ML23156A061

PRM-071-011 - 59FR08143 - U.S. Department of Energy, Receipt of a Petition for Rulemaking

ML23156A061
Person / Time
Issue date:	02/18/1994
From:	NRC/SECY
To:
References
PRM-071-011, 59FR08143
Download: ML23156A061 (1)

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KEYWORD: ADAMS Template: SECY-067 02/18/1994 PRM-071-011 - 59FR08143 - U.S. DEPARTMENT OF ENERGY, RECEIPT OF A PETITION FOR RULEMAKING PRM-071-011 59FR08143 RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

STATUS OF RULEMAKING PROPOSED RULE: PRM-071-011 OPEN ITEM (Y/N) N RULE NAME:

u. S. DEPARTMENT OF ENERGY, RECEIPT OF A PETITION FOR RULEMAKING PROPOSED RULE FED REG CITE:

59FR08143 PROPOSED RULE PUBLICATION DATE: 02/18/94 NUMBER OF COMMENTS: 3 ORIGINAL DATE FOR COMMENTS: 05/04/94 EXTENSION DATE: I I FINAL RULE FED. REG. CITE: 63FR32600 FINAL RULE PUBLICATION DATE: 06/15/98 NOTES ON: PET. REQ. THAT NRC AMEND REGS. ON PACKAGING & TRANSPORTATION OF RA STATUS DIOACTIVE MATERIALS TO EXEMPT CANISTERS CONTAINING VITRIFIED HLW F F Rt7LE: ROM DOUBLE CONTAINMENT REQUIREMENT. [SEE PR-71, 62FR25146] HISTORY OF THE RULE PART AFFECTED: PRM-071-011 RULE TITLE: U.S. DEPARTMENT OF ENERGY, RECEIPT OF A PETITION FOR RULEMAKING PROPOSED RULE PROPOSED RULE DATE PROPOSED RULE SECY PAPER: 97-047 SRM DATE: 04/04/97 SIGNED BY SECRETARY: 02/04/94 FINAL RULE FINAL RULE DATE FINAL RULE ECY PAPER: 98-040 SRM DATE: 04/30/98 SIGNED BY SECRETARY: 05/20/98 STAFF CONTACTS ON THE RULE CONTACTl: MICHAEL T. LESAR CONTACT2: MAIL STOP: P-223 MAIL STOP: PHONE: 492-7758 PHONE:

DOCKET NO. PRM-071-011 (59FR08143) In the Matter of U. S. DEPARTMENT OF ENERGY, RECEIPT OF A PETITION FOR RULEMAKING DATE DATE OF DOCKETED DOCUMENT 12/06/93 02/14/94 04/11/94 05/04/94 05/09/94 -06/03/94 06/08/94 11/30/94 02/14/94 04/05/94 04/28/94 05/03/94 05/17 /94 06/02/94 TITLE OR DESCRIPTION OF DOCUMENT PETITION OF THE DEPARTMENT OF ENERGY REQUESTING EXEMPTION OF VITRIFIED HIGH-LEVEL WASTE CANISTERS FROM REQUIREMENTS OF 10 CFR 71.63(8) FEDERAL REGISTER NOTICE - RECEIPT OF PETITION FOR RULEMAKING COMMENT OF NYE CO. NUCLEAR WASTE REPOSITORY (LES W. BRADSHAW) (

1)

REQUEST FOR EXTENSION OF TIME SUBMITTED BY THE STATE OF IDAHO DEPARTMENT OF HEALTH AND WELFARE (INEL OVERSIGHT PROGRAM) COMMENT OF U.S. ENVIRONMENTAL PROTECTION AGENCY (RICHARD E. SANDERSON) (

2)

NOTICE OF EXTENSION OF COMMENT PERIOD, PUBLISHED AT 59 FR 26608 ON 5/23/94. COMMENT PERIOD EXTENDED TO 6/3/94. COMMENT OF IDAHO DEPT. OF HEALTH & WELFARE (INEL) (STEVEN R. HILL, ADMINISTRATOR) (

3)

State of Idaho DEPARTMENT OF HEALTH AND WELFARE INEL Oversight Program

• 800/232-INELQQCKETEO 1410 N. Hilton
• Boise, Idaho 83706 900 N. Skyline* Idaho Falls, Idaho 83402 June 2, 1994 Secretary U.S. Nuclear Regulatory Commission Attn:

Docketing and Service Branch Washington, D.C. 20555 OSHRC CECIL D. ANDRUS Governor

• 94 JlJN -8 PJ2 :48 JERRY L. Hir':~

OFFICE OF SECRETARY DOCKETING & SERVICE BRANCH@ Dept. of Health and Welfare DAVID L. HUMPHREY Overalght Coordinator STEVE R. HILL Admlnlatrator (B01) 208/334-0498 (FAX) 208/334-0429 (IF) 208/528*2600 (FAX) 208/528-2605 The state of Idaho's INEL Oversight Program submits the following comments on the Department of Energy's petition for rulemaking to the Nuclear Regulatory Commission. The Department has requested an exemption from the double container requirement under the transportation regulations for vitrified high-level waste (10 C.F.R. §71.63). Theoretically, exempting certain high-level materials from the double containment requirement is

not, from Idaho's perspective, at cross purposes with the regulations.

Protection of human health and the environment during transportation may be accomplished for these wastes through single containers. Idaho is concerned that, while the Department advocates a plausible exemption, the details necessary to demonstrate health and environmental protection is missing. In support of the exemption, the Department submitted several reports on the fracture properties of glass and ceramics. Each report is designed to illustrate the similarities between spent nuclear fuel and vitrified high-level waste; therefore, establishing the basis for the currently sought exemption. The Commission's initial exemption for spent fuel was based upon the decision to exempt "nonrespirable" forms of plutonium from the double containment requirement. 39 Fed. Reg. 20,960 (1974). The Department reports seek to establish properties of brittleness, impact absorption and fragment production which reproduce the characteristics of spent reactor fuel. The Department concludes in these reports that vitrified waste could react to transportation incidents in the same manner as spent fuel.

However, missing from the supporting documentation submitted by the Department were the parameters or performance standards that each high-level waste form must meet in order to be classified as vitrified waste eligible for the exemption.

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U.S. Nuclear Regulatory Commission June 2, 1994 Page 2 Specific criteria for the vitrified waste must be established and subjected to public scrutiny prior to applying the exemption to the broad range of wastes found in the Department of Energy complex. Tests to date focus upon specifications for a range of wastes found at the Savannah River Site. Through characterization efforts, the Department's inventory of high-level waste has been shown to be fairly diverse. Adding complication to this varied inventory is the heterogeneous characteristics of high-level waste forms. The full matrix of high-level waste inventory, which dictate the stability of the vitrified form, has not been explored and subjected to testing. For example, treatment technology for the Idaho National Engineering Laboratory high-level waste will not be selected until June 1, 1995. 1 If generally applicable standards cannot be established at this time, the Department should be seeking indi victual, ad hoc exemptions for those waste streams demonstrated to meet Commission criteria for single containment. 2 To date, the Department's petition promises to develop the contents of the high-level waste form and the quality assurance procedures for vitrification. 3 The definitive specifications for the waste form are not available for review. The conclusion that Idaho has reached, and that Commission staff should

reach, is that rulemaking is premature.

Instead, the Department, the Commission and other interested parties should continue to develop and discuss in a public forum the appropriate criteria to define the exemption. The Department is, as stated earlier, free to pursue individual exemptions. The state appreciates the Commission's extension of time to review the technical data provided and submit these comments. In that regard, Idaho requests that the Commission's enhanced 1R.D. Jones to S.R. Hill, Response to High Level Waste Treatment Technology Letter of Dec. 10, 1993, Jan. 28, 1994. 2As stated in the original rulemaking, "The latter category (other plutonium bearing solids that the Commission determines suitable] provides a means for the Commission to evaluate, on a case-by-case basis, requests for exemption of other solid material where the quantity and form of the material permits a determination that double containment is unnecessary." 39 Fed. Reg. 20,960 (1974). 3u.s. DOE, Technical Justification to Support the PRM by DOE to Exempt HLW Canisters from 10 C.F.R. § 71.63(b); Rev. O, p. 15 (Sept. 30, 1993.

U.S. Nuclear Regulatory Commission June 2, 1994 Page 3 rulemaking procedures, including access to the Internet System, be applied to this petition. Thank you for your consideration of these matters. STEVER. HILL Administrator INEL Oversight Program Central Office SRH/lvh cc: Craig Halverson, INEL O.P. Steven G. Oberg, INEL O.P. Teresa A. Hampton, DAG David L. Humphrey, IDHW John Roberts, DOE-ID, OCRWM

( DOCKET NUMBER PETITION RULE PRM "7 / -I I (S-1 Ff< s-/L-13) NUCLEAR REGULATORY COMMISSION 10 CFR Part 71 [Docket No. PRM-71-11] DOCKETED US RC [7590-01-P] -94 JUN -3 PS :07 OFFICE OF s __ cRET7H,Y DOCKET ING & SER~ ICE BRANCH U.S. Department of Energy, Receipt of a Petition for Rulemaking: Extension of Comment Period AGENCY: Nuclear Regulatory Commission. ACTION: Petition for rulemaking: Extension of comment period.

SUMMARY

On February 18, 1994 (59 FR 8143), the Nuclear Regulatory Commission (NRC) published for public comment a petition for rulemaking filed by the U.S. Department of Energy. The petitioner requested that the NRC amend its regulations governing packaging and transportation of radioactive materials to specifically exempt canisters containing vitrified high-level waste from the double containment requirement specified in NRC's regulations. The comment period for this petition for rulemaking was to have expired on May 4, 1994. The INEL oversight Program of the State of Idaho has requested a thirty-day extension of the comment period. In view of the fact that the State of Idaho has received technical background documents regarding the proposed amendment only recently from the U.S. Department of Energy and requests this extension to review these documents and provide comments, the NRC has decided to extend the comment period for an additional thirty days. The extended comment period now expires on June 3, 1994. DATES: The comment period has been extended and now expires June 3, 1994. Comments received after this date will be

it is practical to do so, but assurance of consideration cannot be given except as to comments received on or before this date. ADDRESSES: Send written comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Attention: Docketing and Service Branch. Deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:45 am and 4:15 pm on Federal workdays. For a copy of the petition, write: Rules Review Section, Rules Review and Directives Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555. FOR FURTHER INFORMATION CONTACT: Michael T. Lesar, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Telephone: 301-492-7758 or Toll Free: 800-368-5642. Dated at Rockville, Maryland, this 17'1-day of May, 1994. For the Nuclear Regulatory Commission. J t ting Secretary of the Commission. 2

DOCKET NUMBER PETJJ_!9N RULE PAM 7 /- / ~'1FR~1'1.3) UNITED STATES ENVIRONMENTAL PROTECTION AGENCY WASHINGTON, D.C. 20460 MAY 3 - 1994 Mr. John c. Hoyle Acting Secretary of the Commission Nuclear Regulatory Commission Attention: Docketing and Service Branch Washington, DC 20555

Dear Mr. Hoyle:

In accordance with its responsibilities under Section 309 of the Clean Air Act, the U.S. Environmental Protection Agency has reviewed the Nuclear Regulatory Commission's 10 CFR Part 71 petition for rulemaking, published in the Federal Register February 18, 1994, and has no comments. We appreciate the opportunity to review the petition. If you have any questions, please call ~eon (202) 260-5053 or have your staff contact Ms. Susan Offerda,,i (202) 260-5059. / Siit/} I Richard E. Sanderson Director Office of Federal Activities 'JUL 2 1 199!__ Acknowledged by card..,..,_.. _ Recycled/Recyclable Printed with Soy/Canola Ink on paper that contains at least 50% recycled fiber

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State of Idaho DEPARTMENT OF HEALTH AND WELFARE INEL Oversight Program

• 800/232-INOOCKETED 1410 N. Hilton
• Boise, Idaho 83706 900 N. Skyline* Idaho Falls, Idaho 83402 Secretary of the Commission ATTN:

Docketing and Service Branch U.S. Nuclear Regulatory Commission "94 MAY -4 P 4 :56 OFFICE OF SECRETARY DOCKETING F.,. SERVICE BRANCH One White Flint North Bldg., Rm. 16-H-20 11555 Rockville Pike Rockville, MD 20555 RE: PRM-71-11

Dear Secretary:

CECIL D. ANDRUS Governor JERRY L. HARRIS Director Dept. of Health and Welfare DAVID L. HUMPHREY Overalght Coordinator STEVE R. HILL Administrator (B01) 208/334-0498 (FAX) 208/334-0429 (IF) 208/528*2600 (FAX) 208/528-2605 The state of Idaho, INEL Oversight Program, requests an extension of time for the public comment period in the referenced matter. The state has recently received technical background documents from the Department of Energy regarding the proposed rule. In order to review these documents and provide comments, I am requesting an extension of thirty days from the May 4, 1994 deadline. You may contact me at the above questions concerning this request. this matter. STEVER. HILL Administrator INEL Oversight Program Central Office SRH/lvh address should you have any Thank you for your attention to cc: Teresa Hampton, Deputy Attorney General Steve Oberg, INEL Oversight Program Pnnted on Aeq,cled p..,_,

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DOCKETED USNRC OOt;/ET NUMBER PETITION RULE PRM 7 l-J.J {5C/ FR ~JL/3) (!) NYE COUNTY NUCLEAR WASTE REPOSITORY PROJECT OFFICE -n.t am 11 1\\11 :49P.O. BOX 1767

• TONOPAH, NEVADA 89049

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(702) 482-8183

• FAX (702) 482-9289 April 5, 1994 Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN:

Docketing and Service Branch RE: PRM-71-11

Dear Secretary:

This letter constitutes the comments of Nye County, Nevada on the Department of Energy's Petition for Rulemaking to amend the provisions of 10 CFR 71.63(b) to exempt canisters containing vitrified high-level waste from the double containment requirement specified in that regulation. Nye County, Nevada is the situs jurisdiction for the proposed spent nuclear fuel and high-level nuclear waste repository at Yucca Mountain, Nevada. As the situs jurisdiction, the County exercises oversight responsibility over, and follows very closely, all aspects of the Department of Energy's repository program. We have reviewed the DOE petition, as well as the technical justification which DOE submitted in support thereof. We agree with the rationale and arguments advanced by the DOE. Accordingly, Nye County has no objection to the NRC's granting PRM-71-11, and amending 10 CFR 71.63(b) in the matter requested in that petition. Very truly yours, Ni~y*;;:;t~ Les W. Bradshaw, Manager Nuclear Waste Repository Project Office MAY 1 0 1994', Acknowledged by card............................. _

U.S. Nuclear Regulatory Commission April 5, 1994 Page2 cc: John Roberts, Department of Energy Robert Loux, State Nuclear Projects Office William L. Offutt, Nye County Manager Affected Units of Local Government Phil Niedzielski-Eichner, Governmental Dynamics, Inc. Nick Stellavato, Nye County On-Site Geotechnical Representative Malachy R. Murphy, Lane, Powell, Spears and Lubersky LWB:tp 94040405.ltr

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PR NUCLEAR REGULATORY COMMISSION 10 CFR Part 71 [Docket No. PRM-71-11]

• 94 f-14 P ll : ::: 8 u.s. Department of Energy, Receipt of a Petition for Rulemaking AGENCY:

Nuclear Regulatory Commission. ACTION: Petition for rulemaking; Notice of receipt.

SUMMARY

The Nuclear Regulatory Commission (NRC) has received and requests public comment on a petition for rulemaking filed by the U.S. Department of Energy (DOE). The petition has been docketed by the Commission and has been assigned Docket No. PRM-71-11. The petitioner requests that the NRC amend its regulations governing packaging and transportation of radioactive materials to specifically exempt canisters containing vitrified high-level waste from the double containment requirement specified in NRC's regulations. The petitioner believes such an amendment would permit more cost-effective high-level radioactive waste management by DOE in the geologic repository and would not adversely affect the safety of the transportation package. DATE: Submit comments by (75 days-'f~o/./otig publication in the Federal Register). 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 comme nts to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555. Docketing and Service Branch. Attention:

Deliver comments to 11555 Rockville Pike, Rockville, Maryland, between 7:45 am and 4:15 pm on Federal workdays. For a copy of the petition, write: Rules Review Section, Rules Review and Directives Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555. FOR FURTHER INFORMATION CONTACT: Michael T. Lesar, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, 4lt DC 20555. Telephone: 301-492-7758 or Toll Free: 800-368-5642. SUPPLEMENTARY INFORMATION:

Background

The Nuclear Regulatory Commission (NRC) received a petition for rulemaking dated November 30, 1993, submitted by the U.S. Department of Energy (DOE). The petition was docketed as PRM 11 on December 6, 1993. The petitioner requests that the NRC amend its regulations specified in 10 CFR Part 71 that govern packaging and transport of radioactive materials. Specifically, the petitioner is seeking a specific exemption for canisters containing vitrified high-level waste (HLW) from the requirements currently contained in 10 CFR 71.63(b) regarding special requirements for plutonium shipments. The petitioner notes that current NRC special requirements for plutonium shipments (10 CFR 71.63) specify that all shipments of plutonium with an activity greater than 20 curies per package must meet the double containment requirement in 10 CFR 71.63(b). 2

Under the Nuclear Waste Policy Act of 1982 (NWPA), as amended, DOE is responsible for developing a geologic repository for the disposal of high-level radioactive waste and spent fuel. Shipments of HLW must be approved for shipment through DOE's Civilian Radioactive Waste Management System (CRWMS) for transport to and disposal in the geologic repository.

Also, under the NWPA, all packages used to transport spent fuel and HLW must be certified by NRC.

On June 17, 1974 (39 FR 20960), the NRC published a final rule requiring that shipments of plutonium with activity greater than 20 curies per package meet the double containment requirement of 10 CFR 71.63(b). The petitioner admits that 10 CFR 71.63(b) applies to the shipments of vitrified HLW. However, the petitioner also claims that these shipments should be exempt from the double containment requirement because this material is analogous to spent fuel. As the petitioner notes, the preamble to the final rule states that spent fuel is exempt from the double containment requirement specified in 10 CFR 71.63(b) because these solid forms of plutonium were determined to be "essentially nonrespirable. 11 The petitioner also indicates that the evaluation of the respirability potential of canisters filled with vitrified HLW is based mainly on the results of impact tests. In support of the petition for rulemaking, the petitioner has included a document entitled "Technical Justification to Support the PRM by the DOE to Exempt HLW Canisters from 10 CFR 71.63(b)" (technical justification). The petitioner claims that 3

the tests described in the technical justification demonstrate that the canisters containing vitrified HLW compare favorably with the physical integrity of the metal cladding surrounding the spent fuel pellets in reactor assemblies. The petitioner believes that because canisters containing vitrified HLW are analogous to spent fuel, these canisters should be exempt from the double containment requirement specified in 10 CFR 71.63(b). The NRC is soliciting public comment on the petition submitted by DOE that requests the changes to the regulations in 10 CFR Part 71 as discussed below. The Petitioner Pursuant to the Nuclear Waste Policy Act of 1982, as amended (NWPA), the petitioner is the Federal agency responsible for developing and administering a geologic repository for the deep disposal of high-level radioactive waste and spent fuel. The petitioner proposes to ship the HLW from each of its three storage locations at Aiken, South Carolina; Hanford, Washington; and West Valley, New York, directly to the geologic repository in casks certified by the NRC. The HLW currently exists mostly in the form of liquid and sludge resulting from the reprocessing of defense reactor fuels. The petitioner proposes to solidify (vitrify) this material into a borosilicate glass form in which the HLW is dispersed and immobilized and place it into stainless steel canisters for storage and transport to the geologic repository. The petitioner indicates that it is submitting this petition for rulemaking to amend 10 CFR Part 71 so that it can 4

manage the transportation and disposal of high-level waste in a cost-effective and efficient manner without adversely affecting the safety of the transportation package. Discussion of the Petition The petitioner has submitted this petition for rulemaking because it believes it will be adversely affected by the current regulations that require plutonium shipments with activity greater than 20 curies per package to be shipped in a double containment format. The petitioner's primary concern is that the double containment requirement specified in 10 CFR 71.63(b) will prevent it from effectively performing its responsibility under the NWPA to administer the transportation of canisters containing vitrified high-level radioactive waste for disposal in the geologic repository in an efficient and cost-effective manner. The petitioner states that although the current regulations are appropriate in exempting reactor fuel elements from the double containment requirement specified in 10 CFR 71.63(b), canisters containing vitrified HLW should also be exempted. The petitioner states that spent fuel was exempted from the double containment requirement in 10 CFR 71.63(b) because the fuel pellet itself and the surrounding metal cladding were found to provide adequate protection against the possible dispersion of plutonium particles both under normal transport conditions and during hypothetical accident conditions. The petitioner believes that the tests described in the technical justification provide sufficient technical information to indicate that the 5

borosilicate glass mixture and the storage canisters are analogous to spent fuel that is exempt from the double containment requirement. In the technical justification, the petitioner describes the physical characteristics of the austenitic stainless steel canisters that will house the vitrified HLW and indicates that the packages will pass a 7-meter drop test onto a flat, essentially unyielding surface without a release of its contents. The petitioner emphasized that this test should not be confused with the hypothetical accident tests specified in 10 CFR 71.73, Hypothetical Accident Conditions." The petitioner also clarifies that the 9-meter drop test required in 10 CFR 71.73(c)(l) applies to the entire package, including the cask which must be certified by the NRC used to transport the canisters containing the vitrified HLW. The petitioner provides a detailed comparison in the technical justification between the steel canister that will house vitrified HLW and the reactor fuel elements that are exempt from the double containment requirement in 10 CFR 71.63(b). The petitioner notes that the plutonium contained in reactor fuel elements is encased in solid ceramic fuel pellets surrounded by a sealed, sturdy metal cladding that inhibits dispersion of radioactive particles. The petitioner believes the impact tests performed during the past 20 years on canisters containing simulated HLW glass forms indicate that these canisters qualify for exemption from the double containment requirement. 6

Helium leak tests and dye penetrant tests performed after the impact testing have demonstrated that the vitrified HLW canisters can withstand a 9-meter drop test. The petitioner acknowledges that reactor fuel elements were exempted from the double containment requirement in 10 CFR 71.63(b) because they are considered to be "essentially nonrespirable." The petitioner believes that because the canisters have not been exposed to the high levels of radiation present in a commercial reactor, these e packages will not be subject to molecular-level changes in material properties, such as increased embrittlement, unlike spent reactor fuel cladding. The petitioner concludes that the numerous impact and followup tests on simulated vitrified HLW canisters indicate that the canisters provide, at minimum, protection comparable to that provided by spent fuel cladding. In the technical justification, the petitioner also compares the physical and chemical characteristics of the vitrified HLW glass mixture to spent fuel pellets. The petitioner notes that production of potentially respirable particles from the glass mixture could result from cooldown processes after being poured into the HLW canister, normal handling and transport conditions, and hypothetical accident conditions. Because impact studies of simulated waste glass from the DOE Savannah River site have shown comparable levels of fracture resistance and similar fractions of respirable particles when compared to unirradiated uranium fuel pellets and other potential waste form materials, the petitioner believes that the fracture resistance of simulated HLW glass is 7

comparable to that of uranium fuel pellets. The petitioner asserts that leak tests performed for both normal transport and hypothetical accident scenarios indicate that the quantity of respirable material produced is minute and fully supports the conclusion that the vitrified HLW canister waste form is "essentially nonrespirable" and, therefore, analogous to reactor fuel elements. The petitioner also notes that evaluations show that the total concentration of plutonium in an individual fuel assembly is more than 100 times greater than that in an HLW cannister from the savannah River site. The petitioner indicates that the maximum quantity of plutonium projected for the Hanford and West Valley HLW canisters is much less than that of the Savannah River HLW canisters. The petitioner also notes that canisters containing vitrified HLW will be enclosed within a shipping cask that has been certified by NRC during actual transport conditions. The petitioner concludes that this arrangement will further reduce the potential for canister damage and for a release of respirable particles of radionuclides. The petitioner asserts that proposed disposal criteria would result in a cost-effective option that would not adversely affect public health, environmental quality, the safety of the transportation package, or the safety of workers who handle the transportation package. The petitioner also asserts that the current regulatory limits on radioactivity in the transportation package are intended to protect not only individuals who 8

transport and handle the waste but also the general public if a transportation accident enroute to the geologic repository site results in a release of radioactive material. Adverse Effects on the Petitioner The petitioner believes that it will be adversely affected if the double containment requirement in 10 CFR 71.63(b) is applied to canisters containing vitrified HLW. The petitioner notes that the only alternative would be to design and construct a double containment transportation cask. The petitioner believes that a double containment requirement would add additional handling steps to the loading and unloading of the HLW canister, resulting in an increase of time and expense in HLW shipments. The additional handling process would increase the radiation dose received by workers and create additional contaminated metal hardware, resulting in increased disposal effort and expense. The petitioner also asserts that a double containment requirement for this HLW form would require additional shipments because of a potential decrease in payload capacity of the cask. Additional shipments would create a corresponding increase in risk to affected populations along the transportation route to the geologic repository. The petitioner believes that the double containment requirement would impose an unnecessary and unduly burdensome rule that cannot be justified in terms of any incremental benefits to public health and safety. The Petitioner's Proposed Amendment The petitioner requests that 10 CFR Part 71 be amended to 9

overcome the problems the petitioner has itemized and recommends the following revision to the regulations: The petitioner proposes that S71.63 be amended by revising paragraph (b) to redesignate paragraph (b)(3) as paragraph (b)(4) and adding a new paragraph (b)(3) to read as follows: §71.63. Special requirements for plutonium shipments. (b) Plutonium in excess of 20 curies per package must be packaged in a separate inner container placed within outer packaging that meets the requirements of Subparts E and F for packaging of material in normal form. If the entire package is subjected to the tests specified in S71.71 (Normal Conditions of Transport), the separate inner container must not release plutonium, as demonstrated to a sensitivity of 10-6 A2 per hour. If the entire package is subjected to the tests specified in S71.73 (Hypothetical Accident Conditions), the separate inner container must restrict the loss of plutonium to not more than A2 in one week. Solid plutonium in the following forms is exempt from the requirements of this paragraph: (1) Reactor fuel elements; (2) Metal or metal alloy; (3) Canisters containing vitrified high-level waste; and (4) Other plutonium-bearing solids that the Commission determines should be exempt from the requirements of this section. 10

The Petitioner's Conclusion The petitioner has concluded that the double containment requirement specified in 10 CFR 71.63(b) should not be applied to shipments of canisters containing vitrified HLW because this waste form is analogous to spent reactor fuel elements, which are I exempt. The petitioner believes that impact and leak tests on I the canisters, chemical analyses of spent fuel and simulated HLW borosilicate glass mixtures, and other studies of the levels of radioactivity present in the proposed transportation packages demonstrate that canisters containing vitrified HLW are analogous to spent reactor fuel elements and, therefore, should be exempt f.rom the double containment requirement in 10 CFR 71. 63 ( b)

• The petitioner has proposed an amendment to the current regulations in 10 CFR Part 71 that it believes will permit more cost-effective disposal of high-level waste without adversely affecting the safety of the transportation package, the workers who handle the package, affected populations along the transportation corridor, or the environment.

Dated at Rockville, Maryland, this /L/.,{ day of February, 1994. u l~:ulatory Commission. ilk, ' f the Comm1sion. 11 7

Secretary Department of Energy Washington, DC 20585 November 30, 1993 U.S. Nuclear Regulatory Commission Attention: Chief, Docketing and Service Branch Washington, D.C. 20555

Dear Sir:

DOCKET NUMBER PET, 1 I01~ RULE PRM ,J [ 5q F f{ frl'-f3) DOCKETED The U.S. Department of Energy hereby submits the enclosed petition for rulema.king under the provisions of 10 CFR Part 2.802. We believe that 10 CFR Part 71 should be amended to exempt vitrified high-level waste canisters from the requirements of 10 CFR Part 71.63(b). This rulema.king amendment would help to ensure a more appropriate use of resources. If high-level waste canisters are not exempt from 10 CFR Part 71, it could have the impact of diverting resources from other, more worthy endeavors associated with the Civilian Radioactive Waste Management Program. The subject of this petition has been previously discussed with the Commission's Division of High-Level Waste Management staff and with the Advisory Committee on Nuclear Waste. We would appreciate your consideration and acceptance of this petition. Any questions regarding the petition may be addressed to Mr. John Roberts of my staff on 586-9896.

Enclosure:

Daniel A. Dre f , Director Office of Civilian Radioactive Waste Management Petition of the U.S. Department of Energy for a Rulemaking to Exempt Vitrified High-Level Waste Canisters from the requirements of 10 CFR Part 71.63(b)

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cc: C. Gertz, YMPO T. J. Hickey, Nevada Legislative Committee R. Loux, State of Nevada D. Bechtel, Las Vegas, NV Eureka County, NV Lander County, Battle Mountain, NV P. Niedzielski-Eichner, Nye County, NV W. O_ffutt, Nye County, NV L. Bradshaw, Nye County, NV C. Schank, Churchill County, NV F. Mariani, White Pine County, NV V. Poe, Mineral County, NV J. Pitts, Lincoln County, NV J. Hayes, Esmeralda County, NV B. Mettam, Inyo County, CA

PETITION BY THE U.S. DEPARTMENT OF ENERGY FOR AN NRC RULE-MAKING TO EXEMPf CANISTERS OF VITRIFIED HIGH-LEVEL WASTE FROM THE REQUIREMENTS OF 10 CFR 71.63(b) Docket No. -----

1.0 INTRODUCTION

Title 10 of the Code of Federal Regulations, Part 71 - Packaging and Transportation of Radioactive Material ( 10 CFR Part 71) - specifies special requirements for plutonium shipments. 10 CFR 71.63(b) requires that plutonium in excess of 20 curies per package must be packaged in a separate inner container placed within outer packaging that meets the requirements of Subparts E and F for packaging of material in normal form. In addition, if the entire package is subjected to the tests specified in Parts 71.71 and 71.73, the inner container must meet certain containment requirements for normal and hypothetical accident conditions. However, 10 CFR 71.63(b) exempts solid plutonium in the following forms from these requirements: (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 [i.e. 10 CFR 71.63(b)]. High-Level Waste, generally existing in the form of liquid or sludge, will be solidified (vitrified) in the form of borosilicate glass contained in stainless steel canisters and transported to a geologic repository for disposal. These canisters will contain quantities of plutonium which are in excess of the 20 curie threshold limit, so that 10 CFR 71.63(b) requirements are applicable to the shipment of these HLW canisters. The Department of Energy (DOE). considers that the canisters containing vitrified HL W have containment properties at least comparable to those of reactor fuel elements; therefore, they should be exempted from the requirements of 10 CFR 71.63(b), analogous to the exemption provided for reactor fuel elements. Through this petition for rulemaking(PRM), DOE seeks such an exemption for HL W canisters and to document such an exemption in the form of an amended 10 CFR, Part 71. 63(b). This petition includes the information to be provided for a PRM in accordance with the requirements of 10 CFR 2. 802 ( c). The proposed amendments to the current regulation, 10 CFR Part 71, are included in Section 2, the grounds for and DOE's interest in the action requested are described in Section 3, and a discussion of the specific issues involved, supporting arguments, and relevant information are provided in the attached Technical Justification document.

• Page 1 of 3

2.0 PROPOSED AMENDMENT TO 10 CFR 71.63(b) The following specific amendment to 10 CFR 71.63(b) is proposed. Revise the last sentence of 10 CFR 71.63(b) as follows: "Solid. plutonium in the following forms is exempt from the requirements of this paragraph: (1) Reactor fuel elements; (2) Metal or metal alloy (3) Canisters containing vitrified high-level waste; and ( 4) Other plutonium bearing solids that the Commission determines should be exempt from the requirements of this section. " 3.0 PETITIONER'S GROUNDS FOR AND INTEREST IN THE PETITION This section describes the DOE' s grounds for and interest in the action requested. The DOE will be the licensee for a geologic repository developed pursuant to the Nuclear Waste Policy Act, as amended, for the-disposal of spent nuclear fuel and high level waste. Most of the HL W has been produced as a result of defense-related atomic energy activities; the remainder is HLW from the West Valley Demonstration Project resulting from commercial reprocessing. According to current plans, HLW in the form of canisters containing borosilicate glass will be shipped directly to the geologic repository in casks certified by the Nuclear Regulatory Commission. As such, these casks will be subject to the requirements of 10 CFR Part 71. Specifically, 10 CFR 71.63(b) requires that "plutonium in excess of 20 curies per package must be packaged in a separate inner container placed within outer packaging that' meets the requirements of Subparts E and F. " However, solid plutonium in the following forms is exempt from the requirements of Part 71.63(b): (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. As described in detail in the attached Technical Justification document, a separate inner container is unnecessary for the HL W because of the high degree of confinement provided by the stainless steel waste canister and the nonrespirability of the solid, plutonium-bearing waste form. An exemption has been provided for reactor fuel elements from the double containment requirement. It is DOE's position that the HLW canisters containing the vitrified waste are at least as good as reactor fuel elements in terms of physical integrity of the waste form and containment of its plutonium contents. Imposition of I_>art 71.63(b) Page 2 of 3

requirements on the HL W canisters would provide no significant increases in overall safety

• of the transportation package, and would result in increased program cost and complexity.

Through this petition for rulemaking, DOE is seeking an exemption for canisters containing vitrified high-level waste from the requirements of 10 CFR 71.63(b), analogous to that provided for reactor fuel elements. DOE requests that such an exemption be formally. documented in an amended Part 71 by specifically including vitrified HLW canisters in the list of solid plutonium forms that are exempted from the requirements of Part 71.63(b). 4.0 SUPPORTING INFORMATION A discussion of the specific issues involved in the petition, supporting arguments, and other relevant information is given in the attached Technical Justification.

5.0 CONCLUSION

Because the proposed HL W form provides a comparable level of protection of public health and safety as does spent fuel contained in an NRC-certified cask, the requirement of a separate inner container for HL W represents an unduly burdensome rule. Exemption from the double containment requirement is needed in order for the DOE to efficiently carry out its mission to dispose of HL W in a deep geologic repository (in accordance with the Nuclear Waste Policy Act, as amended). Based on the confinement capability of the canister and the experimentally demonstrated nondispersibility of the waste form, it has been shown that the vitrified HL W canisters are comparable as a waste form to Reactor Fuel Elements, which are exempted from the requirements of 10 CFR 71.63(b). Accordingly, the vitrified HL W canisters qualify for exemption from the requirements of 10 CFR 71.63(b). Based on the information presented, DOE petitions the Commission to amend 10 CFR Part 71 to include canisters containing vitrified HLW canisters in the exempted list of solid forms of plutonium, ~ accordance with the proposed amendment presented in Section 2. Respectfully Submitted, ci:~~

• Office of Civilian Radioactive Dated / 0_

0 ) 3 Waste Management Page 3 of 3

Civilian Radioactive Waste Management System Management & Operating Contractor Technical Justification WBS: 9.2.2.2.3 QA: NIA to Support the PRM by the DOE to Exempt HLW Canisters from 10CFR71.63(b) Rev. 0 September 30, 1993 Prepared for: U.S. Department of Energy Office of Civilian Radioactive Waste Management 1000 Independence Avenue, S.W. Washington, D.C. 20585 Prepared by: TRW Environmental Safety Systems Inc. 2650 Park Tower Drive Suite 800 Vienna, Virginia 22180 Under Contract Number DE-AC0 1-91 RW00134

TABLE OF CONTENTS. Page I. INTRODUCTION................................................. I

2. BACKGROUND...... *........................................... 2

3. PROPOSED RULE................................................ 3

4. DESCRIPTION OF HLW CANISTER AND CONTENTS.................... 4 4.1 HLW CONTENTS............................................. 6 4.2 CANISTER DESIGN........................................... 7

5. WASTE ACCEPTANCE SPECIFICATIONS............................. 15 5.1 BOROSILICATE GLASS WASTE FORM........................... 15 5.2 CANISTER CHARACTERISTICS................................. 16 5.3 FINISHED PRODUCT CHARACTERISTICS......................... 16

6. COMPARISON TO REACTOR FUEL ELEMENTS........................ 18 6.1 HLW GLASS VERSUS FUEL PELLETS............................ 18 6.2 HLW CANISTER VERSUS FUEL ROD CLADDING................... 19

7. CANISTER DROP TESTS......................................... 20 7.1 CANISTER INTEGRITY....................................... 20 7.2 HLW GLASS PARTICLES...................................... 20 7.2.1 MCC-15 Test and Analyses................................. 21 7.2.2 Other PNL Impact Tests and Analyses......................... 21

8. PROGRAMMATIC WSTIFICA TION.................................. 26 8.1 ALTERNATIVES.........,................................... 26 8.2 OPERATIONAL CONSIDERATIONS.. *............................ 26 8.3 ECONOMIC CONSIDERATIONS

................................ 26

9. REFERENCES.................................................. 27 w

LIST OF FIGURES Page

l. Savannah River Site HLW Canister.................................... 13

2. Savannah River Site HLW Canister Oosure.............................. 14

3. Summary of Impact Particle-Size Distribution Results....................... 25 iv

LIST OF TABLES Page

1. Summary of Canister Characteristics.................................... 5

2. Radionuclide Content of Most Highly Radioactive Glass Expected to Be Produced at SRS..................................... 9

3. Projected DWPF Waste Glass Compositions............................. 12

4. Summary of DWPF HLW Canister Drop Tests............................ 24 V

l. INTRODUCTION The Department of Energy (DOE) requests that the regulations in 10 CFR 71.63(b) be amended to exempt canisters containing vitrified high-level waste (I-IL W) from the double containment requirement for the shipment of plutonium with an activity greater than 20 c.uries per package.

Both spent fuel and vitrified HL W -- which will be shipped in the Civilian Radioactive Waste Management System (CRWMS) to a final disposal site in a geologic repository -- contain plutonium in quantities greater than the threshold limit of 20 curies per package. Spent fuel was exempted from the double containment requirement in the original rulemalcing because of its physical integrity and low potential for inhalation of plutonium particles. The fuel pellet itself and the surrounding metal fuel rod cladding were detennined to provide adequate protection against the possible dispersion of plutonium particles under both normal conditions of transport and during hypothetical accident conditions. Vitrified I-aW contained in stainless steel canisters provides a comparable level of safety protection to that provided by fuel elements, and therefore should also be exempted from the double containment requirement. The following sections provide a background on this issue, and provide a detailed description of a representative I-aW canister and its vitrified waste content (bounding worst-case radionuclide concentrations) at the Savannah River Site in Aiken, SC. The Savannah River Site will be the first of the 3 I-aW sites (the others are Hanford, WA and West Valley, NY) to vitrify its HLW into a borosilicate glass waste form suitable for transport and final disposal. In addition, the waste acceptance specifications, analyses, and supporting documentation which must be provided to DOE's Office of Civilian Radioactive Waste Management (OCRWM) is described. This documentation assures that the waste glass is prepared and packaged in accordance with approved standards to meet CRWMS system-level requirements and to verify the quality and consistency of the waste form. A summary of supporting technical analyses, including the results of numerous impact tests of simulated HL W canisters, which confirms the physical integrity and low dispersion characteristics of this waste form is presented in Sections 6 and 7. Finally, the DOE's programmatic justification for the request for exemption from double conIBinment is provided. l.

2. BACKGROUND As provided by the Nuclear Waste Policy Act of 1982 (NWPA), OCRWM has been given the responsibility to develop a geologic repository for the disposal of high-level radioactive waste and spent fuel. The transportation of these waste forms is addressed in the 1987 amendments to the NWP A. All packages used to transport spent fuel and high-level waste must be certified by the Nuclear Regulatory Commission (NRC).

The NRC's regulations governing the packaging and transport of radioactive materials are given in the Code of Federal Regulations - Section 10, Part 71. Specifically, 10 CFR 71.63 imposes special requirements for the shipment of plutonium with an activity greater than 20 curies per package. Since all spent fuel and high-level waste transported in the CRWMS will exceed this threshold radioactivity level, 10 CFR 71.63 is applicable. The waste must be shipped as a solid per Part 7 l.63(a), and it "must be packaged in a separate inner container placed within outer packaging that meets the requirements of Subparts E and F... " (i.e., double containment) in accordance with Part 71.63(b). In addition, if the entire package is subjected to the tests specified in Part 71.71 (Normal Conditions of Transport), no release of plutonium from the inner-most container is allowed. When subjected to the more severe tests specified in Part 71.73 (Hypothetical Accident Conditions), then the separate inner container must restrict plutonium release to specified limits. An exemption is provided, however, in 10 CFR 71.63(b) for spent reactor fuel elements from the double containment requirement of this paragraph. The double containment requirement was originally implemented in 10 CFR 71 through a final rule in 1974 (39 FR 20960, June 17, 1974). The Statement of Consideration accompanying this rulemaking action discusses the basis for the exemption of certain solid plutonium fonns from the double containment requirement. The text of the relevant section of the Federal Register notice states: ".. solid forms of plutonium that are essentially nonrespirable should be exempted from the double containment requiremenL" Spent reactor fuel elements, and plutonium-bearing metal or metal alloys were exempted from the double containment requirement because these solid forms were deemed to be "essentially nonrespirable". The evaluation of the respirability potential of various waste forms is based primarily on the results of impact tests which measure the quantity of respirable fines produced. Respirable particles are considered to be those having an aerodynamic diameter of less than IO µm. Physical integrity of the inner container and low dispersability of the contained plutonium are also key considerations in the evaluation of the transportation package. In this Technical Justification, the vitrified high-level waste will be shown to compare very favorably with the respirability characteristics of spent fuel. The canister containing the vitrified HL W in borosilicate glass will also be compared with the physical integrity of the metal cladding surrounding the spent fuel pellets in reactor fuel assemblies. 2

3. PROPOSED RULE The DOE requests that the regulations in 10 CFR 71.63(b) be amended to exempt high-level waste canisters from the double containment requirement. as currently provided for spent fuel.

The high-level wastes are currently stored at three federal DOE defense-related sites and one former commercial reprocessing

• site. The waste is stored in various interim forms such as liquids, slurries, sludge, calcine, etc.. at each of the sites. The high-level waste will be heated and processed with material components of glass to form a homogeneous mixture which will be poured into a stainless steel canister, sealed and allowed to cool to form a solidified glass waste form. Because of the high degree of confinement provided by the waste canister and the physical characteristics of the solid plutonium-bearing waste glass (i.e., negligible respirable fines), a separate container inside the outer packaging is not considered to be necessary.

DOE therefore proposes that the last sentence of 10 CFR 71.63(b) be reworded to read as follows: "Solid plutonium in the following forms 1s exempt from the requirements of this paragraph: ( 1) Reactor fuel elements; (2) Metal or metal alloy;

• (3) Canisters containing vitrified high-level waste; and

( 4) Other plutonium-bearing solids that the Commission determines should be exempt from the requirements of this section." Added 3

4. DESCRIYfION OF HLW CANISTER AND CONTENTS During the vitrification process, molten borosilicate glass containing high-level radioactive waste will be poured into special stainless steel canisters which are cooled and sealed for eventual shipment of the solidified waste to a geologic repository. Such canisters of high-level waste will be produced at the Savannah River Site (SRS), Hanford (HANF). and at the West Valley Demonstration Project (WVDP). In addition, high-level waste from the reprocessing of DOE and Navy fuels at the Idaho National Engineering Laboratory (INEL) will eventually be vitrified and transported to the repository. Since the first canisters from INEL are not expected to be produced until 2014, the final decision on the waste form to be used has not been made at this time. This waste form will, however, be at least comparable in quality to those described herein. The vitrification process and final waste form characteristics have been defined for the SRS, HANF, and WVDP sites. A detailed physical description of the high-level wastes from each of these sites is provided in a U.S. DOE (OCRWM) publication issued in July 1992: "Characteristics of Potential Repository Wastes", DOE/RW-0184-Rl.1 Supporting data from this document is referenced extensively in the technical details which follow.

Cylindrical stainless steel canisters will be used at each of the three sites to enclose the borosilicate glass waste form. The canisters will be fabricated from 304 or 304L stainless steel which conforms to American Society for Testing and Materials (ASTM) specifications. Depending on the component. the base material may be pipe, plate stock or another form. The canisters will be 24 inches (61 cm.) in diameter and 118 inches (300 cm.) high, filled with borosilicate glass to about 85% of the total canister volume (to minimize the potential for overfilling the canister). "The canister designs for SRS and HANF are identical. The WVDP canister has the same outside diameter and length but has a smaller wall thickness and a wider filler neck."1 All canisters will meet the requirements defined in the Waste Acceptance System Requirements Document (W ASRD) - Ref. 2. A summary comparison of some of the relevant physical and radiological characteristics of the HL W glass waste forms and canisters from each of the 3 HL W sites is given in Table l. The estimates of maximum rac!ic-activity and thermal power are indicated as of the time of filling of the canister; these numbers are based on the most highly radioactive immobilized waste composition currently planned for these sites. The maximum values of these parameters are of prime importance in the design of the repository and transportation system. 1 The latest schedule for the commencement of vitrification operations at each of the three HLW sites is as follows: (1) 1993 for SRS, (2) 1996 for WVDP, and (3) 2000 for HANF. Since the limiting in terms of the maximum plutonium content per canister, the SRS canister and its HLW contents have been selected as the representative waste form package for this supporting study. A detailed description of the SRS HL W canister and its contents is given in the narrative which follows. Similar information for the HLW at Hanford and West Valley can be found in Reference l. 4

v Table 1. Summary of Canister Characteristics West Valley Savannah Demonstration River Hanford Project Site Site (WVDP) (SRS) (HANF) Nominal wall thickness cm 0.34 0.95 0.95 in 0.134 0.375 0.375 Weights, kg Canister 252 500 500 Glass 1,900 1,682 1,650 Total 2,152 2,182 2,150 Plutonium content per canister

• Weight, kg 0.130 0.352

.025 Activity, curies 361 3176 29 Total curies per canister

• 114,700 234,400 298,000

( 1. l 47E+o5) (2.344E+o5) (2.980E+o5) Watts per canister. 342 709 869 "These are estimated maximum values from ORIGEN2 calculations based on radionuclide compositions supplied by the sites. Curies and watts shown are at time of filling the canister, except for WVDP where the values shown are for the end of year 1991. For WVDP, maximum values are assumed to be 110% of average values. Maximum values for SRS and HANF do not necessarily represent initial operations. SOURCE: Ref. 1. 5

4.1 HLW CONTENTS Interim forms of HLW have been produced and stored onsite at Savannah River since 1954 as a result of the reprocessing of defense reactor fuels. These wastes are stored in large underground tanks where they have been allowed to settle and have been neutralized, resulting in the formation of a bottom layer of heavy sludge and a top layer of supernatant liquid. Subsequent evaporation of the lighter top layer has reduced the total volume of 1-Il.W produced by almost 60%, to a total volume of approximately 122,000 m 3 as of the end of 1989.1 The evaporation of the supernatant liquid, which contains almost all of the Cs-137 activity, has produced a saturated salt solution and a salt cake consisting of the salts crystallized out of the saturated solution. The major components of the salt solution and salt cake are sodium nitrate, sodium aluminate, and sodium hydroxide, together with most of the Cs-137. "Almost all of the radioactivity in the salt solution and salt cake is due to Cs-137 and its short-lived daughter Ba-137m."1 The sludge represents approximately 11 % of the total waste volume in the storage tanks. It "is . composed largely of the precipitated hydroxides of iron, aluminum, manganese, and other metals: it contains about 60-65% of the total radioactivity, including most of the Sr-90 and small amounts of actinides (principally isotopes of uranium, plutonium. and curiuin) that were not recovered from the fuel during reprocessing. The largest portion of the actinide radioactivity is due to the plutonium isotopes Pu-238 and Pu-241. The sludge is kept essentially separate from the salt solution and salt cake by storage tank selection and transfer operations."1 "Starting in 1993, the sludge and most of the radioactivity in the salt solution and salt cake will be processed at the Defense Waste Processing Facility (DWPF) complex at the SRS to produce canisters of borosilicate glass in which the HI.. W is dispersed and immobilized. The glass to be produced at this site is referred to as sludge-precipitate glass. It will consist of a blend of (1) washed sludge, (2) washed precipitate made by treating the salt solution"' in order to separate the cesium and smaller quantities of other radionuclides, and (3) glass frit Consistent glass quality will be achieved through the careful monitoring and control of glass waste pour temperature, pour rate, and chemical composition. The borosilicate glass waste form will contain approximately 28 wt% sludge oxides. After the HLW canisters have been filled, cooled, sealed and decontaminated, they will be transferred to interim storage buildings until final shipment to the geologic repository. The radionuclide composition estimated by SRS to represent the most highly radioactive glass likely to be made at this site is shown in Table 2; this is the best current estimate of maximum activity per canister. The data given in this table is based on sludge aged an average of 5 years and a cesium-containing precipitate derived from the supernatant liquid which has aged an average of 15 years. The total activity and decay heat at the time of filling of the canister (based on the maximum limiting radioactivity values given in Table 2) are 234,400 Ci and 709 W per canister. SRS has made a forecast of the radionuclide content of the glass produced during each 6

year of vitrification plant operation. The calculated average radioactivity of canisters produced through the year 2020 is 65,900 Ci per canister, considerably less than the maximum. Finn estimates of the detailed radionuclide compositions of individual feed batches are not expected to be available until about one year before the start of vitrification of each batch.1 Seven reference compositions of HL W borosilicate glasses which span the range of compositions expected to be produced at the SRS arc given in Table 3. Four of these compositions, denoted as Batches 1 - 4, have been projected from the existing HLW inventory. These compositions arc representative of the actual waste material which will be vitrified during the first 10 years of operation of the DWPF. In addition, three hypothetical glass compositions have been projected. The firs~ denoted as Blend, is a mixture of Batches 1 - 4 and represents the glass composition used as the design basis for the DWPF. The Purex and HM glasses are hypothetical extreme ranges of possible. glass compositions. The Purex composition represents the lower design limit on glass viscosity. This is a "worst-case" bounding composition in terms of glass durability. The HM waste glass composition contains a high level of aluminum waste and is representative of the upper design limit on glass viscosity. 4.2 CANISTER DESIGN Design specifications and dimensions of the SRS canister are given in Figures 1 and 2. 'The main body of the canister is made of schedule 20 type 304L stainless steel pipe with an outside diameter of 61 cm. and a nominal wall thickness of about 0.95 cm. The overall length of the canister is 300 cm. (118 in.). The weight of the empty canister is about 500 kg. (1,100 lbs.). Each canister will contain 0.626 m3 of glass, or about 1,680 kg. (3,710 lbs.), when loaded to about 85% of its total volume. The density of the reference glass is about 2.73g/cm3 at a temperarure of 25 deg. C. The total weight of a loaded canister is therefore about 2,180 kg. (4,810 lbs.)." 1 The canister is fabricated and inspected according to the American Society of Mechanical Engineers (ASME) Code (Sections VIII and IX). Procurement specifications and inspection procedures will be documented. All welding operations will be performed in accordance with ASME Section IX - Welding and Brazing Qualifications. After fabrication, but prior to filling and sealing operations, the canister integrity is verified by test to show that leakage is less than 10*1 attn-cc/sec. Pressure testing of the canister to 225 psi is also performed. 3 Immediately after the canister is filled, a temporary seal plug is shrunk-fit into a sleeve in the neck of the canister. Shrinkage of the canister nozzle and sleeve occurs during subsequent cooling operations such that the closure plug becomes tightly sealed. After the canister cools, a helium pressure leakage test is performed to verify that the canister temporary seal leakage rate is less than 2 x 10°"' atm-cc/sec of helium to prevent moisture infiltration into the canister. If the.. seal fails this test. it is removed and replaced with another temporary seal plug. The outer surface of the canister is then decontaminated by blasting with glass frit, and the temporary seal plug and sleeve are pushed down further into the canister neck. The final seal is made by upset resistance welding a weld plug ( 5-in. diameter, 0.5 in. thick, 304L stainless steel material) into the canister nozzle to 7

complete the sealing of the closure. "A force* of 75,000 lb, a current of 225,000 amps. and a voltage of approximately 10 volts is used to make the l.5-sec weld. The technique was chosen after consideration of seven alternative processes, including gas tungsten arc, gas metal arc, plasma arc, Thennit, electron beam, laser beam, and friction welding, because of the high weld quality and relatively simple equipment required. " 4 Tests performed on the seal weld indicate that it is capable of withstanding at least 4,000 psi internal pressure while maintaining a leak tightness of 10..a attn/cc/sec. 8

Table 2. Radionuclide Content of Most Highly Radioactive Glass Expected to Be Produced at SRS (Ref.I). Mass Radioactivity Thermal Power Radionuclide (g/canister) (Ci/canister) (W /canister) Cr-51 l.008E-21 9.312E-17 l.996E-20 Co-60 l.502E-01 l.699E+02 2.619E+OO Ni-59 3.163E-01 2.397E-02 9.519E-07 Ni-63 4.824E-02 2.975E+OO 3.000E-04 Se-79 2.439E+OO l.699E-01 4.232E-05 Rb-87 9.961E+OO 8.719E-07 7.278E-10 Sr-89 l.470E-09 4.267E-05 l.473E-07 Sr-90 3.426E+02 4.675E+04 5.426E+Ol Y-90 8.795E-02 4.786E+04 2.653E+02 Y-91 3.085E-08 7.568E-04 2.715E-06 Zr-93 4.443E+02 l.117E+OO 1.298E-04 Zr-95 4.680E-07 1.005E-02 5.084E-05 Nb-94 5.147E-04 9.646E-05 9.830E-07 Nb-95 5.407E-07 2.l 15E-02 l.013E-04 Nb-95m 3.272E-l0 l.247E-04 1.730E-07 Tc-99 l.816E+02 3.079E+OO l.545E-03 Ru-103 5.217E-13 1.684E-08 5.827E-11 Ru-106 6.729E-01 2.252E+03 l.339E-01 Rh-103m 5.028E-16 l.636E-08 3.761E-12 Rh-106 6.346E-07 2.259E+03 2.167E+Ol Pd-107 2.863E+Ol l.473E-02 8.732E-07 Ag-llOm 2.647E-05 l.258E-0l 2.098E-03 Cd-113 l.472E-01 5.009E-14 8.420E-17 Cd-l 15m 4.763E-14 l.213E-09 4.518E-12 Sn-121m 1.336E-03 7.902E-02 l.581E-04 Sn-123 3.lOlE-05 2.549E-0l 7.951E-04 Sn-126 l.556E+Ol 4.415E-01 5.508E-04 Sb-124 4.071E-12 7.123E-08 9.445E-10 Sb-125 8.226E-01 8.496E+02 2.656E+OO Sb-126 7.365E-07 6.159E-02 l.138E-03 Sb-126m 5.619E-09 4.415E-01 5.622E-03 Te-l26m l.532E-02 2.760E+02 2.320E-01 Te-127 4.555E-08 l.202E-01

• l.622E-04 9

Table 2. (continued) Mass Radioactivity Thermal Power Radionuclide (g/canister) (Ci/canister) CW/canister) Te-128m l.302E-05 l.228E-01 6.597E-05 Te-129 l.457E-19 3.053E-12 l.089E-14 Te-129m l.576E-16 4.749E-12 8.316E-15 Cs-134 2.606E-01 3.372E+o2 3.433E+oo Cs-135 8.633E+-O 1. 9.943E-02 3.319E-05 Cs-136 l.068E-44 7.838E-40 1.066E-42 Cs-137 4.989E+-02 4.341E+o4 4.802E+ol Ba-136m 3.195E-50 8.607E-39 l.040E-41 Ba-137m 7.724E-05 4.155E+o4 1.632E+o2 Ba-140 1.404E-41 l.024E-36 2.853E-39 La-140 7.734E-43 4.304E-37 7.205E-39 Ce-141 l.260E-15 3.591E-11 5.250E-14 Ce-142 4.005E+-02 9.609E-06 0.OOOOE+OO Ce-144 3.093E+OO 9.869E+o3 6.547E+OO Pr-143 1.780E-39 l.198E-34 2.291E-38 Pr-144 1.306E-04 9.869E+o3 7.255E+ol Pr-144m 6.545E-07 l.187E+o2

• 4.063E-02 Nd-144 4.l lOE+-02 4.860E-10 0.OOOOE+OO Nd-147 1.570E-49 1.261E-44 3.038E-47 Pm-147 2.609E+-Ol 2.419E+o4 8.679E+oo Pm-148 4.243E-16 6.975E-1 l 5.364E-13 Pm-148m 4.722E-14 l.009E-09 l.277E-11 Sm-147 8.796E+ol 2.000E-06 2.738E-08 Sm-148 1.916E+-Ol 5.788E-12 6.901E-14 Sm-149 7.420E+OO l.781E-12 0.OOOOE+OO Sm-151 9.418E+OO 2.478E+o2 2.906E-02 Eu-152 2.132E-02 3.688E+oo 2.790E-02 Eu-154 2.295E+OO 6.196E+o2 5.543E+OO Eu-155 1.021E+OO 4.749E+o2 3.455E-01 Eu-156 9.489E-37 5.231E-32 5.392E-34 To-160 9.923E-11 l.120E-06

9. l lOE-09 11-208 3.829E-12 l.128E-03 2.645E-05 10

Table 2. (continued) Mass Radioactivity Thermal Power Radionuclide (g/canister) (Ci/canister) (W /canister) U-232 6.256E-04 l.339E-02 4.301E-04 U-233 1.636E-04 l.584E-06 4.605E-08 U-234 5.485E+OO 3.428E-02 9.875E-04 U-235 7.278E+Ol l.573E-04 4.122E-06 U-236 l.742E+Ol l.128E-03 3.054E-05 U-238 3.122E+04 l.050E-02 2.663E-04 Np-236 l.323E-06 l.744E-08 3.514E-ll Np-237 1.263E+Ol 8.904E-03 2.722E-04 Pu-236

• 2.297E-04 l.221E-01 4.249E-03 Pu-237 7.401E-16 8.941E-12 3.292E-15 Pu-238 8.667E+Ol l.484E+03 4.919E+Ol Pu-239 2.076E+02 l.291E+Ol 3.979E-0l Pu-240 3.809E+Ol 8.681E+OO 2.704E-0l Pu-241 l.620E+Ol l.670E+03 5.176E-02 Pu-242 3.206E+OO l.224E-02 3.616E-04 Am-241 3.210E+OO l.102E+Ol 3.661E-0l Am-242 l.776E-08 l.436E-02 l.628E-05 Am-242m 1.488E-03 l.447E-02 5.709E-06 Am-243 2.902E-02 5.788E-03 l.860E-04 Cm-242 l.057E-05 3.495E-02 l.288E-03 Cm-243 l.078E-04 5.565E-03 2.039E-04 Cm-244 l.329E+OO l.076E+02 3.763E+OO Cm-245 3.910E-05 6.715E-06 2.225E-07 Cm-246 l.739E-06 5.342E-07 l.747E-08 Cm-247 7.116E-09 6.604E-13 2.107E-14 Cm-248 l.614E-10 6.864E-13 8.533E-14 Totals 3.427E+04 2.344E+0S 7.093E+02 11

Major Glass Components Al1O3 B1O3 BaSO4 CaO CaSO4

• Cr1O3 CuO F~O3 FeO Group A" Group Bb K10 Lip MgO MnO N~O N~so.

NaCl NiO SiO1 ThO1 TiO1 U3Os Table 3. Projected DWPF Waste Glass Compositions Source: Ref 1. Constituent Sludge Type (Wt. %) Blend Batch 1 Batch 2 Batch 3 Batch 4 3.98 4.87 4.46 3.25 3.32 8.01 7.69 7.70 7.69 8.11 0.27 0.22 0.24 0.26 0.38 0.97 1.17 1.00 0.93 0.83 0.08 0.12 0.11 0.10 Trace 0.12 0.10 0.14 0.13 0.14 0.44 0.40 0.41 0.40 0.46 6.95 8.39 7.11 7.48 7.59 3.11 3.72 3.15 331 3.36 0.14 0.10 0.14 0.10 0.20 0.36 0.22 0.44 0.25 0.(i() ' 3.86 3.49 3.50 3.47 3.99 4.40 4.42 4.42 4.42 4.32 135 1.36 135 1.35 1.38 2.03 2.06 1.62 1.81 3.08 8.73 8.62 8.61 8.51 8.88 0.10 0.10 0.12 0.10 0.13 0.19 0.31 0.23 0.22 0.09 0.89 0.75 0.90 1.07 1.09 50.20 49.81 50.17 49.98 49.29 0.19 0.36 0.63 0.77 0.24 0.90 0.66 0.67 0.66 1.02 2.14 0.53 2.30 3.16 0.79 "Group A: radionuclides of Tc, Se, Te, Rb, and Mo. HM Purex 7.08 2.89 6.94 10.21

• 0.18 0.29 1.00 1.02 Trace 0.12 0.09 0.14 0.25 0.42 4.95 8.54 2.19 3.78 0.20 0.08 0.89 0.08 2.14 3.58 4.62 3.12 1.45 133 2.07 1.99 8.17 12.14 0.14 0.12 0.09 0.26 0.40 1.21 54.39 44.56 0.55 O.ol 0.55 0.65 1.01 2.89 h(iroup B: radionuclides of Ag, Cd, Cr, Pd, 11, La. Ce, Pr, Pm, Nd, Sm, Th, Sn, Sb. Co, Zr, Nb. Eu, Np, Am, and Cm.

12

cj.. J C b ~ .5" Ref --- SN o.tall Z Surf-"I(" ~-- :M.DCI"' 10.11'" ASME F

• D Head 24"' 00 a 311"" Thi& min 24* R OFD *~**,c11
• 414 ** Sf Fluld for I"' pipe 3/4"" OC!I
• I" SF.

... t<<ial: ASTM A240 Type 304L Hot fo,med °"' 1/111"" Thll a 34-li"" OOPlatt. EAimated We'9h1 1000 Iba Capacity to Top K.R. 25.3 a, h Toul lnaim Vol. 28.1 cu h 24.00"

• 3/1" Wall Pipe

...,.... : ASTM A3l2 Type 304L ASME fi.n.,d and"- D*lfled Heed 1.118"" 24" 00 a 311" 1111n Thie 24" II OfD MIi" IC" J.. Sf .._.: ASTM A240 Type 304L ORNL DWG 90-4 18 DETAILZ I r- --WJ--b~----t------t-~ ,.oo-*

• j~*

5.000" aO. 1.00" ao.o.r*.

---

1.13" Ilia

---

~-t.12"'dit-------

Figure 1. Savannah River Site HLW canister. Source: Ref. 1.

\\ ORNL OWG 90-419 C..,,.,., Weld P~ A....,bly SX400191 -- T Off,l)Oflf'( Cloa,** s,....,~ SX4 00800 Figure 2. Savannah River Site HLW Canister Closure. Source: Ref. I.

5. WASTE ACCEPTANCE SPECIDCA TIO NS The specifications which HL W forms from each of the sites must meet in order to be acceptable for transportation to and disposal in the geologic repository have been defined in the Waste Acceptance System Requirements Document. The form and content of the documentation which each site must submit to OOE/OCRWM to demonstrate compliance with the W ASRD will be defined by DOE's Office of Environmental Restoration and Waste Management (EM) since they are the responsible organii.ation within DOE for HLW form production. As a minimum, the documentation must include a Waste Form Compliance Plan (WCP), a Waste Form Qualification Report (WQR), Production Records, and Storage and Shipping Records.2 The WCP is a detailed description of the methods, analyses and programs which will be put in place. in order to demonstrate compliance with each of the specifications defined in the W ASRD. The results of the waste form testing and analyses defined in the WCP will be given in the WQR. The Production Records will describe the individual HL W canisters and their contents based on samples to be drawn during waste vitrification operations and other production documentation (e.g., welding records).

The Storage and Shipping Records will describe the phy~ical characteristics of each of the HL W canisters and contents. In addition, these records will identify any unusual events which have occurred during either interim storage or transportation. The W ASRD requires the HL W producers to establish, maintain, and execute a quality assurance program which satisfies each of the applicable criteria of the DOE OCRWM Quality Assurance Requirements and Description (QARD) document - Ref. 5. This QA program for the HL W sites will cover all activities from the time of waste form production through waste acceptance. As such, this program governs the HL W glass and canister production p~esses. The collection of QA records which includes the WCP, WQR, Production Records, and Storage and Shipping Records will form the basis for the acceptability of the HL W canisters for disposal in the repository. A summary of the major specifications defined in Ref. 2 which must be adhered to in order to ensure a high quality, consistent HLW glass product and canister is given below. 5.1 BOROSILICATE GLASS WASTE FORM Before production processing of the HL W into borosilicate glass commences, each site must make projections of the chemical composition and radionuclide inventory of the finished glass product. During actual production operations, each site must report the chemical composition and crystalline phase stability for the waste form. In addition, the oxide composition of the waste form must be reported for the oxides of elements present in concentrations greater than 0.5 wt% based on chemical analyses of samples. The estimated total and individual canister radionuclide inventory of the glass must be defined for all radionuclides which constitute more than 0.05% of the total activity of the glass and have half-lives greater than 10 years. The material composition of the waste form must be compatible with that of the canister such that no internal 15

corrosion of the canister talces place which would.adversely affect normal or abnormal handling, storage and transport operations. One of the most important aspects of the HL W vitrification operations which must be demonstrated is adequate control over the consistency of the waste product. Each site must demonstrate control of waste form production by comparing melter batch production samples against the Environmental Assessment (EA) benchmark glass using the Product Consistency Test (Ref. 6) or equivalent In addition, the concentratio_ns of lithium, sodium, and boron in the leachate (after normalization for the concentrations in the glass) must be less than those of the benchmark glass. Finally, in order to preclude a nuclear criticality incident. the waste form shall be designed for criticality safety under both normal and postulated accident conditions such that the calculated effective multiplication factor after applying all uncertainties is below 0.95.2 5.2 CANISTER CHARACTERISTICS The canister enclosing the vitrified HL W glass waste form must be made of austenitic stainless steel with a concentric neck and a lifting flange. Canister dimensions, weight, and glass filling height are based on the requirements of the W ASRD. The canisters described in Section 4 of this document conform with these specifications, and will meet all other requirements identified in the W ASRD. The specific material composition of the canister and its components must be reported to DOE/OCRWM. This shall include the ASTM alloy specification and composition of the fill canister material, canister label material, any filler material used for welding, and the method of fabrication of the canister. Each canister is also required to have a unique alphanumeric identifier which must be clearly visible from the top and side of the canister at least until the end of the retrievability phase at the repository. The cover gases used to provide protection against corrosive processes will be helium, argon, or other inert gases. The leak rate of these gases from the outermost closure of the canister shall be less than 10 atm-cm3 /sec. 2 5.3 FINISHED PRODUCT CHARACTERISTICS The finished product specifications detail a wide range of requirements for the sealed canister, ranging from limits on surface dose rates and internal heat generation to drop tests. The canistered waste form shall not contain explosive. pyrophoric, or chemically reactive materials in an amount that could compromise the repository's waste isolation capability. After closure; the canister shall not contain: (1) free gases other than air, cover, and radiogenic gases; and (2) detectable amounts of organic materials. The canistered waste form must be capable of remote handling using a grapple design specified by each site, and must maintain its dimensions throughout normal handling operations encountered during storage, transportation and repository disposal. At the time of shipment. the canistered HL W form must be capable of withstanding a 7-meter drop onto a flat. essentially unyielding surface without a release of its contents. The results from this canister impact test shall include information on the measured canister leak rates and deformation. 2 16

The requirement of a 7-meter drop test applies only to the canistered HLW. It should not be confused with the hypothetical accident condition tests which are specified in 10 CFR 71.73. The 9-meter drop test "onto a flat, essentially unyielding, horizontal surface" in 10 CFR 71.73(c)(l) applies to the entire package. The cask which is used to transport the HL W canisters must be certified by the NRC to withstand the 9-meter drop tesL 17

6. COMPARISON TO REACTOR FUEL ELEMENTS The canisters containing vitrified HLW are plutonium-bearing solids of the type that are suitable for exemption from the double containment requirement based on the justification given by the Commission in the Statement of Considerations for the exemption of reactor fuel elements.

Reactor fuel elements were exempted because they are considered to be "essentially nonrespirable." The plutonium contained in reactor fuel elements is encased in a solid ceramic fuel pellet matrix surrounded by a sealed and sturdy metal cladding material which inhibits any possible dispersion of the hazardous fission products. 6.1 HLW GLASS VERSUS FUEL PELLETS The canisters of vitrified HL W are at least as good as spent reactor fuel elements with respect to their low dispersability characteristics and their baniers to radionuclide release. The HLW glass form serves as the initial barrier to the release of plutonium and other immobilized radionuclides. Production of potentially respirable particles within the glass matrix could result from these sources: ( 1) cooldown processes after being poured into the I-ll., W canister. (2) normal handling and transport operations; and (3) hypothetical accident conditions. The sources causing the production of potentially respirable particles for reactor fuel elements are identical with the exception of the first source. For reactor fuel elements, irradiation of the fuel rods causes cracking of the 1.0 - 1.5 cm. long U02 pellets and other dimensional changes due to the high operating temperature and pressures. Post-irradiation examination of typical fuel rods have indicated that cracks form in both the radial and longitudinal direction so that a previously whole pellet consists of 20 - 40 interlocking pieces. The sharp corners of the dished fuel pellets are often crushed into many small fragments.20 The durability and fracture resistance of the HL W glass form are desirable characteristics since they limit the production of fine particles potentially available for inhalation and dispersion as a result of mechanical stresses induced during postulated transport conditions. Impact studies of Savannah River simulated waste glass (SRL 131 formulation) have shown comparable levels of fracture resistance and similar fractions of respirable particles when compared to unirradiated UO2 pellets and other solid potential waste form materials. An experimental laboratory-scale brittle fracture study was conducted by Argonne National Laboratory in the early 1980's (References 7 and 8) to measure the size distribution of the impact fragments for several different simulated glass waste forms and ceramics. The impact test consisted of placing a cylindrical specimen of potential waste form on its side between two hardened tool steel plates inside a sealed chamber. Each specimen was impacted by a standard weight from a preselected height so that the available impact energy per unit volume of the test specimen was identical for comparative purposes. The size and number of particles produced by the impact were measured using a variety of techniques which were determined to give the most accurate results. The fraction of respirable particles produced by equivalent impacts (at an energy density of 1.2 J/cm3) was determined to be 0.02 wt.% for the UO2 fuel pellet specimens and 0.016 wt.% for the SRL131 simulated waste glass. Sensitivity studies performed at varying energy densities indicates that the respirable fraction 18

increases linearly with increasing energy densities (drop heights). Although the extrapolation of these laboratory-scale results to larger sizes cannot be justified due to the lack of proven scaling laws, the results are useful in comparing the relative properties of the unirradiated materials. The fracture resistance of simulated HL W glass is therefore comparable to that of UO2 fuel pellets. Another factor which should be considered in evaluating the potential for respirability of plutonium particles is the total quantity of plutonium present in each of the waste forms. As noted in Table 1, a maximum quantity of approximately 350 grams of plutonium will be contained in the Savannah River HLW canister. Since the nominal volume of HLW glass is 6.26 x 105 cm3, the concentration of plutonium in the HL W glass form is approximately 5.6 x 104 grams of plutonium per cm3 of waste. Based on ORIGEN2 analyses of the nuclide composition expected for average burnup PWR fuel which has been allowed to decay for 5-years prior to shipment, the total quantity of plutonium per assembly is on the order of 4000 grams. This yields a plutonium concentration of approximately 7.8 x 10*2 grams per cm3 of spent fuel. Thus, the concentration of plutonium in an individual fuel assembly is more than 100 times greater than that in a HL W canister. This result should be intuitively expected since the HL W was produced as a byproduct of the reprocessing of commercial and military fuels to extract useful plutonium and uranium. In addition, the maximum quantity of plutonium projected for the Hanford and West Valley HLW canisters is much less than that of the Savannah River _HLW canister. 6.2 HLW CANISTER VERSUS FUEL ROD CLADDING The HL W metal canister serves as an additional barrier to the potential release of radionuclides, including plutonium, into the interior of a HL W transport cask. The canister is similar in this respect to the reactor fuel cladding. The structural integrity of the HL W canisters has been demonstrated through numerous impact tests at Pacific Northwest Laboratories (PNL) and Sandia National Laboratories (SNL). The results of these tests are described in the following section. Canisters of vitrified HLW are required to be capable of withstanding a drop of 7-meters onto a flat, essentially unyielding' surface, as specified in the WASRD. The drop test analyses will be documented in the WCP and WQR for each of the HL W sites. Final closure weld controls and procedures for the HLW canisters will also be described in detail in the WCP, WQR, and Production Records which are required to be submitted to DOE/OCRWM (per the W ASRD - Ref. 2) by each interim HL W storage site. Another factor contibuting to the integrity of the HL W canisters, as compared to the spent reactor fuel cladding, is that the metal canisters have not been exposed to the high levels of radiation existing in a commercial power reactor. High levels of radiation have been shown to cause slight molecular-level changes in material properties, including increased embrittledness. The protection that the canister provides should be at least comparable to that provided by spent fuel cladding. 19

7.

CANISTER DROP TESTS A number of drop tests using canisters contammg simulated HL W glass forms have been performed over the past 20 years. These tests have yielded important information on the durability of the canisters and the fracture resistance of the simulated_ HLW glass form. This provides further support for the position that canisters containing vitrified HLW are a waste fonn that qualifies for exemption from the double containment requirement A summary of the results of drop tests of bare scale-model or full-size canisters containing simulated HLW onto an unyielding surface is given in Table 4. 3 The testing program included a number of different canister and glass combinations with variations in the drop orientation and the angle of impact. There was no release of the glass material from any of the canisters as a result of the impact Helium leak tests and dye penetrant tests conducted following the impacts have shown the ability of these canisters to withstand an impact (9-meter drop) greater than that required by the WASRD (7-meter drop) with no penetration of the canister shell. 7.1 CANISTER INTEGRITY A total of at least 13 different canisters made of the reference Type 304L stainless steel have been drop-tested from a height of 9 meters ( ~ 30 feet) in various studies sponsored by the Department of Energy (References 9 - 13). None of the canisters showed any observable evidence of rupture or cracking. Ten of the canisters of the reference dimensions identified in the WASRD were leak-tested with helium following the drop tests with no leakage detected greater than the instrument sensitivity (2.4 x 10*10 std-m3/sec/scale division). In addition, nine of these ten canisters were subjected to a dye penetrant examination in the area of the damage zones to detect for any non-visible cracks or flaws. No defects were observed based on these tests. Two of the studies measured strain levels by the use of strain circles etched on the canisters and measured before and after the impacts. Maximum strain levels of 12 - 16% were observed, well below the 55% strain level at which Type 304L stainless steel begins to exhibit signs of failure. 14 Detailed descriptions of the testing methodology and r'!sults can be found in References 9 - 13. 7.2 HLW GLASS PARTICLES Of critical importance in the scientific/technical basis for the exemption of the HLW canister from the double containment requirement is the evidence that the waste form is "essentially nonrespirable". Several studies have performed detailed analyses on the particle size distribution generated as a result of the maximum mechanical forces that would be imposed on the HLW canister during shipment (i.e.. following the canister drop tests). Despite the mechanical impact forces exerted upon the HLW canister during both normal transport and postulated accident conditions, nearly all the glass waste form remains in a solid form of large shards with limited generation of small particles. 20

7.2.1 MCC-15 Test And Analyses The Materials Characterization Center (MCC) at PNL has been developing standard tests to characterize the performance of nuclear waste forms under both normal and accident conditions. As a part of this effort, the MCC has developed "MCC-15: Waste/Canister Accident Testing and Analysis" (Ref. 11). This test method was developed in order to provide data on m...w canister integrity, deformation, and waste fonn particle size distribution following a free drop impact under standard accident conditions. As a part of the verification of the test method, two separate tests were performed using one full size prototypic Savannah River DWPF canister. The canister was filled according to the DWPF reference process (Ref. 15). The canister was dropped from a height of 9 meters onto its lx>ttom comer at an angle that placed the canister's center of gravity over its comer. Following the impact and subsequent evaluation of canister integrity, the canister was disassembled and the waste form removed to determine the particle size distribution. The results of this screening and sieving process (described in Ref. 11) yielded about 50 g. of particles of respirable size (smaller than 10 µm). Since the DWPF canister contains approximately 1680 kg. of waste glass with a maximum plutonium activity of 3176 Ci., the 50

g. of respirable fines corresponds to a plutonium activity of less than 0.10 Ci. The quantity of respirable particles contained within the intact canister as a result of the impact corresponds to 0.003 wt. % of the total canister waste form mass; this fraction is considerable less than that observed in the laboratory-scale impact tests (0.016 wt. %).

7.2.2 Other PNL Impact Tests and Analyses The MCC at PNL performed additional impact testing and analyses which were completed in December 1988 (Reference 12). Two full-scale DWPF canisters filled with reference simulated borosilicate glass waste were impacted under either normal (0.3 m. vertical drop) or accident conditions (9.1 m. vertical drop). The goals of these series of tests were: (1) to generate data on waste fine generation as a result of the impacts; and (2) to measure particle release through holes which were intentionally drilled into the canisters to estimate the effect of an artificially imposed and arbitrary worst case event 12 Four holes with a diameter of 0.28 cm. were intentionally drilled into the shell of the canister which had been dropped from a height of 0.3 m and the canister was subsequently transported over 2000 miles to simulate actual transport conditions. Three of the holes were located in the canister impact area (canister lx>nom) to provide conservative (maximum) estimates of the mass and size distribution of fines released from potential canister flaws. The quantity and size of the fines released through these manufactured flaws was measured carefully using filter assemblies attached to each hole. The canister was placed in a horizontal orientation inside a wooden box which was placed on the bed of a tractor trailer for the highway round trip from the Hanford Site to Cheyenne, Wyoming and back. During this transportation flaw leak testing, between 0.1 and 260 mg. of glass particles exited each hole. However, only 0.044 to 12.3 mg of these glass particles were in the respirable size range (i.e, with a diameter smaller than 10 µm). 21

Following the transportation flaw leak testing *of the normal condition canister (0.3-m. drop), pressurized flaw leak testing was performed on both impacted canisters. The purpose of this test was to determine the mass and size distribution of glass fines that would exit a design flaw in a pressurized canister following the free drop impact tesL For the canister which had been subjected to the 0.3-m. drop, the four holes that had been created in the transportation flaw leak test were welded closed. For both DWPF canisters, four holes were again drilled into the wall of each canister with all holes located in the bottom 7.5 cm. of the impacted canisters. The holes were then closed and the internal plenum of the canister was pressurized to a predetermined level. The holes were then opened and the canister was allowed to depressurizc. During depresswization, the glass waste particles exiting each of the holes was collected and measured in terms of quantity and size. The variations in testing conditions and results arc as follows: (1) For the normal condition canister, the diameter of the four holes was 0.28 cm.* and the canister was pressurized to 2.0 psig. All particles exiting the artificially created canister flaws were collected and the particle size distribution determined. Only 1.2 to 2.9 mg of glass fines were detected exiting these holes, with only 0.78 to 1.9 mg being less than 10 µmin diameter. (2) For the accident condition canister (9.1-m. drop), the diameter of the four holes was 0.95 cm. and the canister was pressurized to 3.0 psig. An analysis of the quantity and sizes of particles exiting these flaws during depressurization yielded a total quantity of 79 to 333 mg. per hole, with the maximum quantity of respirable particles released through any one hole being 3.04 mg. Upon completing all the leak testing, each of the canisters was disassembled and the particle size distributions of the entire canister were measured. The quantity of particles in the respirable category ( < 10 µm) was 61 g. (0.004 wt.%) for the normal conditions canister drop~ from 0.3 m., and 239 g. (0.014 wL % ) for the accident conditions canister. The total activity of plutonium for the largest quantity of particles produced (i.e., 9.1 m. free drop and fine sizes < 10 µm) corresponding to 239 g. is 0.45 Ci.3 Since the total quantity of respirable fines was approximately 4 times greater for the canister dropped from 9.1 m. and these fines were concentrated in the bottom (impacted) area of the canister, the majority of the respirablc fines for the 9.1 m. impacted canister were a result of the impact. A summary of the particle size distributions for both the PNL normal conditions canister drop and the 9.1 m. (30-ft.) drop test is given in Fig. 3, along with similar results from the MCC-15 test described in 7.2.1 and other past impact test data. The similarity of the results of these different tests confirms the consistency of the particle size results. The quantity of respirable fines produced is minute and provides full support for the classification of the HLW canister waste form as "essentially nonrespirable", analogous to reactor fuel elements. An evaluation of the significance of the total quantity of respirable particles contained within the canister as a result of the 9.1 rit drop can be performed by comparison to the methodology used for a corresponding safety analysis for spent fuel transport. For shipping cask 22

safety assessments, it is assumed that 0.003% o( the spent fuel is considered to be released as a result of cladding failure of a single fuel rod. Furthermore, the portion of the fuel particles ejected from the fuel which are potentially respirable as a result of cladding failure is approximately 10%. 16 These assumptions are based on experiments conducted at Banelle and Oak Ridge National Laboratory. If similar conservative assumptions are made for the HLW canister, then 0.007 g. of respirable HLW glass particles (0.00003

• 239 g.) are assumed to escape from the canister into the interior of a transportation cask. This amount corresponds to a plutonium activity of 1.4 x 10*5 curies, or 5 x 10*3 A2 values. The assumption that 0.003% of the available HLW glass particles escape into the interior of the shipping cask cavity during a postulated breach is borne out by similar magnitude results from the PNL drop tests. A total of 0.0013% of the respirable size glass particles escaped from the holes which were intentionally drilled into the canister body (3.04 x 10"3g/239 g). Since the HLW canisters are robust, it is reasonable to conclude that any HL W glass particles that are produced in the glass matrix will be retained within the canister. A number of independent drop tests have confirmed the structural integrity of the canisters following extra-regulatory impacts. During actual transport conditions, the HL W canister will be enclosed within a shipping cask -- resulting in reduced canister damage and minimal production of HL W glass fines.

23

Table 4. Summary of DWPF HLW Canister Drop Tests TEST RESULTS CANISTER TEST DROP WEIGHT GLASS DYE HELIUM TEST REPORT NUMBER SCALE NUMBER AND ORIENT A TION PROPERTIES PENETRANT LEAK Peterson and Alzheimer 1-4,(18-21) 1/1 I 9 M. Bottom Comer Ref. Borosilicate PNL-5250 Each of 4 2 I M. Side Puncture Glass Waste canisters 3 9 M. Nozzle Corner (1983) No Cracks No Leaks tested 3 times. Peterson, Alzheimer, 1 (4) 1/1 1 9 M. Nozzle Corner Ref. Glass No Leaks Slate 2 (5) 1 9 M. Bottom Comer PNL-5251 2 1 M. Side Puncture No Cracks No Leaks ~ Slate, Pulsipher, Scott 1/1 9 M. Bottom Comer Ref. Glass No Cracks No Leaks PNL (Waste Management 1985 Paper) Famswonh and Mishima 1 (A27) 1/1 1 0.3 M. Bottom Frit 165 (Ref.) No Cracks No Leaks PNL-6379 2 (AIO) 1/1 1 9 M. Bottom @ Pour Rate No Crdcks No Leaks of 240 lb/hr. Uncapher, Madsen, Stenberg (SAND87-2516)

• Tested Bare I

1/l 1 9 M. Bottom Frit 165 @ Visual Inspection Revealed No Flaws 2 Iii. I 9 M. Top at -20oF Ref. Pour Rate Visual Inspection Revealed No Flaws

• Tested Inside 0

1/l 1 9 M. Bottom Visual Inspection Revealed No Flaws 1/l-Scale Cask 1 l/l 2, 3 9 M. Top Puncture Visual Inspection Revealed No Flaws 2 l/l 4, 5 9 M. Side Puncture Visual Inspection Revealed No Flaws 3 1/l 6, 7 9 M. C. G. Bottom, Side Visual Inspection Revealed No Flaws

a.... 'JI' f'N L

• lo f=T' Oc.u t-o,.. L* NOCM 1,, £,o.. o,TIOH\\

I o... on OtLW C:.nille, I r.. o.. 10 100 Figure 3. Summary of Impact Particle-Size Distribution Results 25

8.

PROGRAMMATIC JUSTIFICATION 8.1 ALTERNATIVES The only alternative would be to initiate efforts to design a double containment transportation cask. Technically, this could be accomplished. However, it is unnecessary from a safety perspective since the plutonium embedded in the solid glass waste form is essentially nonrespirablc and a level of containment is provided comparable to that provided for spent fuel elements (which only require single containment). The technical basis for a double containment requirement for the borosilicate glass HL W logs during transport from the HL W producer sites to the geologic repository for permanent disposal is unsupported. A double containment requirement for HLW in the form of borosilicate glass contained within sturdy metal canisters would impose an unnecessary and onerous burden on the DOE as discussed below. 8.2 OPERATIONAL CONSIDERATIONS The design of a cask for double containment would add additional handling steps to the loading and unloading of the HLW canister, increasing the time (and resultant expense) required to return the cask for the next shipmenL The operating efficiency of the CRWMS transportation program is reduced by these unnecessary handling steps. In addition, this would increase the total radiation dose received by workers loading and unloading the cask. One of the goals of the CRWMS is to keep radiation exposure to a level that is ALARA (As-Low-As-Reasonably-Achievable). The addition of a second level of containment would also create more activated metal hardware which must be disposed of. Another major factor which must be considered is the potential reduction in payload capacity caused by the extra volume and weight of a second containment level. If more than one 1-Il... W canister is to be transported in a single shipping cask (as is likely for a rail cask), then the payload of double containment 1-Il... W canisters would likely be reduced. This, in tum, would negatively impact the operating efficiency of the CRWMS by creating the need for more shipments and a corresponding increase in risk to affected populations along the transportation corridor. 8.3 ECONOMIC CONSIDERATIONS - Life Cycle Cost Impact If the double containment requirement is enforced for this waste form, the total system life cycle cost of transporting these HLW canisters will be affected due to the added cost of: (1) the material comprising the added containment barrier, (2) the labor resulting from the extra handling steps in the loading and unloading of the canister from the transportation cask; and (3) the added number of shipments caused by a potential decrease in payload capacity of the cask. The added costs incurred by the inclusion of an additional level of containment can not be justified in terms of any incremental benefits to public health and safety. 26

9. REFERENCES

l.

U.S. Department of Energy, July 1992. Characte~istics of Potential Repository Wastes, DOE/RW-0184-Rl, Vol. 1 and 3, Office of Civilian Radioactive Waste Management, Washington, D.C.

2.

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