * . . * . . . . . i EEG Staff . . . . . . . . . . * . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii Executive Swnrnary. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i i i List of *rables . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ix List of Figures. . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . xi
- 1. IN'l"RODUCTION * . . * * * . . . . * * * * * . * * * . . . . . * * * . . * * * * . . * * * . . . * . 1
- 2. DOUBLE CONTAINMENT * * . * . * * . * * . * * * * * * * * * * * * . * * . * * * * * * . . . . 7 2.1 Statement of I s s u e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7
- 2. 2 Regulatory Considerations...................... 7 2.2.1 Regulations and History.................... 7 2.2.2 Bases for Exemption Mechanism *..**..*..*.** 13 2.2.2.1 Applicability of 1974 Regulations .... 13 2.2.2.2 Respirability * . . . . * . . . . . . . * . . . . . . . . . . 14 2.2.2.3 Comparison With Spent Fuel Shipmen ts . . . . . . . . . . . . .* . . . . . . . . . . . . . 1. 4 2.2.2.4 Amount of Plutonium Per Shipment . . . . . 15 2.3 Possible Risks and Consequences *.. : *...*......* 15 2.3.1 Radiological Considerations - Incident Free Transportation . * . . . . * . . . . . . * . . * . . . . . 16 2.3.1.1 External Radiation ****..*..*......... 16 2.3.1.2 Design Effect on Radiation Level . . . . . 17 2.3.1.3 Occupational Radiation Exposure . . . . . . 19 2.3.2 Radiological Considerations - Accidents .... 20 2.3.2.1 Fractional Releases From Accidents ... 20 2.3.2.2 Expected Number of Accidents ...****** 25 2.3.2.3 Radionuclide Releases From Accidents.......................... 26 2.3.2.4 Radiation Doses From Accidental Releases . . . . . . . . . . . . . . . . . . . . . . . . . . . 29
. 2.3.2.5 Radiological Contamination * . * . . . . . . . . 30
- 2. 3. 3 Radiological Healt1'1 Effects From Transportation. . . . . . . . . . . . . . . . . . . . . . . . . . . 32 2.3.4 Non-Radiological Risks ..*.*.**.*......*...* 35 2.3.5 Trading-off Radiological and Non-Radiological Risks . . . . . . . . . * . . * * . . . . . 36 2.3.5.1 Projected Radiological and Non-Radiological Risks.* . * . . . . . . . . . . 37 2.3.5.2 Is Trading Off.Appropriate? . . . . . . . . . . 39
-vii-
TitJ.e
- 3. CONTINUOUS VENTING AND GAS GENERATION . . . . . . . . . . . . . . . . . . 44 3.1 Statement of Issue . . . . . . . . . . . . . . . . . . . . . . . . * . . . . 44 3.2 Regulatory Considerations . . . . . . . . . . . . . . . . . . . . . . 45 3.3 Gas Generation in TRU Wastes . . . . . . . . . . . . . . . . . . . 47 3.3.1 Gas Generation Processes . . . . . . . . . . . . . . . . . . . 49 3.3.2 Radiolysis in THU Waste . . . . . . . . . . . . . . . . . . . . 50 3.3.3 Controlling Pressure and Hydrogen Buildup .. 61 3.3.4 Modeling of Hydrogen Gas Buildup . . . . . . . . . . . 64 3.3.5 Alternative Control Strategies . . . . . . . . . . . . . 71
- 4. HIGH-CUFlIE CONTENT . . . . . . . . . . . . . * . * . . . . . . . . . . . . . . . . . . . . 78 4.1 Statement of Issue . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 78 4.2 Possible Risks and Consequences . . . . . * . . . . . . . . . . 79 4.2.1 Comparison With FEIS; . . . . . . . . . . . . . . . . . . . . . . 79 4.2.2 Comparison With Other Radioactive Materials Shiprnents. . . . . . . . . . . . . . . . . . . . . .
4.2.3 Non-Radiological Considerations . . . . . . . . . . . .
4.3 Operational and Economic Considerations . . . . . . . .
- 4. 3 .1 Re-PacJc:aging. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 4. 3. 2 Number of Shipments. . . . . . . . . . . . . . . . . . . . . . . .
BO 84 84 B4 86
- 5. REDESIGN O:fi' TRUPACT . . . . * * . . . . . . . * . . . . . . . . . . . . . . . . . . . . . . B7 5.1 Modifications Being Considered . . . . . . . . . . . . . . . . . B7 5.2 Possible Radiological Impacts . . . . . . . . . . . . . . . . . . BB
- 6. CONCLUSIONS AND RECOMMENDATIONS . . . . . . . . . . . . . . . . . . . . . . . . 90 6 .1 Conclusions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 6.1.1 Double Containment . . . . . . . . . . . . . . . . . . . . . . . . . 90 6.1.2 Continuous Venting and Gas Generation . . . . . . 91 6.1.3 High-Curie Shipments . . . . . . . . . * . . . . . . . . . . . . . 92
- 6. 2 Reconmenda tions. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 93 REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . Ill **************** 95 APPENDICES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . o * * *
- 99 A. Modeling Hydrogen Generation and Dissipation in THU Waste Packages B. "Discussion of Propargyl Ethers as Hydrogen Getters with Respect to Nuclear Waste Disposal"
-viii-
LIST 01 TABLES
- 1. Double Containment (DC) Hequirements For Shipments Greater Than 20 Ci Pu .. B
- 2. Estimated Occupational Radiation Doses From Loading & Un.loading Various THUPACT Designs. 21
- 3. Fraction of Radionuclides Released as Respirable Aerosols From Transportation Accidents . . . . . . . . . . . 22
- 4. Expected Number of Accidents Involving Radionuclide Releases During Truck Transportation to WIPP . . . . . . . . 26
- 5. Average Amount of Radioactivity Being Transported to WIPP . . . . . . * . . . . . 27
- 6. Quantities of Radionuclides Released to the Environment From Truck Accidents . . * . . * . * * . . * . 28 7* Radiation Population Doses From Expected WIPP Transportation Accident Releases . * . . . . . . . . . . . . . . . 30 B. Estimated Population of Individual Radiation Doses From Severe Transportation Accidents . . . . . . . . . . . . . . . . . . 31
- 9. Expected Latent Cancer Fatalities From Transportation to WIPP . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34
- 10. Estimated Latent Cancer Fatalities From Severe Transportation Accidents .. 34 11.. Non-Hadiological Fatalities Expected From Shipment of CH-THU Wastes to WIPP by Truck ... 35
- 12. Possible Latent Cancer Fatalities From Leaving Stored Waste at INEL . . . . . . . . . . . 40 1.3. Estimated Deaths Expected to Occur From
\I the WIPP Project . . . . . . . . . . * . . . . . . . . . . . . . . 42
- 14. Venting of 'l'HUPACT.................................... 46
_t5. Major Gas Generation Processes and Rates . . . . . . . . . . . . . .
1.6. SARP Maximum TRUPACT Loadings . * . . . . . * * . . . * * * . . . . . . . . . . 78
-ix-
LIST OF TABLES
- 1. 7. Comparison of Releases Between FEIS & Chapter 2 . . . . . . . BO 1.8. Comparison of Spent Fuel and TRUPACT Radiological Toxicity, . . . . . . . . . . . . * . . . . . . . . . . . B2 V
1.9. Releases From Truck Accident Involving Spent Fuel and *.rRUPACT. * . * . . . . * . . . . . . . . . . . . . * . . . . . . . . . 83
-x-
LIST OF FIGURES Page
- 1. Points of Origin and Principal Destinations of THU Waste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
- 2. TRUPACT-I Transport Package . . . . . * . * . . . * . . . . . . . . . . . . . . . 6
~- Hydrogen Generation in Experimental Waste Matrices . . . . . . D ********* e *********************** 51.
- 4. G (gas) as a Function of Time . . . . . . . . . . . . . . . . . . . . . . . . . 53
- 5. G (gas) as a Function of Integrated Dose . * . . . . . * . . . . . . 55
- 6. Two Component G (gas) Time-Variant Function . . . . . . . . . . . 56
- 7. Model Predicted G (gas) Variation With Integrated Dose . . . . . . . . . . . . . . . . . . . . . . . . * . . . . . . . . . 58 B. Modeled and Observed Time Varying G (hydrogen) in a Sealed RFP Drum of TRU Waste . . . . . . . . . . . . . . . . . . . . . 59
- 9. Observed Time-Varying G (hydrogen) in a Sealed Rfo'P Drum of THU Waste. . . . . . . . . . . . . * . . . . . . . 60
- 10. Filter Vent Concepts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63
- 11. Schematic of Drums Inside TRUPACT . . . * . . . . . . . . . . . . . . . . . 65
- 12. Model Verification Results. . . . . . . . . . . . . . . . . . . . . . . . . . . . 67
- 13.
14.
15.
Hydrogen Buildup in Vented Drums . . . . . . . . . . . . . . . . . . . . . . 68 Vented 'I'RUPACT with 36 Drwns (20 Ci/drum)
Rockwell Hanford Operations Catalyst Bed and Vent for Drums of Radioactive Waste . . . . . . . . . * . . . . . 74 70
- 16. Effects of Getter on Hydrogen Concentration in Drums. . . . . . * . . . * . . . . . . . . . . . . . . . . . . . . . 76
-xi-
- 1. INTRODUCTION The Waste Isolation Pilot Plant (WIPP) Mission is to provide a research and development facility to demonstrate the safe disposal of fadioactive waste resulting from the defense activities and programs of the United States (Ref 1). During the WIPP Project 6.250.000 cu ft of defense transuranic waste (THU) will be disposed of in a repository 25 miles east of Carlsbad. New Mexico ln a bedded salt formation at .a depth of 2150 feet. The THU wastes will be shipped from the Los Alamos National Laboratory (LANL) in Los Alamos. New Mexico. Idaho National Engineering Laboratory (INEL) in Idaho Falls. Idaho.
Rocky Flats Plant (RFP). Rocky Flats. co .. Hanford National Laboratory. Hanford. WA .. Oak Ridge National Laboratory (ORNL).
Oak Ridge. TN .* Savannah River Plant (SRP). Aiken. South Carolina. the.Mound Laboratories in Miamisburg. Ohio. the Nevada Test Site (NTS) and.Lawrence Livermore National Laboratory (LLNL) in California (Ref 2).
The Department of Energy (DOE) has developed a Type B packaging system known as the TRUPACT (Transuranic Waste Package Transporter) to transport the THU waste to WIPP. The present 36 drum design (TRUPACT-I) will require about 24.000 shipments over a 20 year period beginning October 19BB. The relative fraction to be shipped via truck and railroad has not been determined.
Figtire 1 shows the generation and.storage sites of the THU wastes that will be transported to WIPP via truck or rail.
Figure 2 shows a schematic diagram of the TRUPACT.
1
This report specifically evaluates the 36-drum THUPACT-I design.
although reference is made in places to a possible 48-drum design. Two units are being built to the THUPACT-I design and i t is EEG's understanding that DOE certification will be sought to transport THU wastes in these units.
V
~bile this report was being prepared. the DOE announced plans in May 1986 to try to redesign the THUPACT to include double containment and eliminate venting. Subsequently. the Albuquerque Operations Office DOE funded the American National Standards Institute to establish a panel to make an independent review of waste packaging issues. The Statement of Work specified. 0 This task will initially consider the need for separate inner containment for plutonium packagings and the nonradioactive gas venting from packages containing transuranic wastes ... The Panel's work will be completed by September 30, 1986. Since DOE appears to be still questioning the technical need for these requirements. EEG believes i t is necessary to publish our analyses arid conclusion on these heal th and saf_ety issues related to the transportation of THU Waste to WIPP.
There are four interrelated sets of safety regulations governing the packaging of radioactive materials transported in the U.S .
(Ref 2a). The Department of Transportation (DOT) is responsible for regulating safety in transportation of all hazardous materials. including radioactive.materials, and its packaging requirements are given in 49 CFR Part 173. The Nuclear Regulatory Commission (NRC), under the Atomic Energy Act of 1t954, as amended. also regulates the transportation of r:'adioactive materials. Through a memorandum of understanding w'ith the DOT. t.pe NRC r~views and approves packages used by its commercial licensees for radioactive materials exceeding Type A quantities and fissile material. NRC's packaging and 2
transportation regulations are provided in 10 CFR Part 71. The Department of Energy, except for special cases legislated by Congre~s. is not subject to NRC regulations. DOE packag.ing requirements. which are applicable to its contractors. closely parallel the provisions of 10 CFR Part 71 and are contained in DOE Orders. The packaging requirements of all three agencies have been brought into conformance, more or less. with the transport recommendations of the International Atomic Energy Agency (IAEA). in which the U.S. is an active participant. The IAEA transportation recommendations are given in IAEA Safety Series #6 (Ref 3).
The issues addressed in this report are whether the existing design of the shipping container (TRUPACT) meets minimal regulatory requirements relating to the safe transport of radioactive materials issued by the U.S. Department of Transportation and the U.S. Nuclear Regulatory Commission and what the health and safety consequenees are (if any) of not meeting these regulations.
Transportation regulations that have been issued by the U.S.
Department of Transportation (49 CFR 173.7 (d)) permit the U.S.
Department of Energy to evaluate. approve and certify its own packages. provided the regulations are equivalent in safety to those specified by the U.S. Nuclear Regulatory Commission in 10 CFR Part 71~ This agreement has been in effect since 1973 (Ref
- 4)
- Congressional authorization of the WIPP mission was contained in the December 1979 Appropriations Act for the national security programs and functions of the DOE for FY 19B0 (PL 96-164). The express purpose is to provide a research and demonstration facility to demonstrate the safe disposal of radioactive wastes 3
resulting from the defense activities and programs of the United States exempted from regulation by the Nuclear Regulatory Commission.
Although the DOE was exempted from NRC transportation regulations. ten months later the Department of Energy issued the Final Environmental Impact Statement (FEIS) for WIPP in October 1980 which stated. "The transportation of radioactive wastes to WIPP will comply with the regulations of the U.S.
Department of Transportation (DOT) and the corresponding regulations of the U.S. Nuclear Regulatory Commission (NRC)"
(Ref 1). Nothing was said regarding the use of DOE transportation regulations in lieu of those issued by NRC or DOT.
While exemptions to regulations are acceptable mechanisms to demonstrate conformance to a standard. the DOE did not indicate in the WIPP FEIS that their commitment to comply with the regulations of the DOT and NRC was through exemptions to be i's sued by either the DOT or the DOE.
4
,I V,
LANL ,,____
fo:ndia Lovelacto
'--l..r-~--*...J WIPP 0 Generators of TRU
/}, Stora~ s;ite R8D dispc,801 facility
- - Shipments J'igure ta Points of origin and princip:Ll destinations o:f TRU waste.
Filtered vents Outer door
!mer door Figure 2Q TRUPACT-I transport package *
- 'I
- 2. DOUBLE CONTAINMENT 2.1 Statement of Issue The TRUPACT was designed in 1978 for single containment (Ref 5).
Federal regulations in existence at that time. as well as today.
required a double containment design (Ref 6 and 7) for shipments in excess of 20 curies of plutonium. Most of the shipments to WIPP will have more than 20 curies of plutonium.
2.2 Regulatory Considerations
- 2.2.1 Regulations and History A chronological history of the significant regulatory requirements follows and is also shown on Table 1. In August 1973 the U.S. Atomic Energy Commission (AEC) issued a notice of proposed rule making (NPR) to require special packaging conditions for shipments of plutonium in excess of 20 curies.
In June 1974. the AEC issued regulations (10 CFR 71) (Ref 6) requiring shipments of plutonium in excess of 20 curies to be in
- a solid form and doubly contained.
of the contents." They also stated.
The AEC noted that after studying the comments on the August 1973 NPR. the effect of their amended provisions "is still to requ~re double containment "The Commission considers i t most important that solid form plutonium be doubly contained and that both barriers in the packaging maintain their integrity under normal and accident test conditions." In 1978 the Transportation Technology Center of the Sandia National Laboratories designed the TRUPACT with single containment.
In December 1979 the DOE commented to the NRC on the double containment requirement of the NRC and specifically requested 7
TABLE 1.
DOUBLE CONTAINMENT (DC) REQUIREMENT FOR SHIPMENTS GREATER THAN 20 Ci Pu 0 Aug 1.973 AEC issued NPR to require DC (FR).
o June 1.974. AEC in 10 CFR 71.42 requires DC and solid form effective in 1978 (FR).
0 1.978 Sandia designs TRUPACT with single containment.
0 Dec 1.979 DOE and Sandia letters request NRC to exempt shipments of Pu contaminated wastes from DC.
0 Dec .1979 WIPP authorized by Congress.
(X) o Oct 1980 WIPP FEIS commits to meet NRC trans. regs.
o May .19 DOE Orders (Regulations) require compliance with NRC 1.0 CFR 7.1 DC (DOE 5480.1.. Chg 3. p III-6).
o June 1.9B2 DOE Peer Review in Aug 19 notes that design does not meet 10 CFR 71 NRC DC requirement (SAND ~2405).
o Jan 1.9B3 Sandia response ignores issue of regulatory requirements and exemption for DC (SAND 82-1.493).
o March 19B3 DOT requires all Type B packages to be designed and constructed to meet applicable requirements of NRC 10 CFR PART 71 (49 CFR 173. 41.3).
o July 1.9B3 DOE issues Draft Order with exemption mechanism from DC {DOE 54B0.1.A Chg 3 Draft 7-29-83) .
0 Aug 19B3 NRG confirms requirement for solid form and DC in 1.0 CFR 71.63 and notes that i t turned down request for an exemption to solid form and DC requirements for waste since the general consideration was that the Pu must be in non-respirable form.
0 Aug 1983 EEG identifies non-compliance with DC 10 CFR 71 regulation (EEG-24).
0 Dec 1984 DOE Draft SARP claims justification for single containment.
0 July 1.9B5 EEG states justification for exemption inadequate and recommends DC design.
0 July 1.9B5 DOE establishes exemption mechanism from DC for shipments of plutonium contaminated wastes (Draft DOE Order 54B0.3. page 7. 1.0-10-B5).
that shipments of plutonium contaminated solid was.te materials be excluded from the double containment requirement. The reason cited (Ret 8) was that the provisions were inconsistent with requirements by the IAEA and the DOT. DOE recommended resolution by conformance with the IAEA.provisions.
Sandia National Laboratory also commented to the NRC on the
~roposed rule in December 1979 and urged the NRC to exempt plutonium contaminated waste materials from the double qontainment requirement or at least include in the regulations the guidelines upon which the Commission would base its determination for an exemption (Ref 9). The Sandia and DOE requests to exclude waste were subsequently rejected by the NRC in its 19a3 revision of 10 CFR 71.
As noted earlier the DOE WIPP FEIS stated that the transportation of wastes to WIPP would comply with DOT and NRC regulations. The TRUPACT design was proceeding without the double containment requirement. Since the Department of Energy has the authority to issue its own regulations on the transportation of radioactive materials that are exempt from NRC licensing. DOE issued orders in May 19 to all its staff and contractors involved with the shipment of radioactive material to meet the NRC regulations for double containment as well as all other requirements contained in NRC"s regulations 10 CFR t1.31 - 71.42 "that as p~esently set forth provide a reasonable set of technical standards" (Ref 10). There were no caveats or exemption ~echanisms identified in the DOE Orders. Thus the design was then in apparent violation of the Departments own Orders.
In August 19 a peer review of the TRUPACT preliminary design was convened by.Sandia. The peer review committee*s report {Ref 11) 10
published in June 19ff2. recognized that the design failed to meet NRC regulations and stated as follows: "The TRUPACT designers are faced with a dilemma regarding single or double containment. The regulations specify that packaging for shipments of plutonium in excess of 20 curies. with certain exemptions. must be designed for double containment. The preliminary TRUPACT design (single containment with planned application for exemption from double containment) could fulfill the regulations. provided the exemption is granted. However, if the exemption is not granted, an additional effort later in the program would be required. In assessing the various alternatives for the TRUPACT design. the issue of single versus double containment for CH-THU should be addressed in the near term to provide the necessary guidance for design purposes."
The failure to meet NRC and DOE design requirements was again recognized in the report's Executive Summary in stating "The overall design approach appears to be satisfactory except for resolution to the regulatory requirements for double containment or exemption therefrom."
The peer review stated. "Double containment for shipments of
- plutonium in excess of 20 curies per package is required in 10 CFR 71.42. with certain exceptions. The TRUPACT design strategy is to apply for exemption from this double-containment requirement. due to the low risk inherent in CH radioactive waste. It is recommended that the designers secure an early determination of this exemption from the U.S. NRC Transportation Certification Branch or else commence designing for the possibility that double containment will be required." In ,July 19B5 EEG also urged DOE to submit the design to NRC for their evaluation of an exemption (Ref 13).
11
The subsequent January 19B3 Sandia response to the peer review comments (Ref 12) ~as to ignore the double containment issue and merely state . . . TRUPACT is being designed with a single level of containme~t in the packaging ... There was no discussion of the need to obtain an exemption. Distribution of the report was limited to DOE and its contractors. other federal agencies.
selected railroads. the American Trucking Association and the American Association of Railroads with the proviso that no one was authorized to further disseminate the information without permission. EEG did not learn of the existence of either the peer review report or the Sandia response reports until 19B5 when they were referenced in the DOE draft Safety Analysis for Packaging (SARP). In July 19B3 the DOE issued a draft Order that provided the Department an exemption mechanism from the requirements of double containment (Ref 15).
an exemption was not identified.
The basis for such On August 5. 19B3. the NRC reaffirmed the need for double containment for shipments in excess of 20 curies (10 CFR 71.63.
Ref 7) and *said that the request was justified when imposed by the AEC in 1974 and the NRC considers that the need for this requirement still exists.
NRC noted in the Supplementary Information to its Federal Register promulgation that i t had received a request to exempt plutonium contaminated solid waste from the requirements for solid form and double containment or alternatively to specify the criteria that would qualify for that exemption. The Commission commented that the plutonium must be in non respirable form. exemption must be considered on a case*by case basis and that some solid waste forms undoubtedly would not qualify as being sufficiently nonrespirable. The issues are not new. EEG pointed out the failure of the TRUPACT design to conform with the NRC standard in August 19B3 (EEG-24. Ref 14) 12
In December 1984 the DOE claimed that the Draft SAHP contained justification for single containment. At a meeting with DOE devoted to TRUPAGT in May 1985, EEG stated that the justification was inadequate and on July 29, 1985 suggested several alternatives to DOE that would be acceptable to EEG (Ref 1.3) . They included:
1.. Obtain an exemption from NRG:
- 2. Redesign the TRUPAGT for double containment
- 3. Provide approved type Binner containers in the TRUPAGT.
- 4. Meet the NRG A /week release limits to the inside of 2
the TRUPAGT.
In July 1985 DOE promulgated an Order that exempts plutonium bearing wastes from the DOE ma.ndated 10 CFR 71. requirements of double containment. provided that the Office of Operational Safety of the Department approves. No basis is identified for approval despite DOE's urging 2.5 years earlier that NRC list criteria for such an exemption.
2.2.2 Bases for Exemption Mechanism Regulatory agencies generally provide mechanisms whereby exemptions can be sought from the provisions of regulations issued by those agencies. Since DOE is self-regulating i t is the responsibility of DOE's Albuquerque Operations Office (ALO) to demonstrate that an exemption should be provided. The following addresses some of the possible justification for an exemption.
2.2.2.1 Applicabil~ty of 1974 AEG Transportation Regulations to Waste Shipments: It is ger.erally agreed that the original 13
motivation for the 1973 Notice of Proposed Rulemaking of the AEC was concern for reducing hazards from accidental releases of shipments of liquid plutonium nitrate for reactor fuel by requiring solid form and extra packaging for the shipment of plutonium fuel. Therefore. one could argue that since the regulation was never intended to apply to THU waste and was not addressed in the rulemaking procedure. i t is improper to apply such regulations to the shipment of waste to WIPP. However, enclosure A of the AEC's 1974 rulemaking procedure (Ref 16) specifically noted that plutonium contaminated waste would not be included in the list of exempted materials but would be considered for possible exemption on a case by case basis. NRC reaffirmed this position in 19B3. Hence, the regulations were intended to apply to waste.
2.2.2.2 Respirability: The double containment requirements were established to take into account that the plutonium may not be in a "nonrespirable" form. However, the DOE WIPP Waste Acceptance Criteria permits up to 1% of the waste (by weight) to be respirable. Thus, shipment of 36 drums of average plutonium concentration (Ref 1) could have 1.5 to 2 Ci of plutonium in respirable form present in the THUPACT even if the plutonium concentration was not enriched in th~ respirable particles.
Some heat source plutonium shipments would.exceed that amount in each drum. Also, the wastes average 25% combustible material and are constantly undergoing radfolytic decomposition (see Chapter 3). EEG does not believe i t is prudent to consider such wastes as stable and non respirable.
2.2.2.3 Comparison with Shipments of Spent Fuel: Shipments of spent fuel do not require double containment. If one. could show that the inhalation hazard from the release of THU wastes following accidents were equal to or lower t.han risks following 14
the spent fuel following spent fuel accidents. an argument could be made for the adequacy of single containment. This issue is addressed later in the report.
2.2.2.4 Amount of Plutonium per Sh.ipment: One might argue that the amount of plutonium in a shipment of waste is a small quantity in comparison to fuel shipments. AEC defined a large shipment of plutonium as 20 curies or more. Tlle average CH waste shipment identified in Ref 1. is 1.50 Ci plutonium per shipment. Hence. shipments to WIPP wer.e always consio.E:?red to be large shipments .
- Under the new DOT terminology (Hef 4). "highway route controlled quantities" apply to shipments of more then 6 curies Pu-239.
curies Pu-238. and 24 curies ~--24.1. All. shipments to WIPP 9
would be included under this definition.
2.3 Possible Risks & Consequences The purpose of pac~aging certification is to insure tpat packages carrying radioactive materi~ls will have sufficient integrity so that the radiological implications of releases from rough handling and severe ~ccidents will be acceptable. Since the quantity and relative toxicity of a container's contents directly effect the consequences of an accident. the requirements of a package increase when more radioactive or more toxic raclionuclides are in the container. However. the procedure for d.etermining an acceptabl.e package design. while based partially on test data or analyses. also involves qualitative considerations and engineering judgment.
The double containment requirement was set in a qualita.t.ive manner as being practical or reasonable without quantitative
.15
determinations being made of the increments of safety being obtained and the cost of attaining the increments from designs of varying stringency. A qualitative approach also makes the determination of what is "equivalent" to the required design a subjective one. EEG believes that a serious effort should be made to quantify the incremental health and safety benefits that might be obtained from more stringent designs. This quantification may not be conclusive but will be attempted below.
2.3.1 Radiological Considerations - Incident Free Transportation Occupational workers who load. unload. and transport the contact handled transuranic (CH-TRU) waste will receive radiation doses from WIPP related transportation in the TRUPACT. Doses can occur from external radiation during routine handling and transportation and from releases during accidents. Internal doses could occur from resuspension of surface contamination or from releases of radioactive material following failures in the Type A packaging and the Type B TRUPACT. Releases from the TRUPACT would probably occur only following a severe accident .
The most probable (and largest) internal radiation doses would occur from inhalation of respirable sized particles. although, ingestion of particles through water or food is also possible.
2.3.1.1 External Radiation: There will be radiation doses received by persons along the routes to WIPP from accident-free transportation. These doses are not projected to be large to any individual nor to the total population (Ref 2). but since they have a 100% probability of occurrence they represent virtually all of the expected population dose.
16
The transuranic elements emit very little gamma radiation and all. of the emissions are low energy. The most predominant gamma 241 ray is the 0.06 Mev x-ray from the decay of Am. This "soft"
..... x-ray is relatively easily attenuated. with <1.% being transmitted through the walls of TRUPACT. Most of the remaining THU radionuclides have x-rays of 0.1. Mev or greater that occur 4
with a frequency of <10- per disintegration. However. since 241 many shipments will contain very little Am and a 0.1. Mev gamma ray is much less attenuated in the TRUPACT wall i t is considered conservative and prudent to assume this higher energy in shielding calculations .
- There are minor amounts of fission and activation products present in the THU waste inventory that emit higher energy gamma radiation. F"or example. in the wastes stored at INEL there are an estimated 6.2 Ci of 60 co. 6.1. Ci of 137 cs. 56 Ci of mixed activation products and 130 Ci of mixed fission products. There 232 are also about 20 Ci of gamma emissions from the decay of u 233 and u in the waste (Ref 1.8). Although these radionuclides comprise less than 0.1.% of the total amount of radioactivity stored at INEL. an evaluation by EEG indicated that about 15% of the radiation escaping the TRUPACT would be due to these higher energy gamma radiations. These higher energy gamma radiations will be ignored in the following analysis because the assumption of 0.1 Mev photons is believed to add adequate conservatism.
2.3.1..2 Design Effect on Radiation Level: The final design of the TRUPACT will have a substantial effect on the amount of radiation that is attenuated within the TRUPACT and its walls.
Three factors influence this: (1) the density of material 2
(g/cm) within the packages: (2) the g/cm 2 of material in the TRUPACT walls: and (3) the specific materials present in the shielding. A doubly conta~ned TRUPACT would have a greater mass 17
between the waste and e*xterior and thus a. reduced external dose rate. .The possible effect of .double cont.a.i.nment on the average external ra.di.atio*n level is discussed below ..
Th~ DOE has stated infornial,ly that double containment m+/-ght add about 4. 000 pounds .to
- the weight e>f
- the TRUPACT.
- A 4. 000 pound inner type B steel container would have a weight of about 4.4 .J 2
gm//i:::m and would reduce the average exte_rior radiation level to 60% or less of the level with the present design. Even if double containment limits the number of drums in the TRUPACT to the present 2 wide configuration this would result in a raqiation level per drum 10% less than would result from the 3 dr~m wide configuration that is planned with the 48-drum TRUPACT-II design. (The average INEL drum would have 99+% of the radiation coming from the fxrst row of drums and even a conservative. low-density load would deliver about 93% of the dose from the first row).* The difference in dose rate between a 2 drum wide doubly contained TRUPACT and a 3 drum wide singly contained TRUPACT would be greater than 10% if the mass of Kevlar and steel in the walls is reduced in the TRUPACT-II design.
The population radiation dose delivered by TRUPACT transportation (assuming 100% by truck) to WIPP has been estimated to be 3.3 person-rem/yin New Mexico (Ref 2). The collective dose to people in other states was not estimated but from mileage extrapola~ions would be about 5.4 person-rem.
These annual doses are based on a shipment rate of 318.000 3
ft /y. The total population dose. in-state and out. estimated to be delivered during the repository lifetime (6.2 million cubic feet of waste) would be about 170 person-rem. Thus double containment would result in a dose reduction ranging from 17 to 67 'person-rem. depending on whether the TRUPACT dimensions are altered.
18
2.3.1..3 Occupational Radiation Exposure: Persons involved in the loading. unloading and transporting of wast.es in the TRUPACT will rece~ve so~e external radiation dose from .the packages they are handling and internal doses from inhaling air containing resus~end~q ~ont~mination. An evaluation was made of the t.
effects of vari_ous 'l'RUPACT designs on t.he annual occupational radiation, dose.
The backg,round do~ument used in this evaluation was "PreJ_iminary Raqiatio~ Dose Ass~ssment to WIPP Waste H~ndling Personnel".
WTSD-~-00,Q. February 198,5 (Ref 1.9)
- This report included a ste1;>-bY,-step time and m.otion study of all operations involved in rece.i v ~'?,g an.d. unload.i,ng a loaded TRUPAC"F. and shipping out the empty 'I'.~UPACT. Although the report is considered preliminary an~. has 9ot be~n critically re~iewed by EEG we believe i t is thorough en_oug__h to be us.ed as the basis to estimate the effect of di~f~rent designs on occupational radiation doses.
A number of adp:iti_,ona.1 assumptions were necessary in order to compare esti_m,ated! doses from different designs. A 4,000 pound inner liner was assumed for the double-contained design. It was also a~sumed* that a: 48,druw TRUPACT design would' have the same
- wal_l, t,hick;nes~ as the T:RUPACT!-r design and that weight limits would. stil __l J)ermit al.l 4f:}-dru!}1 TRUPACTS to carry a full load.
Both of these assun;iptions are non-conservative. i.e .* they would lead, to lpwer estimated;occupat~onal doses than the most likely dose. T.h,e f,ol,lpwing,: detaj..led assumptions had, to be made on each 1.,. Wh.eth,er the time re,qµi;i::eq: f.or. the workers to do the su,b,-task is, depen,dent on the number of* TRUPACTS received or the number of six-packs handled:
- 2. Whether. the. shielding effect of the TRUPACT walJ_s would be a factor:
-19
- 3. Wheth~r the exposure from the door was to a 2 x 2 drum stack (36-drum design) or to a 3 x 2 drum stack (4B-dru~ design). ~
It was decitjed to assume the time required to unbolt and bolt the inner TRUPACT door was the same per TRUPACT in all designs and that the dose for releasing tie-downs and removing dunnage was per six-pack handled. The average air concentration used in determining internal doses was taken from Table 6.2-4 in the WIPP Safety Analysis Report (Ref 20). The results are shown in Table 2.
The conclusion to be drawn ~rom Table 2 is that a double contained 36-drum TRUPACT will result in a lower occupational radiation dose than either the current design or a 4B-drum TRUPACT design. For example. a 4B-drum TRUPACT would expect to deliver an additional dose of about about 22 person-rem over the 20-year project lifetime compared to a 36-drum. double contained design.
2.3.2 Radiological Considerations - Accidents 2.3.2.1 Fractional Releases From Accidents: Projections have been made in the Preliminary Transportation Analysis the expected frequency and severity of accidents that could cause releases from the TRUPACT and the fraction of (PTA) of
- radionuclides released for each accident severity category (Ref 2). These release fractions are compared with those estimated in two other documents in Table 3.
Both NUREG-0170 and the RADTRAN II User Guides predict a considerably greater release (factors of 2 to 500) for severity categories VI-VIII than the PTA. but bracket the PTA numbers for 20
TABLE 2 ESTIMATED OCCUPATIONAL RADIATION DOSES FROM LOADING & *UNLOADING VARIOUS TRUPACT DESIGNS (Person-Rem/Year)
Radiation Doses Design Externa1 Internal(a) Total Incr~men t_al (c)
TRUPACT- 9.8 1.. 3 1.1. .1. + 1.. 2 36 Drums with D.C. 8.6 1.. 3 9.9 36 Drums with D~C. + 1.0(b) 8.9 1.. 4 1.0.3 + 0.4*
48 Drums without D.C. 1.0. 1.. 0 1.1.. 0 + 1. .1.
(a) 50-year effective dose equivalent.
(b) assumes 1.0% more TRUPACT shipments are necessary.
(c) with respect to double contained design.
TABLE 3 FRACTION OF RADIONUCLIDES RELEASED AS RESPIRABLE AEROSOLS FROM TRANSPORTATION ACCIDENTS Severity Category Documents Where Estimated PTA (a) NUREG-01.70(b) RADTRAN(c)
I 0 0 0 II 0 0 0 III 5-9d 0 2.5-5 IV 5-B 0 2.5-4 V 5-7 0 2.5-3 VI 5-6 1.-4 2.5-3 VII *5-5 5-4 2.5-3, VIII 5-4 i-3 2.5-3 (a) Reference 2. *
(b) Reference 21. (Table 5-B for 1.975 plutonium shipments).
(c) Reference 22 (page 71.. large loose powder in Type B c:ontainer). *_
9 (d) 5-9 = 5 X 1.0 .
categories III-V. It"s not obvious which of these sets of assumptions is more realistic. The NUREG-01.70 values are based on tests at Sandia NaCional Laboratory of containers commonly used to ship plutonium in the mid-1.970 period. The bases for the release values in the RADTRAN II User Guide were not referenced. An earlier (1.983) draft version of the PTA (Ref 23) explained the basis of the PTA release fractions as a footnote to Table D-3: "These data are based on design basis criteria for the TRUPACT and the *projections in Reference 1. * [NUREG-01.70]
for typical packages put into service after 1.985. The projected performance of the TRUPACT is several orders of magnitude better than indicated in this table". The predicted 1.985 releases in NUREG-01.70 were: zero for category I-VI: 1.0 -4 for Category VII:
-3 '
and 10 for category VIII.
22*
The test data used in NUREG-0170 probably was not for containers similar to the TRUPACT. The other references provide fewer details. Besides the uncertainty of container design. only a fraction of the WIPP waste form (which is very heterogeneous) fits any of the waste categories assumed in NUREG-0170 or the RADTHAN II User Guide. Since there were release tests conducted with the full-scale testing of TRUPACT Unit- . i t is well to consider how these compare with the above estimates.
TRUPACT Unit-0 was tested after being loaded with 36 drums simulating various waste forms that will be shipped to WIPP.
Each type of drum was tagged with a unique tracer so that releases from each waste form could be estimated. The observed release fractions (from the drums to the inside of the TRUPACT cavity) from the full series of hypothetical accident tests 3 -6 averaged 1.25 X 1 - for total particles and 2.40x10 for aerosolized and respirable particles. The total fractional
-6 release ranged from 3.3 X 10 for soft wastes on top drums away from the door to 6.7 X 1 -
3 for hard wastes on bottom drums near the door (Ref 24).
The 9 meter drop test is considered to be at the lower limit of Severity Category III (Ref 21). However. the total fractional releases mentioned above are the sum from all tests (a 0.3m drop. two 9m drops. four 1m puncture tests. and a thermal test)
,, which is probably equivalent to a single test of higher severity category. Release fractions quoted above are to the inside of the TRUPACT while estimates of accident consequences are based on releases to the environment. No attempt was made to measure the quantity of tracer that was released from the TRUPACT, but since Unit-O had both door seals and filters fail as a result of the thermal test. i t is pos~ible that the loss was about 30% of 23
the amount that was aerosolized and respirable. This would be a 7
fractional release of about 7 X 1.0- . which is equivalent to a severity category greater than V with the PTA and NUREG-01.70, and greater than II with RADTRAN II assumptions. From the above considerations. i t is considered reasonable to assume that the present modified TRUPACT design. which passed the 1986 thermal test without loss of door seal or filter integrity, will have the release fractions estimated in the PTA up through a severity category VI accident and will have the release fractions 4
estimated in NUREG-01.70 for category VII (5 x 1.0- ) and the
-3 RADTRAN II value for category VIII (2.5 x 1.0 ) . The reason for estimating higher release values for categories VII and VIII is based on the design with vents which could release more radio-activity in a more severe accident.
VI, 1.0
-4 '
for category VII, and 1.0
-3 A doubly contained TRUPACT would be assumed to conform to the NUREG-01.70 estimate for a 1.985 plutonium shipping container (i.e., zero through category for category VIII).
Particulates greater than 1.0 µm also need to be considered because, if released, they would contaminate the environment and require clean-up. It is expected that the mass of particulates associated with particles >~0 µm will be much larger than the m~ss associated with <1.0 µm. For example, the Waste Acceptance Criteria permits 1.5% of the waste to be particle sizes <200µm.
Al.so, in the Unit-0 tests the total release from drums was about 525 times the mass of <1.0 µm particles suspended in the TRUPACT cavity. For accidents with Severity Categories I-VI, i t is assumed that no particulates >1.0 µm will be released from the TRUPACT. This is based on the hypothesis that leakage paths through the filters and seals would be small enough so that larger particles would be discriminated against (as observed with the aerosol sampling train used during the Unit-0 full scale tests). However. in very severe accidents. there could be 24
a major failure of filters and/or door seals and possibly the release of contaminated particles generated by fire. A sampling indicated that the average ratio of particulate mass <210 µm to that <10 µm was 138 (Ref 25). It will be arbitrarily assumed that with the present TRUPACT design the ratio of <210 µm mass released from the TRUPACT will be 3 times the <10 µm mass for a Category VII accident and 6 times for a category VIII Accident.
For a doubly contained TRUPACT the ratio is assumed to be 3 times the <10µm mass for a category VIII accident.
2.3.2.2 Expected Number of Accidents: The PTA uses New Mexico State data on frequency of truck accidents per kilometer on specific routes and national data for rail accidents. The accident frequency rate and the number of kilometers per year traveled in New Mexico is then used in the RADTRAN II Model which incorporates the fraction of accidents in each severity category and the related fractional releases with meteorological and dose conversion data to calculate population doses from accidents. The model does not directly calculate the expected accident frequency in each severity category or the releases (and consequences) resulting from individual accidents. The expected annual and total number of truck accidents for all states in each severity category are shown below in Table 4.
Rail accidents will not be tabulated because current expectation is that only a small percentage of shipments will actually be made by rail. Also. calculations indicate that releases per TRUPACT shipment by truck will be slightly greater than releases from rail shipments.
The projections in Table 4 indicate that if the TRUPACT releases some radionuclides for all accidents of severity categories>
III. there will be more than 12 accidents with the release of radioactive materials during the lifetime of the WIPP Project 25
TABLE 4 EXPECTED NUMBER OF ACCIDENTS INVOLVING RADIONUCLIDE RELEASES DURING TRUCK TRANSPORTATION TO WIPP
'I'. '
Severity Category Per Year Lifetime Total Lifetime. Urban(a)
III 0.49 9.6 5.5 IV 0 . 1.1. 2.2 1.. 3 V 0.023 0.44 0.1.7 VI 0.01.0 0.1.9 0.043 VII 0.000B5 0.01.7 0.0032 VIII 0.0001.7 0.0032 0.00057 TOTAL 0,63 1.2. 7.0 (a) Includes both urban and suburban accidents.
and 7 of these would be in urban or suburban areas. If integrity could be maintained for all accidents with severi.ty category~ VI there would be only 0.02 accidents involving releases.
2.3.2.3 Radionuclide Releases from Accidents: The number of accidents with releases and the release fraction in each severity category can be combined with an average TRUPACT load to obtain the quantity of radioactivity expected to be released during the WIPP lifetime. The average number of curies per TRUPACT load to be shipped from each generating site and the overall average was derived from data in Reference 26. It was necessary to make several assumptions in deriving these averages since all data were not internally consistent. The results are summarized in Table 5.
26
TABLE 5 AVEBAGE AMCUNI' OF" HADIOACrIVTTY BEilG TRANSRRI"F.D TO WIFP (Curies of Alpha Radiation)
GenP..rating Presently St.ored Waste Newly Generated Waste Average Ci
... Site Volume 3
(m ) Cur.ies Volume Curies per T.FUPACT Hanfor:tl 1.3.700 44-.600 21.400 42.300 1.7 .1.
INEL-RFP 35.700 205.000 74.300 247.000 30.B LANL 6.180 1.51..000 6.070 1.52.000 1.85.
CP.NL 490 21..800 545 5.050 1.94.
(.
SHP 3.900 597.000 1.0.600 2.030.000 1.360.
(a)
TCJI'ALS 60.000 1..020.000 1.1.6.000 2.480.000 1.49.
(a)
The average mi.leage wei.ghted load is 1.84 Ci/TRUPACT -
mile.
'fhe quantity of radionuclides associated with respirable sized materials (<1.0 µm) that might be released from different severity accidents for the average. the average SRP. and the maximum permitted shipments and the expected quant.i.ties that would be released during the operating lifetime of WIPP (considering probabilities) are shown in the following TabJ.e.
As mentioned above. the amount of radionuclides associated with particles >.1.0 µmare be assumed to be 2 times the respirable fraction for a Severity Class VII accident and 5 times for a Severity Class VIII accident.
From Table 6 i t is estimated double containment would reduce the expected quantity of radionuclides released from accidents to lB% of that with the current design. Also the doubly contained design would limit the curies released in the class VIII accident to 40% of that with the current design. This would be 27
TABLE 6 qJANITTIES OF RADICNUCLIDES nELEASED TO 'IHE ENVJJU~fENI' FH.:M TRUCK ACCIDENTS (millicuries. particles .S 10 ;an MAD)
Severity Total No. Present Design Releases Doubly Contained Releases Class Releases Avg/Ace SRP/Acc Ma..,x/ Ace Expected Avg/Ace SRP/ Ace Yiax/ Ace_ &']:JeCted III 9.6+0 9.3-4(a) 6.B-3 5.3-2 B.9-3 IV 2.2+0 9.3---3 6.8.2 5.3-1 2.1-2 V 4.4-1 9.3-2 6.B-1 5.3+0 4.1-2 N
Ct:J VI 1.9-1 9.3-1 6.B+O 5.3+1 1.9-.1 VII 1.7-2 9.3+1 6.8+2 5.3+3 1.5+0 1.9+1 1.4+2 1.1+3 3.1-t VIII 3.2-3 4.6+2 3.4+3 2.7+4 1.5+0 1.9+2 1.4+3 1.1+4 6.0-1 TOTAL 1.2+1 3.3+0 9.1-1 (a) 9.3-4 = 9.3X1- 4
a reduction in respirable sized particles from 3.4 Ci to 1.4 Ci for the average SRP waste shipment and from 27 Ci to 11 Ci for the maximum proposed load of 10.700 alpha curies.
2.3.2.4 Radiation Doses From Accidental Releases: Several different types of radiation doses are important and will be estimated. These are:
- 1. Population doses from the amounts of radiation expected to be released during the operating lifetime of WIPP:
- 2. Population doses from the more severe accidents that have a low probability of occurrence: and
- 3. Maximum individual doses '(50 year dose commitment and first year dos~) from more severe accidents. Also, the possible health effects from these accidents will be estimated.
The estlrnated average and minimum atmospheric dispersion values (X/Q) in the 22.5 degree downwind sector were taken from Table
- 33. Appendix Hof the WIPP Final Environmental Impact Statement (Ref 1). Key assumptions inciuded population densities of 619 2 2 persons/km in suburban and urban areas and 2 .persons/km in r~ral areas. releases occurring over a one hour period. and an 3
individual breathing rate of 1.2 m /hr.
Table 7 indicates that about 23 person-rem are expected from accidental releases from the TRUPACT. This is <14% of the expected external rad:iati.on dose to the population along the routes during normal operations and <11% of the expected occupational doses from loading and unloading the TRUPACT.
Aiso, the calculated decrease +/-n expected dose due to double containment is 17 person-rem.
29
TABLE 7 BADIATION POPULATION DOSES FROM EXPECTED WIPP THANSPOBTATION ACCIDENT RELEASES (50-Year Dose Commitment In Person-Rem)
ORGAN DOSE Effect.ive Condition Dose Equiv. Lung Bone Present Design urban/suburban 21. 75. 220.
rural 2.1 _ 7. 5 22.
total 23. 83. 240.
Double Con ta.inmen t urban/suburban 5.5 20. 57.
rural 0.6 2.1 6. :.'\
total 6 .1. 22. 63.
The numbers in Table B show that substantial population and individual dose commitments could result from a Severity Category VII or VIII accident in an urban area. The probability of one of these accidents occurring during the lifetime of the project ~s about U.4%. which is low. but certainly not incredible. F*urthermore. the maximum individual doses are signi.ficant ror an average SBP load or for the maximum permitted loa.d. For a 48-drum TRUPACT design, the average doses to populations and individuals would be 33% higher than the numbers in Table B.
2.3.2.5 Radiological Contamination: A significant radionuclide rel.ease from a TRUPACT accident would result in considerable environmental contamination. Contamination beyond a permissible 30
J *
- TABLE 8 ESTIMATED POPULATION & INDIVIDUAL RADIATION DOSES FROM SEVERE TRANSPORTATION ACCIDENTS (50-Year Effective Dose Commitment in Person-Rem and Rem) (e)
TRUPACT LOADING CONDITION Dose Recipient Average Average SRP Maximum VII (a) VIII (a) VII VIII VII VIII Present Design w
..... Population ( ) 3.1.+3 1..6+4 .2.3+4 1.. 1. +5 1..8+5 8.9+5 Max Individ c 2.7+0 1..3+1. 1..9+1. 9.7+1. 1..5+2 7.6+2 Double Containment*
Population 6.1.+2 6.1.+3 4.5+3 4.5+4 3.6+4 3.6+5 Max. Individ 5.3-1. 5.3+0 3.9+0 3.9+1. 3.0+1. 3.1.+2 (b)
Probability Annual ~-7-4(d) 2.9-5 1..9-5 3.4-6 Total 3.2-3 5.7-4 3.7-4 6.6-5 (a) Severity category VII and VIII accidents (b) Probabilities are for occurrences in urban & suburban area*s (c) 5% of the time ma~tmum doses would be double the values (d) 1.07-4 = 1.7 X 1.0 (e) The fraction of the 50-year effective dose commitment delivered in the first year is 0.1.0.
level would have to be decontaminated or the area would have to be quarantined for certain uses. The required remedial action could be very expensive especially if the action occurs in an urban or suburban area. Also. all radionuclide releases. not just those that are qssociated with respirable sized particles.
will contribute to the level of contamination.
No attempt will be made here to specifically estimate the cost of cleanup that might be typical along WIPP routes. However.
from the curve in Figure 3-2 of the Urban Study (Ref 27) and adjusting for New Mexico urban/suburban density (about .043 of.Manhattan's density) and for 1986 prices (about 1.5 times 1979 prices) i t is estimated the cost would be about $16 million for a category VIII accident for an average SRP load without double containment and $5 million with double containment. So double containment would result in significant economic savings from*a very severe accident.
An additional advantage of double containment would be the drastic reduction (from 12.5 to 0.02) in the expected number of release accidents during the WIPP campaign (see Table 4). While most of these additional accidents would be small and not involve significant cleanup costs they would require monitoring costs and a great deal of public explanation.
2.3.3 Radiological Health Effects from Transportation The relationship between the amount of radiation received and the expected health effects has been studied extensively by national and international organizations as well as by individuals. Correlations between dose and effect involve a number of variables including type of radiation. organ being irradiated. age at time dose is delivered. sex of the person 32
receiving the dose, and in some cases the rate at which the dose is delivered. The conversion factors determined by different investigators vary considerably and in many cases a range is reported rather than a single number. This report will use a range of 100-250 latent cancer fatalities (LCF) per million person-rem of 50-year effective dose equivalent and external whole body radiation. This range encompasses the values used in the 19B0 BEIH report (Ref 17) and the suggested values in the HADTHAN II code. Other health effects. such as genetic and life-span shortening will not be estimated here .
Tables 9 and 10 indicate that a double contained THUPACT is expected to result in fewer latent cancer fatalities than either the present design or a 48-drum design both from routine transportation and from releases following severe accidents.
However, the expected LCF are low in all cases and the differences between designs are not enough to justify one design over others.
The justification for double containment rather than single containment is based on the increased safety in case of accidents. The drastic reduction in the expected number of accidents with radionuclide releases will significantly reduce costs of monitoring. quarantine and decontamination and have a positive benefit on public perception of transportation safety.
As shown in Table 10. the decrease in estimated latent cancer fatalities due to double containment is substantial for Class VII and VIII accidents. We believe the additional protection against low (0.1-1.0%) probability of accidents that can be obtained by double containment already warrants its incorporation into the design of the TRUPACT.
33
TABLE 9 EXPECTED LATENT CANCER FATALITIES FROM TRANSPORTATION TO WIPP Incident Free Transeortation TRUPACT MODEL Population Occupational Expected Accidents Total LCF Present Design .01.7-.042 .022-.054 .002-.006 .041.-.10 Double Contain .010-.025 .019-.048 .001-.002 .030-.075 D. Contain + 10% .010-.025 .020-.050 .001.-.002 .031.-.077 48 Drum .011.-.028 .022-.054 .002-.006 .035-.088 TABLE 10 ESTIMATED LATENT CANCER FATALITIES FROM SEVERE TRANSPORTATION ACCIDENTS Average Load Average SRP Load TRUPACT Model Class VII Class VIII Class VII Class VIII Present Design 0.31.-0.77 1.5-3.8 2.3-5.6 1.1.-28.
Double Contained 0.06-0.15 0.61.-_t.5 0. 45-1.1 4. 5-1.1.
'1:8 Drum 0.41.-1.0 2.0-5.0 3.0-7.5 15.-33.
Non-Radiological Risks The transportation of material by truck or train also involves risks unrelated to the nature of the cargo. The principal risks come from vehicle accidents that cause injuries and deaths.
There also are latent cancer fatality deaths that would be expected from motor vehicle emissions. Non-radiological unit risk factors presented in SAND B3-0B67 (Ref 2B) are used in Table 11 to estimate non-radiological risks from shipment of CH-THU to WIPP by truck.
Table 11 lists expected non-radiological fatalities from truck shipments that are about two orders-of-magnitude greater than the expected Latent Cancer Fatalities from radiation exposure.
This could lead one to contend that non-radiological safety is a more important concern in package and system development than is radiological safety.
TABLE 1.1 NON-RADIOLOGICAL FATALITIES EXPECTED FROM SHIPMENT OF CH-THU WASTES TO WIPP BY THUCK Total Round Jrip Latent Area distance (10 km) Fatalities Injuries Cancer Fatalities Rural 6B. 4.6 56.
Suburban 3.B .06 1.4 Urban O.B .00B 0.4 .08 Totals 73. 4.7 5B. .OB It should be noted that the high non-radiological to radiological fatality ratios estimated for 100% truck shipments 35
to WIPP are not estimated for rail shipments. There are several Leasons for this difference:
1.. Fatal accident rates per kilometer for trucks average about 3-1/2 times those for a railcar:
- 2. A railcar will hold 2 TRUPACTS, therefore only half as many shipments are required:
- 3. Rail shipments move at a much slower average speed, partially because an average train shipment is stopped most of the time. This increases the routine radiation dose to the public along the route. Using the assumptions in Referenc:e 28 for all wastes that could be physically shipped to WIPP by rail leads to the prediction that there would be about 1.0 accidental deaths, 0.1 non-radiological latent cancer fatalities, and 0.8 latent cancer fatalities from incident free radiation exposure.
2.3.5. Trading Off Radiological and Non-Radiological Risks In prepared testimony to the NM Radioactive Materials Legislative Committee on September 25, 1985, Joint Integration Office
- DOE, (JIO) the Di_rector of the Albuquerque Operations Office.
stated that an appropriate justification for using a TRUPACT design that contains only single containment is that the number of J_ives that could be saved from non-radiological risks would greatly exceed the expected increase in radiological deaths. Two aspects of this argument need to be evaluated:
- 1. Is the contention factually correct?
36
- 2. Is i t appropriate to trade-off radiological and non-radiological health and safety risks?
~~ese two issues are discussed separately below.
2.3.5.1 Projected radiological and non-radiological risks: The analysis above indicates that for 100% truck shipments the expected non-radiological deaths are about two orders of magnitude greater than the expected radiological deaths.
Therefore. for this condition i t seems reasonable to expect that the possibilities of reducing total deaths by changes in the transportation system would be most likely in the non-radiological area. JIO has contended that non-radiological deaths are directly related to vehicle miles and that since double containment would reduce the payload. require more shipments. and increase vehicJ.e miles. i t would result in more total deaths.
Many steps can be taken to reduce death per vehicle mile (e.g .*
better driver training. more rigid safety checks of vehicles.
routing and timing of transportation). However. these steps could (and should) be applied rigidly to whatever transportation system is chosen. Consequently. we agree that total vehicle-miles is still the most appropriate index to estimate non-radiological deaths.
For transportation by rail the radiological and non-radiological risks are similar (the above estimate gives a non-radiological to radiological risk of about 1.4. which is probably within the error of the estimate) and the minimization of total risk would require consideration of both types of risk. Also. truck shipments are expected to result in 2.7 times the total deaths as rail shipments. This 1c.**.1.ggests that the most efficient action 37
that might be taken to save lives would be to ship all wastes by rail. if rail access is available. Present information from JIO is that most shipments are expected to be by truck.
Double containment would result in extra vehicle miles if the change in design reduced the number of drums or boxes that could be carried or if the extra weight of the TRUPACT required a decrease in the number of containers per shipment. The ,JIO indicates that double containment would result in a 30% increase in vehicle-miles. No analysis has been presented to justify this figure. though i t is believed to be simply the ratio of the net payload in the present TRUPACT (1B.200 pounds) to that which might exist with double conta.inment. There are two reasons why this figure is probably too high:
- 1. From limited available data (197B data from INEL only - Ref 29) i t appears that most shipments will not be weight limited. The average weight of drums would amount to only 11,900 pounds per TRUPACT and a load of 2 Rocky Flats boxes would average only 5.600 pounds.
If a large number of drums were processed At Idaho National Laboratory (INEL) in the Process Experimental Pilot Plant (PREPP) loads could become weight limited since these drums weigh about 1.200 pounds each.
Extensive processing would also drastically effect the efficiency DOE believes could be attained with a 48-drum TRUPACT design.
- 2. Preliminary data suggest that. with proper load
-management. a large number of TRUPACT loads would not exceed 20 curies of plutonium and could be shipped with single containment (Ref 29).
3B
If all shipments to WIPP are by truck in a doubly contained TRUPACT. we estimate an increase in the expected non-radiological deaths by 5-10% and this increase would be greater than any expected decrease in radiological deaths. The estimated non-r~diological deaths would increase by 0.48 from 10% greater-mileage and radioiogical deaths would decrease by 0.02. However, i{ the i~tent is to minimize total expected non-radiological deaths. the WIPP Project Office (WPO) should ship all wastes by rail from those storage or ~eneration sites that have ~ail access. Maximizing rail shipments would save an expected 2.$ lives.
2.3.5.2. Is trading off appropriate? The concept of balancing act:ivit:ies i~vplvi~g Fadiation fisks so t~at the total expected health and saf~tY effects from both radiological and non-
~owever, we do not believe t~is "trade-oft" apppoac~ ~as eve~ been u~e~ i~ setting
' ' ' ,* . ' * *
- I stapdarq~. wr~t:~P~ regu+at~ons. or in making radiation prot:ec,...ion anq. waste managernen~ qecisions. Furthermore. i t appears that even in transportation of CH-THU wastes to WIPP
, ' *
- I * *, , ,,. I! : ! , , ; . ', I *', ; * ,, , , ' ; ,
this p~iJosop~y is' not b~ipg .
appli~tj consistently.
~ f i t were.
all possil:>le shiprnenr,s would pe by rail. The principal philosophy , .,
behind ' .
radiation p~otection regulations
. . and decision-
'. ~ " . ,
makinQ appears to be twofold:
- 1. To be certain that expectetj radiation doses to individuals~~~ populat~ons meet stan4ards that have been ~evelope9:
- 2. To offer atjditional protection agains~ the higher conseq4ence - lower propabili.ty accide;nt. These high conseque~ce effects are hidden when they are co~bined with probabili~y and presented only as expected doses.
39
DOE did not use the least expected fatality concept for decision-making in either the WIPP Final Environmental Impact Statement or in the various draft Environmental Assessments for the first repository candidate sites (Ref 30).
Appendix N of the WIPP FEIS concludes that leaving presently stored wastes at INEL would result in no expected health effects. would cost only $600.000 per year. and would have a danger of latent cancer fatalities from three low-probability scenarios. These are shown in Table 12.
Table 12 POSSIBLE LATENT CANCER FATALITIES FROM LEAVING STORED WASTE AT INEL Scenario LCF Comments Explosive Volcano 0.48 - 4.4 Volcanic Lava Flow 2.4 - 22. Dose cornmitment calculations for this scenario subject to large uncertain-ties.
Human Intrusion 0.04-0.38 Greater confinement disposal at INEL was estimated in Appendix N to reduce these possible LCFs by a factor of one hundred for a capital cost of 1.9 to 21 million dollars and a $600.000/year surveillance cost.
40
~~e FEIS did not compare these low probability LCFs with the expected and low probability deaths that might result from constructing and opera ting the WIPP site. The .low probability LCFs in Table 12 can be compared with those from Class VII or VIII accidents in Table 10 and one can speculate on the relative probabilities of the various scenarios. There clearly will be expected deaths from WIPP construction, transportation and operation. These are estimated in Table 13.
Table 13 indicates that the expected fatalities that will occur from shipping INEL and RFP wastes to WIPP will be 4.7 if all shipments are by truck and 3.0 if rail shipments are optimized .
- Thus, the decision to dispose of INEL & HFP TRU wastes at WIPP traded off 3.0 expected deaths from non-radiological causes in order to preveht several low probability events from occurring.
This trade-off also involved the expenditure of over $0,5 bLllion more than would have been necessary to monitor the wastes at INEL and introduced the possibility of low probability transportation accidents. The DOE's current plans to ship all wastes by truck. would result in an additional 1.B expected deaths .
- We conclude that the original decision to build the WIPP Project was made because.of the desire to protect against low probability radiological doses and environmental contamination and did bot consider minimizing either non-radiological deaths or costs. Furthermore. the DOE_claim that double containment is undesirable because of the extra highway deaths that would occur is inconsistent with plans to ship 100% by truck and thereby increase the expected deaths by about 6 times that due to double conta.inment.
4.1
TABLl<: 13 ESTIMATED DEATHS EXPECTED TO OCCUR FROM THE WIPP PROJECT Source of Expected Deaths Comments Death Total INEL&-RFP At Site(a)
Radiar_ion 0.03 . 01. non-TRUPACT related occupational exposure o.2o 6 Construction assume 4x1.0 person-hours 6
Surface Op. 0.48 assume 4x1.0 person-hours Underground Op. 2.00 6 assume 2x1.0 person-hours Other Employees 0.24 6 assume 1.0xlO person-hours Total Site 3.0 1.. B Transportation AJ.l truck rad .OB .04 non--rad total 4.B 2.9 Total 4.9 2.9 Max Rail rad .80 .50 non-rad 1.. 0 .61:1 Total .1.B 1.2 (a) Estimates of deaths per person-hour taken from pages 4-45 and 5-29. Reference 30.
(b) One fatality has already occurred.
42
The DOE Office of Civilian Radioactive Waste Management (OCRWM) estimated the costs and the radiological and non-radiological risks of transporting high-level wastes to the various proposed repository sites. These differences are substantial. e.g.,
shipment to Richton Dome by rail was $0.98 billion and 12.3 deaths less than shipment to Hanford. Truck shipments are estimated to cause about 3.3 times the total deaths as rail shipments. Yet under the grouping of environment.
socioeconomics. and transportation Hanford was ranked first and Richton fourth. It appears that OCRWM does not consider either cost or expected deaths from transportation to be a very important criteria in repository siting. However. OCRWM"s present preference is toward rail shipment even though costs are similar to truck (from +14% at Richton to -14% at Hanford). So, unlike the WIPP Project, OCRWM is favoring the transportation mode that results in the least deaths.
EEG concludes that using the trade-off of expected non-radiological deaths with expected radiological deaths has little or no precedent in waste management decisions and has not been applied elsewhere in the WIPP project. even in the transportation area. We believe invoking this principle to
- argue for an exemption to double containment is inconsistent with prior decisions, unprecedented and inappropriate.
43
- 3. CONTINUOUS VENTING AND GAS GENERATION 3.1 Statement of Issue The incorporation of venting in the TRUPACT raises the following concerns: Type B packages must be designed to pass rigorous tests for leak-tightness so that even in severe accident conditions only extremely small quantities of particulate radionuclides could escape. At the same time. i t reduces the probability of failure due to changes in internal pressures from causes such as changes in elevation during transport or gas generation in the waste. These latter two conditions suggest potential advantages to continuously venting the TRUPACT in order to control pressure buildup. A third concern is that the gases being generated in the waste include hydrogen and oxygen.
which can form a potentially flammable or explosive mixture at concentrations above 4 or 5 volume percent. Department of Transportation regulations prohibit shipment of wastes in packages subject to formation of explosive mixtures of gases.
Venting might be considered a preventive measure i f i t could be shown to be effective for controlling flammable mixtures of hydrogen and oxygen in both the TRUPACT and the Type A packages.
However. in the regulatory experience to date. there is no evidence that filtered venting to prevent the buildup of explosive mixtures of hydrogen has ever been an NRC accepted design alternative to purging of containers followed by controlled shipment-time. or to using catalytic recombiners to limit radiolytic hydrogen buildup when large quantities of hydrogen might be generated.
DOE has contended gas generation is of little concern in causing an increase in pressure that could result in package failure.
But extremes in altitude variation and environmental 44
temperatures could cause a 7.5 psig pressure differential in a sealed TRUPACT. If there were frequent cyclicaJ_ pressure changes of this magnitude DOE has -stated that this might shorten the operational life of the TRUPACT packaging as a resuJ.t of inner frame weld joint fatigue (Ref 3.1). A detailed engineering analysis of pressure-induced weld joint failure has yet to be published by DOE. so the full details of the contributing design factors or the probability of such a failure mode cannot be commented on here.
The issues for the TRUPACT are whether venting is both needed
- and permissible to preclude fatigue failure and formation of flammab_le or explosive mixtures of hydrogen gas .in the shipping container.
means.
or whether these conditions can be avoided by other 3.2 Regulatory Considerations A chronological history of the more significant regulatory requirements is shown on Table .14 *
- In relationship to the TRUPACT design.
especially significant. In 1979.
several events are the IAEA issued non-obligatory regulations that permitted both continuous and intermittent venting. In .19B.1 Sandia designed the TRUPACT for continuous venting. although DOE Orders (Ref .10) had prohibited such a feature in May .198.1. Although NRC issued regulations in August 19B3 intended to conform to the draft IAEA regulations.
continuous venting during transport was specifically banned [.10 CFR 7.1.43 (h)]. The demonstration of compliance with the permitted release limits cannot depend on filter performance [.10 CFR 7.1. 45.1 (b) J.
45
Table 14 VENI'Iln OF 'lH.JPACT 0 1973 IAEA does not pennit continuous venting for Type B packages.
IAEA Safety Series No. 6, 1.973 Edition.
0 Jan 1878 NRG prohibits direct venting to atnosphere [1.0 CFR 71..35 (c)].
0 1978 Sandia begins 'IHJPACT design.
0 1979 IAEA prohibits continuous venting for type B packages. IAEA Safety Series No. 6, Revised E.dition 1.979.
0 May 1981 IXlE prohibits direct venting to atnosphere. DOE 5480.1., chg 3, III-1.2.
0 1.981. Sandia designs 'IHJPACT for continuous venting.
0 Aug 1.981. Sandia Peer Review does not discuss issue (SAND -2405) Published June 1.982.
0 Dec 1982 IX>E convenes major meeting to address hydrogen gas generation problem in transportation.
0 Aug 1.983 NRC prohibits continuous venting 1.0 CFR 71.. 43 (h) . Ganpliance with pennitted release limits cannot depend on filters, 1.0 CFR 71.51. (b).
0 Aug 1.983 EEO issues report based on sealed TIUPACT (EED-24).
0 Dec 1.984 IDE Draft SARP cla:ins justification for continuous venting.
- o 1.985 IAEA regulations permit intermittent venting tut prohibit the use of filtration to comply with release limits. IAEA. Safety Series No. 6. 1985.
0 July 1.985 EErJ. states justification for e,~emption inadequate and recoomends IXlE apply to NRC for e,~emption .
Noting that continuous venting was b~nned by NRC, EEG issued a report in August 1983 (Ref 14) with gas generation calculations predicated on the TRUPACT being sealed. EEG has repeatedly pointed out at meetings with DOE that the design with continuous venting violates DOT regulations as well as those issued by NHC and DOE (Ref 13).
After extensive draft revisions the IAEA published regulations in 1985 (Ref 3) permitting intermittent venting of type B(M) packages during transport. provided that the operational controls for venting are acceptable to the relevant competent
- authorities. Since NRC does not permit intermittent venting. i t would not apply to the U.S. However. NRC has committed. in the supplementary information accompanying its final 1983 rule.
conform with the anticipated IAEA revisions (1985).
to Nevertheless. the 1985 IAEA revisions continue to impose a ban on filtration for B(U) packages. The 19B5 IAEA regulations do not contain any overt statement on continuous venting but appear to preclude such a feature by not permitting a pressure relief system from the containment system. Hence the design appears not to conform with the IAEA regulations .
'
- 3.3 Gas Generation in THU Wastes The generation of gases from the degradation of defense Transuranic waste forms has been under investigation for the past decade. A number of reviews and summaries of data generated by these investigations have been prepared during this time (e.g .. Molecke and Clements. references 32 and 33) to assist in the development of Waste Acceptance Criteria for WIPP and the designs of the TRUPACT. Most of the early work focused on overpressurization effects of (largely inert) gas generated after wastes are emplaced in the repository.
47
During this study process, the specific concerns about gas generation have changed, to the present emphasis on hydrogen gas buildup in shipping containers. The 1.981. decision to vent the TB.UPACT has been reconsidered several times in the recent past.
In May 1986 the Albuqu 7 rque Operations Office announced that they were recommending that a redesigned TRUPACT (TRUPACT-II) be vented during shipment.
The aim of the discussion in this chapter is to examine the suitability of the present plans for the design of Type A containers and the TRUPACT transportation package to deal with gas generation related problems.
The chief concerns related to gas generation are:
Type A packages, the TRUPACT, 1.) the production of flammable or expl.osive concentrations of gases in or in the repository itse1.f; the release of part1culate contamination with carrier gases in 2)
Type A or TRUPACT packages: and 3) the long-term pressurization of the repository (post-closure). Only the first two are relevant to the present discussion. The first issue can be translated into more specific package design issues based on the strategy adopted to prevent the formation of flammable or explosive mixtures. Until recently, DOE strategy favored the use of venting both Type A and TRUPACT packages to achieve control. There is evidence that venting (RFP bung filter vent, or Hanford vent clip) will control hydrogen concentrations to below flammable levels in drums or boxes containing modest alpha curie loadings and low average G-values when in storage. There are no data for such drums in either a sealed or vented TRUPACT.
However. i t is questionable whether venting of the TRUPACT can be depended upon to maintain hydrogen concentrations below flammable levels when carrying a high curie load. It is also not clear whether either Type A or B packages can be certified with continuous venting. These considerations are pursued below.
48
3.3.1 Gas Generation Processes There are a number of gas generation processes in TRU wastes:
bacterial, thermal. radiolysis, and corrosion. Current data (Table 15) indicate that bacterial degradation of wastes has the potential for the greatest gas generation rate (moles/yr/drum) provided the right conditions exist (temperature. substrate.
presence or absence of oxygen. etc.). However. bacterial action does not appear to be significant for the short-term.
transportation phase of THU waste handling.
Table 15 MAJOR GAS GENERATION PROCESSES AND RATES Process Material Mole/yr-drum Bacterial Composi~e. aerobic 0.9-1.2 Decomposition Composite, anaerobic 1.2-32 Radiolysis Cellulosics 0.002-0.012 composite 0.002-0.006 PVC 0.01-0.0B
- Corrosion From Reference 32 Mild steel (anoxic conditions) 0.0-2.0 Of the remaining two processes. radiolysis is the more significant in the majority of cases. although corrosion has been proposed to explain the apparently unexpectedly high hydrogen gas production rates in certain RFP wastes under anoxic, wet container conditions (Ref 33). As a result, the debate over the need for. and advisability of venting Type A and 49
TRUPACT packages to achieve control over hydrogen generating wastes is based on the current data and understanding of hydrogen generation by radiolysis in TRU waste forms rather than any other mechanism.
3.3.2 Hadiolysis in THU Waste Alpha irradiation of waste matrices often results in higher gas yields than beta-gamma irradiation apparently due to the high percentage of energy deposited and possibly the high density of ionization associated with alpha tracks. Empirically, the alpha radiolysis process can be described by the number of gas molecules released for each 100eV of alpha energy deposited .
The gas generation parameter is called G (gas). For G (gas) =
241 . 4 1.0, each decay of Am should yield 5.48 X 10 gas molecules, 239 4 while for Pu the yield per disintegration would be 5.14 X 10 molecules assuming 100% of the energy is deposited in the waste.
G(gas) is not an intrinsic property of the material in which a given transuranic radionuclide is mixed, although some waste matrices do clearly show tendencies toward higher G-values than others. Work by Zerwekh (Ref 34) has shown that cellulosics and polyethylene evolve more gas than do rubber compounds during radiolytic decomposition. While some researchers have been tempted to conclude that gas yields in a small sample of typical THU waste (Fig 3, which shows hydrogen generation rate as a function of watts deposited per kg of waste) consistency within each waste category show satisfactory (Ref 33). others have observed a wide range in G (gas) values within various waste categories.
The gas yield (G) has been observed to vary with time (or.
equivalently, integrated dose) for a given TRU waste. This aspect has very ~mportant implications for the prediction of gas 50
16-------------------------
14 12 10 4
- 2 lsoprene 0.2 0.3 0.4 0.6 W/KG WASTE MATRIX Figure 3. Hydrogen generation in experimental waste matrices (from ref. 33).
51
formation. In the Clements and Kudera study of gas yield. for example. an average G value was calculated for a number of waste forms over a 13-week period (Ref 33). Over a 13-week period.
large changes in G (gas) may occur in some waste forms.
Averaging over such i long time tends to smooth peak and low values. There may be very significant consequences resulting from even short-term high gas yields. Although the causes for changes in gas yield are not completely understood. the most likely explanation for the decrease of G (gas) with time in most waste forms is matrix depletion. Matrix depletion may result from changes in contact between contaminated surfaces and organics in the waste. transformation of the matrix due to radiation effects. and loss of suitable hydrogen bond sites within the range of alpha particles from contaminant sites. An example of extreme matrix change brought about by radiolysis is the observed formation of a fine powder by radiolytic degradation in cellulosic waste forms and neoprene drybox glove material (Ref 34). The powder contained approximately 50% of the THU contaminant that was added originally. Powder formation may contribute to the changes in G (gas) in such wastes. but this has not been demonstrated. Few other waste matrices showed similar degradation products.
In six experimental studies. long halftimes of decay of G (gas) have been observed.
plastics. and rubber In the case of a mix of cellulosics.
(Fig 4-a).
water-soaked cellulosics the halftime is 630 days.
(Fig 4-b).
For the halftime is nearly 10 years (3465 days). On the basis of the data from these studies.
i t is tempting to conclude 'that any change in G (gas) value would not be of interest as far as implications'for transporta-tion are concerned (30-60 days). and that an average G (gas) determination over a period of several months could suffice to quantify the amount of gas generated.
52
A. ORY CELLULOSICS s------------------- v =
T 2 630 days 4
3
~
.. ~
\.:)
2 O-t-----------.-------------.1 0 500 1000 1500 2000 ELAPSED TIME (DAYS)
B. WATER-SOAKED CELLULOSICS 5------------------- T 1/2 = 3465 days 4
2 0-+---------------------11 0 500 1000 1500 2000 ELAPSED TIME (DAYS)
Figure 4. G (gas) as a function of time (from ref. J4). G = gas molecules per 100 eV.
53
]
However. as a result of our recent reviews of data on gas generation in a sample of HFP waste forms (Ref 33) and models of gas generation in sealed and vented packages (Ref 36). a different perspective on G (gas) has emerged.
The initial gas generation rate G (gas) can be much higher than the average as shown by the post-closure data of the TRU waste package (Ref 35). Figure 5 is a plot of G (gas) against energy released (eV). Note that a non-standard use of the term "dose" occurs in this literature. Here i t means energy released rather than the usual energy deposited per gram. For short times (low dose) the G (gas) value fs nearly 3. and later decreases to 1.
similar pattern of initial short duration. high G (gas) value.
A followed by a nearly constant long-term G (gas) value. has emerged from our analysis of the recent RFP study data. Using a hydrogen diffusion model described below (also see Appendix A).
the data for hydrogen buildup in vented drums was modeled. The best overall fit was obtained assuming a two component model of G (gas) as a function of time:
G (gas) = G Initial
+ G constant where A is a relatively short term decay constant.
there are other possible models that would fit the data.
No doubt particularly a double exponential model where the constant term is replaced by a constant plus exponential.
The resultant G (gas) parameter. shown in Figure 6. has the short-term declining G-value and a long-term constant G-value.
The figure also illustrates that the function is consistent with a measured 13-week average G-value which is much smaller than 0
Initial" 54
3.2-------------------------------.
2.8 2.4 2.0 I. 2 0.8 0.4-'------------------------.,--------ii 0 20 6b eo 100 140 DOSE (~V x 10-23)
Figure 5. G (gas) as a function of integrated dose (from ref. 35).
55
6----------------------------..
G=GI (EXP(-KT))+G2 5
4 Avg. G = 2.91 Measured G ='3.5
~
"l:i.
\:, 3
\:,
2 Constant o ......-----....--...----....... -"1--.. . .----...------.....----.. . .
0 2 3 4 5 6 7 8 9 K) 11 12 13 14 TIME (WEEKS)
Figure 6, Two component G (gas) time-variant function, 56
As partial confirmation of the general correctness of this approach the dose, again, in the non-standard sense of energy released into the waste, for one of the cellulosic waste drums was computed, and G (gas) predicted by the model vs dose was plotted (Fig 7). Although the predicted G (gas) was higher for this case. the overall shape is quite similar to the curve from the experimental data. The G (gas) and G (hydrogen) parameters were measured in the RFP experiments by successive one-week determinations of the changing hydrogen partial pressure in sealed drums. A separate set of measurements was made when these drums were vented. Therefore a limited comparison between modeling predicted G (hydrogen) behavior and observed G (hydrogen) changes in the sealed drum can be made. As seen in Figure 8, there is reasonably good agreement between these two estimates. However. as Figure 9 illustrates. in some cases the apparent initial G (hydrogen) is very much larger than the average. There is considerable uncertainty about the exact time the first reading was taken post-sealing ("time-0" in the data set). which has a large effect on estimated G (hydrogen) during the first week. Only more detailed measurements of the initial phase of hydrogen generation will resolve the question of whether a large initial G (hydrogen) occurs and if that results in a rapid filling of the drum void with hydrogen.~
" There are a number of potentially viable alternative explanations for the observed rapid initial rise in hydrogen concentration. One possibility. applicable to waste consisting of a number of separate sealed plastic bags of waste in a drum, is that G(gas) is nearly constant. but when purging occurs before sealing a drum. a large remnant hydrogen concentration remains in the several bags. which then quickly d~ffuses into the drum void post-sealing. In monolithic waste forms hydrogen diffusion from the core may cause a similar short term response.
57
CELLULOSIC G (GAS) VALUES VS. INTEGRATED DOSE 4
2
.e~---.--....--.......--.,...---,.--~-.....,.---,--...,.--T----,
0 2 3 4 5 6 7 8 9 10 II DOSE ( eV x ,o-23)
Figure 7. Model predicted G (gas) variation with integrated dose (from ref. J2).
58
G (HYDROGEN) IN DRY COMBUSTIBLE WASTE
( DRUM 24545 )
6
..........,..........,Predicted 2
o..,____________________________......,
0 I 2 3 4 5 TIME (WEEKS)
Figure 8. Modeled and observed time varying G (hydrogen) in a sealed RFP drum of TRU waste.
59
G (HYDROGEN) IN COMBINED SLUDGE (DRUM 29258) 9-,-------------------------
8 7
3 2
o....,_________________________.
0 2 4 6 8 12 TIME (WEEKS)
Figure 9. Observed time-varying G (hydrogen) in a sealed RFP drum of TRU waste.
60
Thus the process of generation of gases in THU wastes.
particularly flammabl.e or explosive mixtures of hydrogen. is quite complex. particularly in the initial period following purging of the drums and installation of vents. It has been shown here that the G (hydrogen) parameter can be described as a two-component exponential function of time. The long-term component may be a constant. or have a large half-time (2-10 year half life) compared with the short-term component (O.b - 3 week half life) The short term component is at least sometimes associated with an apparently very large G (hydrogen) compared with the long term average value. As mentioned in the footnote to the previous paragraph. the apparent high G (hydrogen) may be due to other processes at work. It is the effect. of course, which is of real concern.
Given these characteristics ot* G (hydrogen) and G (gas). the next question to address is how the formation of flammable or explosive mixture in shipping containers and the TRUPACT can be avoided.
3.3.3 Controlling Pressure and Hydrogen Buildup Given that some wastes will rapidly evolve large quantities of hyd~ogen gas. and the obvious desirability of controlling pressure and flammable gas buildup in transport packages. i t is clear that some form of control is needed. There are four principal options to consider:
- 2. Using getters to trap the hydrogen gas:
61
- 3. Venting the containers and the THUPACT: and
- 4. Management control.
Management control includes purging of the inner waste packages with inert gases just prior to shipment, and controlling the shipping time so that concentrations of hydrogen and oxygen remain below 4% in transit. The NRC approach to regulating transport of hydrogen generating low-level wastes involves management control [Ref 40].
The first option. recombining hydrogen and oxygen, has been used on a large number of drums of transuranic waste at Hanford, and has been used with the Three Mile Island (TMI) hydrogen generating wastes (Ref 37). A disadvantage is that corrosion or other oxidation processes may compete for oxygen, leaving an unacceptably high hydrogen concentration. The use of hydrogen getters is apparently an untried option at Hanford and at TMI at this time and will be discussed further below.
The third option. venting. has received the most attention by DOE. The concept which has been most thoroughly investigated involves venting the waste packages through rugged high efficiency filters or permeable gaskets and vent clips so that gases are released. Particulates are supposed to be retained in the containers even under severe transport conditions. Figure iO (a) illustrates the small filter (RFP bung filter) being considered by DOE for Type A packages, and iO (b) the filter design which has been proposed for the TRUPACT (if i t is vented). The prime consideration is whether venting will provide the needed control of hydrogen concentrations in Type A packages and the Type B TRUPACT under the actual transport conditions of the waste forms and THU concentrations anticipated 62 I.
A. ROCKY FLATS PLANT SMALL BUNG FILTER CONCEPT HOLE AREA = 0.079 cm2 EFF. THICKNESS= 13.24 cm 0
.... Gasket
.. 11 1/8 vent
- 8. TRUPACT I PROPOSED FILTER
- Well for filter cartridge
- Filter cartridQe
,~------+-----1=1r-Door Figure 10. Filter vent concepts.
63
in defense wastes to be shipped to WIPP. DOE has supporLed a number of research efforts aimed at understanding how venting of Type A packages may serve to control hydrogen buildup. As a result, there is a growing body of data on the performance of venting devices such as the RFP bung filter under storage conditions, which indicates that at relatively low curie loadings, venting will maintain drum concentrations below 4-5 percent hydrogen (Ref 33). Although there are some experimental vent performance data for higher drum loadings, none are available for a fully loaded TRUPACT. Thus computer modeling must be used to provide performance predictions for the TRUPACT.
There is a definite need for confirmatory data for these model predictions; 3.3.4 Modeling of Hydrogen Gas Buildup A recently developed computer model of hydrogen dissipation in sealed or vented. nested transport packages by SAIC (Ref 36) provides a tool for accomplishing these performance predictions.
EEG has made a number of modifications to this model which have made i t possible to use the approximate diffusion properties of the filters in the model instead of empirically developed effective diffusion coefficients. The EEG modified hydrogen gas buildup model approach to modeling filtered vents parallels Kazanjian's 1983 work (Ref 38) and is described below.
The TRUPACT container geometry is shown in Figure 11. The inner volume represents the Type A packages containing THU waste. In the case of a fully loaded TRUPACT with 36 55-gallon drums, the inner volume represents one of these drums and the outer volume represents 1/36th of the TRUPACT void when loaded. Each volume is assumed to have a filtered vent, with characteristic thickness and area. Vents can be modeled as sealed as well as 64
I FX 2 Inner filter II %
Void Void FA 2 I I
Outer filter Inner volume I
Outer volume Figure 11. Schema.tic of drums inside TRUPACT.
65
open. For simplicity. gas sources are assumed to exist only in the inner volume.
Using the previously described two-component model of G (gas).
and the described dimensions of the RFP bung filter (Ref 39). G (gas) values were fit to observed hydrogen buildup data (Ref 33) for a number of cases. For the case of dry cellulosics. values of G (gas) were found which fit the observed vented drum data (Fig 12b) quite well. A maximum G (gas) of 4.5 was required.
and an average of 3.5. The observed G (gas) for this case was
- 3. 7. Using these same values for G (gas). the sealed drum case was simulated. again with good resu1. ts (Fig 12a) . These are independent data sets, and thus provide verification that the general modeling strategy is sound. More precisely. i t should be said that this is a reasonable representation of the phenomena. even if the physical mechanism is the diffusion of hydrogen out of sealed inner voids.
Other cases were simulated, illustrating that the model can be used to predict hydrogen ingrowth in cases where the initial G (gas) is low (Fig 13a). as well as when i t is quite high and of short duration (Fig 13b).
The experiments at RFP with actual drums of TRU waste have shown that venting with the RFP bung filter does limit the accumulation of hydrogen to level.s below those found when the drums were sealed. On the basis of these experiments. then venting alone will maintain concentrations of hydrogen below 4%
if the product of G (H ) and a-curies is below about 40 (Ref 2
33). Unfortunate-ly. as was noted in the RFP experiment. no drums were tested for hydrogen generation rates of 20 to 60 a CiG(H ) to confirm their prediction.
2 66
A. SEALED DRUM HYDROGEN ACCUMULATION IN DRY COMBUSTIBLES (DRUM 24545)
.4
.. ~ .3
~
~
~ .2
~
~
);:
T1/2 = 504 hr
- t:: Gmax. = 4.5
.I GavQ. = 1.5 Gmeas. = 3.5 o .....
0 - - - - - - . . - - - - 3 - - - - - - - - - - - 6 ---
TIM£ ( W££KS)
B. VENTED DRUM HYDROGEN ACCUMULATION IN DRY COMBUSTIBLES (DRUM 24545)
.30----------------------
~
.20 .
.,., / r---,_ ___ ___ /
.,,,,......._'-\
-- \
' . Observed \
--/
T 1/2 = 504 hr.
Gmax.= 4.5 Gavg. = 1.5 Gmeas. = 3.5 0 ....
0 ---- 2 -- 3 -- 4 -- 5 -- 6 -- 7 - -10 - ~II TIM£ (W££KS)
Figure 12. Model verification results.
67
A. HYDROGEN BUILDUP IN GREASE WASTE (DRUM 31403)
.50..----------------------..
.40 ,, .,,,,,, .,,,,.,,---
Observecy/
/
/
I L
T 1/2 = 1344 hr.
Gma,t. = I
.10 GavQ. = 3.25 Gmeos. =22.5 o....,_____________________.,. Alpha Ci= 0.2 0 2 3 4 5 6 7 8 9 IO II TIME (WEEKS)
B. HYDROGEN BUILDUP IN GREASE WASTE (DRUM 31254) 1.20----------------------
1.00 /
~ .80 Cl)
I I
I '" .___. ____
Observed
~
---- /
~
~ .60 I
~
I}: I e.40
- t:: I T1/2 = 12 hr.
I Gmax.:: 88 GavQ. = 2
.20
~ Gmeas. 15 =
=
Alpha Ci 0.35 0
0 2 3 4 5 6 7 8 TIME ( WEEKS)
Figure 1J. Hydrogen buildup in vented drums.
68
By modeling. we have attempted to examine the predicted efficacy of venting. If the case of dry combustibles where G (hydrogen)
= 2.1 is extended to the maximum controllable loading by the Clements and Kudera method. the allowable loading wou.ld be 20 curies and the 4% limit would be exceeded in about 3 weeks.
Furthermore, if 36 of these drums were loaded into a vented TRUPACT. concentrations would reach 4% hydrogen in 6 weeks.
(See Figure 14).
It has been suggested (Kazanjian) that at higher curie loadings a larger filter would be required to limit hydrogen concentrat.i.on. The results of our modeling. however. indicate that a 2B-fold increase in filter area would be required to
- achieve a 30% reduction in hydrogen concentration. A filter this large would risk a reduction of containment integrity.
Our perception of the venting process at this time is that during the initial post-closure period following purging. the relatively rapid buildup of hydrogen concentration either due to high initial G(hydrogen) or the presence of hydrogen in sealed packages diffusing into the void. or both. quickly displaces air from the void space without a large loss of hydrogen. since the initial hydrogen concentration gradients are smaLL. But then as the hydrogen concentration bui.lds. even though the hydrogen contributions from various sources may drop to modest levels.
the hydrogen concentration can rise to the flammable or explosive limit if the curie level is high enough. A critical
£actor in the process just outlined ~s the occurrence of an initial high influx of hydrogen. even for a few hours or days.
While such initial high rates of hydrogen gas input can be expected. confirmatory experimentation is definitely needed.
Based on data developed thus far. and our current modeling 69
4-r------------------------
/ / /
/
/
/
/
/
/
/
/ Drums
/
/
/
/
/
/
/ TRUPACT
/
/
2 3 4 5 6 TIME (WEEKS)
Figure 14. Vented TRUPACT with 36 drums (20 Ci/drum).
70
results. we conclude that venting of the Type A and Type B (TRUPACT) waste packages to achieve hydrogen control is not likely to result in the desired effect for a significant portion of the present de.fense THU inventory and thus should not be relied upon as a control mechanism. particularly i f a relatively inexpensive and effective alternative control practice is available.
3.3.5 Alternative Control Strategies Fortunately there are alternatives to the venting of transport and storage packages in order to achieve control over the formation of flammable or exp:Losive mixes of gases. Many of these are currently in use in the handling and transport of h+/-gh level wastes. both in the defense (Ref 37) and commercial sectors. These p~actices all rely on the outer (Type B) package not being vented, which has the obvious advantage of conforming wit-:h NRG and DOT regulations.
A strategy for the storage, preparation for shipment, and transport within controlled time limits following sealing of waste packages must be developed by DOE to properly implement a sealed TRUPACT shipping system compatible with gas generating
- wastes.
Components of such a strategy are expected to include:
- 1. Identification of wastes requiring special handling to control gas generation:
a) Methods for computing H and o generation rates in 2 2 various waste forms. particularly short-term high hydrogen evolution rates. based on waste forms.
curie content. internal packaging. etc.
71
b) Methods for confirming gas generation buildup (through QA programs).
- 2. Treatment of Gas Generating Wastes:
a) Venting gases from drums which have been in storage.
b) Installation of filtered vents, permeable gaskets, or other systems which will allow drums to continue to vent during storage and transport.
c) Dilution of drum voids with inert gases prior to sealing.
d) Introduction of hydrogen-oxygen recombiners or, perhaps better, hydrogen getters in the drum void.
- 3. Provision of Administrative Controls:
a) Identification of special problem wastes.
b) Creation of control system to track storage and shipment times after closure of the containers and the addition of getters or recombiners to assure wastes can be handled and transported to WIPP without the buildup of excessive levels of hydrogen.
c) Creation of a data base on waste forms. G-values, alpha-curie content. etc. for predictive purposes and QA.
Regarding the first of these components. NRC's Office of Nuclear Materials Safety and Safeguards (NMSS) has provided some guidance on how to deal with shipment 6f wastes subject to hydrogen generation (Ref 40). This should provide valuable guidance for the DOE TRU wastes as well. The generic requirements specify that for gas generating wastes i t must be 72
determined by "tests and measurements of a representative package" that hydrogen and oxygen concentrations do not exceed 5% by volume durin'g *a period of time that is twice the expected shipment time. More recently NRC has recognized that an analytic approach can be effective as a means for determining gas generation (Ref 41). Thus a valuable tool for control is a flexible. well-tested. and peer-reviewed hydrogen generation and control assessment methodology and associated data base. The analytic approach involves determ.ining the hydrogen generated in the waste by radiolysis during a period of time after closure and twice the shipping time. This requires determining well the properties of waste influencing gas generation by suitable tests and measurements on representative waste forms (such as those
- reported by Clements and Kudera in Reference 33). A valuable refinement of this modeling approach would be the provision of capability to estimate hydrogen contributions from sealed inner packages as an alternative. or contributor to. observed high G (hydrogen).
The second component. venting. has been extensively discussed above. but H -o control by recombiners or getters warrants 2 2 further discussion. Catalytic recombiners remove hydrogen and oxygen in the ratio of Z-to-1. However. when oxygen is being scavenged by oxidation of the drum or waste components. excess hydrogen can build up . If oxygen is sufficiently limited. there
.is not a high hazard from flammability. but there is a potential for ignition upon venting to the atmosphere. Catalytic recombiners seem to be most appropriate under conditions of re1.atively short term storage post sealing and purging of drums.
An existing individual package recombiner packet design is shown in Figure 15. Hydrogen and oxygen diffuse to the catalyst where they recombine to form water vapor. The vapor condenses on colder surfaces in the system. A combination of Engelhard 73
A
.....-----6 in.------
1/2 in.3 of Engelhard Catalyst "o" Spot welds
.,L A--
Ai'Fi, --A 6in.
CROSS SECTION
~
Stainlns ltfft t
2 in. Fits on drum flange beneath lid l
CATALYST ASSEMBLY VENT CLIP Locate in top section of waste drum F.lgure radi-e
- 15. Rockwell Hanford operations catalyst bed and vent for drums of waste.
- Dextro D and silicon coated catalysts have been found to be effective under both dry and wet conditions (Ref 37).
Hydrogen getters. in contrast to recombiners. selectively remove hydrogen by chemical reaction regardless of oxygen concentration. and thus do not have the limitation of susceptibility to competing processes for removal of oxygen.
One potential getter is propargyl ether mixed with a metallic catalyst. Details on this getter are described by Neary in Appendix B. Others are described by Trujillo and Courtney (Ref
- 42) . One disadvantage of getters is that they are consumed in the gettering process. Thus. careful consideration has to be given to the total amount of hydrogen expected to be generated during the storage and shipment period so an adequate quantity of getter can be provided. Both recombiners and getters must be properly placed in the transport package. If they were used in the TRUPACT. then there may be some material and labor savings over the construction and placement of individual packets for placement in the drums or boxes. However. a compeJ..ling argument in favor of placing the control materials directly in individual drums is that the interaction of hydrogen and getter surface occurs sooner and more efficiently in the drum than in TRUPACT.
If the removal process is limited primarily by diffusion of hydrogen to the active surface. eCfective control can be anticipated by placing a hydrogen getter (Fig 15) in the drum.
It may be possible to spray getter material in sufficient quantity on the inner surface of Type A packages to effectively control hydrogen for the handling and shipping period. A computer model simulation of such a process is shown in Figure 1.6. A simple representat.ion of the removal by gettering was assumed with only a limited number of sites ava_ilable (0. 5 hydrogen moles equivalent). Without the use of getters, a 75
4 I
I I
I
~3 I (I) I
~ I
~2
~
~
'!Ii.:
- t:: I I
/ Without getter I
l I
0 0 2 3 4 5 6 7 8 9 10. II 12 TIME (WEEKS)
~gure 16. Effects of getter on hydrogen concentration in drums.
76
flammable level of hydrogen is reached in less than 3 weeks.
However, with getters. an add~Lional 3 weeks of very low hydrogen concentrations are realized before the getter is consumed. and several more weeks during which ingrowing hydrogen concentrations are still below four percent.
Si.milar simulations with the getter assumed in the TRUPACT du not suggest that effective control of hydrogen would occur. In this configuration. hydro-gen must accumulate and then diffuse out of the drums and reach the getter or recombiners in the THUPACT before any removal occurs. Further modeling and experimentation are needed to establish the best control strategy. but placement of getters in each drum appears to be
- the best control option.
An added advantage of placing the getter or recombiner materials in the Type A package instead of the TRUPACT is that model si_mulations indicate that where the TRUPACT is sealed, but the drums are vented, flammable mixtures can accumulate in the drums even though the TRUPACT void levels are acceptably low. If the getter is placed in the Type A packages also. control of both containers is ach~ev~d. which should be the only acceptable condition for transport and receipt at WIPP .
- The third component, development of a administrative controls. is critical to the successful control strategy. If the option of using getter materials to control hydrogen buildup is adopted.
i t would appear that the more detailed auditing of waste matrix form. curie content. inner packaging characteristics, etc. that would otherwise be required, could be avoided. However. this is an area which requires detailed evaluation by DOE. The issue of special problem waste, particularly high curie content waste.
raises other concerns which will be discussed in Chapter 4.
77
- 4. HIGH-CURIE CONTENT 4.1 Statement of Issue The draft THUPACT Safety Analysis Report for Packaging (SARP) conclusions on the quantities of various THU waste forms that can be transported in a THUPACT load are shown in Table 16 (Ref 31).
TABLE .16 SARP MAXIMUM TRUPACT LOADINGS Waste Type Normal Weapons Total Ci 4,450 Total PE-Ci 840 (a)
Limiting Criteria criticality control Am-Enhanced .12,020 4.340 heat generation (360W)
Heat Source .1.1,200 .14.200 heat generation 239 (a) PE-Ci= Equivalent curies of insoluble Pu based on inhalation toxicity (Ref 46).
DOE estimated the bounding consequences that might occur from accidents while transporting THU wastes to WIPP in Chapter 6 of the Final EIS on WIPP. These consequences assumed a total radioact.ivity loading on a rail car (containing .126 drums in 3 TRUPACTS) of 79.5 PE-Ci of insoluble THU wastes. This loading assumed all drums contained an average quantity of THU wastes.
The release fractions and other scenario assumptions used in the FEIS were similar to those used in NUREG-0.170 and are considered 78
typical for nuclear materials transportation. EEG believes that the assumptions were reasonable. but slightly unconservative.
The FEIS also stated that the maximum radioactivity in a drum would be 25 times the average value. A railcar carrying two TRUPACTs could contain ZB.400 PE-Ci. This value is 357 times that used to calculate bounding consequences in the FEIS and 14 times the implied upper limit in the FEIS.
The key issues are whether:
- 1. Such a drastic increase in the PE-Ci load of the TRUPACT has such a substantive change in the predicted consequences from Chapter 6 in the FEIS that i t should not be permitted without an amendment to the FEIS:
- 2. The potential hazards of these proposed maximum shipments are excessive compared to other radioactive material shipments.
4.2 Possible Risks and Consequences 4.2.1. Comparison with FEIS There are numerous differences between the calculations in
- Chapter 6 of the FEIS and Chapter 2 of this report besides the number of PE-Ci being transported. These include the assumed fTactional releases and dose conversion factors (see Table 17)
The PE-Ci of radionuclides released shown in Table 17 is a better measure of the comparative risks estimated in the FEIS and in this report than the dose received by the maximum individual because the FEIS doses were calcuiated using older dose conversion factors which are not directly cornparabJ.e to those calculated in Chapter 2. The Table 17 comparison shows 79
Table 17 COMPARISON OF RELEASES BETWEEN F'EIS &- CHAPTER 2 (PE-Ci)
Fractional Average Railcar Load Maximum Railcar Accident Release Total Released Tot.al Released FEIS 1.. 75-4 (a) B.0+1. 1..4-2 2.0+3 3.5-1.
Category VII 5.0-4 4.7+2 2.3-1. 2.8+4 1..4+1.
Category 2.5-3 4.7+2 1..2+0 2.8+4 VIII 7.1.+1.
(a) 1..75 - 4 = 1..75 X 1. -4 that a maximum Category VII accident releases about 40 times the amount predicted in the FEIS. The Category VII release from an average truck shipment (the most probable mode) is 8 times the projected FEIS release from an average rail shipment. Another comparison (see Table 1.9) is that a Category VII accident with the average Savannah River Plant truck Shipment (1.0 PE-Ci/TRUPACT) would release 2-1./2 times that released in the implied maximum rail accident in the FEIS. EEG believes that these estimated releases from a Severity Category VII accident (2% probability of occurrence during WIPP lifetime) amount to a substantive change in the expected impacts of the project.
4.2.2. Comparison with other Radioactive Material Shipments.
Most transuranic waste has so little penetrating radiation that they can be handled without shielding (hence the name contact-handled). Since all high level wastes and spent fuel, as well as some low level waste, require shielding for safe handling.
80
th<.~re is a tendency to think of THU waste as a benign form of radioactive waste. However. inhalation following an accidental release is a more important exposure pathway than external gamma radiation.
Some of the contact-handled TRU waste shipments coming to WIPP may be as hazardous or more hazardous than shipments of spent fuel or defense high level waste following an accidental release for the following reasons:
j_
- The THU radionuclides are much more toxic per microcurie inhaled (which is the more likely pathway
- 2.
resulting from an accident during transportation or operation) than are fission products:
Much of the CH-THU waste being shipped to WIPP will not be as immobilized as spent fuel encased in zircaloy or steel cladding, or defense high level waste (bHLW) fused in borosilicate glass within a st.eel canister. Thus. a severe accident involving THU waste could reiease a higher fraction of the THU waste container contents:
- 3. Some of the shipments that may come to WIPP will have an inhalation toxicity inventory number of Annual Limits of Intake) a spent f~el assembly (as measured by the simi.la:r to tha:t of (see Table 1$). For example. a TRUPACT load of heat source waste at SRP has an average toxicity inventory of about 95% of a spent-fuel assembly and the inventory would require about 970 THUPACT loads if i t .is all shipped to WIPP. There would also be some high-curie loads from other laboratories. primarily Los Alamos. The Defense High 81
Level Wast.e t*rom SHI? has a toxicity i.nventory less than either of the above: one DHLW glass canister is 231:l only 49% of the average Pu shipment (Ref 43).
TABLE 1.8 COMPARISON 01" SPENT F"UEL AND TRUPACT RADIOLOGICAL TOXICITY WASTE PE-Ci/LOAD THU/SF Spent ¥uel, Rail 26,800(a)
Truck 3.830 Defen8e High Level Waste. Truck 1.,760 THU PACT Maximum, Rail 2B,500(b) 1.06 Maximum, Truck 14.200 3.72 Wil?P Average 233(c) .06 LANL Average 222(c) .06 SRI? Overall Average 1. 0 (c) .47 SRP Heat. Source Average 3.BOO(c} .95 (a) Reference 45 (7 Assemblies/Cask for Hail. 1. for Truck)
(b) Reference 31.
(c) Reference 26 for 1. TRUPACT (Truck}.
Combining the PE-Ci per shipment for various wastes and t.he re.lease fractions from SAND 80-21.24 (Ref 44) for spent, f'ue:l and C.h.:1.pter 2 for THU wast.es in the TRUPACT leads to t:.he anticipated releases shown in Table 19. These values indicate that in a severe accident even the average THU waste shipment to WlPP could be expected to release a much more toxic quantity of radioactivity than a spent fuel shipment involved in a similar accident. A doubly contained design is projected to eliminate any release from a Category VI accident (19% probability of 82
oc~urrence during WIPP lifetime) and significantly decreased releases from more severe accidents.
TABLE 19 RELEASES F'ROM TRUCK ACCIDENTS INVOLVING SPENT FUEL AND TRUPACT Shipment Load Release Release (PE-Ci) It'raction (PE-Ci)
Spent Fuel 3830 2 X 10-6(a) 0.00'17 6 VI(c)
(b)3600 5 X 10-TRUPACT 0.01.B 4
5 X 1.0- ~II 1.. 8 2.5 X 1.0- VIII 9.0 Double Contained 3600 0 VI 0 TRUPACT 1. X 1. -4 VII 0.36
- 1. X .1 -3 VIII 3.6 (a) Reference 44 (credible worst-case accident).
(b) Average PE-Ci load from SRP heat source wastes.
(c) Reference 2. Roman numerals refer to accident severity category.
The gas generation problem is an additional factor to consider if high-curie loads are to be shipped. As discussed in detail in Chapter 3. there is considerable uncertainty in the ability to predict gas generation rates and to control concentrations of hydrogen to below the 4% threshold for flammability. The potential gas generation problems increase with increasing curie content in a container or in a TRUPACT load for similar waste matrices.
f:J3
4.2.3. Non-Radiological Considerations If there were a reduction in the maximum number of PE-Ci that could be transported in a TRUPACT leading to an increase in number of shipments there would be a corresponding increase in non-radiological injuries and deaths as discussed in Chapter 2.
EEG questions whether a significant reduction in the limit would result in a significant increase in the number of shipments (see discussion below). At any rate we do not believe the selective trading-off of radiological versus non-radiological risks should be used to justify TRUPACT design and operation criteria.
4.3 Operational and Economic Considerations 4.3.1 Re-Packaging There would be some costs and additional occupational radiation exposure incurred if i t were necessary to repackage.currently stored waste in order to comply with a significant reduction in the permitted PE-Ci load in a TRUPACT. Otherwise. re-packaging would not be difficult or unprecedented: some drums have been opened and inspected at most generating sites in order to verify drum contents with records and assay results.
EEG believes that little or no re-packaging would be required if the permissible load limit were set slightly above the average PE-Ci content of a generator's waste. The proposed 1,000 PE-Ci limit in the waste acceptance criteria for a drum or box (which EEG believes should be lowered) will require at least one constraint. There are an estimated 250 drums that contain greater than 330 alpha Ci of heat source waste (which would be
>400 PE-Ci if the radionuclide were all 238 Pu) but we are not aware of any greater than 800 PE-Ci (Ref 46). There are two implications of these data on high-curie drums:
B4
- 1. It clearly would be possible to load a TRUPACT with greater than 14.200 PE-Ci;
- 2. It should be possible by load management to mix these high-cUrie drums with weaporis waste and lower curie heat source waste drums in order to hold the total TRUPACT load to less than 2,000 PE-Ci at SRP. and much lower elsewhere.
Ahother possibility. which DOE is considering. is to dispose of the high-curie drums via incineration of waste and incorporating the residue with DHLW for disposal in a i-ILW repository .
- A positive idad management prograhl to ~iniinize the totai PE-Ci load in a TRUPACT should not be particularly cbstly because the coritairiet-s in~~t be assayed separately for PE-di c6rttent and adherence to other waste acceptance criteria. Following assay.
the PE-Ci content is known and the containers can be assembled for (more-or-less) average THUPACT ioads. These average ioads would also be preferable at the WIPP site for handiing. loading*
on the hoist. and einplacihg iri Uhderground storage rooms.
The DOE ha.s refUseq tQ comrnit to a positive load management
- program. but they have assumed that random probability would preclude two or more above-the-average PE-Ci drums froin being involved in several of their transportation. operatlon. and post-closure scenarios. Since high-curie containers tend to be stored together at the waste generating sites. EEG believes that without a positive program i t is not ~rUdent to assume the occurrence of .high-curie drums in a TRUPACT or at WIPP is purely random.
85
4.3.2 Number of Shipments The number of TRUPACT loads shipped to WIPP should not be increased if a maximum limit is chosen which is slightly above the average TRUPACT load at each generating site. Also, since some wastes will be processed at most sites i t may be possible to reduce the average concentration per container while carrying out operations that would be done for other reasons.
The average shipment from SRP (1800 PE-Ci) is estimated to carry an accident risk, with the TRUPACT-I design, similar to the credible worst-case spent fuel accident for a Category VI accident and about 2 orders-of-magnitude greater for a Category VII accident. A double contained. average SRP loaded TRUPACT is estimated to be safer than the worst case spent fuel accident in a Category VI accident (4% occurrence probability in an urban/suburban area) and to release over one order-of-magnitude more radionuclides in a Category VII accident (0.3% probability in an urban/suburban area).
From the above considerations i t appears that a doubly contained TRUPACT could be permitted to carry the average SRP THU waste shipment without incurring a significantly greater hazard than would occur from shipping spent fuel by truck. Therefore.
limiting the maximum load in a doubly contained, non vented TRUPACT to slightly above the SRP average load should, be acceptable and could be accomplished without increasing the number of shipments.
U6
- 5. REDESIGN OF TRUPACT 5.1 Modifications Being Considered A value engineering analysis by DOE concluded that potentially significant total system economies would be possible by making major design changes to the present TRUPACT-I design. This anal.ysls assumed the THUPACT--II design would have single containment and continuous venting (Ref 47). Changes bei.ng considered lnclude:
(1) Revising the overall dimensions of ~TIUPACT to increase the capacity from 36 to 48 drums {the number 3
of 112 ft boxes that can be carried is not increased);
(2) Drastically reducing the weight of the empty TRUPACT in order to increase payloads. This is done by replacing the roller floor with a slip-plate system; using conventional steel banding or plastic stretch-wrap material rather than steel frames to hold 6 drums together in a "6-pack": reducing the thickness of the inner liner and Kevl.ar puncture shield: and reducing the amount of dunnage:
(3) Changing the method of applying foam insulation
- - between the inner liner and the outer skin.
The WIPP Project Office has stated verbally that the full-scale tests conducted on Unit-0 will be applicable to the new design and additional full-scale tests may not be necessary.
Also, i t is believed that only an amendment to the THUPACT-1 SARP will. be required. The present schedule is to have a draf't B7
amendment to the SARP in the fall of 19B6. a final SARP in March 1987. and a certified design in October 1987. First delivery of operating units will be prior to October 19BB. when first waste shipments are scheduled. Present plans are to build only two operating units (Units 2 and 3) of the TRUPACT-I design if the new design is accepted.
5.2 Possible Radiological Impacts This report specifically evaluates only the TRUPACT-I design which is expected to be recommended by the Albuquerque Operations Office for certification by DOE in the third quarter of calendar year 19B6. Also, i t is not certain that a redesign will be recommended and i t is not known what specific changes would finally be incorporated. However. since a redesigned TRUPACT appears likely and if construction is implemented as much as 90% of the WIPP fleet could be TRUPACT-II units. EEG believes i t worthwhile to point out some preliminary concerns.
Some of the advantages and disadvantages of a 48-drum TRUPACT were evaluated and discussed in Chapter 2. Radiological effects from routine operations are slightly worse for the 48-drum design and accidental releases from an average load would also be greater.
EEG believes that some of the proposed changes in the design are substantive and that not all the results of evaluating and testing Unit-0 of the TRUPACT-I design can be transferred to the new design. Questions that arise include:
- 1. Do the significant changes in dimension~ of the TRUPACT really result in a package that is structurally stronger for all drop orientations as DOE claims?
BB
- l.. How will the thinner inner liner and Kevlar punct.ure plates hold up under the full-scale drop and puncture tests imposed on Uni t--0?
- 3. How will the decreased amount of dunnage (compared to the Unit-0 test where voids were carefully packed with considerable dunnage) affect integrity of the inner containers during drop and puncture tests?
- 4. Will the new method of applying insulation foam during construction avoid the problems of uneven density that occurred initially with the old method of application?
EEG believes that DOE must rigorously evaluate the effect of any proposed changes and should realize that full-scale tests may be necessary in order to prove the adequacy of the TRUPACT-II design .
89
- 6. CONCLUSIONS AND RECOMMENDATIONS 6.1 Conclusions Although DOE stated in the WIPP Final Environmental Impact Statement (FEIS) that the transportation of wastes to the WIPP would comply with the regulations of the U.S. Department of Transportation and the corresponding regulations of the U.S.
Nuclear Regulatory Commission, TRUPACT-I was designed in violation of NRC packaging regulations (10 CFR 71) on two specific counts:
.1.. Double containment was not provided as specified in 10 CFR Part 71.63 for solid material containing more than 20 Ci of plutonium:
- 2. The package was designed to provide continuous venting (through HEPA fil.ters) from the storage cavity to the environment which is prohibited in 10 CFR Part 71.43(h) as well as in 49 CFR 173.413. A principal part of the venting issue is the problem of hydrogen gas generation in THU wastes.
An additional issue is the DOE intent to allow shipment of up to 12,020 Ci of CH-THU Waste in a TRUPACT.
6.1.1 Double Containment EEG estimates that the lack of double containment will increase the external radiation dose to the public and occupational workers by about 30% during normal transportation. Although the decreased population dose resulting from double containment was not large (about 90 person-rem during the project lifetime) it 90
is an incidental benefit that would accrue from meeting the regulation.
The principal advantage to double containment is in drastically reducing the latent cancer fatalities (LCF) that would occur if a Severity Category VII or VIII accident were to occur. For example, an average Savannah River Plant (SRP) shipment involved in a Category VIII accident would result in about 20 LCF with the current design and only about 8 LCF with double containment.
Also. with single containment the maximum individual dose from a Category VIII accident involving the maximum proposed load could lead to early acute health effects.
Another advantage in double containment is the drastic decrease (from 12 to 0.02) in the expected number of radionuclide release accidents. All release accidents incur significant monitoring costs and the larger releases can cost millions of dollars for decontamination and waste disposal. Also, any release accident will cause an increase in public perception of transportation accident risks. even if there are no significant public doses received.
o .1. 2 Continuous Venting and Gas Generation Continuous venting was incorporated into the TRUPACT design in
. 19 for the expressed purpose of eliminating possible package fatigue failure due to cyclical pressure changes. However.
continuous venting compromises the integrity of a CH-*TRU package because i t provides a pathway for release of radionuclides to the environment in event of filter malfunction. In addition.
the package may be more susceptible to failure around the vents if a severe transportation accident occurs.
91
Most of the CH--TRU wastes destined for WIPP produce some gas through radiolysis and processing the waste into a concrete matrix does not eliminate hydrogen generation. Therefore. some gas producing wastes will be shipped to WIPP.
There are uncertainties in predicting gas generation rates in individual Type A packages and in determining how the rates decrease with time after purging. However. experimental data produced to date indicate that venting alone will maintain hydrogen concentration below 4% in only very-low-curie content packages. Modeling results also suggest that a vented TRUPACT would not reach a 4% hydrogen concentration with such low curie packages within a reasonable shipping time. However. modeling data also suggest that a substantial number of the existing waste packages could not maintain hydrogen concentrations below 4% and i t is questionable if the TRUPACT with high-curie loads could be transported in 30 days without exceeding this level.
Alternate strategies for controlling gas concentrations exist.
It appears that a combination of administrative controls and use of hydrogen-oxygen recombiners or hydrogen getters in the waste package is probably a more reliable system for hydrogen control than venting.
6 . .1.3 High-Curie Shipments The proposed TRUPAC'l' maximum load of 12. 020 Ci of americium--
enhanced wastes and 11,200 Ci of heat source waste contains 357 times the plutonium-equivalent curies used in determining the "bounding" transportation accident consequences in the WIPP FEIS. This leads to estimated releases of 40 (Category VII accident) to 200 (Category VIII accident) times those projected in the FEIS.
92
A TRUPACT shipment with the maximum heat source load could contain about a.7 times the inhalation toxicity of a spent fuel assembly being transported by truck. A Category VII accident is estimated to release 230 times the PE-Ci of a credible worst case spent fuel accident. A double contained TRUPACT would release only about one-fifth as much.
The proposed maximum loads are not necessary to ship CH--TRU wastes to WIPP. By proper load management i t would be possible to ship all. Savannah River Plant wastes with a maximum load of about 2.000 PE-Ci. Maximum loads at other facilities could be much less.
6.2 Recommendations
- 1. The present TRUPACT-I design should not be certified for transporting CH-THU wastes to WIPP.
- 2. The TRUPACT-I design, without continuous venting. should be certifiable for transporting up to 20 Ci of plutonium per shipment. This limit would give a PE-Ci release in a Severity Category VII accident similar to that from a spent fuel shipment.
- 3. The TRUPACT should be redesigned to include double containment and eliminate continuous venting. Our understanding is that the current DOE proposal for the TRUPACT-II design incorporates these recommendations.
- 4. DOE should continue research to better define the gas generation problem and investigate the application of available technology for hydrogen gas control by hydrogen-oxygen recombiners and by hydrogen getters. A more 93
positive administrative control system should als6 be developed.
- 5. The maximum permitted load in a doubly contained TRUPACT should be set at about 2.000 PE-Ci. This limit would allow, by load management. the shipment of all stored wastes at all of the storage sites in 36 drum (or more) shipments and would reduce the estimated release in a Category VII accident to about 25 times that expected from a credible worse case spent fuel accident.
- 6. DOE should amend their 9/9/B3 Order 5480.3 and require the shipment o.f plutonium bearing waste to meet the NRC 10 CFR 71 requirements of double containment.
94
References
- 1. U.S. Department of Energy. "Final Environmental Impact Statement Waste Isolation Pilot Plant," DOE/EIS 0026. UC-70 October 1980 Vol I and II.
- 2. Tappen, J .. Fredrickson, c .. and Daer. G .. "Preliminary
.. Radiological Analysis of the Transportation of Contact Handled Transuranic Waste Within the State of New Mexico."
WTSD-TME-002. Revision 1. June 1983.
2a. Cohen. s .. "A Note on the Implications of DOE Exceptions to 10 Clt~R Part. 71 in the Design of the TRUPACT-I Shipping Container." report .to EEG by SC&-A, Inc .. April 1985.
- 3. IAEA Safety Standards. Safety Series No. 6, "Regulations for the Safe Transport of Radioactive Materials." 1985.
- 4. 49 CFR Part 171. "Requirements for Transportation of Radioactive Materials." Code of Federal Regulations Title 49.
- 5. Eakes. R.G., et al .. "THU Waste Transportation Package Development." SAND 80-0793, TTC-0085.
- 6. 10 CFR Part 71 Federal Register Vol 39. No 117. June 17.
1974. Title 10. Part 71.
- 7. 10 CFR Part 71. "Rule to Achieve Compatibility with the Transport Regulations of the International Atomic Energy Agency (IAEA) ." Code of Federal Regulations. Federal Register Vol. 48. No. 152. August 1983.
- 8. December 17. 1979 letter from Ruth Clausen, Assistant:
Secretary for Environment. US DOE to Lee v. Gossick.
Executive Director for Operations. US NRC.
- 9. December 18. 1979 letter from R. B. Pope. Sandia Laboratories. to Secretary of the Commission. US NRC.
- 10. U.S. Department of Energy Order 5480.1 Change 3, May 19.
- 11. "Andersen. J.A., et al .. "Peer Review of the Preliminary Design and Program Interface for the Transuranic Waste Package Transporter (TRUPACT) . " SAND-2405 ,Tune 1982.
Specified External Distribution Only.
95
- 12. Pope. H.B .. et al., "Design Team Response to Peer Review of the Preliminary Design for the Transuranic Package Transporter." SAND 82-_t48. Specified External Distribution Only.
- 13. July 29, 1985 letter Robert H. Neill. EEG to Joseph McGough, DOE.
- 14. Neill. R.H. and Channell, J.K .. "Potential Problems from Shipment of High-Curie Content Contact-Handled Transuranic Waste to WI!?P." EEG-24. August 1983.
- 15. U.S. Department of Energy Draft Order 5480.lA Change 3.
Jul.y 28. 1983.
- 16. U.S. Atomic Energy Commission 4/.18/84 Enclosure A. Title 10, Part 71. Chapter .l.
- 17. National Academy of Science. National Research Council.
"The Effects on Populations of Exposure to Low Levels of Ionizing Radiatio~: 1980." July 1980.
- 18. EGG, Idaho. Inc. "INEL THU Waste Presentation to the Environmental Evaluation Group from the State of New Mexico." November 1983.
- 19. Harvill. J.P., "Preliminary Radiation Dose Assessment to WIPP Waste Handling Personnel," WTSD-TME-009. February 1985.
- 20. U.S. Department of Energy. Waste Isolation Pilot Plant Safety Analysis Report.
- 21. U.S. Nuc.lear Regulatory Commission "Final Environmental Impact Statement on the Transportation of Radioactive Material by Air and Other Modes." Office of Stahdards Development. NUREG-0170, Vol. I. December 1977.
- 22. Madsen. M. . Wilmot. E. . and Taylor. V. . "RADTRAN II User Guide." Sandia National Laboratory. SAND 82--26. February
.1983.
- 23. Woolfolk. s.w .. "Preliminary WIPP Transportation Analyses." W'rSD-TME-002. April 1983.
- 24. Sandoval. H.P .. Apple. M.A. and Grandjean. N.R .. "The Fraction of Waste Contents Released from 55-Gallon Drums to the TRUPACT-I Cavity During Type B Package Testing,"
SAND 84-2645 (TTC-0537). May ~985.
96
- 25. C.lements. T.L. and Kudera. D.E .* "THU Waste Sampling Program: Volume I - Waste Characterization," An Informal Heport. EGG-WM-6503. September 19B5.
- 26. U.S. Department of Knergy. "Spent Fuel and Radioactive W~ste Inventories. Projections and Characteristics."
DOE/RW-0006 Revision 1. December 19B5.
.. 27 . Finley. N.C .. et al .* "Transportation of Radionuclides in Urban Environs: Draft Environmental Assessment."
NUREG/CR-0743 (SAND 79-0369). July 1.980.
- 28. Wilmot, E.L .. et al .. "A Preliminary Analysis of the Costs and Risk of Transporting Nuclear Wastes to Potential Candidate Conruercial Repository Sites" SAND 83-0Ub7, June 1.983.
- 29. Shefelbine. Henry c .. "Preliminary Evaluation of the Cl1aracteristics of Defense Transuranic Wastes." SAND 78-1850. November 1978.
- 30. U.S. Department of Energy. "Draft Environmental Assessment Davis Canyon Site. Ut~h." DOE/RW-0010. December 1984.
31.. Burgoyne. R.M. et al .. "THUPACT Draft Safety Analysis Report for Packaging (SARP)," SAND B3-7077/GA-A16B60.
Novernller. 1.984.
- 32. ~olec~e. Martin A .. "Gas Generation from Transuranic Waste Degradation: Data Suh~ary and Interpretation." SAND 79-1.245. December 1979.
- 33. Clements. T.L. and Kudera, D.E .* "THU Waste Sampling Program: Volume II - Gas Generation Studies." EGG-WM-8503, September 1985.
- 34. Zerwekh. Al. "Gas Generation from Radiolytic Attack of THU Contaminated Hydrogeneous Waste." LA-7674-MS, JU;ne 1979.
- 35. Kosiew+/-cz. Stanley T., et al .. "Stud~es of Transuranic Waste Stor~ge Under Conaitions Expected +/-n the Waste Isolation Pilot Plant (WIPP). LA-7931-PR Progress Heport.
January 1980.
3H. Science Applications International Corporation. "A Theor~tical Model for Hydrogen Buildup and Dissipation,"
Draft Report. November 1985.
97
- 37. Henrie, James 0 . , et al .. "Hydrogen Control in the llandling, Shipping and Storage of Wet Radioactive Waste,"
RHO-WM-EV-9-P.
- 38. Kazanjian, A.R., "Gas Generation Results and Venting Study for Transuranic Waste Drums," RFP-3739.
- 39. Kazanjian. A.R .. "Radiolytic Gas Generation in Plutonium Contaminated Waste Materials." RFP-2469. October 1.976.
- 40. U.S. Nuclear Regulatory Commission. "Clarif.ication of Conditions for Waste Shipments Subject to Hydrogen Gas Generation." IE Information Notice No. 84-72. September 1.984.
41.. U.S. Nuclear Regulatory Co~nission, Transportation Certification Branch Approval Record. Combuutible Gas Mixture. May 22, 1.985.
42.
43.
Trujillo. R.E. and Courtney, R.L .* "Organic Hydrogen Getters." Journal of Materials Science. 12(1977)937-943.
Baxter. Richard G .. "Description of Defense Waste Processing Facility Reference Waste Form and Canister,"
Savannah River Plant. DP-1.606 Revision 1.. August 1.9B3.
- 44. Wilmot, Edwin L .. "Transportation Accident Scenarios for Commercial Spent F"uel." SAND B0-21.24. February 1.9.
- 45. U.S. Department of Energy. "Draft Environmental Impact Statement - Waste Ioolation Pilot Plant," DOE/EIS-0026-D.
April 1.979.
- 46. U.S. Department of Energy. "Assessment of Transuranic Activity Linlits for WIPP THU Waste." WTSD-TME-062, April 1.9U5.
'17. Halverson. T.W. and Co.Le. L.T .* "Optimization of Waste Operations at WIPP," Waste Management 'BB. Tucso~. AZ, March 1.986.
- 48. Ziegler. D.L. and Wilkinson, F.D .. "An Assessment of Radiolytic Gas Generation in Waste Containers For Transportation Considerations," RFP-3735. September 1.9B4.
98
APPENDICES 99
APPENDIX A Modeling Hydrogen Generation and Dissipation in THU Waste Packages
APPENDIX A Modeling Hydrogen Generation and IHssipation in THU Waste Packages liydrog1:.~n: gene.ration by radiolys+/-s in the waste matrix of THU wtiste ~~ckages c~rt lead to the formation of potentially flammable concentrations in the void spaces unless properly controlled. At tbe pr~~ent time there appears to be an ina<lequgte experimental data b~se covering a wide range of waste categories, ctirie loadihgs. and varieties of waste packages on w.hic.h t.o build programmatic and regulatory planning. Under
- these circU~stances i l is necessart to rely on modeling the behrivior of hyd~ogen ih enclosed volumes to extend th~ present experimental data base to include other possible combinations of w~st~s fdr~. cu~ie loading, hydrogen getters. pack~ge design, etc.
T.he EEG mode.Ling effort is based on a generalized model of THU waste container hydrogen production and 1removal developed by SAIC for DOE (Ref 36). The SAIC model was modified to accept input of sp~bific verit characteristics (effective vent hole radius and filt~r thickness) and flow through the vent was presumed to be diffusion dominated. The geometry of the containers was restricted to two volumes for simplici~y. The general mathematical formulation of the model follows the SAIC strategy except foL the venting aspect and the specific represerttatibn of a decaying G (gas) due to matrix effects.
For an exhaustive discussion of the mathematical formulation of the model, reference should be made to the SAIC report (Ref 3G).
Here, an abbreviated discussion will be given, with emphasis on aspects of the EEG model which are different from the SAIC version.
The EEG two-region model assumes an inner Type A waste container with a given void volume placed inside the TRUPACT, which has its own specific void volume dependent on the number of drums and dunnage volume used in loading (typically 13.6m3
- but could be as little as 4m 3 ). For simplicity i t is assumed that each drum releases hydrogen (if vented) into a proportionate share of the avaiiable TRUPACT void. The gases produced in the waste are assumed to quickly migrate to the accessible void of the waste container and then diffuse into the TRUPACT. and then to the outside if both are vented.
The rate of production of hydrogen and other gases is dependent on the alpha-curie loading of the waste and the G(gas) and G(hydrogen) values. Since a two-component model of hydrogen generation as a function of'time was found to be indicated by our review of the data. our model has the form
-Kt H(t) = H e 0
+ H .(moles/hr) 1 Where H is the production rate at time t = 0 and K is the decay 0
constant for gas generation. A similar expression describes the production of other gases such as co 2
- once released to the void v*olume. the hydrogen. concentration is computed as a molar fraction of the total number of moles in the void.
C(t) = N(t)/M(t)
Where N(t) is the number of moles of hydrogen and M(t) is the total number of moles in the void at time t. The addition of one mole of hydro.gen to a particular volume increases both N ( t) and M(t) by one. but the addition of a molecule of another gas increases only M(t) by one. If the void is vented so that the
.. inventory is constant. then the addition of a mole of any gas will result in a mole being released. The probability that the released mole is a mole of hydrogen is given by the relative concentration of hydrogen. C(t). Clearly. this assumption .is reasonable only if complete and instantaneous mixing always occurs (at least to the level of resolution of the smallest time step in the calculation. about one hour}.
The flow of hydrogen out of a vented container is presumed to occur through a vent filter. Rather than assuming "plug" flow (i.e .* a volumetric rate defined by a hole area and average velocity). i t is assumed that the process is diffusion dominated at the pressures and flow rates antici-pated. The hydrogen flux through a f.ilter is represented by the relatibn:
Where
'I.
- P = Pressure in container R = Ideal gas constant T =Temperature.deg K.
FA= Filter area FX = Filter equivalent thickness (C2-C1) = Hydrogen concentration differences D = Diffusion coefficient for hydrogen in air
The equivalent thickness is estimated following the approach of Ziegler (Ref 4B). based on the characteristics of the vent Where FX1 = Hole thickness FA1 = Hole area Fl= Hole porosity FX2 = Filter thickness FAZ= Filter area F2 = Filter porosity In the case of sealed containers. the pressure is calculated at each time step in the calculation by averaging changes in temperature and total gas inventory. and converted into estimated changes in concentration using the ideal gas law.
th In general. the time rate of change in hydrogen in the i container is given by dN. = H 1.. ( t) -R 1.. ( t) + [ V .1.-
- 1. ( t) - V . ( t) ) +Q * ( t) 1 .1. .1.
dt Where H, (t)
.1.
= Hydrogen generation rate R. (t)
.1.
= Hydrogen removal rate by absorbers (if present)
V. (t) = Hydrogen flux due to
.1.
diffusion through vents Q. (t) = Hydrogen flux due to temperature and
.1.
pressure changes
APPENDIX B
- Discussion of Propargyl Ethers as Hydrogen Getters with Respect to Nuclear Waste Disposal
DISCUSSION OF PROPARGYL ETHERS AS HYDROGEN GETTERS WITH RESPECT TO NUCLEAR WASTE DISPOSAL by M.P. Neary. PhD June 30. 1.986
Experience to date with the model indicates that by using actual filter characteristics for the Rocky Flats Plant small bung filter and the reported percent void, hydrogen fraction and curie loading for a set of experimental drums. i t is possible to approximately match the reported hydrogen concentration changes with time in both vented and unvented cases. The "free" variable in this approach is G(gas). It was as a result of such a fitting-process that the two-component decaying G(gas) concept emerged. An alternative approach based on a fixed G (gas) concept and another time varying parameter may possibly also be found to explain the observed data. But the present approach offers the considerable advantages of having successfully predicted independent observed time-varying G(gas) and requiring
- a minimum of ad-hoc parameter value choices in the initialization of the model.
A BASIC language version of the model used in these simulations will be available to interested parties .
INTRODUCTION Considerable concern by the New Mexico Environmental Evaluation Group is centered on radiolyticly produced hydrogen in the TH.lJPAC'l' shipping containers which are scheduled to be used to 1
transpo~t transu~anic waste to WIPP. < > It is not only possible but probable thgt radiolytic or catalytic hydrogen will be produced by combination of certain transuranic waste and other organic ch~micals abun~ant with hydrogen. This would be a proble~ if solutions. ~queous or organic, of alpha-emitting actinides were allowed in WIPP storage containers. According to (2) one source. a build-up of hydrogen gas to 4% by volume or more in the containment system constitutes an explosive hazard .
NRC has done work to confirm the older lower explosive limit shown above. Their findings show that 10 to 12% by volume hydrogen in air is a more practical lower limit for explo-
. (3) sion. Given either limit i t is certainly true that a violent explosion can result from low concentrations of hydrogen in air.
Explosions occurring in this way would probably cause little direct damage to humans: however. the accidental dispersal of transuranic wastes could cause considerable indirect losses.
Means of removing gaseous hydrogen from a mixture of gases exist
- and are sufficiently efficient when intelligently used to obviate concern for the generation of explosive levels of hydrogen within nuclear waste transportation and storage con t.ainers. Such means include: electrical recombiners.
catalytic recombiners. and organic getters. Because the first two produce water their use would be forbidden. The subject to be considered here is organic getters and. in particular. the gettering properties of propargyl or acetylenic compounds.
First. some background information on the explosive character of hydrogen will be considered.
BACKGROUND Of the diatomic gases. hydrogen is the sma1lest (occupies the least volume per mole). has the greatest mean free path (largest distance or longest time between collision) and the greatest velocity at STP.*< 5 > The diffusion rate of hydrogen in air.
which is related to the square root of the inverse ratio of the densities of hydrogen and air. is the greatest of all diatomic gases. Because of these physical properties. hydrogen is relatively fast to be uniformly distributed throughout a volume when driven by diffusion alone. Mixing processes driven by heat or agitation serve to hasten or maintain uniform distribution.
Mixtures of hydrogen and a variety of other gases are flammable/explosive. They include oxygen. halogens. and nitric 4
and nitrous oxides. < >
The terms .,flammability" and "explosive limits .. are generally loose. Flammability may refer to the relative ability of the material to burn exothermally in the presence of oxygen. From this viewpoint. pure hydrocarbons are more flammable than hydrocarbons containing oxygen which. in turn. are more flammable than those containing halogen. Alternatively.
flammability may refer to the volatility of a compound.
Flammability may be influenced by explosive limits of mixtures of air and combustible gases. Thus. a mixture of n-pentane in air will explode only when the percent by volume of pentane is between ~.5 and 7.5. At higher or lower concentrations no explosion will take place on application of spark or flame. or
~gnition. At the other extreme. hydrogen is explosive in the range of 4 to 74 percent by volume in air! (Z.S) Ignition is required for both combustion and explosion. hence ignition
- standard temperature and pressure
temperatures relate to the ease .of initiation of either combustion qr explosion. The ignition temperature of hydrogen within the explosive limits cited above is 530°c in air. <4 >
Hence. the activation energy for the formation of water from hydrogen and oxygen (the ignition of hydrogen in air) is fairly high. taking the:ignit.i.on temperature as a measure of the
- activation energy. *Active surfaces of certain metals may greatly lower the activation energy and* hence the ignition tempe~at~re. (S)
- Most workers agree that the difference between a conflagration and an explosion of gas-air mixtures is related to the burning velocity e~pressed in centimeter~ per second. The maximum burning velocity of hydrogen-air mixtures of b~tween 4 and 74 percent by volume is 4,o cm/sec. the greatest or nearly so of any combustible gas-air mixture by a factor of ten. By comparison. n-pentane. which forms a flammable/explosive mixture with air at .1.5 to 7.5 percent-by volume. has a maximum burning velocity of 43 cm/sec! It can be concluded that hyd~oien-air
. .. (6).
mixtures *can explode with unusual violence.* _
BACKGROUND
SUMMARY
The minimum explosive limit of hydrogen is very low. The activation energy for hydrogen ignition can be drastically lowered by adsorption of hydrogen onto certain metal surfaces.
Ignition of hydrogen-air mixtures within the explosive limits t results in a particularly powerful. and therefore destructive.
explosion. The radiolytic generation of hydrogen from nuclear
- Most hydrogen research laboratories have either blowout walls or a roof that is not fastened.
waste within containers is expected in amounts that could reach explosive levels.
The use of a hydrogen getter that operates continuously for long periods of time. that does not form water or pyrophoric compounds. that is effective. efficient. generally inert and nontoxic, and that is small in size and inexpensive is highly desirable.(?)
PROPARGYL HYDROGEN GETTERS Gettering Mechanism Generally. an unsaturated organic compound can take up (getter) hydrogen and its isotopes when an active metallic surface is present. Such metals are those found in Group VIII of periodic table. (B) If a dry mixture of the getter and active metal were suddenly introduced into a gaseous mixture of hydrogen and air, and the volume percent of hydrogen were within the explosive limits, ignition and explosion would occur without significant gettering. This is due to the vast difference in the rate laws for gettering and the competing explosive reaction. The explosion occurs because of the presence of active metal surfaces. However, if the hydrogen is slowly introduced into a mixture of getter, active metal, and air. the getter reaction will limit the buildup of hydrogen. thus keeping the overall r
volume percent of hydrogen below the lower explosive limit. The specific pathway by which gettering proceeds is specified below.
Although hydrogenation (gettering) is an exothermic process. the reaction does not take place spontaneously because the amount of energy required.to break a pi bond in the olefin or propargyl compound, or a sigma bond in hydrogen. is too large. The function of the active metal (catalyst) is to lower this
activation energy stepwise so that the activation energy of each is much lower than that required for thermal breaking of the pi or sigma bonds.<B.B)
... Metals such as platinum. pallaq+um. silver. nickel. a~d copper strongly adsorb hydrogen and unsaturateq molecules. The atoms in the metal surface have unpaired electrons which can interact wit~ the electrons in the relatively exposed sigma orbital of the hydrogen m9l~cule and the pi orbital of the double or triple bond. Hydrogen thus a~sorbed can dissoci~te. yielding adsorbed hydrogen atoms. This is due to a great reductio? of activation energy for sig~a ~ond ~reakin~ of adsor~ed hydrogen. Th~ alkene or alkyne can form an adsorbed free bi-radical on such a surface. For the olefin. '
reaction of the free bi-radical and two hydrogen atoms leads to a saturated molecule and desorbtion.
- ., ' l * *.
For the a~ky~e. f~ur ~Ydfogen ~to~~ react before saturation and desorbtion. Beca~se of th~ various step~ in the reaction involving unpaired electrons and weak bonds. none has a high activa~i~n ~~~rg;*. CB* B) , . .~ . . "* , *i '
In order that reaction occur between the adsorbed molecules.
they must approach each other closely and be properly oriented.(~) Not only the size.an~ structure of the reactants but also the crystal structure o~ the surface of the catalyst determines these space relationships.
~ . ' . - '
The reverse reaction is not possible in view of both energetic. entropy and ster~o
' I * * ( B . 1.J.) .. * . . , ' ' . ..
considerations. ,* * ~tis evi~e~t t~at the o~timum conditions t and type of catalyst will v~ry for ev~ry different pair of reactants. Fortunately. ~rdro~~~ation catalysts have bee~
developed which show high activity for a wide range of propargyl compounds: hence catalytic hydrogenation is an eminently
. . . . .. CB) . . .* . .
practical process *. ,
Hydrogenation Catalysts A few of the most effective metal catalysts for hydrogenation of propargyl compounds are listed below:
Heterogeneous Hydrogenation Catalysts (7.B.1.1.)
Platinum black (unsupported)
Platinum black/carb_on Platinum black/calcium carbonate Platinum black/asbestos Platinum black/alumina Palladium black (unsupported)
Palladium black/carbon Palladium black/calcium carbonate Palladium black/asbestos Palladium black/alumina Homogeneous Hydrogenation Catalysts (7.9.1.0)
Noble metal chelates Organometallic complexes (i.e ** dichloro-bis (triphenyl-phosphine) platinum or palladium
~or the supported catalysts listed above under "Heterogenous Hydrogenation Catalysts". the term "black" refers to the most finely divided form of element. The elements* percent by weight r'
supported on the various substrates ranges from 1.% to 20%:
however. 5% by weight gives the best results. (1.1.) Even though other metals in Group VIII of the periodic table can be used as catalysts for hydrogenation. platinum and palladium are usually preferred because of the rapid hydrogenation reactions they catalyze. Other less expensive metals from Group VIII may provide sufficiently rapid catalysis. In any case. the catalyst-propargyl compound weight ratio is in practice adjusted
to provide both the desired capacity and a hydrogenation rate that exceeds the production by a margin of safety. For example.
a propargyl ether-catalyst formulation between 60 and 65% by weight of organic gives 90% hydrogenation in 60 minutes at a rate of 14.4 min H per mole of organic getter per sec. The 2
catalyst used was 5% by weight supported on calcium
- (7 9) carbonate. ' The homogeneous catalysts listed above require hydrogenation re~ctions to be carried o~t in solution. The best advant~g~ of such ari approach would be realized 6nly when the hydrogen bearing gas mixture is passed or bubbled through the 11 solution. < > This means of limiting hydrogen in a closed voluine will not be discussed further here .
- Propargyl Organic Compounds Numero~s off-t~e-~helf propargyl ~ompounds are available.
range in physical state from gas to liquid to solid. And as They their molecular weight .increases the compounds te'nd to soli.ds and t6 act less ~y~o~horically in a hydrogen and oxyjen atmosphere. Likewise. flammability. toxicity. and other irritatin~ ~roperties diminish as molecular weight increases.
The overall ~oxicity 6f pro~ar~yl ~ompounds generally depends more on substituent groups than on the acetylenic character. In t
- general propargyl compounds are unreactive alone unl*ess in the presence of a catalyst. Solid propargyl compounds generally are more versatile in the subject application.< 11
- 12 )
The reactivity of propargyl c~~pounds is divided into two categories. one concerned with the acetylenic character and the other concerned with the substituent groups.*< 12 > A third
$substituent groups are those chemical moieties introduced on the starting materials or later to either make the synthesis easier or impart specific physical properties to the product.
category could be considered in which the effect of substituent groups on the acetylene group is considered. For purposes under discussion. the last two categories are the most important because a wide range of substituted propargyl compounds are available. Thus. side reactions involving substituent groups and the environment can be avoided by selection of the appropriate propargyl compound. Substituent groups near (i.e .*
within a carbon atom) the propargyl group usually reduces its capacity for and rate of hydrogenation.< 7 . 9 > This is not surprising in view of the adsorption and bi-radical formation step described above. The propargyl group does not react with fixed gases such as oxygen. nitrogen. carbon monoxide. carbon dioxide. and methane except under extremes of temperature and pressure (i.e .. greater than 150°c and 2 Atmospheres).
Therefore. these gases do not compete or interfere with hydrogen uptake in a mixture. <11 > Likewise. a moist. acidic. or corrosive atmosphere will not react with a propargyl compound such as diphenyl propargyl ether (DPPE). particularly if the DPPE-catalyst solid mixture is not immersed in such a liquid.
0 At elevated temperatures (ca 120 C) many propargyl compounds will crosslink. <9 >
The three selection rules for the appropriate propargyl compound are: low or no substituent reactivity. a solid over the temperature range of use. and the lowest molecular weight with the greatest molar capacity for hydrogen uptake.
compound that has been most useful is the dimer of 1.6-The propargyl
- diphenoxy-2.4-hexadiyne or diphenyl propargyl ether. DPPE. DPPE is a solid up to B0°c. and when combined with a hydrogenation catalyst. may be used with equal efficiency to getter hydrogen
. l pressure as low as i4 0- 6 atmosp h eres an d up a t a h y d ro g en par t ia to 2 atmospheres. Whether or not DPPE may be used at low temperatures depends on the rate of hydrogen generation (i.e .*
,,,,. 0 :1 if the rate is low. DPPE can be used to -5 0 c) .< 9 > Even though DPPE melts at about ao c 0
and cross linkage may be initiated at about 120°c. hydrogenation still occurs. At 150°c further hydrogenation is limited by complete cross linking. The maximum efficiency for hydrogenation is obtained between -4°c and 71 cc. C7.9.Hl.11)
Because DPPE is a solid below ao 0 c. it has virtually no vapor pressure beiow that temperature and no flammability. When exposed to direct flarrie. however. the compound will burn. It is estimated that DPPE mixed with the hydrogenation catalyst will be effective for io years at. 50°c and lose less than .10% of the propargyl dompourid due to vaporization or side reactions with impurities. <9 >
Formulation The DPPE and catalyst are usually combined in a suitable soivent so that DPPE is dissoived. The resulting slurry can be dried in a vacuum oven. painted onto a surface and dried or adsorbed onto another substrate. as desired. The DPPE coating on the catalyst thus forms a barrier which reduces or obviates the hydrogen-oxygen reaction at the catalyst surface. Because hydrogen easily diffuses through the coating and oxygen does not. very t
- little or no water is* t h us f orme d . (7.11)
The surface area of the coated catalyst affects the ini~ial rate of hydrogenation anct*has little to do with the total.capacity.
In fact. for DPPE. 65% by weight on catalyst (5% palladium black on calcium carbonate) hydrogenates* to .100%. <7
- S) The uptake rate of this formulation is .14.4 mm H /mole of. DPPE/sec. (S) 2 Hence. if ~he hydrogen partial pressure is increasing at 14mm/hour * .1/3600 of a mole of.DPPE-catalyst would hydrogenate
at a rate equal to the production rate. Given the production rates of hydrogen. simple calculations predict the quantities of DPPE-catalyst needed. bearing in mind that the uptake is 100%
efficient and 4 moles of hydrogen are taken up per mole of DPPE 3
(MW=262 and molar volume= 1B3 cm ).
Cost The off-t.he-shelf prices of catalyst and DPPE are generally not high (i.e. . DPPE costs approximately $1. 00 / gram and pa.lladium black on activated carbon (5% by weight) costs approximately
$l. 50 / gram) .
However. i t is expected that economy of scale will reduce both costs substantially. In the case of DPPE, a low price of
$0.25/gram could be anticipated along with $0,75/gram for palladium black on activated carbon (5% by weight). Other less expensive metals which catalyze gettering. albiet at a lower rate. may sti. l l b e appropriate
. ( si. l ver, f.or examp l e ) . ( 11. 13)
Use Once fabricated, the DPPE-catalyst solid mixture can be disposed in a variety of ways. Coatings on surfaces in the container and/or loose placement in a dry container is acceptable.< 11 )
The mixture can be disposed between two porous plugs or filters and fixed in the top of the storage drums or the TRUPACT vent.
Because vented containers are expected to "breathe". by locating the getter near or in the vent. effective gettering is expected.
Whether or not the getter should be disposed at various locations in the TRUPACT cavity depends on the nature of the load of storage drums and how they are vented and if a getter is disposed within them. Clearly. if each drum that is likely to
produce hydrogen is equipped with a getter. no further gettering should be required in the TRUPACT. However. if the barrels are vented into the TRUPACT and are not equipped with getters~ the TRUPACT can and should be so equipped with an appropriately scaled getter system.
summary Propargyl getters are effective in maintaining a very low (less than 1 ppm) hydrogen concentration in a closed space. Their use requires no power. generates no water. occupies a very small volume. and last 10 years at 50°c. Their cost is modest. they are no-toxic and non-pyrophoric. The above characteristics recommend propargyl getters in most circumstances.
l
BIBLIOGRAPHY
- 1) Neill. R.H. and J.K. Channell. "Potential Problems from Shipment of High-Curie Content Contact-Handled Transuranic
.... (CH-THU) Waste to WIPP Report EEG--24. Environmental 0
Evaluation Group. Environmental Improvement Division.
Health & Environment Department. State of New Mexico.
August 1983.
- 2) Hodgman. C.D., Editor-In-Chief. Handbook of Chemistry and Physics 42nd Edition The Chemical Rubber Publishing Co.
(1960)
- 3) Neill. R.H .* private communication
- 4) Dean. John A .. editor. Lange's Handbook of Chemistry 12th edition, McGraw-Hill Book Company. 1979 .
- 5) 6)
The Merck Index, 9th edition, Merck & Co .. Inc .. Rahway, N.J .. USA, 1976.
Noller. Car J_. Chemistry of organic Compounds. W. B.
Saunders Co .. Philadelphia, 1957.
- 7) Andersoft~ D.R .. et al. U.S. Pateftt *3.896,042 (1975).
- 8) Pearce, R .. et al. Catalysis and Chemical Processes.
Halsted Press. John Wiley and Sons. New York. 1981.
- 9) Trujillo. R~E .. et al J. Mater. Sci .. 12(1977)937.
- 10) Courtney. R.L .. et al J. Mater. Sc+/- .. 12(1977)175.
1:.1) Neary. M.P., Los Alamos National Laboratories Classified Data . .1980-81.
1.2) Streitwieser-. Andrew, Jr. a:nd Clayton H. Heathcock.
Introduction to 0rganic Chemistry. Macmi.llan Publishing Co. * :Dnc. New York. .1'98:f.
I..
1:3) chemical: Dynamics Corp. So Plainfield', NJ 07080
Environmental Evaluation Groups Reports (Continued)
EEG-11 Channell. James K., Calculated Radiation Doses From Radionuclides Brought to the Surface if Future Drilling Intercepts the WIPP Repository and Pressurized Brine, January 1982.
EEG-12 Little. Marshall s .. Potential Release Scenario and Radiological Consequence Evaluation of Mineral Resources at WIPP. May 1982.
EEG-13 Spiegler, Peter. Analysis of the Potential Formation of a Breccia Chimney beneath the WIPP Repository. May 1982.
EEG-14 Zand. Siavosh M ** Dissolution of Evaporites and Its Possible Impact on the Integrity of the Waste Isolation Pilot Plant (WIPP) Repository. (Draft).
Bard, Stephen T .. Estimated Radiation Doses Resulting if an Exploratory Borehole Penetrates a Pressurized Brine Reservoir Assumed to Exist Below the WIPP Repository Horizon, March 19B2.
EEG-16 Radionuclide Release. Transport and Consequence Modeling for WIPP. A Report of a Workshop Held on September 16-17, 19, February 1982.
EEG-17 Spiegler, Peter, Hydrologic Analyses of Two Brine Encounters in the Vicinity of the Waste Isolation Pilot Plant (WIPP) Site. December 1982.
EEG-18 Spiegler, Peter, Origin of the Brines Near WIPP from the Drill Holes ERDA-6 and WIPP-12 Based on Stable Isotope Concentrations of Hydrogen and Oxygen, March 1983.
EEG-19 Channell, James K ** Review Comments on Environmental Analysis Cost Reduction Proposals (WIPP/DOE-136) July 1982. November 1982.
EEG-20 Baca, Thomas E .. An Evaluation of the Non-radiological Environmental Problems Relating to the WIPP, February 19B3.
EEG-21 Faith, Stuart, et al .. The Geochemistry of Two Pressurized Brines From the Castile Formation in the Vicinity of the Waste Isolation Pilot Plant (WIPP) Site. April 1983.
Environmental Evaluation Groups Reports (Continued)
EEG-22 EEG Review Comments on the Geotechnical Reports Provided by DOE to EEG Under the Stipulated Agreement Through March 1.
1983. April 1983.
EEG-23 Neill. Robert H ** et al .* Evaluation of the Suitability of the WIPP Site. May 19B3.
EEG-24 Neill. Robert H. and James K. Channell. Potential Problems From Shipment of High- Curie Content Contact-Handled Transuranic (CH-TRU) Waste to WIPP. August 1983.
EEG-25 Chaturvedi, Lokesh. Occurrence of Gases in the Salado Formation. March 1984.
EEG-26 Spiegler. Peter. Environmental Evaluation Group's Environmental Monitoring Program for WIPP. October 1984.
EEG-27 Rehfeldt. Kenneth. Sensitivity Analysis of Solute Transport in Fractures and Determination of Anisotropy Within the Culebra Dolomite. September 1984.
EEG-28 Knowles. H.B .. Radiation Shielding in the Hot Cell Facility at the Waste Isolation Pilot Plant: A Review. November 1984.
EEG-29 Little. Marshall s .. Evaluation of the Safety Analysis Report for the Waste Isolation Pilot Plant Project, May 1985.
Dougherty. Frank, Tenera Corporation. Evaluation of the Waste Isolation Pilot Plant Classification of systems, Structures.
and Components. July 1985.
EEG-31 Ramey. Dan. Chemistry of the Rustler Fluids. July 1985.
EEG-32 Chaturvedi. Lokesh and James K. Channell. The Rustler Formation as a Transport Medium for Contaminated Groundwater.
December. 1985.
ENVIRONMENTAL EVALUATION GROUP DOCKETED USNRC
- - - - - - - - - - - - - - - - - - - - - A N E G l U A L O P P O R T U N I T Y / AFARMATIVEACTIONEMPLOYER - -
- 7007 WYOMING BOULEVARD, N.E.
SUITE F-2 *9a APR 27 P3 :36 ALBUQUERQUE, NEW MEXICO 87109 (505) 828-1003 FAX (505) 828-1062 ..
.. , r r-u DOCKET NUMBER TAFF April 22, 1998 PETITION RULE PRM .. ~
( 1,:3 FR S'3f,,.;L) @
Secretary U.S. Nuclear Regulatory Commission Attn: Chief, Docketing and Service Branch Washington, DC 20555-0001
- Re: Petition for Rulemaking Pursuant to Provisions of 10 CFR 2.802 Regarding the " Special Requirements for Plutonium Shipments", 10 CFR 71.63, NRC Docket No. PRM-71-12.
Dear Mr. Secretary:
The following comments are provided on the subject petition by Mr. Frank Falci, former DOE employee who is now President, International Energy Consultants, Inc., to relax the requirements for transportation of plutonium by deleting 10 CFR 71.63 in its entirety.
- 1. Plutonium 239 is an extremely hazardous material as evidenced by the annual occupational limits imposed in ICRP-26 (1979) of 5 x 102 Bq/year for inhalation. This amounts to 0.22 x 1o-6 grams Pu/year. Stated more graphically, a worker has received his annual limiting intake of plutonium if he inhales more than one fourth of one millionth of a gram of plutonium. ICRP 48 (1986) updated data on the metabolism of Pu and either left values of f1 unchanged or increased by a factor of 10. More recent plutonium ALI quantities published by EPA (Federal Radiation Council Report #11 , 1988) are slightly lower than the ICRP values making them even more stringent.
- 2. The WIPP Project, which is expected to receive 30,000 shipments of CH-TRU and 8,000 shipment ofRH-TRU during its lifetime, will be the principal activity impacted by the proposed petition. The WIPP Project already has the NRC certified TRUPACT-11 package for CH-TRU waste ready for use and they have not indicted that this package was placing an unreasonable burden on them.
- 3. The claims of unreasonable cost from double containment are unsubstantiated. They sound familiar to those made by DOE in the early 1980s in defense of the single contained TRUP ACT-I design . DOE contended that double containment would increase costs and decrease the payload thereby resulting in increased fatalities and injuries from highway APR 3 0 1998 Acknowledged by card ,uuuuamWU.lUUHAHl.1.--
Providing an independent technical analysis of the Waste Isolation Pilot Plant (WIPP),
a federal transuranic nuclear waste repository.
U.S. NUCLEAR REGULATORY COM !ISS!ON RULEMAKINGS &ADJUDIC, JIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Dori ,ent 'at:.,l1cs
U.S. NRC Page 2 April 22, 1998 accidents due to the extra shipments required. TRUP ACT-II turned out to be more economical, easier to inspect for maintenance, and have a bigger payload than TRUP ACT-I (42 drums vs. 36 drums). We believe the TRUPACT-II design is an indication that the current rule has adequate technical flexibility. Modifications to the TRUP ACT-II design which have been certified by NRC (the pipe containers) or where certification is being requested (the half-pack) are further indications that the current system can be adapted as necessary.
- 4. The total life cycle cost of WIPP is $19 Billion. Transportation is estimated to cost $1 .6 Billion (Ref SEIS-11 for WIPP). The cost of building the TRUPACT is $250,000. It is estimated that 60 TRUP ACTS will be required at a cost of $15 million. That is less than
- 0. 1% of the total project cost. Hence the cost of the double containment is a trivial fraction of the project's cost.
S. The petitioner also slips in the suggestion that 71.63(a) be deleted for the sake of consistency. This section requires that plutonium in excess of20 curies (>318 FGE) must be shipped as a solid. Thus, approval to ship unlimited quantities of plutonium as a liquid is being requested. This strikes us as a reckless suggestion. Being in liquid form in single containment raises additional questions of criticality and much increased chances of liquid contamination even with non-accident transportation. The WIPP Waste Acceptance Criteria has always considered liquid in transuranic waste containers as undesirable and limits WIPP containers to less than 1% free liquids even for containers that contain less than I curie of plutonium.
- 6. Current plans to ship residues from RFETS to WIPP have required DOE to obtain certification for pipe containers that can contain 200 Fissile Gram Equivalents (FGE) of plutonium within a 55-gallon drum and 2800 FGE in a TRUPACT-11. DOE is seeking variances to Safeguard Termination Limits that would allow residues containing up to 10 wt% plutonium to be put in the pipe containers. Thus it appears that much more concentrated waste will be shipped to WIPP than was envisioned in 1987 when it was decided that all shipments to WIPP should be in a double-contained, non-vented, NRC certified package. To abandon these safety features now with wastes becoming more concentrated is a backward step.
- 7. The petition claims a great importance for preserving the sanctity of the A2 system. We view the A2 values as old approximations of a relative hazard. Other values would be used today ifNRC thought it important to update the A 1 and A2 lists. We believe the A2 values are appropriate as used but are unrelated to the special provisions for plutonium in 71. 63.
U.S. NRC Page 3 April 22, 1998 Hence no reasons have been presented in Mr. Falci's petition on the basis of health effects, unreasonable restraints on transportation or an economic burden to warrant watering down a standard that has been met.
Sincerely,
~~*
{..I Robert H. Neill Director RHN :JKC:js cc: C.R. Chappell, SFPO/NMSS
.ANN R. SMITH DOCKE TED 5.J.J4 .J.ZnaSt. NW March 20, 1998 USNRC Wtrsliinpm, DC 20015-1.151 DO KET NUMBER *9a NAR 25 A10 :54 PE I' 10 RULE PR 1 I - I~
Secretary ( (,J>F~B'3t,,:l) OFFIC~_.\ :i f:_q~;:>1 i!'Y U.S . Nuclear Regulatory Commission RULt IC _ ,, I ,I'**' r.**10 ADJUDICA LU, S STAFF Washington, DC 20555
Dear Sir:
This letter is regarding Docket No. PRM-71-12. I am opposed to changing the special regulations for transportation of plutonium. I do not believe that you should eliminate the regulations for a separate inner container for plutonium. I also do not
- believe that plutonium should be transported as a liquid, and I believe that the regulations must specifically forbid that.
I strongly believe that the regulations for transporting plutonium must not be changed, and that the changes that have been proposed result in reducing the safety of shipments of plutonium. If you change these regulations, I believe that the potential for an accidental spill of plutonium is increased and that there would be public health consequences of such an accident.
Soon there will be a lot more shipments of plutonium, since the plutonium that was generated for nuclear weapons will be shipped for disposal. I believe that it is imperative that the special regulations for these shipments should not be eliminated, so that these shipments are as safe as possible. It is hard to believe, as stated by the petitioner, that this regulation is doing such damage to our national budget--just see what one large accident would do to the national budget and the trust that the nation has placed in your agency.
Thank you for your serious consideration.
Sincerely, AnnR. Smith 5334 32nd Street, N.W.
Washington, DC 20015 AR 2 6 1998 ckn edg
,$. UCLEAR REGULATOR VV1V11YnovlON RULEMAKI GS&ADJUDICATI S STAFF OFRCE OF THE SECRETAR OF THE COMMISSIO DocumGrit Statistics
__,;;;;;,/.-.~ ~ g--
1 2 ~ "R.tD..S
DOCKETED
(!)
s~ March 20, 1998
$3.)4 .J2,u/St 9{_ 'W USNRC
'W111~on i>C 2D01$-1J$1
'96 MAR 24 P3 :27 Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Rulemakings and Adjudications Staff
Dear Sir:
This letter is regarding Docket No. PRM-71-12. I oppose changing the existing special requirements for plutonium shipments. The petitioner wants to eliminate these safety regulations because there are a large number of waste drums to be shipped! That is exactly why the regulations should be retained, and should be strengthened. Perhaps, as pointed out by the petitioner, additional radionuclides should require the special requirements of a separate inner container. I would strongly support requiring other radionuclides to require separate inner containers.
You must not consider for a second allowing liquid plutonium to be transported on the nations highways and rails! I hope that you consider the safety of these shipments, and do whatever you can to assure that the best containers, including containers with a separate inner container, are used for these shipments.
Thank you for your consideration.
Sincerely, Frank Smith 5334 32nd Street, N .W.
Washington, DC 20015
U.S. UCLEAR REGULATORY ISSIO RULEMAKINGS &ADJUDICATI S STAFF OFACE OF THE SECRETARY OFTHECOMMISSI Docllnent Statistlrs Postmark Date ~ oYl a/.2lJ/q f - n.o ~ J..,$., ,.A')'"l ~
ecelw<' _ _.,_ _,___
r Reprt iuced Special istribution ~
~ - - -~ /
DOCKETED USNRC
'98 FEB 12[7~~!~ -PJ 4 NUCLEAR REGULATORY COMMISSIOf'bFFI( ,*- r:. r ,c:,-:r ) . r .,
10 CFR Part 71 A~JUULoE,:['.~i ;i!::r_ ~-1~t~~
,~.- k *'
ll
~ :Af- F
[Docket No. PRM-71-12]
International Energy Consultants, Inc.; DOCKET NUMBER Receipt of Petition for Rulemaking I I RUL E PAM 11- I~
( ~3 Fl< 8 31o:L)
AGENCY: Nuclear Regulatory Commission.
ACTION: Petition for rulemaking; Notice of receipt.
SUMMARY
- The Nuclear Regulatory Commission (NRG) has received and requests public comment on a petition for rulemaking filed by the International Energy Consultants, Inc.
The petition has been docketed by the Commission and has been assigned Docket No.
PRM-71-12. The petitioner requests that the NRG amend its regulations that govern packaging and transportation of radioactive material. The petitioner believes that special requirements for
- plutonium shipments should be eliminated.
~ 51 l'i'i8' DATE: Submit comments by (75 eaye f-ello11\11F1g puelieatie" in u,e Federal Rogj&ter).
Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given except as to comments received on or before this date.
ADDRESSES: Submit comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Attention: Rulemakings and Adjudications Staff.
Deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:30 am and 4:15 pm on Federal workdays.
For a copy of the petition, write: David L. Meyer, Chief, Rules and Directives Branch, Division of Administrative Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.
You may also provide comments via the NRC's interactive rulemaking website through the NRC home page (http://www.nrc.gov). This site provides the availability to upload comments as files (any format), if your web browser supports that function. For information about the interactive rulemaking website, contact Carol Gallagher, 301-415-5905 (e-mail:
CAG@nrc.gov).
FOR FURTHER INFORMATION CONTACT: David L. Meyer, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 . Telephone: 301-415-7162 or Toll Free: 800-368-5642 or e-mail: DLM1@nrc.gov.
SUPPLEMENTARY INFORMATION:
Background
The Nuclear Regulatory Commission received a petition for rulemaking submitted by Frank P. Falci on behalf of the International Energy Consultants, Inc. in the form of a letter addressed to the Secretary of the Commission, dated September 25, 1997. The petitioner believes that 10 CFR 71.63(b) should be eliminated. As an option, the petitioner believes that 10 CFR 71.63(a) should also be eliminated. This option would totally eliminate 10 CFR 71.63.
The petitioner made the same recommendation in a letter dated July 22, 1997, which he provided as a comment in the Commission's proposed rulemaking amending 10 CFR 71 .63(b) 2
to remove canisters containing vitrified high-level waste from the packaging requirement for double containment.
The petition was docketed as PRM-71-12 on October 22, 1997. The NRC is soliciting public comment on the petition. Public comment is requested on both the petition to eliminate 10 CFR 71.63(b), as well as the option to eliminate 10 CFR 71.63 totally, as discussed below.
Discussion of the Petition NRC's regulations in 10 CFR Part 71, entitled "Packaging and Transportation of Radioactive Material," include, in§ 71.63, special requirements for plutonium shipments:
§71.63 Special requirements for plutonium shipments.
(a) Plutonium in excess of 20 Ci (0.74 TBq) per package must be shipped as a solid.
(b) Plutonium in excess of 20 Ci (0.74 TBq) per package must be packaged in a separate inner container placed within outer packaging that meets the requirements of subparts E and F of this part for packaging of material in normal form. If the entire package is subjected to the tests specified in §71. 71 ("Normal conditions of transport"),
the separate inner container must not release plutonium as demonstrated to a sensitivity of 10-6 A,jh. If the entire package is subjected to the tests specified in §71.73
("Hypothetical accident corirJitions"), the separate inner container must restrict the loss of plutonium to not more than A 2 in 1 week. Solid plutonium in the following forms is exempt from the requirements of this paragraph:
( 1) Reactor fuel elements; (2) Metal or metal alloy; and (3) Other plutonium bearing solids that the Commission determines should be exempt from the requirements of this section.
3
The petitioner requests that§ 71.63(b) be deleted. The petitioner believes that provisions stated in this regulation cannot be supported technically or logically. The petitioner states that based on the "Q-System for the Calculation of A 1 and A 2 Values," an A 2 quantity of any radionuclide has the same potential for damaging the environment and the human species as an A 2 quantity of any other radionuclide. The petitioner further states that the requirement that a Type B package must be used whenever package content exceeds an A 2 quantity should be applied consistently for any radionuclide. The petitioner believes that if a Type B package is sufficient for a quantity of a radionuclide X which exceeds A 2 , then a Type B package should be
- sufficient for a quantity of radionuclide Y which exceeds A 2 , and this should be similarly so for every other radionuclide.
The petitioner states that while, for the most part, the regulations embrace this simple logical congruence, the congruence fails under§ 71.63(b) because packages containing plutonium must include a separate inner container for quantities of plutonium having an activity exceeding 20 curies (0. 74 TBq). The petitioner believes that if the NRC allows this failure of congruence to persist, the regulations will be vulnerable to the following challenges:
(1) The logical foundation of the adequacy of A 2 values as a proper measure of the potential for damaging the environment and the human species, as set forth under the Q-System, is compromised; (2) The absence of a radioactivity limit for every radionuclide which, if exceeded, would require a separate inner container, is an inherently inconsistent safety practice; and (3) The performance requirements for Type B packages as called for by 10 CFR Part 71 establish containment conditions under different levels of package trauma. The satisfaction of these requirements should be a matter of proper design work by the 4
package designer and proper evaluation of the design through regulatory review. The imposition of any specific package design feature such as that contained in 10 CFR 71.63(b) is gratuitous. The regulations are not formulated as package design specifications, nor should they be.
The petitioner believes that the continuing presence of § 71 .63(b) engenders excessively high costs in the transport of some radioactive materials without a clearly measurable net safety benefit. The petitioner states that this is so in part because the ultimate release limits allowed under Part 71 package performance requirements are identical with or
- without a "separate inner container," and because the presence of a "separate inner container" promotes additional exposures to radiation through the additional handling required for the "separate inner container." The petitioner further states that "... excessively high costs occur in some transport campaigns," and that one example " .. . of damage to our national budget is in the transport of transuranic wastes." Because large numbers of transuranic waste drums must be shipped in packages that have a "separate inner container" to comply with the existing rule, the petitioner believes that large savings would accrue without this rule. Therefore, the petitioner believes that elimination of § 71.63(b) would resolve these regulatory "defects."
As a corollary to the primary petition, the petitioner believes that an option to eliminate
§ 71 .63(a) as well as § 71 .63(b) should also be considered. This option would have the effect of totally eliminating § 71.63. The petitioner believes that the arguments propounded to support the elimination§ 71.63(b) also support the elimination of§ 71 .63(a).
The Petitioner's Conclusions The petitioner has concluded that NRC regulations in 10 CFR Part 71 which govern packaging and transportation of radioactive material must be amended to delete the provision regarding special requirements for plutonium shipments. The petitioner believes that a Type B 5
package should be sufficient for a quantity of radionuclide Y which exceeds the A 2 limit if such a package is sufficient for a quantity of radionuclide X which exceeds the A2 limit. It is the petitioner's view that this should be true for every other radionuclide including plutonium.
Dated at Rockville, Maryland, this I/[!: day of February 1998.
For the Nuclear Regulatory Commission.
l__
- oy I retary of the Commission.
6
INTERNATIONAL ENERGY CONSULTANTS, INC.
8905 COPENHAVER DRIVE POTOMAC, MARYLAND 20854 (301 ) 340- 1047 (301) 340-2229 FAX DO".: 'ETED
'AND t.r STAFF September 25, 1997 ..; ** RO DOCKET NUMBER Secretary PETITION RULE PAM 7 /- I~
U.S. Nuclear Regulatory Commission IP3F;e i31,,:2)
Attn: Chief, Docketing and Service Branch Washington, DC 20555-0001 Re: Petition for Rulemaking Pursuant to Provisions of 10 CFR 2.802 Regarding the "Special Requirements for Plutonium Shipments", 10 CFR 71.63
Dear Mr. Secretary:
I have previously communicated with you about the NRC Rulemaking on Requirements for Shipping Packages Used to Transport Vitrified High Level Waste, Docket Number 71-11, by my letter to you dated July 22, 1997. Receipt of my letter was acknowledged by your office in a notice postmarked July 25, 1997.
One of my earlier comments was:
"The Notice of Proposed Rulemaking referred to a NRC Staff Requirements Memorandum (SRM) dated October 31, 1996. This SRM directed the staff to initiate and expedite the Notice of Proposed Rulemaking and this has been accomplished. The SRM also stated, 'In the longer term, the staff should also address whether the technical basis for 10 CFR 71.63 remain valid, or whether a revision or elimination of portions of 10 CFR 71.63 is needed to provide flexibility for current and future technologies.' This broader look at 10 CFR 71.63 is to be completed by September 26, 1997. I recommend that pursuit of the total elimination of 10 CFR 71.63 be undertaken forthwith."
In keeping with the intent of my earlier recommendation, I hereby petition the U. S. Nuclear Regulatory Commission (NRC) for a rulemaking under the provisions of 10 CFR 2.802. I request that 10 CFR 71.63 (b) be deleted in its entirety.
This action is needed to remove from Nuclear Regulatory Commission regulations, provisions which cannot be supported technically or logically. Based on the "Q-System for the Calculation of A1 and A2 Values", an A2 quantity of any radionuclide has the same potential for damaging the environment and the human species as an A2 quantity of any other radionuclide. Therefore, the requirement that a Type B package must be used whenever package content exceeds an A2 quantity should be applied consistently for any radionuclide. That is, if a Type B package is sufficient for a quantity of radionuclide X which exceeds A2 , then a Type B package should be sufficient for a quantity of radionuclide Y which exceeds A2
- And this should be similarly so for every other radionuclide.
ll S. UC EAR REGULAT-ORY COMMISSION RU MAKINGS &ADJUDICATIONS STAFF OFFICE OF THE SECRETARY OF THE COMMISSION Document Statistics Postmark Date Cf ;; rf / q1 Copies Reoolved _ _....J Add'I Cqli88 Reproduced _
special DIStlriOl.dion 1:120 es~ r'i 1 '-
U IQ. n
U.S. Nuclear Regulatory Commission September 25, 1997 While the regulations do embrace this simple logical congruence for the most part, the congruence fails under 10 CFR 71.63 (b) wherein packages containing plutonium must include a separate inner container for quantities of plutonium having a radioactivity exceeding 20 curies (with certain exceptions).
If the NRC allows this failure of congruence to persist, these regulations will be vulnerable to the following challenges:
(I) The logical foundation of the adequacy of A2 values as a proper measure of the potential for damaging the environment and the human species, as set forth under the Q-System, is compromised.
(2) The absence of a radioactivity limit for every radionuclide which, if exceeded, would require a separate inner container, is an inherently inconsistent safety practice.
(3) The performance requirements for Type B packages as called for by 10 CFR 71 establish containment conditions under different levels of package trauma. The satisfaction of these requirements should be a matter of proper design work by the package designer and proper evaluation of the design through regulatory review. The imposition of any specific package design feature such as that contained in IO CFR 71.63 (b) is gratuitous. The regulations are not formulated as package design specifications, nor should they be.
In addition, the continuing presence of 10 CFR 71.63 (b) engenders excessively high costs in the transport of some radioactive materials with no clearly measurable net safety benefit. This is so in part because the ultimate release limits allowed under IO CFR 71 package performance requirements are identical with or without a "separate inner container", and because the presence of a "separate inner container" promotes additional exposures to radiation through the additional handling required for the "separate inner container".
Further, excessively high costs occur in some transport campaigns. One instance of such damage to our national budget is in the transport of transuranic wastes. Large numbers of transuranic waste drums must be shipped in packages which have a "separate inner container" to comply with the existing rule. Large savings would accrue without this rule.
The elimination of 10 CFR 71.63 (b) in its entirety would resolve these regulatory defects.
I recognize that the arguments prcsenkd here aiso re{kct :1poH 10 CFR 71.63 (a) because;: 10 CFR 71.63 (a) also relates only to a limited group of radionuclides. I request that NRC consider this matter as a corollary to the basic petition.
Sincerely, President FPF/aw