ML20140H364

Staff Evaluation of U.S. Department of Energy Proposal for Monitored Retrievable Storage

ML20140H364
Person / Time
Issue date:	03/31/1986
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
References
NUREG-1168, NUDOCS 8604040103
Download: ML20140H364 (143)

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NUREG-1168

! Staff Evaluation of i

U.S. Department of Energy Proposal

for Monitored Retrievable Storage 1

U.S. Nuclear Regulatory Commission Office of Nuclear Material Safety and Safeguards i

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i NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

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1. The NRC Public Document Room,1717 H Street, N.W.

Washington, DC 20555

2. The Superintendent of Documents, U.S. Government Printing Office, Post Office Box 37082,

Washington, DC 20013-7082 j 3. The National Technical Information Service, Springfield, VA 22161 Although the listing that fo!!ows represents the majority of documents cited in NRC publications, l

it is not intended to be exhaustive.

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NUREG-1168 Staff Evaluation of U.S. Department of Energy Proposal for Monitored Retrievable Storage I Manuscript Completed: March 1986 Date Published: March 1986 Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission r Washington, D.C. 20b55 y ~..,

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DISCLAIMER In large measure this document was preparet by the staff of the U.S. Nuclear Regulatory Commission from draft documentation provided by the U.S. Department

{ of Energy (DOE). Although the staff has every reason to believe that the in-formation is accurate, because of the limited time available for staff review, more recent additions, deletions, or corrections to the DOE proposal documentation may not be reflected accurately in this evaluation. Accordingly, the staff reserves the right to reevaluate and modify its conclusions to incorporate any additional information that may be submitted.

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ABSTRACT As directed by the Nuclear Waste Policy Act of 1982, the U.S. Department of Energy has prepared a proposal for the U.S. Congress for a facility that can

be used for the monitored retrievable storage of spent fuel from commercial users. This report describes the evaluation performed by the staff of the U.S. Nuclear Regulatory Commission of the design concepts for the monitored retrievable storage facility proposed by the Department of Energy. On February 5,1986 the NRC submitted its principal comments to the Department Energy in the letter shown on the following pages.

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? ! 1 NUCLEAR REGULATORY COMMISSION g g WASHINGTON, D. C. 20555

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..... February 5, 1986 l CHAIRMAN l

Mr. Benard C. Rusche, Director Of fice of Civilian Radioactive Waste Management U.S. Department of Energy Washington, D.C. 20585

Dear Mr. Rusche:

We are pleased to provide our comments to the Department of Energy (DOE or the Department) on its proposal to Congress for monitored retrievable storage (MRS). Our comments are based on the information provided to us by the Department in our consultative role as required by the Nuclear Waste Policy Act of 1982 (NWPA). This information has been provided primarily in the form of conceptual design information for the MRS.

Some comments stem directly from the Commission meeting at which you briefed us on the MRS proposal on January 23, 1986.

The review schedule issued by DOE did not accommodate an extensive review of the environmental assessment information provided. We note, however, that two of the three specific sites selected by the Department for consideration for the MRS have previously been subject to environmental analysis and evaluation for nuclear power plants by our agency in accordance with the National Environmental Policy Act of 1969.

In addition, the NWPA stipulates further environmental evaluation should the Congress approve the DOE proposal for an MRS.

Our comments are limited to our role as a regulatory agency.

In this regard the principal regulation governing the licensing of an MRS is 10 CFR Part 72. We are currently considering modifications to that regulation to clarify procedures and requirements the Department will be following if the Congress approves the proposal.

With respect to our review from a regulatory standpoint, we offer the following observations.  ;

1. Sitina - The preferred site identified by DOE for the MRS l is the site of the former Clinch River Breeder Reactor l Plant, which has already been shown to be a qualified l site from the standpoint of public health and safety for a nuclear power plant. Moreover, based on present information, the staff knows of no information which would disqualify the alternate sites. DOE, however, has recognized the need for further investigations and evaluation of the designated site as related to the particular characteristics of the MRS design.

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Mr. Senard C. Busche  ;

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2. Design - The MRS conceptual design appears reasonable

.from the standpoint of public health and safety.

Although an in-depth review would be required before the facility could be licensed, it appears from the conceptual design that each requirement in 10 CFR Part 72 can be met.

3. Cask Certification - DOE must design a safe and reliable transport system, including durable transport equipment.

You have indicated that transport casks developed under NWPA for tranFporting commercial spent fuel to a repository will be certified by NRC. Based on experience to date, spent fuel can be moved safely in NRC-certified casks.

4. Demonstration of Consolidation - The consolidation of spent fuel needs to be adequately demonstrated to assure that this operation can be performed on the production-scale contemplated for the MRS. To date, a few spent 1 fuel assemblies have been taken apart and the rods consolidated, and a significant number of fuel assemblies have been reconstituted (i.e., the rods have been removed and replaced within assemblies). In this sense the consolidation process is feasible. You have indicated in the Design Verification Plan (Appendix C to the Program Plan) your intent to test and demonstrate disassembly / consolidation equipment, principally at the Idaho National Engineering Laboratory.

5.. Safeguards - The NRC staff's analysis of the MRS safeguards provisions at the conceptual design stage indicates that all NRC safeguards requirements can be met, In addition to the above considerations having a bearing on l the health and safety of the public, our observations are of fered on the procedures and institutional relationships to be followed by the Department,

a. License Application - For DOE to meet its planned schedule, the license application you submit to NRC would have to be complete and technically nound, meeting all l NRC requirements. The NRC staff will continue to consult i with DOE during the preparation of its application.

b. Coordination with Repository Organizations - Because the MRS would prepare spent fuel to be compatible with repository requirements, DOE must closely coordinate efforts with each candidate repository organization.

Your schedule indicates submittal of a license application in 1989 for the MRS, approximately two years

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prior to the selection of the first repository site from the slate of candidate sites. The materials required for the disposal packages produced at the MRS might be vi

Mr. Benard C. Rusche i different for each candidate repository site according to the different physical and chemical properties of each 1 repository environment. There fore, the application for the candidate MRS facility will need to show how DOE will be able to accommodate each design. Another essential aspect to the preparation of the package is the close coordination required between the repository and MRS organizational quality assurance programs to assure that the repository requirements are met. At this time, the )

staff foresees no impediment at the MRS that would l foreclose repository options for package requirements.

c. Transportation Requirements - You have clearly indicated that you intend to use NRC-certified casks, but there remains a degree of uncertainty regarding your commitments to other NRC transportation requirements.

The DOE Draft Transportation Institutional Plan states that, "Further, when shipping commercial waste to facilities developed under the NWPA, the DOE has made the commitment to comply with DOT and NRC regulatory requirements that pertain to the transportation of nuclear materials." However, except to the extent required by NWPA, DOE is exempt from NRC transportation regulations. At the January 23 meeting, you indicated your intention to follow all NRC transportation regulations applicable to the commercial sector.

Therefore, you should clarify your transportation plans accordingly in the MRS proposal.

The above comments relate to the NRC staff's technical evaluation of the MRS proposal. There are, however, some difficulties and uncertainties associated with the procedural approach the MRS would follow. For example, integration of the Commission's NEPA responsibilities with its licensing responsibilities presents some conceptual dif ficulties. The DOE proposal assumes that DOE would submit an environmental report with its MRS application, and the NRC would prepare the environmental impact statement (EIS). While Section 141(d) of the NWPA relieves the Commission of the responsibility for considering the need for the facility, it is silent concerning alternative sites, the NEPA comment process, and cost-benefit analysis. This creates an anomalous situation where the Commission would be considering such factors af ter the Congress had approved the MRS and, perhaps, DOE's preferred site. How these procedural matters are resolved will significantly affect whether the 30-month licensing schedule j suggested by the DOE MRS proposal is reasonably achievable. ,,

We suggest, and you agreed at the January meeting, that i Congress could address the NEPA issues in legislation I authorizing the MRS. '

In addition to the above, the NRC staf f is preparing a report '

which provides additional detail on its evaluation of the design concepts for the MRS, principally from the perspective )

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Mr. Benard C. Rusche i of the requirements of 10 CFR Part 72. If the MRS is to be constructed and operated, it would be licensed pursuant to this regulation. For this re,-son, the staff used 10 CFR 72 as its primary guidance. Minor changes to this rule are being proposed to explicitly cover licensing of an MRS, should it be authorized by Congress. This evaluation refers to the current version of 10 CFR 72; however, the changes the Commission has under consideration may affect a few requirements. These areas are noted within this report. The report, which will soon be available, should be useful to DOE in developing its definitive design, if Congress approves its proposal.

Sincerely,

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Nunzio J. Palladino Chairman

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TABLE OF CONTENTS Page DISCLAIMER ............................,................................ 11 ABSTRACT ........,..................................................... iii 1 SUMARY AND CONCLUSIQNS ............................................ 1-1 1.1 Need ..........................................................

1.2 Feasibility ...................................................

1-1 1-1 1.3 Siting ........................................ 1-2 1.4 Design ........................................ ............... ............. 1-2 1.5 Cas k Ce rti f i ca t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

. 1-2 1.6 Demonstration of Consolidation ................................ 1-2 1.7 Safeguards ............. ...................................... 1-3 1.8 License Application ........................................... 1-3 1.9 Coordination with Repository Organizations .................... 1-3 .

1.10 Transportati on Requi rements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-3 1 2 INTRODUCTION AND BACKGROUND ........................................ 2-1 2.1 Nuclear Waste Policy Act of 1982 .............................. 2-1 2.2 Purpose and Characteristics of the MRS ........................ 2-2 2.3 Brief Physical Description of Facility ........................ 2-3 3 EVALUATION ......................................................... 3-1 3.1 Siting ........................................................ 3-1 3.1.1 Clinch River Site ................................... 3-3 3.1.2 Hartsville Site ..................................... 3-3 3.1.3 Dak Ridge Reservation Site .......................... 3-4 3.1.4 Conclusions ......................................... 3-5 3.2 Gene ral Des i gn C ri te ria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-5 3.2.1 Quality Standards ................................... 3-5 3.2.2 Protection Against Environmental Conditions and 3.2.3 Natural Phenomena .................................. 3-6 Protection Against Fire and Explosions .............. 3-14 3.2.4 Sharing of Structures, Systems, and Components ...... 3-20 3.2.5 Proximity of Sites .................................. 3-21 3.2.6 Testing and Maintenance of Systems and Components ... 3-22 3.2.7 Emergency Capability ................................ 3-22 3.2.8 Confinement Barriers and Systems .................... 3-26 3.2.9 Instrumentation and Control S 3-50 3.2.10 Control Rooms ...............ystems .................

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TABLE OF CONTENTS PAGE 3.2.11 Utility Services .................................... 3-53 3.2.12 Criticality Design and Control ...................... 3-60 3.2.13 Radiological Protection ............................. 3-63 3.2.14 Spent Fuel and Radioactive Waste Storage and Handling 3-73 3.2.15 Waste Treatment ..................................... 3-90 3.2.16 Decommissioning ..................................... '3-98 3.3 Accident Analysis ............................................. 3-99 3.3.1 Requirements ........................................ 3-99 3.3.2 DBA Description ..................................... 3-100 3.3.3 Description of Accident Analysi s . . . . . . . . . . . . . . . . . . . . 3-100 3.3.4 Discussion .......................................... 3-101 3.3.5 Conclusion ...................................... ... 3-102 3.4 Safeguards .................................................... 3-102 3.4.1 Objectives and Review Approach ...................... 3-102 3.4.2 MRS Site Protection ................................. 3-103 3.4.3 Shipment Protection ................................. 3-106 3.4.4 Safeguards Contingency Planning ..................... 3-108 3.4.5 Material Accountability Record Keeping .............. 3-109 3.4.6 NRC Experience in Applying Requirements ............. 3-111 3.4.7 Reassessment of Certain NRC Requirements ............ 3-112 3.4.8 Safeguards Conclusions............................... 3-112 3.5 Repository Interface ............................... 3-113 3.6 Transportation ................................................ ........... 3-114 3.6.1 Conclusion .......................................... 3-116 4 REFERENCES ..........'............................................... 4-1 FIGURES 2.1 Layout of MRS R&H building .................................... 2-4 3.1 Artist's rendition of the MRS at Clinch River site, Roane County, Tennessee ............................................ 3-2 3.2 Sealed storage cask ........................................... 3-29 3.3 Details of sealed s torage cask arrangement . . . . . . . . . . . . . . . . . . . . 3-30 3.4 D rywel l s to rage a r ra ngeme nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-31 3.5 Schematic diagram of R&H building filtration and exhaust ,

systems ....................................................... 3-37  ;

3.6 Schematic diagram of SF, HLW, and RHTRU flowpath in the R&H  !

building ..................................................... 3-77 '

3.7 MRS isometric rendering ....................................... 3-78 3.8 Liquid radwaste streams and processing ........................ 3-92 3.9 Solid radwaste sources and processing ......................... 3-94 j

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3.1 Laboratory / hot cell tests involving Zircaloy clad fuel rods .... 3-44 ';

3.2 Ory storage demonstrations involving Zircaloy-clad fuel ........ 3-45 1

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ACRONYMS ACI American Concrete Institute ACS American Concrete Society "

AISC American Institute of Steel Construction ALARA as low as reasonably achievable ANSI American National Standards Institute ASME American Society of Mechanical Engineers BWR boiling water reactor CFR Code of Federal Regulations CHTRU contact-f:andled transuranic (waste)

CRDRP Clinch Rivur Breeder Reactor Plant DBA design-basis accident DCS distributed control system DE design earthquake DOE U.S. Department of Energy 00T U.S. Department of Transportation EPBAX electronic private branch automatic exchange FES Final Environmental Statement HAW high activity waste HEPA high efficiency particulate air HLW high level waste HVAC heating, ventilation, and air conditioning ISFSI independent spent fuel storage installation LLW low level waste LWR light water reactor i

MOU memorandum of understanding MRS monitored retrievable sturage NEPA National Environmental Policy Act of 1969 NFPA National Fire Prntection Association NRC U.S. Nuclear Regulatory Commission NWPA Nuclear Waste Policy Act of 1982 OCRWM Office of Civilian Radioactive Waste Management PWR pressurized water reactor i

QA quality assurance i

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ACRONYMS (Continued)

RAD Regulatory Assessment Document RG regulatory guide

R&H receiving and handling RHTRU remote handled transuranic waste RO repository overpack SER Safety Evaluation Report ,

SF spent fuel SNM special nuclear material -

TRU transuranic waste TVA Tennessee Valley Authority UHF ultra-high frequency UPS uninterruptible power supply '

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SUMMARY

AND CONCLUSIONS In preparing this evaluation, the Nuclear Regulatory Commission Staff (NRC, staff) reviewed the information provided by the U.S. Department of Energy (DOE) in its proposal (00E/RW 0035, Vols I-III) to the Congress for an integrated monitored retrievable storage facility (MRS) for high lovel radfuactive waste and spent nuclear fuel. The information included an environmental assessment, a conceptual design report, a licensing plan, and a program plan. Neither the depth of information presented not the time available for its review were suf-ficient to allow the staff to reach conclusions that could be cersidered com- i parable with those in a staff safety or environmental evtluction related to a licensing decision. However, the level of detail was sufficient for the staff to provide preliminary judgments that are independent of tho.e reatned by DOE.

1.1 Need d

The NRC is neutral on the question of the need for an MRS. Whether or not the facility is needed is primarily a business decision within the overall waste management system being developed by DOE to implement the Nuclear Waste Policy Act of 1982. However, the packaging functions that would be provided by the MRS in preparing material for disposal in a repository are needed whether or not there is an MRS. The NRC supports the Congressional finding (Section 141(a)(5))

that disposal of high level waste in a repository should proceed, regardless of the action regarding the MRS.

1.2 Feasibility From the standpoint of public health and safety, the MRS is feasible. The natur( of the operations involved- passive storage and relatively simple mechanical processes--indicates that the MRS would create a limited potential for accidents or adverse consequences.

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The preferred site identified by the DOE for the MRS is the site of the fonner Clinch River Breeder Reactor Plant, which has already been shown to be a quali-fied site from the standpoint of public health and safety foi a nuclear power plant. Moreover, based on present informaticn, the staff knows of no informa-tion which woyld disqualify the alternate sites. DOE, however, has recognized the need for further investigations and evaluation of the designated site as related to the particular characteristics of the MRS design.

1.4 Desian The MRS conceptual design appears reasortable from the standpoint of public health and safety. Although an in-depth review of the detailed design would be required tefore the facility could be licensed, it appears from the conceptual design that each requirement in 10 CFR 72 can be met.

1.5 Cask Certification 00E must design a safe and reliable transport system, including durable trans-port equipment. 00E has indicated that transport casks developed under NWPA for transporting commercial spent fuel will be certified by HRC. Based on experience to date, spent fuel can be moved safely in i:RC-certified casks.

1.6 Demonstration of Consolidaticn The consolidation of spent feel needs to be adequately demonstrated to assure that this operation can be performed on the production scale contemplated for the MRS. To date, a few spent fuel assemblies have been taken apart and the rods consolidated, and a significant number of fuel assemblies have been recon-stituted (i.e., the rods have Leen removed ,and replaced within assemblies).

in this sense the consolidation process is feasible. DOE has indicated in the Design Verification Plan (Appendix C to the Program Plan) its intent to test and demonstrate disassembly / consolidation equipment, principally at the Idaho National Engineering Laboratory.

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1.7 Safeguards The NRC staff's analysis of the MRS safeguards provisions at the conceptual design stage indicates that NRC safeguards requirements apparently can be met.

1.8 License Application For DOE to meet its planned schedule, the MRS license application would have to be complete and technically sound, meeting at NRC requirements. The staff will continue to consult with DOE during the preparation of its application.

1. 9 Coordination with Repository Organizations Because the MRS would prepare spent fuel to be compatible with repository requirements, 00E must closely coordinate efforts with each candidate repost-tory organization. The DOE schedule indicates submittal of a license applica-tion in 1989 for the HRS, approximately 2 years prior to the selection of the first repository site from the slate of candidate sites. The materials required for the disposal packages produced at the MRS might be different for each candidate repository site according to the different physical and chemical properties of each repository environment. Therefore, the application for the candidate MRS facility will need to show how 00E will be able to accommodate each design. Another essential aspect to the preparation of the package is the close coordination required between the repository and MRS organizational quality assurance programs to assure that the repository requirements are met. At this time, the staff foresees no Impediment at the MRS that would foreclose repository options for package requirements.

, 1.10 Transportation Requirements '

DOE his clearly indicated that it intends to use NRC-certified casks, but there ,

remains a degree of uncertainty regarding DOE's commitments to other NRC trans-portation requirements. The DOE Draf t Transportation Institutional Plan states  !

that, "Further, when shipping ccmmercial waste to facilities developed under  !

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the NWPA, the DOE has made the commitment to comply with DOT and NRC regulatory requirements that pertain to the transportation of nuclear materials." However, except to the extent required by NWPA, DOE is exempt from NRC transportation regulations. At the January 23, 1986 meeting of the Commission the Director of the DOE Office of Civilian Radioacitve Waste Management indicated DOE's intention to follow all NRC transportation regulations applicable to the commercial sector. The Commission commented in its letter to the Director that this point be clarified in the MRS proposal to the Congress.

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i 2 INTRODUCTION AhD BACKGROUND 2.1 Nuclear Waste Policy Act of 1982 The Nuclear Waste Policy Act of 1982 (Act) requires that DOE " complete a de-tailed study of the need for and feasibility of, and shall submit to the Congress a proposal for, the construction of one or more monitored retrievable storage facilities for high-level radioactive waste and spent nuclear fuel." The ab-breviation MRS has become the common designation for a monitored retrievable storage facility and is so used in this report. The MRS is considered by the Congress to be an " option for providing safe and reliable management of... waste or spent fuel."

The Act also provides that "the Secretary [of the Department of Energy] shall consult with the Commission and the Administrator [of the United States Environ-mental Protection Agency), and shall submit their comments on such proposal to the Congress at the time such proposal is submitted." The Congress also recognized the Commission's licensing responsibility pursuant to section 202(3) of the Energy Reorganization Act of 1974.

That section states that "the Nuclear Regulatory Commission shall...have licensing and related regulatory authority" for " facilities used primarily for the receipt and storage of high-level radio-active wastes."

The Nuclear Waste Policy Act of 1982 does not specify when or whether an MRS will be constructed. That decision is to be made separately by the Congress.

When Congress decides to go forward with an MRS, DOE may then apply for a license under the Commission's regulatory authority. At the moment, the Com-mission's role--and consequently the role of this report--is advisory.

This report states bases for the Commission's comments on the proposal, as well as information on safety matters related to this type of facility. In its 2-1

review, the staff has drawn heavily from its experience with the regulation of  ;

non-reactor nuclear facilities, particularly waste and spent fuel storage I facilities.

The principal regulations that pertain to this type of facility are set forth j in 10 CFR 72, " Licensing Requirements for the Storage of Spent Fuel in an  ;

Independent Spent Fuel Storage Installation (ISFSI)" (February 28, 1985). If the MRS is to be constructed and operated, it would be licensed pursuant to this regulation. For this reason, the staff used 10 CFR 72 as its primary guidance. Minor changes to this rule are being proposed to explicitly cover ,

licensing of an MRS, should it be authorized by Congress. This evaluation  !

refers to the current version of 10 CFR 72; however, the changes the Commission has under consideration may affect a few requirements. These areas are noted i within this report.

In preparing its proposal for the MRS, DOE developed design concepts that pro-vide a suitable basis for cost estimates and a general comprehension of tech-niques to be used, including safety features. However, if DOE submits an application for a license for the MRS, the application would have to include considerably more detailed information on facility design and administrative controls. This present staff review, then, reflects only the depth of avail-able information. As may be discerned from the staff's evaluation, DOE has anticipated many of the questions and concerns consequent to this type of evaluation and is committed to providing the type of information expected dur-ing subsequent design phases.

2.2 Purpose and Characteristics of the MRS The MRS would receive, package, and store spent fuel and solidified high level radioactive waste until a long-term Federal repository is ready to receive it.

The spent fuel and waste would be monitored during storage and would be retrieved for shipment to the Federal repository.

The fuel assemblies used in light water power reactors consist of long rods of fuel held together in parallel arrays, with uniform spacing between the rods.

In the reactor, water flows in the spaces between the rods to carry heat away.

After the fuel assemblies have been used, and after initial radioactive decay, 2-2

the space between rods is no longer needed for heat dissipation. (This is particularly true if the fuel assemblies have been removed from the reactor for more than 5 years and many radionuclide:. have decayed.) In preparation for disposal at the repository, the rods in each assembly can be consolidated into a tightly packed array. Such consolidation would be dore at the MRS.

In addition to the consolidation of the spent fuel at the MRS, the fuel would be packaged in canisters that would be compatible with package requirements estab-lished for final disposal in a geological repository. Other operations at the MRS might include packaging of the high level waste (received as borosilicate glass " logs") into a final disposal canister and treating and packaging for final disposal the secondary wastes produced as a result of the consolidation and packaging operations.

These operations would be, for the most part, mechanical operations. In addi- l tion, liquid chemicals might be used occasionally to decontaminate and otherwise l clean plant areas and equipment. Most of the daily aci.ivity would involve tech-niques that are standard practice in facilities handling radioactive materials. I 2.3 Brief Physical Description of Facility Chapter 3 of this report provides a detailed description of the MRS. The MRS would be a " hot cell" complex consisting of a receipt and handling (R&H) build-ing, a large outside storage area, and many supporting structures. Both rail and road access would be available, and staging areas would be available should material arrive at a high rate.

All spent nuclear fuel and high level waste would arrive in specially designed transportation casks.

These casks would be constructed of a dense material to shield against radiation and would'be tightly sealed. The casks would be moved into the R&H building and stationed vertically beneath any one of the four pro-cess cells. With special shielding arrangements in place, the lid of the cask would be re:noved. The spent fuel assemblies would be withdrawn from above by cranes and placed in the process cell. Figure 2.1 shows the general layout of the R&H building.

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In the process cells, the assemblies would be prepared for disassembly and consolidation and placed in a horizontal position; all of the rods in several assemblies (three to seven at a time) would be removed laterally. The rods would contain all of the nuclear fuel material. They would be formed into a circular cross-section and pushed through an aperture in the cell wall, into a horizontally positioned canister. This horizontal canister would be sealed against the wall in another large cell on the opposite side of the wall. After the wall aperture on the side of the process cell was closed, the canister full of spent fuel rods would be moved to another area of the large cell. The can-ister air would be evacuated; the canister would be immersed in a special inert gas mixture (argon with trace helium); and a lid would be seal welded to close the canister opening.

The sealed, inerted canister could then be placed in a temporary lag storage vault built into the floor of the large cell or transferred to either of two handling areas of the cells. In these handling areas, the canister could be either loaded into a transport cask for transfer directly to a repository for final disposal (along with other similar canisters) or loaded into a reinforced concrete cask for transfer to longer term storage.

In addition to the R&H building, there would be other structures at the site.

A contact-handled transuranic waste (CHTRU) structure would be adjacent to the R&H building for storing only these types of wastes.

i As an alternate to cask storage, the canisters could be stored in drywells.

These drywells would be augered holes lined with concrete and steel; one canister would be placed in each hole. Both the casks and drywells would be monitored for temperature and leakage of canister gas. The drywell lid will be seal welded.

When the Federal repository is ready to receive the waste material stored in the casks or drywells, the canisters would be removed, inspected, and loaded into transport casks. The canisters would be moved to the repository on trains dedicated to that purpose only.

2-5

3 EVALUATION 3.1 Sitina On April 25, 1985, DOE announced that it had identified a preferred site for the MRS and two alternative sites. The preferred site is the site of the former Clinch River Breeder Reactor Plant (CRBRP). The alternative sites are the Tennessee Valley Authority (TVA) Hartsville nuclear plant site and a location on the Oak Ridge Reservation. The three proposed sites are in Tennessee. These locations were selected after a screening process that identified a geographic region for the site primarily on the basis of transportation requirements (Holter and Braitman,1985).

Figure 3.1 is an artist's rendition of the MRS at the Clinch River site.

I By its very nature, the MRS is not nearly so dependent on the physical char-acteristics of a site as is a repository. Moreover, two of the three sites proposed by DOE have already been subject to detailed NRC staff review in I connection with reactor license applications, and the third is near a site for which the staff developed review information. The paragraphs below address each site in some detail.

In its earlier reviews of the sites, the staff noted that the impact on the environment was being considered in accordance with the National Environmental Policy Act of 1969 (NEPA); in each pertinent instance, an environmental impact statement was completed. The Nuclear Waste Policy Act of 1982 provides that, except for the question of need, the full environmental review process shall be followed. Therefore, a new environmental impact statement would be prepared to fulfill the requirements of NEPA.

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3.1.1 Clinch River Site The NRC staff issued a Final Environmental Statement (FES) for the Clinch River site in February 1977 (NUREG-0139) and issued a supplement to that statement in 1982. A Site Suitability Report in the Matter of Clinch River Breeder Reactor Plant (NUREG-0786) was issued on March 4,1977, and revised in June 1982. The staff concluded that the " site is suitable for a facility of the general size and type as the CRBRP from the standpoint of radiological health and safety considerations." Because certain aspects of the CRBRP design were not complete, the staff could not reach a final position. However, these aspects were related to differences between a breeder reactor and a light water reactor and would have no bearing on the issue at hand. Indeed, the staff's judgment on that issue was that those differences were not likely to be significant.

In its earlier review, the staff reviewed, evaluated, and accepted all of the Clinch River site's physical characteristics related to natural phenomena.

These included meteorology (dispersion factors and strong storm effects),

hydrology (maximum probable flood and groundwater flow), and geology / seismology (local maximum earthquake acceleration and foundation support). In applying for a license for the MRS, DOE would have to review the analysis in the Clinch River FES and its supplement to determine what limitations, if any, remain applicable to the MRS. In addition, the detailed site investigations to be undertaken by DOE, if the Congress approves this proposal, must include a

, thorough assessment of the potential for solution cavities and their importance t'o foundation stability.

3.1.2 Hartsville Site The NRC staff issued an FES related to the construction of the Hartsville nuclear plants in June 1975 (NUREG-75/039). A Safety Evaluation Report (SER) for those plants was issued in April 1976 (NUREG-0014). (None of the Hartsville plants has been completed.) Neither the FES nor the SER for the Hartsville plants identified any siting topics as outstanding issues. In those reviews of meteorology, hydrologic engineering, and geology and seismology, the staff concluded that (1) the onsite meteorological data provide a reasonable, con-servative basis for preliminary calculations of atmospheric dispersion; (2) the flooding potential of the site has been properly identified and the elevation 3-3

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of the plant grade precludes flooding; and (3) for the safe shutdown earthquake, the appropriate reference ground acceleration for the seismic design of struc-tures at the Hartsville sita is 0.2 g. As with the Clinch River site, DOE would I have to review the Hartsville FES to determine what limitations, if any, re- l main applicable to the MRS.

3.1. 3 Oak Ridge Reservation Site Unlike the other two car.didate sites, the Oak Ridge Reservation site has not been selected and modified for another purpose; no safety evaluation reports or environmental statements are available that deal directly with the site.

However, it is near a site that was reviewed by the staff, the site proposed by EXXON Nuclear in 1977 for its Nuclear Fuel Recovery and Recycling Center.

Although the staff did not complete its review of the EXXON application (because the application was withdrawn), the staff did reach some conclusions that are pertinent to the site proposed as an alternative site for the MRS.

The following observations can be made from internal NRC memoranda prepared as part of the review of the EXXON application (circa 1977-78).

(1) The design-basis earthquake for the Oak Ridge Reservation site would have a 0.25 g zero period peak acceleration, identical to that of the CRBRP and EXXON sites.

(2) Ninety mph is a reasonable design " fastest mile" wind velocity for the site.

(3) The Region I tornado intensity as defined in Regulatory Guide (RG) 1.76 (maximum speed of 360 mph) should apply to the Oak Ridge Reservation j

site. '

(4) The Oak Ridge Reservation site is probably well above maximum probable flood stage.

However, although the staff has some data on the Oak Ridge Reservation site, considerable effort would be necessary to obtain data comparable to those al-ready gathered and evaluated for the other two sites. To complete its evaluation 3-4 l

the staff would need data from borings and site surveys and from the mete-orological tower for the exact site location.

3.1.4 Conclusions Subpart E of 10 CFR 72 gives the siting evaluation criteria pertinent to the MRS. At this time, there is no technical reason that any of the proposed sites would be unacceptable for the MRS.

3.2 General Design Criteria Subpart F of 10 CFR 72 provides the criteria for the design, fabrication, construction, testing, and performance of structures, systems, and components important to safety in a storage facility. The following sections (1) describe how the MRS would meet each of the criteria and (2) evaluate the status of de-velopment as related to the assurance of public health and safety.

3.2.1 Quality Standards 3.2.1.1 Requirements 10 CFR 72.72(a) states:

Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance to safety of the function to be performed.

3.2.1.2 Description The Conceptual Design Report (Parsons, MRS-11, September 1985, Vols I and II) describes the standards that would be used in constructing the MRS. Except where noted in the detailed discussion that follows, the staff finds these selected standards adequate. No attempt is made to describe each standard in j this section; however, if the Congress authorizes the MRS, the staff will review the applicability of each standard in detail.

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3.2.1.3 Discussion 1

Subpart G of 10 CFR 72 involves a quality assurance program that, in its essentials, is identical to that in Appendix B to 10 CFR 50, the Commission's regulations for nuclear power reactors. DOE committed to the requirements of Standard NQA-1 of the American National Standards Institute and the American Society of Mechanical Engineers (ANSI /ASME), which implies full compliance with the Commission's quality assurance requirements.

3.2.1.4 Conclusions and Recommendations A commitment to ANSI /ASME NQA-1 at this stage of design is sufficient. If Congress authorizes the MRS, the initial NRC formal regulatory action would be a pre-application quality assurance evaluation and inspection.

3.2.2 Protection Against Environmental Conditions and Natural Phenomena 3.2.2.1 Requirements 10 CFR 72.72(b) gives the requirements for protection against environmental conditions and natural phenomena as follows:

(1) Structures, systems and components important to safety shall be designed to accommodate the effects of, and to be compatible with, site characteristics and environmental conditions asso-ciated with normal operation, maintenance, and testing of the ISFSI, and to withstand postulated accidents.

(2) Structures, systems and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, lightning, hurricanes, floods, tsunami, and seiches, without impairing their capability to perform safety functions. The design bases for these structures, systems and components shall reflect (i) appropriate consideration of the l most severe of the natural phenomena reported for the site and surrounding area, with appropriate margins to take into account the limitations of the data and the period of time in which the l

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data have accumulated, and (ii) appropriate combinations of the effects of normal and accident conditions and the effects of natural phenomena. An ISFSI need not be protected from tornado missiles but should be designed to prevent massive collapse of building structures or the dropping of heavy objects on to the stored spent fuel as a result of building structural failures.

(N.B.: The Commission is considering modifying this sentence because the protection afforded oy the large mass of water in the wet storage (deep pools) case may no longer exist for dry storage.)

(3) Capability shall be provided for determining the intensity of natural phenomena that may occur for comparison with design bases of structures, systems, and components important to safety.

(4) If the ISFSI is located over an aquifer which is a major water resource, measures shall be taken to preclude the transport of radioactive materials to the environment through this potential pathway.

10 CFR 72.3(w) defines " structures, systems and components important to safety" as those features of the ISFSI whose function is (1) to maintain the con-ditions required to store spent fuel safely, (2) to prevent damage to the spent fuel during handling and storage, or (3) to provide reason-able assurance that spent fuel can be received, handled, stored and retrieved without undue risk to the health and safety of the public.

3.2.2.2 Description 3.2.2.2.1 Items Important to Safety i

The MRS Facility Regulatory Assessment Document (RAD) (Parsons, MRS-11, September 1985, Vol II) designates structures, systems, and components as impor-tant to safety if, on the basis of both analyses and engineering judgment, it is estimated that an accident (1) can result in an off-site release greater than

! 3-7

that allowed by 10 CFR 72.68 (5 rems to the whole body or to any organ for an individual located on the nearest boundary of the controlled area) or (2) will cause a nuclear criticality event. The RAD states that "except for the main-tenance of a geometrically safe configuration of spent-fuel storage, only those structures, systems, and components that are required to maintain a containment barrier are considered important to safety." As stated in the RAD, the Func-tional Design Criteria and the MRS Quality Assurance Program Plan provide the criteria for determining the appropriate designation and classification of structures, systems, and components important to safety:

(1) Prevent undue risk to the health and safety of the public.

(2) Maintain nuclear criticality safety.

(3) Provide for safe shutdown.

The staff infers from the RAD that the following would be considered structures, systems, or components important to safety:

(1) the storage unit structures (the sealed storage casks or the drywells)

(2) the CHTRU facility structure (3) the following operational areas of the R&H building: cask unloading, shielded process cells, in process lag storage, shielded canyon cells, i in-cell lag storage vaults, loadout and decontamination, shipping loadout, remote-handled equipment maintenance, high activity radwaste cell, crane maintenance for the process cells and canyon cells, remote handled air  !

filtration cell, exhaust filter rooms for Zones 1 and 2, uninterruptible power supply (UPS) rooms, battery rooms, and control rooms '

(4) the following systems, subsystems, or components within the R&H building:

the heating, ventilation, and air conditioning (HVAC) exhaust system por-tion housing the first and second stage high efficiency particulate air (HEPA) filtration, including the ductwork between contaminated areas and the first and second stage filters, the spray system, the tornado valve, and the final backdraft damper; the HVAC supply system dampers at the air supply penetrations into contaminated areas and tornado valves; the stack monitor; criticality monitors; liquid effluent monitors; and seismic monitors

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The RAD states:

[C]omponents, systems, and structures arn designated Category I if, dur-ing or following an extreme environmental load (including Design Basis Earthquake [and] Design Basis Tornado, they must perform either of the following safety functions:

(1) Prevent or mitigate the consequences of an uncontrolled release of radioactivity with potential radiological consequences in excess of the limits in 10 CFR 72.

(2) [ Prevent a] nuclear criticality [ event].

3.2.2.2.2 Seismic Category I Designation Of the items listed above as considered important to safety, all except the storage unit structures (sealed storage casks and drywells) are designated seismic Category I. Although the standby generator building is not included on the list, the analysis in the RAD for this building impI;es that it also is considered seismic Category I.

No seismic category is designated for tre spent fuel storage canisters, but the RAD indicates that they would be designed to survive an accidental drop. For the_ sealed storage casks, the RAD indicates that the combined design parameters for the canister and for tr ' cask shiciding result in a " cask configuration (that) is capable of maintaining a safe canister geometry and surviving all natural phenomena events," with only the localized loss of shielding that could come from concrete spalling caused by tornt.do missiles. The RAD states that the sealed storage casks would have to be at.alyzed for the effects of natural phenomena. It gives the results of analyses for seismic response, the use of energy methods to check the overturning of the cask, and differential pressure loads caused by a tornado pressure drop.

The RAD indicates that because of their in ground placement and top shielding, the drywells would be " capable of surviving all natural phenomena events except 3-9

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l tornado generated missiles." It also states that such missiles could result in only localized loss of shielding, with no undue risk to the public. The analy-ses indicated in the RAD are for seismic response and tornado pressure differen-tial load analyses.

3.2.2.2.3 Site Characteristics and Environmental Conditions The site characteristics and environmental conditions utilized for the MRS de-sign include the general site conditions given in the Functional Design Criteria (PNL,1985) and the more specific site characteristics provided for the three proposed sites discussed in Section 3.1. Specific designs were based on the " preferred" site, the CRBRP site.

The general site conditions include (1) seismic zone of 2 or less, as defined by the Uniform Building Code (2) negligible potential for surface faulting or for ground movement due to liquefaction, subsidence, or landsliding (3) 4000 to 15,000 psf soil bearing capacity for static loads (4) negligible potential for flooding from nearby surface water bodies or surface runoff (5) design temperatures of 95 F dry bulb and 75*F wet bulb summer and 2*F winter (6) highest wind (non-tornado) speed of 100 mph (7) maximum precipitation of 3 inches in an hour or 9 inches in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (8) snow loads in accordance with ANSI A58.1, Section 7 (9) tornadoes with 290 mph tangential speed, 70 mph translational speed, pressure drop of 3 psi, and a rate of pressure drop of 2.2 psi /second.

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DOE has indicated that the MRS at the CRBRP site would be located well above the 100 year flood plain, the probable maximum flood level, and the flood surge level.

3.2.2.2.4 General Design Conditions Loading conditions considered for the design of all MRS structures include the Zone 2 seismic loadings plus applicable dead loads, live loads, vehicle loads, snow loads, impact force or unusual vibration loads, thermal loads, wind loads, lateral earth pressure loads, and air pressure differential (due to ventilation system operation) loads. DOE has used load combinations and design methods in accordance with (1) the strength design method described in Standard 318 of the American Concrete Institute (ACI) for reinforced concrete; (2) the working stress method and Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings of the American Institute of Steel Construction (AISC) for steel structures; and (3) Chapter 24 of the Uniform Building Code and ACI 531, Building Requirements for Concrete Masonry Structures for masonry.

Lightning protection systems would be provided for all major MRS buildings and all systems and components important to safety in accordance with National Fire Protection Association (NFPA) Standard 78 and the National Electric Code.

3.2.2.2.5 Seismic Category I Design Conditions The MRS structures, systems, and components designated seismic Category I were designed to withstand a maximum horizontal ground acceleration of 0.25 g and a 0.167 g vertical acceleration. Response spectra were developed in accordance with RG 1.60, and dampings were used in accordance with RG 1.61. The external surfaces of the seismic Category I structures were designed to withstand the design basis tornado. Tornado generated missiles were also considered, inclu-ding a 12-inch-diameter,15-foot-long steel pipe, which was the critical mis-sile for penetration and scabbing. Concrete would be at least 20% thicker than the calculated scabbing thickness (by the Modified National Defense Committee formula). The tornado wind effects were determined in accordance with NRC 3-11

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Standard Review Plan Section 3.3.2 (NUREG-0800) and ANSI A58.1. Loading com-binations and design methods utilized for Category I structure design were in accordance with ACI 349.

3.2.2.2.6 Monitoring of Natural Phenomena Intensity A seismic monitoring system for detection and recording of strong local earth-quakes would be provided in the R&H building, and a mechanical peak accelera-tion recorder that operates without electric power would be centrally located in the storage facility area and in the CHTRU area. The R&H building monitoring, recording, and playback systems would be in accordance with RG 1.12.

The MRS would include meteorological instrumentation system measuring, at four locations, wind speed and direction, ambient temperature, barometric pressure, relative humidity, and precipitation. The system also would collect and trans-mit the information to a microprocessor-controlled system in the R&H building and site services building control rooms.

3.2.2.2.7 Prevention of Radioactive Material Releases to Underlying Aquifers As described in Section 3.2.15 below, liquid radioactive wastes generated within the R&H building would be treated and stored on the site rather than being re-leased to the environment. Liners and monitored sumps would be provided on floors for containing liquid radwaste and any free liquids. All non-radioactive effluents would be monitored for contamination before they are released.

The CHTRU structure, a below ground compartmentalized reinforced masonry struc-ture, would have a leakage and seepage collection system. Drains would be pro-vided in each compartment, and the drain effluent would be collected and piped through a double contained drainage piping to sumps. The liquid effluent would i be transferred subsequently by tank truck to the low level radioactive waste treatment system located in the R&H building.

Within the storage units in the storage facility, the seal welds on the liners of the casks or drywells would prevent releases from contained radioactive materials. The drywell liners that would be grouted in the soil would be 3-12

l designed to maintain their integrity throughout the design life of the storage facility, taking site-specific soil conditions into account.

The RAD states that DOE has committed to obtain the specific-site data needed to i provide "a design basis for a monitoring system to measure... potential releases to underlying aquifers." '

1 3.2.2.3 Discussion At the beginning of the definitive design, DOE should use analyses and engineer-ing judgment to clearly show which structures, systems, and components should be designated important to safety. The importance to safety of the spent fuel canisters as confinement barriers should be indicated, and either the canisters or the storage units (casks or drywells) should be designated seismic Category I (see also Section 3.2.8 below). The RAD does not describe the difference between the designations " surviving all natural phenomena events" and " seismic Category I" for the casks and drywells, but the analysis descriptions presented imply the differences are minor.

DOE has used site characteristics and environmental conditions, load defini-tions and combinations, and methods of analysis that are reasonable at the con-ceptual level of design. These appear to adequately address the most severe natural phenomena loadings, including tornado missiles for Category I structures.

However, DOE, in any future safety analysis report, must provide additional detailed information on possible effects of failure of non-seismic Category I structures during severe natural phenomena, as well as information on either the adequacy of equipment designs or the consequences of failure for equipment within Category I structures during severe natural phenomena. For Category I equipment and substructures located within Category I structures, DOE will have to provide separate analyses and perhaps design features to demonstrate that the equipment and substructure can withstand the severe natural phenomena trans-mitted through the structure. (Examples of such features could include seismic restraints or collision or impact barriers on critical components, or seismic rail and hook clamps on fuel / canister handling cranes.) As noted in the dis-cussion of the Clinch River site, there is a potential for the existence of solution cavities in the carbonate bedrock at the site, and this possible condition must be addressed during site-specific investigations.

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The proposed capabilities for determining the intensity of natural phenomena appear adequate, as do the proposed measures to preclude transport of radio-active materials to the environment through an aquifer located under the MRS.

Underlying aquifers can be expected in limestone rock regions such as the CRBRP site. In developing the final design, DOE would have to consider specific site conditions to determine the need for and extent of a monitoring system to mea-sure potential releases to underlying aquifers.

3.2.2.4 Conclusions and Recommendations In an application for a license for the MRS, DOE would have to provide an explicit and justified designation of structures, systems, and components important to safety. DOE also would have to include a definition of severe natural phenomena survival requirements for the important-to-safety confine-ment barriers--the spent fuel canisters, sealed storage casks, and drywells.

Severe natural phenomena effects within seismic Category I structures and effects of failure of non-Category I structures also would have to addressed.

In addition, 00E would have to further consider whether a system for measuring and controlling potential radioactive material releases to aquifers underlying the MRS is required.

3.2.3 Protection Against Fire and Explosions 3.2.3.1 Requirements 10 CFR 72.72(c) gives the requirements for protection against fire and explosions i as follows:

l Structures, systems and components important to safety shall be de-signed and located so that they can continue to perform their safety functions effectively under credible fire and explosion exposure con-diticns. Noncombustible and heat resistant materials shall be used wherever practical throughout the ISFSI, particularly in locations vital to the control of radioactive materials ara to the maintenance l of safety control functions. Explosion and fire detection, alarm, l and suppression systems shall be designed and provided with sufficient 3-14 l

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capacity and capability to minimize the adverse effects of fires and explosions on structures, systems, and components important to safety.

The design of the ISFSI shall include provisions to protect against ad-verse effects that might result from either the operation or the failure of the fire suppression system.

3.2.3.2 Description The design criteria used for fire protection are in the Conceptual Basis for Design (Parsons, October 1985). The fire safety provisions that meet these cri-teria, as described in the Conceptual Design Report (Parsons, MRS-11, September 1985, Vol I), are as follows:

(1) central fire station on the site (2) automatic wet pipe sprinkler systems in most buildings and dry pipe systems in some buildings (3) fire alarms in all buildings tied to MRS central fire station and security building (4) fire hydrants around the facilities  :

(5) fire-resistant bottled gas storage facilities I

(6) appropriately placed fire dampers in ventilation ducts (7) spray nozzle within the final HEPA filter plenum l

l (8) Halon 1301 system in the control room and electrical equipment rooms (9) standpipe and hose cabinets as secondary water supply for the sprinkler system (10) automatic dry chemical system inside glove boxes and areas with potential flammable 1Iquid fires 3-15 J

(11) portable fire extinguishers (12) use of noncombustible construction materials (13) separation of fire areas (14) separation of paint shop and solvent storage facilities from other facilities and use of explosion proof fixtures The source of water for the fire protection system would be a pipeline from the local supply. The pipeline would supply a storage tank from which water would be pumped to the yard loop. The tank and yard loop also would supply the potable and process water systems. The water supply syste.n is described in Section 3.2.11, below, lhe fire protection system has been designated non-seismic Category I, quality assurance (QA) Level II because this system would not be required to prevent release of radionuclides to the environment or to safely shut down the facility.

However, the spray system and associated water supply tank for the HEPA filter plenums have been designated seismic Category I, QA Level I because the poten-tially elevated temperature in the plenums could result in loss of confinement of radionuclides.

3.2.3.3 Discussion The requirements of 10 CFR 72 can be separated into the four basic topics:

(1) Design and location of st'metures, systems, and components important to safety will be chosen so that these structures and components can continue to perform their function under design-basis fire and explosion conditions.

(2) Honcombustible and heat resistant material will be used where practical throughout the facility, especially in areas important to the control of radioactive materials.

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(3) Fire and explosion alarms and suppression systems will be designed with sufficient capacity to minimize the effect of a fire or explosion.

(4) The design of the facility will include provision for dealing with the effects of operation or failure of the fire suppression system.

Most MRS structures (especially those that contain equipment important to safety) would be constructed of noncombustible materials. These structur'es would be predominantly concrete and steel.

With at least one exception, the facility design for each of the buildings on the MRS site includes provisions for one or more fire suppression systems that use sprinklers, fire hydrants, Halon 1301, dry chemical extinguishers, or portable fire extinguishers. An exception is the system provided for the shieldad hot cells (the process cells and the shielded canyon cells) in the R&H building.

Parsons (October,1985) gives the basis for the design of this portion of the building as follows:

A Fire Suppression System is not required in the shielded hot cells. '

This is due to unoccupied area, massive wall thickness of the cells (fire separation), proper equipment spacing and negligible amounts of combustible materials within the cells. Criticality considerations preclude the use of [a] water suppression system. Contamination and HVAC requirements make impractical the application of [a] total flooding Halon 1301 system.

The concerns over criticality and the Halon system must be considered, and the amount of combustible materials in the R&H hot cells must be studied. As noted in Section 3.2.14 below, the cutting and dismantling of fuel assemblies may create Zircaloy particles (" fines"). Zircaloy particles can display pyro-phoric,* flammable, or explosive characteristics, depending upon particle size, quantity, geometry, and ambient conditions (temperature, moisture, etc.). In powder form (particle sizes less than 62 microns), Zit'caloy is pyrophoric and .

explosive. Coarse powders with particle size between 125 and 840 microns are "Hawley (1977) defines pyrophoric material as any liquid or solid that will ignite spontaneously in air below 54.5'C (130*F).

3-17

flammable. Thus particle sizes up to 840 microns present fire and explosion risk. Chips and pieces of Zircaloy less than 0.013-cm thick and less than

! 0.159-cm wide are easily ignitable (Rockwell Hanford, 1984). (These data have been developed from experiments with unirractated zirconium that may not be fully applicable to irradiated Zircaloy resulting frcm spent fuel cutting.)

j Thus, DOE must provide a further determination of particle size and the combus-tion properties of lircaloy resulting from laser or other cutting of spent fuel assemblies, so the staff can fully assess the potential fire and explosion haz-ard in the hot cells.

The shredding of fuel assembly skeletons adds to the fire and explosion hazard concerns because some non-fuel-bearing ccmponents of the fuel assenblies are

( made of Zircaloy (guide tubes in pressurized water reactor (PWR) fuel assem-blies and spacer grids in both PWR and boiling water reactor (BWR) fuel assem-blies). As currently planned, these non-fuel-bearing components would be col-lected, shredded, and placed in drums within the hot cells. DOE must evaluate the potential for fire hazard resulting from the presence or creation of Zirca-loy particles in the scrap container. (The hot cells might also contain small amounts of other combustible materials such as electrical insulation and lubri-cating material (in the overhead cranes, consolidation systems or other equipment).)

A second area of concern deals with separation of equipment in the systems necessary to ensure confinement for stored fuel or fuel in process in the R&H building. As noted in Section 3.2.8.2 below, it is possible that not all of the equipment in the ventilation and off gas system of the R&H building and the on-site power supply system is adequately protected against common-cause failures resulting from fires.

In the ventilation system (Parsons, MRS-11, September 1985, Vol III, Dwgs H-3-56844 and 45), all of the exhaust fans for Zones 1 and 2 would be located in a single room that is connected to a room that contains all of the exhaust fans for Zones 3 and 4. A fire in the exhaust fan room could disable all of the ventilation systems for the facility. Similarly, three final exhaust filters would be in each of the four final exhaust filter rooms (two for Zones 1 and 2 and two for Zones 3 and 4). In the MRS design, four of the six filters for each pair of zones would be required to adequately perform the 4

3-18

exhaust function. Thus, a fire in one of the exhaust filter rooms could reduce the capacity of the exhaust filter system to below the design basis requirement, The CHTRU facility design provides for sampling the air and monitoring the j interior of the waste compartment for temperature and radiation (PNL,1985), I but there are no provisions for fire detection or suppression. PNL (1985) gives the reasons as non-combustible compartmentized construction and lack of I

combustible materials except as encapsulated in grout and packaged in steel '

drums. The review of the waste streams in Parsons (MRS-11, September 1985, '

Vol III) indicates that combustible materials (shredded wood frames from HEPA filters) are stored in the facility. Therefore, a smoke detection and alarm system, as a minimum, are advisable.

3.2.3.4 Conclusions and Recommendations The philosophy DOE is following in regard to protection against fire and explo-sions appears to be reasonable. However, the staff will make its determination on the adequacy of the design at a later stage.

As noted above, no fire protection system is provided for the shielded hot cells because the cells would have very thick wallc, there would be negligible amounts of combustible materials within cells, and the equipment would be properly spaced.

Moreover, DOE has indicated that the use of a water suppression system is not '

desirable because criticality considerations and HVAC requirements would make use of the Halon 1301 system impractical.

The reactive nature of finely divided zirconium or Zircaloy with water is an important aspect of fire protection. In a number of previous licensing cases, the staff has requested detailed analysis and design information to ensure that safe conditions are always maintained for the cutting, collection, and disposal of the metal. These can be accomplished through controlled oxidation, mass limitation, and/or dilution with inert material. In the Design Verification Plan (00E/RW 0035), DOE has indicated that testing and analysis will develop the necessary designs for the safe handling of zirconium fines. The staff should be informed about the results of these tests. The staff intends to thoroughly evaluate the techniques developed at the licensing review stage, if Congress approves DOE's proposal for the MRS.

3-19

Finally, alternate system components may have to be separated to ensure that a fire in a single fire zone will not defeat an entire system important to safety. The design of the exhaust filter and fan rooms in the R&H building

, should consider whether or not this separation is necessary.

, 3.2.4 Sharing of Structures, Systems, and Components 3.2.4.1 Requirements 10 CFR 72,72(d) gives the requirements for sharing of structures, systems, and components as follows:

Structures, systems, and components important to safety shall not be shared between an ISFSI [MRS] and other facilities unless it is shown that such sharing will not impair the capability of either facility to perform its safety functions, including the ability to return to a safe condition in the event of an accident.

3.2.4.2 Description and Discussion At present DOE does not plan to have the MRS share any structures, systems, or components with another facility. If the MRS were to be sited on Federal land with other facilities, it is possible that some supporting services would be shared. The intent of the criterion is to ensure that such sharing does not dilute the capability to meet safety needs. Because most of the operations and abnormal events are of a passive nature, most sharing would not be critical.

If any MRS were to be sited with existing facilities, the MRS would be reevalu-r:ted for compliance with this criterion.

3.2.4.3 Conclusions and Recommendations On the basis of its review of the present design, the staff finds that the sharing of structures, systems, and components important to safety would offer little difficulty in review.

3-20

3.2.5 Proximity of Sites 3.2.5.1 Requirements t 10 CFR 72.72(e) give the requirements for proximity of sites as follows:

An ISFSI [MRS] located near other nuclear facilities shall be de-signed and operated to ensure that the cumulative effects of their combined operations will not constitute an unreasonable risk to the health and safety of the public.

3.2.5.2 Description and Discussion The cumulative effects of the combined operations of the MRS and any other nuclear facilities located nearby would be considered during the licensing review. The greatest potential for release of radioactivity in airborne effluents from the facility would be in puffs of asKr resulting from the fail-ute of fuel rod cladding during handling operations. Krypton-85 is a weak beta-ray emitter and inert chemically. It is a noble gas, formed in the fuel .

during the fission process, and it decays with a half-life of about 10.7 years.

In all probability, only a single rod would fail at any given time, releasing less than 10 curies of the gas at one time. This release would dilute in the atmosphere and would not be detectable in the environment. Because of the extensive use of high efficiency filters planned for the R&H building where the fuel would be handled, particulate radioactivity would not add significantly {

L to the normal effluents.  ;

l The staff recognizes that accidents could occur during the operation of the MRS. f The major potential event would be the rupture of all fuel rods in three assem- l l

blies resulting from a drop. A conservative estimate of the maximum dose to a nearby individual would be about 60 millirems whole body (see Section 3.3 below).

3.2.5.3 Conclusion and Recommendations On the basis of its consideration of the very limited radiological impact from operation of the MRS, including accidents, the staff concludes that the MRS 4

3-21

l would not be a significant contributor to radiological dose to the public; l from that stand pint, its proximity to other nuclear facilities is inconse-quential. ,

3. 2. ti Testing and Maintenance of Systems and Components )

3.2.6.1 Requirements 10 CFR 72.72(f)-gives the requirements for testing and maintenance of systems i and components as follows: 1 T

l Systems and compdnents that are important to safety shall be designed j to permit inspection, maintenance, and testing. j 3.2.6.2 Description and Discussion 5

The purpose of this criterion is to ensure that at the design stage a prospec-tive applicant considers that inspection, calibration, and testing of systems 4 l

and components important to safety may bp necessary during the operation of the

~

j facility. The DOE design documentation indicates that DOE is considering the  !

i applicationofthisciiterion. DOE is giving particular emphasis to ensuring that radiation exposure during maintenance operations is in keeping with the NRC guidelines to keep such exposure "as low as reasonably achievable" (ALARA),

3.2.6.3 Conclusions and Recommendations I

\, O l If Congress approves the MRS proposal, DOE (in preparing its application) should first identify all structures, systems, and components important to safety and then ensure that each is designed to allow inspection, testing, calibration, maintenance, or any other services that might be necessary. I 3.2.7 Emergency Capability  :

1 3.2.7.1 Requirements 10 CFR 72.72(g) states:

3-22

7_

s Structures, systems and components important to safety shall be designed for emergencies. The design shall provide for accessi-bility to the equipment of onsite and available offsite emergency facilities and services such as hospitals, fire al'd police depart-ments, ambulance service, and other emergency agencies.

i 10 CFR 72.19 states:

An application to store spent fuel in an ISFSI [MRS] thall include plans for coping with emergencies. These plans shall contain the elements that are listed in Section IV, " Content of Emergency Plans" of Appendix E to Part 50 of this chapter. (N.B. This requirement is being considered for modification by the Commission in the context of recent staff analysis on emergency planning.)

3.2.7.2 Description The MRS design includes the following facilities, equipmeat, and systems

• for response to emergency conditions on site:

(1) fire alarm and detection system (2) public address system l

h (3), intercommunication system (4) telephone system (5) door alarm and access control system (6) ultra-high frequency (UHF) radio system (7) building warning and evacuation system l

1 Two alarm monitoring stations have been provided in the design: a primary station in the protected area gatehouse and an alternate station in the security building. Both stations would be below grade in hardened and bullet-resistant structures.

• Parsons, MRS-11, September 1985, Vols I and III; October 1985.

3-23

The telephone and UHF radio systems would provide onsite and offsite communica-tion. The EPABX (electronic private branch automatic exchange) would be located at the site services building. The UHF system would allow wireless, individual paging of key personnel inside and outside buildings; telephone equipment would be installed in all offices, laboratories, the control rooms, operating galleries, and the main electrical room.

The building warning and evacuation system would provide separate and distinct audible and visual warnings during emergencies, including process emergencies, and security, fire, and building evacuation. These alarms would be initiated either automatically or manually from the control room and are displayed in the alarm monitoring stations.

All fire alarms in the MRS would be monitored by a fire alars monitoring compu-ter located in a separate, onsite fire station. Signals would be transmitted to the alarm monitoring stations.

Provisions for mitigating emergencies resulting from natural phenomena, criticality, effluent release, and power outages are discussed elsewhere in this report.

Access to the structures or facilities important to safety would be by paved roads, as shown in the following site arrangement drawings (Parsons, MRS-11, September 1985, Vol III):

H-3-56727, H-3-56734 Clinch River H-3-56742 Oak Ridge H-3-56747 Hartsville These drawings indicate that clear paved roads would provide emergency re-sponse capability for all structures important to safety (R&H building, CHTRU facility, and spent fuel (SF), high level waste (HLW), and remote handled transuranic waste (RHTRU) storage facility), as well as to other structures that would be located on the site.

3-24

The R&H building would be the only structure important to safety that would be continuously occupied. The overall ground level plan, shown in Drawing H-3-56758 (in Parsons, MRS-11, September 1985, Vol III), shows emergency exits on the east and west sides of the facility leading to paved access roads. Access to the upper level floor would be provided by seven separate stairwells, as well as by a passenger elevator and a freight elevator.

First aid stations would be provided in the fire station and the site services building, and health physics capability would be provided in the R&H building.

Paramedic facilities would be located in the fire station. The site ambulance would provide emergency medical evacuation or, if required, an offsite helicopter could land at the MRS heliport.

At this stage of design, 10 CFR 72.19 does not require DOE to prepare an emergency plan.

3.2.7.3 Discussion The design of onsite MRS structures and systems appears to provide adequate capability for emergency response within the facility. The design is in ac-cordance with ,the fonctional design criteria (PNL, 1985). All facilities important to safety would be accessible by clear roads, and the onsite/offsite communication and alarm systems would provide for potential emergencies (particularly when coupled with the provisions for mitigating the effects of natural phenomena, fire and explosion, breach of security, criticality, effluent release, and power outage).

RG 3.42 (September 1979) identifies the type of emergencies to be con-sidered and what actions must be addressed for emergencies. The types of emer-  !

gencies to be considered are as follows:

l (1) personnel emergency (2) emergency alert I (3) plant emergency 1

I 3-25

(4) site emergency (5) general emergency The emergency actions considerations are as follows:

(1) assessment action (during and after the accident)

(2) corrective action (to ameliorate or terminate emergency)

(3) protective action (after an uncontrolled release)

(4) determination of population at risk (5) recovery action (6) protective action guides (7) emergency action levels (8) drill (9) exercise 3.2.7.4 Conclusions and Recommendations The MRS design provisions for emergency response appear adequate. The concep-tual design stage is too early to prepare an emergency response plan in accordance with RG 3.42. Once the design is definitive, DOE can prepare a plan that will embody detailed knowledge of the site.

3.2.8 Confinement Barriers and Systems 3.2.8.1 Requirements 10 CFR 72.72(h)(1) and (3) give the requirements for confinement barriers and systems as follows:

(1) The fuel cladding shall be protected against degradation and gross ruptures. (N.B.: This sentence is being considered for modification by the Commission, because the use of storage or disposal canisters may provide an additional degree of protec-tion not originally foreseen.)

3-26

(N.B. In addition to the above requirement the Commission is consid- I ering further modification to include monitoring and retrievability provisions.)

(3) Ventilation and off gas systems shall be provided where necessary to ensure the confinement of airborne radioactive particulate materials during normal or off-normal conditions.

3.2.8.2 Description The MRS provides confinement barriers and systems for the following:

(1) spent fuel, high level waste (HLW), high activity waste (HAW), and RHTRU -

storage facility (2) 16g storage vaults in the R&H building (3) ventilation and off gas systems for the hot cells and all other areas where airborne radioactive particulates may be generated 3.2.8.2.1 Spent Fuel, HLW, HAW, and RHTRU Storage The MRS is designed to provide long-term storage (up to 100 years) for spent fuel, HLW, HAW, and RHTRU in one of two modes: sealed storage casks or open I field drywells. DOE defines HAW as material that may exceed Class C specifi-cations for low-level waste and may or may not contain some transuranic material.

After being prepared for storage (as described in Section 3.2.14), the storage canisters would be placed in the storage units of one of these two storage modes.

Either type of unit could also accommodate spent fuel as it was received in metal storage / transport (dual purpose) casks (Parsons, MRS-11, September 1985, Vol I).

The storage casks would be seal-welded, self shielded, dry storage units for intact spent fuel assembly canisters, consolidated fuel rod canisters, drums of non-fuel-bearing components, HLW canisters, and drums and canisters of RHTRU.

The sealed storage casks would be cylindrical reinforced concrete structures 3-27

with a carbon-steel-lined cavity in which the canisters or drums would be stored.

A cylindrical concrete shield plug would fit into the open top of the cavity, and a steel cover plate would be seal-welded (with three passes) to the liner flange to close the cask; a security seal would be included in the second seal pass. A nondestructive examination and a leak-check inspection of the seal weld would be performed. Sealed storage casks are a passive method of storing spent fuel and nuclear waste products, the steel and concrete would provide radiation shielding. The heat from radioactive decay would be conducted through the steel and concrete, and it would be removed by atmospheric convection and thermal radiation. Figures 3.2 and 3.3 illustrate the sealed storage cask concept.

. The drywells would be welded pipes (liners) located in borings just beneath the soil surface. They would form cavities for storing intact spent fuel assembly canisters, consolidated fuel rod canisters, drums of non-fuel-bearing components, HLW canisters, and drums and canisters of RHTRU. The liners would be seal-welded after loading.

The liner would have a top flange that rests on the concrete pad that would be poured at the soil surface before the hole was bored. After the liner was in place, the annulus between the liner and hole would be filled with grout up to the underside of the liner flange, and a temporary cover would be placed over the top of the liner. Wen the drywell was ready for loading, the temporary cover -ould be removed, the canister lowered in the liner cavity, and a shield plug placed on top. A final drywell cover would then seal-welded (with three passes) to the liner flange; a security seal wuld be included in the second weld cover pass. A nondestructive examinttion and leak-check inspection of the seal weld would be performed. Drywells also are a passive method of storing spent fuel and nuclear waste products; the surrounding soil would provide i

radiation shielding and dissipate the beat from radioactive decay. Figure 3.4  ;

illustrates the drywell concept.

Transportable metal storage (dual purpose) casks would be sealed, self-shielded, dry storage units for intact spent fuel assemblies or for consolidated fuel rods.

The transportable casks, employed as both the shipping and storage containers, would be loaded in the reactor spent fuel pool, the water removed, and the l

3-28

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casks sealed before they would be shipped by rail to the MRS. The casks would be transported on and off the site horizontally; they would be stored horizon-tally on reinforced concrete cradles. For this design, DOE has assumed that a transportable cask (such as an REA-2023 cask) would be received, handled, and stored. The designs of equipment, handling, and storage for this cask were considered representative of those required for other transport / storage casks.

The storage units would be designed to protect the spent fuel from natural and human-induced events. The units would be designed to take the defined seismic Category I loads. The containment barriers provided would consider the charac-teristics of each type of waste.

For the spent fuel, two containment barriers would be provided. The seal-welded storage canister would provide the primary containment barrier, and the storage unit liner and seal-welded cover would provide the secondary containment barrier.

To protect fuel cladding against degradation and gross rupture, the fuel cladding  ;

surface is maintained below 375*C in an inert gas environment (PNL, 1985).

l For high level waste, one containment barrier would be provided by the storage i unit liner and welded cover. The steel canister around the HLW would provide a second barrier. The radioactive waste would be immobilized in the solidified glass matrix in the canister, providing a third barrier. The surface of canis-ters containing solidified HLW would be maintained below 375*C, and the center-

line waste (borosilicate glass) would be maintained below 500 C. This require-ment would be imposed to prevent crystallization of the glass within the canister during storage.

Temperature monitoring would be provided for a statistically representative number of spent fuel and HLW casks or drywells to ascertain that the thermal performance of the storage units was adequate. Temperatures would be measured on the cask or drywell liner, and could be compared to predictions to assess thermal performarce. Liner temperature readings below established limits would be used to determine if the spent fuel cladding ano HLW canister temperatures remained be' low their respective limits.

4 3-32

The storage unit containment barriers would be designed to allow integrity monitoring throughout their design life. Gas sampling ports would be installed between the canister and the storage unit liner to take internal atmosphere samples and permit checking of internal pressures. Analysis of the gas samples would verify fuel canister integrity, and pressure decay (due to long-term cooling) checks wculd verify the leak-tight integrity of the cask and drywell.

For RHTRU and for the waste generated on the site from consolidation of spent fuel (non-fuel-bearing components), a single containment barrier would be pro-vided by the storage unit liner and welded cover. If RHTRU were received from off the site, it would be in sealed canisters or drums, and the non-fuel-bearing components generated on the site would be installed and sealed in drums. These containers would contain the radioactive materials from a contamination-spread standpoint, and, although they would be sealed, no credit would be taken for them in the prevention of release of stored radioactive materials when in stor-age. The materials contained in the RHTRU waste and non-fuel-bearing components would be solids.

3.2.8.2.2 Lag Storage Vaults The MRS would include provisions for lag storage of spent fuel canisters in the R&H building before they were moved to the onsite storage facility units or were shipped directly to the repository. The lag storage would be a buffer that would permit flexibility in accommodating variations in waste receipt, handling, processing, onsite transportation, and offsite shipping, The lag storage would consist of concrete-shielded vaults covered with removable shield plugs and equipped with a rack system that would support the spent fuel canisters. The lag storage vaults would be divided into compartments, and each compartment would be provided with its own air supply and exhaust channel for removal of heat generated by the stored spent fuel (Parsons, MRS-11, September 1985, Vol 1). The vaults would also be designed to maintain spent fuel cladding temperature below 375*C. If this forced ventilation system is shut down because l of power failure or other cause, heat generated by the stcred spent fuel canis-ters would be removed by conduction and radiation to the concrete walls and by 3-33 1

natural air convection within the vault compartments. Parsons (MRS-11, September 1985, Vol IV) presents studies indicating that more than 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> of such loss of forced ventilation is required before fuel in canisters reaches 375*C.

3.2.8.2.3 Ventilation and Off-Gas Systems 3.2.8.2.3.1 General Considerations The R&H building is expected to be the only building in which significant airborne radioactive particulate materials would be generated during operations. There-fore, a ventilation and off gas system would be provided for the necessary confinement and personnel health and safety during normal and off-normal con-ditions in this building.

The ventilation system for the R&H building would be designed assuming that an active ventilation system is needed under normal operating conditions to pro-vide the appropriate confinement barrier. The design would not provide an isolating containment to retain releases within the building, but would remove and collect the airborne radioactive particulates from the exhaust stream as close as possible to their source, before any remaining effluents were released to the environment.

R&H building ventilation (Parsons, MRS-11, September 1985, Vol I) would be pro-vided by a "once-through" system (as opposed to a recirculating system). The system would supply prefiltered and heated / cooled outdoor air to the building.

After this air performed its heating, cooling, and ventilating functions, it would be filtered for particulates through several stages of HEPA filters, and then exhausted through a stack to the atmosphe m. In addition, the system would be designed to operate in a " cascading" mode, the air always flowing from the less contaminated areas to the more contaminated areas. This flow would result because set pressure differentials would be maintained between various areas of the building, keeping the most hazardous areas at the greatest negative pressure (with respect to the ambient pressure).

l PNL (1985) states that the R&H building would be divided into four ventilation l confinement zones based on the relative potential for expcsure of personnel to airborne radionuclides. The four zones are l 3-34

Zone 1 process zone (highly contaminated area, restricted access zones)

Zone 2 restricted access zone (potentially contaminated areas)

Zone 3 operating zone (not normally contaminated)

Zone 4 unrestricted access zone The static pressures (with respect to ambient atmosphere) maintained in each of I

the zones to ensure air flow from the least hazardous area to the most hazardous area would be Zone 1 -2 to -4 in, wg*

Zone 2 -0.75 to -1.5 in, wg Zone 3 -0.5 to -0.7 in. wg Zone 4 0.05 to -0.1 in wg In addition, incremental zones would be provided within each same zone to distinguish betwaen " dirty" hot cell operations and " clean" hot cell (1-in. wg pressure differential) and between " clean" hot cells and the " clean" lag storage vaults (1.5-in. wg pressure differential).

All shielded hot cells and canyon cells would be designated Zone 1; various remote handling area and glove boxes also would be designated Zone 1. Zone 2 would include hot cell service galleries, low level radwaste treatment facility, final and remote-handled HEPA filter rooms, and other potentially contaminated areas. Zone 3 would include operating galleries, lag storage vaults, personnel

decontamination rooms, and other areas not normally contaminated. Zone 4 would include all unrestricted areas such as the administrative area, control room, l

and the receiving and inspection area.

3.2.8.2.3.2 System Description **

The main ventilation system for the R&H building would be served by two multi-unit, once-through, cascading airflow systems. In addition seral independent

• wg: water gauge

• • PNL,1985; Parsons, MRS-11, September 1985, Vols I and II.

3-35

I ventilation systems would serve miscellaneous Zone 4 areas and the lag storage vaults. Additional standby ventilation systems would be provided for the main control room and the uninterruptible power supply (UPS) rooms.

Figure 3.5 is a simplified schematic diagram of the major exhaust systems.

Main Once-Through System The main once-through system consists of the following subsystems:

(1) supply system (2) exhaust system (3) between-zones filtration system (4) exhaust stack The supply system (Parsons, MRS-11, September 1965, Vol III, Dwg H-3-56835) would contain supply fans with inlet dampers; cooling and heating coils; heat recovery coils; prefilters and filters; isolation, backdraft, and tornado dampers; airflow measuring station; controls for airflow and pressure, and all associated supply air ductwork.

j The final exhaust systems (Parsons, MRS-11, September 1985, Vol III, Dwgs H-3-56836 I and 56837) would collect and remove airborne radioactive particulates generated within the facility before the air is released from the facility. The system would be divided into two subsystems: one for Zones 1 and 2, and one for Zones 3 and 4.

The final exhaust system for Zones 1 and 2 would contain a common exhaust air plenum; six final filter plenums (four operating and two standby) connected in parallel; four exhaust fans (two operating and two standby) connected in paral-lel; isolation valves; airflow stations, backdraft dampers, and heat recovery j coils (four operating and two standby); all associated controls; and all asso- l ciated ductwork.

Each final filter plenum would include a prefilter, two stages of HEPA filters, space for a future carbon filter, a cooling water spray chamber with spark arrestor demister, and aerosol test sections for in place testing of each stage 3-36

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of the HEPA filters. The water to the sprays would be supplied by a dedicated system that would include c water tank pressurized with compressed air.

The final exhaust fan system would be designed with 100% redundancy; the final filters would be designed with 50% redundancy. A master differential pressure controller would reset the control point of other submaster flow controllers in the supply air system to maintain a constant pressure differential at the final exhaust system inlet plenum.

The exhaust subsystems for Zones 3 and 4 would consist of equipment similar to that in the subsystem for Zones 1 and 2, with two exceptions: there would be only three exhaust fans with 50% redundancy, and a flow control station

, would be provided at each fan discharge duct to maintain constant exhaust airflow.

, In addition to the exhaust systems, miscellaneous exhaust fans would be provided for certain Zone 4 areas, such as chemical storage area, chiller room, diesel fume exhaust, and toilets. These fans would discharge air directly to the outside without passing it through HEPA filters or the exhaust stack.

A "between-zone filtration" system would be configured as follows: a single-stage testable HEPA filter in the airstream of all transfer ducts and openings between Zones 1 and 2 and between Zones 2 and 3; a two-stage, remote-handled, non-testable HEPA filter at the air exit from selected Zone 1 areas (e.g., hot cells); a single-stage, remote-handled, testable HEPA filter inside the exhaust ducts from the same selected Zone 1 areas; and a single-stage, testable HEPA filter at the air exit from the other Zone 1 areas.

The exhaust from the major portion of the R&H building would go through a single stack designed for design earthquake and tornado conditions. The ducts from the final exhaust and from the lag storage vault would have radiation monitoring stations.

Standby HVAC Systems 1

The control room and the two UPS rooms each would have a standby HVAC system.

These systems would include air-handling units equipped with a supply fan, 3-38 i l

cooling coil, prefilter, and 85% efficiency final filter; indoor-type air-cooled condensing units including refrigerant piping and controls; all ductwork including supply and return; dedicated outside air intake and er.haust both equipped with tornado dampers; and back draft dampers located in the discharge duct of each air-handling unit.

The control room and the two UPS rooms normally would be supplied with air from the main once-through system. In the event of a power failure, when this system would not operate, the standby HVAC systems (which would be connected to standby power) would be activated to supply and exhaust air through an independent ductwork loop.

Lag Storage Vault Cooling System Lag storage vault cooling would be provided by a once-through, 100% outside air, ventilating system. The system would consist of four air-handling supply units (three operating and one standby) operating in parallel, six exhaust filter plenums (four operating and two standby) each equipped with fiEPA filters, four exhaust fans (two operating and two standby) connected in parallel, an airflow measuring statior) for each fan and ecch vault, tornado and backdraft balancing dampers, and all associated ductwork.

Because of the standby equipment, the system would have 33% redundancy for supply air fans, 50% for HEPA filtration, and 100% for exhaust fans.

3.2.8.2.3.3 System Operation  ;

Normal Operation During normal operation, the supply fans would bring 100% filtered but otherwise untreated outside air into the facility. This air would be cascaded through the different areas. After being filtered as described in Section 3.2.8.2.3.2 above, the air would be exhausted through the stack. The air balance and pres-sure differentials would be controlled by flow monitors, flow controllers, and differential presst.re controllers. The air exhaust would be maintained constant while the supply air would be throttled down to ensure the pressure in the 3-39 i

l building was negative. Automatic dampers between zones would be activated by a pres- j sure differential control system to maintain the required pressure differentials l between zones. In the lag storage vaults, a constant exhaust flow rate would be maintained. The pressure differential within each vault would be controlled on the supply side by differential pressure controllers. (System operation during accidents and abnormal conditions is discussed in Section 3.3.)

Operation During Fire When a fire alarm would sound, the supply air damper to an individual affected area would close, while the exhaust would continue to operate to remove smoke.

If a fire were co cover a major portion of the building, all of the supply air could be cut off and only the exhaust fans would operate, with outside air drawn in through the duct, bypassing the supply fans. When there was a high tempera-ture reading in the exhaust plenums, the firewater spray would be activated.

This spray water could not drain into cells that contain spent fuel. Fires in the lag storage vault would not be anticipated; however, if a fire were to occur in the ventilation equipment room serving the lag storage, the affected portion of the system would be shut down. (The event of a major fire causing loss of the entire cooling system is covered in Section 3.3.)

Operation During Tornado The ventilation system for the R&H building system and lag storage would be designed to withstand tornado missiles and air pressure transients. Missile barriers and pressure-sensing tornado valves, as well as tornado dampers, would be included in the design.

Operation During Power Failure During a power failure, the exhaust fans for the main once-through system would be switched to the standby power generator, as would the solenoid valves and instrument air. All supply fans would stop, and air would be drawn through the supply air bypass duct. Should the power supplying the controls fail, all control dampers would remain open.

3-40

Operation During Earthquake The first stage of the Zone 1 system and the ductwork leading to it from con-taminated areas would be designed to withstand the design earthquake (DE).

Similarly, the building area housing this critical equipment as well as the dis-charge stack would be designed to withstand the DE.

3.2.8.3 Discussion 3.2.8.3.1 Storage Units The storage units (concrete casks or drywells) would be similar to storege units used in dry storage demonstration tests and actual installations in the United States and Canada (Johnson and Gilbert, 1983). Concrete cask (or silo) demonstrations have included PWR storage over a 4 year period at the Nevada Test Site; electric heated casks at heat generation rates of 5.0 and 10.0 KW for over a year at Hanford, Washington; and several examples of Canadian fuel storage over periods of 8 years or more. Drywell demonstrations have included PWR storage in granite for longer than 3 years and in soil for longer than 5 years at the Nevada Test Site, and electric heated drywells in basalt at heat genera-tion rates up to 9 KW for about 2 years and in soil at a heat generation rate of 1 KW for 1 year at Hanford, Washington. Despite the success of these demon-strations, the data base is still somewhat small. If drywells were selected, DOE should consider tests using site-specific materials.

Storage demonstrations to date have not used consolidated fuel with respect to drywell and cask heat loadings. Some of the electrically heated drywell tests have been conducted at heat generation rates greater than those expected for the MRS drywells storing consolidated fuel. However, for the MRS concrete casks with a capacity for 12 spent fuel canisters, heat generation rates about twice those for the highest electrically heated casks tested could be expected.

DOE has committed (in the MRS Design Verification Plan) to demonstrate concrete cask performance under such higher heat loads and the resulting higher tempera-tures for extended time periods.

The double containment that would be provided for stored spent fuel (consisting of the seal-welded canisters within the seal-welded liner of the concrete 3-41

storage casks or drywells) conforms to the NRC concept of " defense-in-depth."

However, the canisters and the concrete storage casks and drywells have not been designated seismic Category I (although they would be analyzed for seismic response, cask overturning, and tornado differential pressure loads.)

Even with a gross failure of the canister and the liner, a loss of containment integrity would not be likely to have significant consequences for public health and safety, because the system would be passive with no high energy driving forces to disperse the contained radioactivity. Such gross failure should be detected by monitoring. In preparing its license application, DOE should con-sider the safety of onsite personnel, a major consideration in recovering from such failure.

A failure mechanism to be considered would be some form of corrosion, possibly accelerated by internal stresses or nonuniformity of materials. Monitoring the air space between the canister and the liner could provide early warning of a canister failure that Muld develop over time by a mechanism such as corrosion.

Monitoring could als, be useful in detecting a failure of the liner. DOE should evaluate the efficacy of proposed monitoring of the storage units with respect to sensitivity and required frequency, for failure detection in all storage units.

3.2.8.3.2 Nuclear Fuel Although 10 CFR 72.72(h)(1) requires that the fuel cladding be protected against degradation and gross ruptures, the double barriers to radioactivity release discussed in Section 3.2.8.3.1 above might preclude the need for such maintenance of the clad barrier. NRC is considering modifications to 10 CFR 72 to reflect this approach (NUREG-1092). However, the MRS limits of 375'C for clad tempera-tures for fuel sealed within the canisters would appear to ensure that the protection against cladding degradation and gross ruptures will be achieved in the inert gas (argon) atmosphere within the canister.

According to A. B. Johnson, Jr., of PNL, experiments and research in the United States and overseas have indicated that intact Zircaloy-clad spent fuel assem-blies store <1 in inert atmospheres at initial peak temperatures up to 380*C have developed no cladding defects, or if the cladding has had defects, the defects have not propagated.

3-42

These tests are summarized in Tables 3.1 and 3.2. Many of these tests are con-tinuing, but according to Johnson, no significant changes in the results have occurred.

None of the tests conducted to date have been performed on groups of consoli-dated fuel rods, which could remain at higher temperatures as compared to single rods or rods in an intact fuel assembly. However, the data do appear to allay concerns about gross ruptures when the cladding is kept below 375*C in the inert atmosphere within the canisters, especially considering that the tests indicated temperatures of 765*C to 800 C were required to drive rods to rupture (Johnson and Gilbert, 1983).

These tests and experiments have indicated that the U0 2 of the fuel pellets can oxidize if it is exposed to air at elevated temperatures. The oxidation is accompanied by expansion of the fuel matrix. Therefore, any defects in the fuel cladding that would permit an oxidizing agent access to the fuel material could result in expansion of the fuel, enlargement of the defect, and poten-tially degradation and gross rupture of the cladding. Data generated in NRC-sponsored tests with artificially defected commercial fuel rods in air (NUREG/CP-0057) indicate that cladding can crack around pre-existing defects at temperatures at least as low as about 217*C. This phenomenon is explained by the formation of U3 0s at this temperature with its attendant expansion.

At the MRS, fuel would be exposed to air during in process lag storage and consolidation until the canisters were sealed. The fuel might also be exposed to air in storage, if a canister leaked. Because fuel with reactor-induced cladding defects would be received at the MRS and cladding defects could develop during fuel handling, especially during consolidation, the maximum fuel cladding temperature before the fuel is put in a canister should be kept below 217*C, These limits could change, permitting an increase in this temperature, at least for short periods, pending the development of more data from on going tests worldwide. The functional design criteria (PNL, 1985) require a limit of 220*C.

The actual peak cladding temperatures for fuel rods in the canisters--whether in lag storage in the R&H building or stored within the outside storage units--

would be determined by calculations based on the outside temperature of the 3-43

- -. - _ - _ ~ _.

Tabla 3.1 Labor tiry/ hot call tr.ts inv21ving Zirecioy-clad fuel rods i

Rods Tcst Surveillante methods cnd r;sults Organt- Clad- Test Temp., Burnup time, Country zation ding type *C Cover gas mwd /tiTU BWR PWR days Type

• No.** End date Results Canada AECL/ Zirca- Rod 220-250 Air 7000-8000 -

13 30 3,4,6 -

In Clea defects OH foy-4 capsule PHWR progress elongated

• • • In progress Zirca- Whole 150 Noist air 5000 -

8 1000 loy-4 rod 150 Ory air 5000 -

8 1300 IRG KVU Rod 100 He + Is 0 - -

550 - - -

No failures capsule 300 He + Ir 0 - -

Zirca- Rod 350-450 Helium To 40,000 -

20 40-170 - - -

In prog *ess loy-4 capsule Zirca- Whole 350-400 Helium -

6 *180 1-3 -

In -

loy-4 rod 7,8 progress FRG/ Ispra/ Zirca- Whole 400 Helium PWR:33,000 14 9 1801 1-3 15 3/82 No clad' failures Italy NUKEM loy-2 rod Zirca- 430 Helium 8WR:24,000 14 9 15019 1 15 2/83 No clad failures loy-4 450 Helium 14 9 30111 1 6 7/83 No clad failures (in progress)

USA BCL/ Zircz- Whole 480t Ltd air, 28,000 -

3 190 1-5 1 1979 No clad failures HEOL loy-4 rod helium -

3 190 510 Ltd air, 28,000 -

3 325 1-5 1 No clad failures helium 3 325 w $70t Ltd air, 28.000 1 30,73 1-5 1 No clad failures e helium -

1 30,73

$ Zirca- Whole 325 Ltd air, 28,000 -

3 '87 2-5 -

1983 No clad failures loy-4 rod argon -

3 87 2-5 -

1983 No clad failures USA BCL/ Zirca- Whole 325 Rpind air, 17,000 1 -

87 1-5 s5 1983 Defects spread I PWL loy-2 rod argon 24,000 1 -

87 1-5 s5 1983 No defect spread USA PWL Zirca- Whole 765-800f Ltd air 0 -

3 1-2 '2-4 -

1982 Pinhole defects loy-4 rodtt at 765+800*C -

USA EPRI/ Zirca- Rod 250-350 Air 28,000 -

48 N - - -

In progress HEOL/ loy-4 capsule spect-TVA/ANL mens USA NRC/ Zirca- Whole 230 Unitd 12,000 4 -

248 1-3 -

8/83fM 1 clad defect HEDL/ loy-2 rod air, argon elongated EG&G Zirca- Whole Unitd 31,000 - a 248 1-3 -

8/83N# No clad failures loy-4 rod air,arger:

• 1-fission gas analyses; 2-nondestructive rod tests; 3-visual inspection; 4-destructive analyses; 5-sipping; 6-weighing; 7-H2 O trapping; 8-fission product migration.

" Number of cover gas samples analyzed; *** Scheduled fo- 1984.

tat 400*C; Stat 430*C; 111at 450*C.

tisothermal for 190 days; then four rods were transfe ..ed to 570*C.

f fUnirradiated rods; however, under test conditions, radiation would have been fully annealed.

[ # increasing rasp to rod rupture; NIn progress; N# Phase 2 is in progress.

4 Source: Table 5 of Johnson and Gilbert,1983.

4 m

t I

Table 3.2 Dry storage demonstrations involving Zircaloy-clad fuel Maximum clad Fuel toep., *C Wet No. of Dry fuel burnup, Storage storage Integrity verifications anses- storage Cover Current. No.

Country Site type N1d/MT11 Cladding concept time bifes time gas Init, final of rods Type

• No.** Datet Results Caseda WNRE WR-1 6,000 Zr-2.5 Mb Silo 0.5-4 yr +138 1975

• Helium 120 *4 2484 - - -

Canada WNRE PWWR 6,000 Zircaloy Silo 0.5-4 yr 360 1976 + Nettum 120 -4 6820 2-4 -

1982 No rod failurest?

l FRG Mnc PWR 33,000- Zircaloy-4 Test 10 mo 1 60 days Noist 300 250 180 1,5 3 1981 No rod failures 43,700 module Ns FRG NWO PWR 33,000 Zircaloy-4 Test 4.5 no 1 12 no Moist 400 330 180 1-3,5 11 1982 No rod failures module N ~

FRG kw SWR 28,000 Zircaloy-4 Metal 1 yr 2/82-16 Heltes 385/ 250/ 896 Itti 9 5/83 No rod failures cast 2/84 180 120 FRG RK5/ PWR 35,000 Zircaloy-4 Metal 2/83-Karlsruhe 10 mo 4 Helium 430/ 340/ 816 1,2,3 5 7/83 No rod failures cask 2/85 200 160 Y Switz- Wuerent- Test 17,000 Zircalcy-2 Metal 6-10 yr 3 MTU Nay 1983 Helium 180 - -

2 - - -

g erland ingen reector cask to final (est) disposal USA EM4D/hTSt PWR 28,000 Zircaloy-4 Surface 2-3 yr 4 1/78 + Helium 170 12C 816 1-511 6 drywell 7/83 No rod failures USA EM43/NT51 PWR 28,000 Zircaloy-4 Silo 2-3 yr 1 11/78- Helium 150 100 204 1-5 8/62 1 6/82 No rod failures usa EM40/NT51 PWR '8,000 Zircaloy-4 vault 2-3 vc 12 12/79 + Helium 220 -

2448 1-5 5 7/83 No rod failures USA Climau/ PWR 28,000 Zircaloy-4 coep dry 2-3 yr 11 To 4/83 Helium 230 s160 2244 1-5 NTS well 11 7/83 No results to date USA EMAD/NTS PWR 28,000 Ifrcaloy-4 FTT 2-3 yr litt 3/83 Air 275 260 202 1-5 7 9/83 One failed rod 1

0 -fission gas analyses; 2-nondestructive tests on rods; 3-visual Inspection; 4-destructive analyses; 5-sipping; ** Number of cover gas samples analyzed.

t0 ate of last cever gas analysis; ttRoos in one basket were examined after 4 years; titPrecharacterization: sipping; profilometry; eddy-current; and oxide thickness measurements on 10 rods.

15ee App +ndia C of for details of Turkey Point fuel assembly histories; ttone of 18 asseeb11es analyzed at hot cell; 5 assemblies at NTS have pre-enacer.arized rees; only interie visual examinations and fission gas analyses have been conducted stece the NTS demonstrations began; ettThree PWR assemblies 1

have been tested in the fuel temperature test (FTT) module, one at a time.

Source: Table 6 of Johnson and Gilbert,1983.

i I

l l

l W_ - - .

l 1

l 1

canister rather than by direct measurement. The calculations would be based on certain assumed geometric arrangements and on expected heat transfer pro-perties and mechanisms inside the liner. In practice, consolidated fuel rods might not have a predictable geometric arrangement and might have different heat transfer mechanisms of importance. This could lead to large uncertainties in the calculated peak cladding temperatures, a fact recognized by DOE in the development of its computer codes and in its planning of full-scale temperature tests of consolidated fuel rods in metal casks. The tests are scheduled for Idaho National Engineering Laboratory in 1987-1988 (Gertz, 1985). The results should be applicable to validation of thermal models for spent fuel storage in the MRS. These tests are discussed in DOE's Design Verification Plan.

3.2.8.3.3 High Level Waste The high level waste centerline temperature limit of 500*C and canister surface teraperature limit of 375*C (PNL,1985) used for the MRS appear adequate with r respect to ensuring that the expected waste glasses would resist crystallization 1

and maintain their "as poured" form.

Crystallization (devitrification) is undesirable for high level waste glasses intended to be emplaced in a geologic repository because it may result in a loss of leach resistance. However, experiments (Rockwell, 1982; Bickford and Jantzen, 1983, and Chick et al.; 1984); indicate that devitrification typf-cally occurs only between 500*C and 950*C, with crystallization below 500*C diffusion-limited because of the high viscosity of the glass matrix. For annealing of samples of typical waste glasses formulations, devitrification has not been observed at temperatures less than 500*C. However, under long-term post-emplacement conditions in a repository, slow partial devitrification may occur even at waste centerline temperatures less than 300*C (Rockwell, 1982).

This or other factors (limits on thermal cycling) concerning the interactions between the waste form and the geologic repository may result in different requirements or restrictions for waste glass storage at the MRS. Whether these considerations will modify the MRS final design or operation will have to await results of ongoing research.

3-46

1 3.2.8.3.4 Ventilation System The staff has evaluated the ventilation system for the R&H building with respect to its capability to confine airborne radioactive particulate materials.

Design features that help meet the confinement requirements include (1) means to collect and remove airborne radioactive particulate materials before they are released to the atmosphere (2) provisions for ventilation control zones and pressure differential (3) provisions for air filtration and backflow prevention between ventilation zones (4) manual controls, as well as automatic controls that can be manually bypassed (5) the capability for automatic shutdown of portions of the ventilation system, if they can limit major accidents (6) the design of equipment to fail in a safe mode (7) provisions that failure of any single ventilation system component does not jeopardize the ability of the system to confine radioactive materials (8) provisions for isolating the exhaust systems that serve highly contaminated or potentially contaminated zones from zones that are not normally con-taminated and unrestricted zones (9) the design of the exhaust control system to be interlocked with the air supply control system to maintain the prescribed pressure differentials and air flows (10) provisions for exhaust system redundancy (11) the capability to cool the lag storage vaults 3-47

l (12) the design, materials, and layout of ductwork to provide for decontamina-tion and to avoid contamination of ducts serving non-contaminated areas (13) equipment arrangement so the exhaust fans are outside the zone being exhausted (14) provisions for air sampling (15) provisions for radiation monitoring of the ventilation system, including the exhaust stack effluent (16) provisions for MRS standby electrical power and localized backup power for critical areas (lag storage vaults)

(17) provisions for system operability during off-normal events (fire, earth-quake, tornado)

In evaluating the ventilation system design, the staff found that the large majority of these features have been included (see Section 3.2.8.2.3 above).

However, certain ventilation system aspects important to safety may require further analysis during the final design. These are: ,

(1) Based on Parsons (MRS-11, September 1985, Vol III, Dwgs H-3-56836, 56837, 56838, and 56840) the only radiation monitoring provided on a continuous basis would be just before the exhaust air enters the exhaust stack (one area for each of the three major exhaust systems, Zones 1 and 2, Zones 3 and 4, and the lag storage vaults). Radiation sampling ports would be provided in the filter banks for the exhaust system for Zones 1 and 2, and none would be provided for the other systems.

1 1

(2) The only fire protection system provided in the exhaust air system would be in the final exhaust filter bank plenums. Thus a fire could propagate through the entire system before it was intercepted. Furthermore, the final exhaust filters would be interconnected at a common plenum, which could propagate the fire and disable several or all filter banks (see also Section 3.2.3 above).

3-48

(3) The design does not sufficiently implement the concept of physical separa-tion of the final exhaust system equipment for ensuring continuing opera-tion during off-normal conditions. All filters for Zones 1 and 2 would be in one room, as are the exhaust fans. The filters and fans for Zones 3 and 4 would be similarly arranged. A fire in one of these rooms would disable the entire exhaust system. In addition, there would be communica-tion between the filter rooms of Zones 1 and 2 and Zones 3 and 4, as well as between the exhaust fans rooms.

(4) As discussed in Section 3.2.3, airborne Zircaloy particles would be generated in the process cell during spent fuel consolidation. Althougn the conceptual design includes a specific system for collecting these  ;

particulates at the source (a dedicated dust collection system), the staff l questions the capability of the design to confine the particles. This approach has two potential disadvantages: (1) it could present a fire hazard at the filters and (2) HEPA filters could be rapidly clogged by the airborne particulates.

3.2.8.4 Conclusions and Recommendations The proposed design of the storage system for spent fuel and HLW appears to provide the means for confining the radioactive materials safely. However, the staff recommends that DOE (1) test the performance of the storage units (casks l and drywells) under expected MRS conditions, (2) use a seismic Category I de-sign, (3) study the galvanic corrosion potential between the carbon steel storage unit liner and spent fuel canisters of dissimiliar materials, and (4) monitor a statistically justified percentage of storage. Although the ventila-tion system appears to provide adequate means for confining airborne radioactive particulate materials, additional provisions are recommended.

One factor in preventing the oxidation of spent fuel during handling and storage is the ability to keep the cladding and fuel cool. The proposed criterion to limit cladding temperatures to 375*C in an inert atmosphere appears adequate for safety, although it is based on a limited body of data. Zircaloy cladding meet-ing this criterion would not be expected to deteriorate. However, some remain-ing uncertainties are being investigated (Proceedings of the International 3-49

Workshop on Irradiated Fuel Storage, 1984), and such investigations should be continued to provide useful supporting information, although the information is not crucial to the dry, inerted atmosphere concept. The uncertainties include (1) behavior of spent fuel in other than a dry inert atmosphere; (2) methods to predict maximum cladding temperatures, especially for arrays of consolidated fuel rods; (3) behavior of cladding at elevated temperatures for extended time I periods, as in arrays of consolidated fuel rods; and (4) behavior of stainless steel cladding under dry storage conditions.

Full-scale tests currently planned for consolidated fuel rods in casks should proceed. The fuel should be typical of light water reactor fuel to be stored

at an MRS. Testing should also continue to determine an acceptable temperature regime for storage of spent fuel in air. Continued and accelerated efforts should be devoted to the development and validation of computer programs for predicting fuel clad temperatures in consolidated fuel rod arrays.

Regarding storage of high level radioactive waste, a 500*C limit on borosilicate glass appears adequate. However, until further research is completed on high level waste package performance in a repository environment, it is not known whether other constraints should be placed on the MRS design or operation.

Although DOE is not required to do such research as part of the MRS effort, the results of such research could affect certain features of the proposed MRS design.

For the ventilation system, additional provisions for fire protection of the final exhaust system should be evaluated (as discussed in Section 3.2.8.3). DOE must also give additional consideration to the separation of equipment in the final exhaust system. For the process cells, the final design should demonstrate the performance and adequacy of an airborne dust / particulate collection system at the spent fuel consolidat. ion station.

3.2.9 Instrumentation and Control Systems 3.2.9.1 Requirements 10 CFR 72.72(i) gives the requirements for instrumentation and control systems as follows:

3-50

l Instrumentation and control systems shall be provided to monitor systems that are important to safety over anticipated ranges for normal operation and off-normal operation. Those instruments and 1

control systems that must remain operational under accident condi- I tions shall be identified in the Safety Analysis Report.

I 3.2.9.2 Description Instrumentation and control systems for the ventilation system are described and discussed in Section 3.2.8 above. Instrumentation and control systems for radiation protection, fire protection, and criticality are described and dis- l cussed in Sections 3.2.13, 3.2.3, and 3.2.12, respectively.

i Another system that would have some importance to safety is the distributed control system (DCS). This system would use independent control consoles, one in the R&H building control room and the other in the separate site services building control room. The system would provide supervisory monitoring and control of all operations and employs closed circuit television. This system would perform analytical functions such as trending plots and provides reports on various topics such as special nuclear material inventory. The DCS would be connected to the standby power system through the UPS.

3.2.9.3 Discussion I

l During normal operation, much of the use of the control system would be directed toward production rather than safety. However, maloperation of the controls could result in less safe conditions. For example, a programming error could conceivably result in a laser cut into the active portion of a fuel rod, which

! would result in the release of fission products and fuel into the handling cell environment. The DCS could enhance safety by monitoring many conditions con-tinually, such as fire water reserve, stored fuel temperatures, and the location of special nuclear material. The staff expects that a complete testing of this system would be performed during the preoperational phase, before operation with radioactive materials, for a full demonstration of the enhancement of safety. DOE has committed to this aspect in its Program Plan.

i 3-51 i ... - - .- - _. -_

3.2.9.4 Conclusion The staff's conclusions related to instrumentation and control systems for ventilation, fire protection, radiation protection, and criticality are discussed in Sections 3.2.12 and 3.2.13. In regard to the distributed control system, the staff concludes that it would have the potential for enchancing safety through a thorough preoperational testing program.

3.2.10 Control Rooms 3.2.10.1 Requirements 10 CFR 72.72(j) gives the requirements for a control room as follows:

A control room or control area shall be designed to permit occupancy and actions to be taken to monitor the ISFSI [MRS] safely under normal conditions, and to provide safe control of the ISFSI [MRS] under off-normal or accident conditions. (N.B. This sentence is being con-i sidered for modification by the Commission with respect to the need for a control room or area in all instances of spent fuel storage.)

3.2.10.2 Description As discussed above, there would be one control room in the R&H building, and an alternate control room in the site services building. Each of these control rooms would house the center for a DCS, as described in the previous section.

The primary control room would be in a seismic Category I structure, but the I alternate control room would not be.

l 3.2.10.3 Discussion I

The relationship of the control rooms to safety is not clear. 10 CFR 72.72 states that occuparecy of the control room (or its alternate) may be necessary.

If detailed analysis would show that certain functions important to safety must be performed from either control room, the criterion would apply. In a previous analysis, the NRC staff determined that complete evacuation of a storage site may be permissible, under remote circumstance, without any radiological impact 3-52

on public health and safety. This is primarily because of the passive nature of the storage and handling operations.

3.2.10.4 Conclusion The Conceptual Design Report (Parsons, MRS-11, September 1985, Vol I) indicates that the MRS would have both a well qualified control room and an independent alternate control room. This approach is very conservative and probably more than adequate. Detailed analysis of credible events would be necessary to establish the need for constant occupancy and use of the control room.

3.2.11 Utility Services 3.2.11.1 Requirements 10 CFR 72.72(k) gives the requirements for utility services as follows:

1 (1) Each utility service system shall be designed to meet emergency l

conditions. The design of utility services and distribution systems that are important to safety shall include redundant systems to the extent necessary to maintain, with adequate capacity, the ability to perform safety functions assuming a single failure.

l (2) Emergency utility services shall be designed to permit testing of the functional operability and capacity, including the full opera-tional sequence of each system for transfer between normal and emer-gency supply sources; and to permit the operation of associated safety systems. l (3) Provisions shall be made so that in the event of a loss of the primary electric power source or circuit, reliable and timely emer-gency power will be provided to instruments, utility service systems,

. the central security alarm station, and operating systems in amounts sufficient to allow safe storage conditions to be maintained and to permit continued functioning of all systems essential to safe storage.

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3.2.11.2 Description The utility systems that would support the MRS include the water supply (pota-ble, process, and fire water), and power supply, fuel oil, as well as systems that would supply plant air, instrument air, breathing air, decontamination solutions, freon, steam, and inert gas.

Water Supply The water supply for the MRS (the CRBRP site) is common for potable, process, and fire water systems. The source of water is the Oak Ridge Gaseous Diffu-sion Plant water treatment plant. A storage tank on the MRS site would provide potable water (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' worth), process water (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' worth), and fire water reserve (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />' worth). The water supply would be provided by three electric pumps (one spare) connected to normal and standby power. In case of fire, an electric fire pump would be activated and take over the distribution for potable, process, and fire protection water. This pump would be connected to standby power and would have a diesel pump backup in case the power failed.

t In normal operation, the two water distribution pumps would supply water on the site via a common loop (for potable, process, and fire water). A standpipe in the tank would not permit the water distribution pumps to draw from below a level that ensures that the 4-hour fire water reserve would be maintained.

The water distribution pumps could accommodate the combined peak cemand of all ,

MRS water users plus water release for one or two sprinklers to cover a minor fire.

If, during a fire of larger magnitude, the water distribution pumps could not maintain the mainline pressure at a minimum of 100 psi, the fire pump would start and the water distribution pumps would be shut off.

• Parsons. MRS-11, September 1985.

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After a major fire, when water below the level of standpipe would have been used, the fire pumps would be shut off manually, the water distribution pumps returned on line, and the water storage bypass valve opened. The tank would thus be refilled while still providing some water supply to the facility.

After such a major fire, the MRS would not restart operations until the fire water reserve was restored.

In the RAD, the water supply system for the MRS has been designated as non-Category I, QA Level III, except fire water supply, which is designated as QA Level II.

Power Supply The power supply for the MRS includes normal electrical power, standby power, and uninterruptible power supply (UPS). As indicated by their nomenclature, each of these systems is intended to supply power to the .MRS under different circumstances.

Normal electrical power would be provided by two feeders, routed on separate steel towers, from the nearest commercial power supply. The towers would be seis-mically designed and segregated laterally so that a tornado generated missile would not cause simultaneous failure of transmission lines. A quarter of a mile outside the MRS, the transmission lines would be routed underground. The main substation would include an outdoor, high-voltage switchyard; power trans-formers; and an indoor medium-voltage switchgear located in the hardened switch-gear room of the standby generator building. Power would be distributed from the medium-voltage switchgear to unit substations via an underground duct bank system, and the cables would be segregated for each switchgear assembly. Dual l feeders would be provided to each double-ended unit substation (those that also would be connected to standby power) and to each primary selective substation.

Standby power would be generated by onsite diesel-engine-driven generators.

Four generators would be provided, each with a continuous rating of at least '

33% of the maximum standby load. Power would be distributed from the standby switchgear to standby unit substations in the buildings connected to standby 3-55

power. If one normal power source were to fail, the second normal power trans-former would assume the MRS load. Should the second transformer fail, all diesel generators would concurrently start and synchronize, and transfer automat-ically to the standby bus. Standby power would be provided to such systems as the HVAC filtered exhaust fans, UPS battery chargers, and selective lighting.

The facility would be in a shutdown mode during use of standby power.

A UPS would be provided for selected services that must be permanently available, such as radiation monitoring, criticality monitoring, and fire alarms. The UPS systems would operate from 480-volt, 3 phase input to the battery charger and 480Y/277-volt, 3 phase output from the inverter. There would be four battery charger / inverter units, each rated at 50% of the maximum UPS load. In the' RAD, this system has been designated as non-Category I, QA Level II, except for the UPS, which is designated as QA Level III.

Fuel Oil System i

The fuel oil system would supply the emergency generators and the standby engine-driven fire water pump, as well as some equipment that is not essential for containment. The fuel oil would be stored in an above ground storage tank from which oil would be pumped to the individual storage tanks of the systems listed above. In the RAD, this system has been designated as non-Category I, j

QA Level II.

3.2.11.3 Discussion Water Supply The RAD states that the water supply system is not classified as a system important to safety because it is not required to prevent a release of radio-nuclidas to the environment or for safe shutdown. However, design provi- I sions would be made to ensure the availability of the system for off-normal conditions, except for the design earthquake, when some of the equipment and components might be disabled. These design provisions would include I

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(1) preserving the fire water reserve by a standpipe in the water tank that would prevent use of the reserve unless a fire were in progress (2) providing redundancy for the fire water pump with a 100% capacity standby diesel-driven pump that would have its own fuel oil storage tank in case i both normal and standby power to the electric-motor-driven pump would fail (3) winterizing the water storage tank by providing an electric water heater that would prevent freezing during winter months (4) including a separate water tank designed seismic Category I (air pressurized) to supply fire water for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to the sprays protecting the final HEPA filters for the exhaust plenums for Zones 1 and 2 (5) installing fire protection equipment in Category I structures designed for the design earthquake (6) preserving the fire-distribution network by providing a looped system sized to allow one main to be out of service and still provide adequate fire protection (7) providing for routine maintenance, periodic calibration, and testing of 1

i instruments serving the water supply system These measures appear to provide adequate assurance that fire water would be available in most instances, except for an earthquake, during which the water supply system might be lost. The final MRS design should be analyzed to determine if, during a major fire, radioactive materials could be released from the R&H building. The staff must analyze the final HEPA filter sprays to deter-mine if they would be effective if they became wet or if they would continue to form a buffer to the outside environment if the building became pressurized as a result of a major fire.

Power Supply The power supply system is not classified as a system important to safety.

However, design provisions would be made to ensure that the system would be 3-57

I available for off-normal conditions, except for the design earthquake, when some of the equipment and components might be disabled. These design provi-sions would include (1) providing the normal power supply system with two redundant and segregated i feeders for the site (2) routing transmission lines underground at a quarter-mile distance from the i MRS (3) locating the medium-voltage switchgear in the hardened switchgear room of the standby generator building (4) providing a standby power supply for selected equipment and components (5) providing redundant equipment for the standby power supply system (6) providing fail-safe features for equipment and components to ensure safe shutdown of the MRS in case of loss of normal power (7) providing for a UPS to selected services important to safety in case of failure of both normal and standby power (8) providing redundant equipment for the UPS system (9) locating the UPS system in a seismic Category I structure designed for the design earthquake These measures appear to provide adequate assurance that power would be avail-able in most instances, except for a large earthquake, during which all  ;

sources of power, including the UPS, might be lost.

The final design of the MRS should be analyzed to determine if the UPS system classification as non-Category I is adequate in terms of the consequences of its not monitoring safe storage or vital facility alarms. In addition, the standby power system classification and loads connected to it should be reviewed to l 3-58 l

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i l

consider (1) the effects of loss of power on operational safety in the process j cells, and (2) the effects of potential long-term loss of cooling capabilities for the lag storage vault and fuel in process within the R&H building.

Fuel Oil System The RAD states that the fuel oil system is not classified as a system important to safety because its failure would not result in a release of radionuclides to the environment or prevent a safe shutdown. However, design provisions would be made to ensure the system would be available for off-normal conditions, except for the design earthquake, when some of the equipment and components might be disabled. These design provisions would include (1) locating the fuel oil storage tank for the standby generator underground, close to the standby generator building (2) locating the fuel oil day tanks serving the operation of the lag storage standby generators and the standby engine-driven fire water pump in Category I structures designed for the design earthquake DOE (Parsons, MRS-11, September 1985) has not presented enough information for the staff to determine if this system is adequate to ensure the availability of fuel oil to the standby generators and to the backup fire pump. The final design should be analyzed to determine (1) the availability of fuel at startup of this equipment if the site fuel oil tank were lost and (2) the survival of the fuel oil day tanks and associated piping and instrumentation during a large earthquake.

3.2.11.4 Conclusions and Recommendations The utility services supporting significant functions for systems important to safety appear generally adequate. However, the staff will analyze the final design to establish the effects on equipment important to safety if fire water, UPS, or fuel oil to the backup fire pump were lost after a design earthquake.

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3.2.12 Criticality Design and Control i

3.2.12.1 Requirements 10 CFR 72.73 gives the requirements for nuclear criticality safety as follows:

(a) Spent fuel handling, transfer, and storage systems shall be de-signed to be maintained subcritical and to prevent a nuclear criticality accident. The design of handling, transfer, and storage systems shall include margins of safety for the nu-clear criticality parameters that are commensurate with the un-certainties in the handling, transfer, and storage conditions, in the data and methods used in calculations, and in the nature of the immediate environment under accident conditions.

(b) The design of an ISFSI shall be based on either favorable geome-try (spacing) or permanently fixed neutron absorbing materials (poisons). Where solid neutron absorbing materials are used, the design shall provide for positive means to verify their con-tinued efficacy. In criticality design analyses for underwater storage systems, credit can be taken for the neutron absorption of rack structures and the water within the storage unit.

3.2.12.2 Description The MR$ would use a dry process. In a limited number of areas, moderator materials (water, solutions used for decontaminating facilities and equipment) might accidentally come in contact with the fuel or equipment containing fuel materials. These areas would include the following:

(1) process cell lag storage (2) canyon cell lag storage ]

(3) drywell storage 1 (4) cask storage (5) liquid radwaste processing 3-60 i

Short-term storage for unconsolidated PWR and BWR spent fuel assemblies would l be provided in the process lag storage area. The canyon cell lag storage area '

would be provided for storing PWR and BWR fuel assemblies, as well as consolidated rods from these assemblies. The assemblies and rods (but not mixtures of the two) would be stored in seal-welded canisters. The only liquid that could accidentally enter the lag storage cells would be decontamination solution.

The dry well and cask storace of sealed canisters containing either fuel assemblies or fuel rods normally would be dry. Criticality analyses have been 4

made in which each canister in the drywells was accidentally filled with water and a single canister (out of a total of 12 canisters / cask) in each storage cask also was accidentally filled with water.

The liquid radwaste system--which would consist of a high activity liquid waste collection vessel, a liquid evaporator, and an evaporator bottoms slurry tank--

might also contain fissionable material. This system would be the only one in the MRS that would normally contain liquid.

3.2.12.3 Discussion If the facility were dry during all operations--from the receipt of the fuel assemblies until the casks or drywells containing canisters of consolidated rods or fuel assemblies (including all penetrations) were sealed--there would be no possibility for accidental criticality outside the liquid radwaste system.

The k, is less than unity for unmoderated uranium, in any form, enriched to less than 5 weight % of U-235. However, because the presence of water or other moderator rr.aterials cannot be ruled out (by accident in the lag storage areas or in the canisters and under normal operations in the radwaste processing area),

special design and process features to avoid accidental criticality must be considered.

The minimum number of PWR fuel assemblies that might be made critical under optimum water moderation and reflection is two; for BWR assemblies, the minimum is five. The rods from a single PWR assembly (depending on enrichment and rod diameter) might become critical at optimum water moderation (where there is no control of spacing between rods) and reflection. Therefore, all steps in the consolidation operation from the time of receipt of fuel assemblies to their 1

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final consolidated storage must be reviewed to determine steps in the opera-tions where moderation is possible. The potential for inadvertent criticality must be determined in all steps where moderation is possible. All degrees of possible water moderation must be considered, from very low density to full density. It has been shown that the maximum k,77 for unpoisoned arrays of LWR fuel materials occurs at very low density water moderation. To eliminate the possibility of adding moderator to the arrays of fuel rods and/or assemblies,

in the proposed MRS operations no liquid materials would be used to fight fires in the hot cells. For example, if a motor were to catch fire, the plan would be to let it burn out. The decontamination solution would be the only source of liquids in the cells. The casks used to store canisters of consolidated rods would have sufficient shielding to guarantee neutron isolation from the rods in adjacent casks. Neutron isolation between canisters in adjacent drywells is also guaranteed (because they will be 16 feet apart center-to-center).

In evaluating the nuclear criticality safety of the lag storage areas, the MRS designers assumed that the maximum credible accident would have the areas flooded with decontamination solution to a depth of 6 inches. However, the

[ designers should also assume that the space between canisters would be flooded

! with a spray of decontamination solution (from leaking decontamination solution lines). One way to eliminate the possibility of a spray within the canister -

array would be to preclude the routing of decontamination solution piping over the array.

Although the liquid radwaste system normally would contain liquids, and the possibility of accidental criticality is very remote because of the low concen-tration of fissionable material, the nuclear criticality safety in the system operation was examined during design. The analysis appears to be reasonable.

If 00E submits an MRS license application, the staff will review the issue.

The design of the facility and its planned operations would be reviewed for compliance with RG 3.4, " Nuclear Criticality Safety in Operations with Fission-able Materials Outside Reactors. RG 3.4 is essentially an endorsement of the related ANSI standard, which currently is ANSI /ANS 8.1-1983. However, use of the standard is not a substitute for detailed nuclear criticality safety analyses for specific operations. 00E should provide the details of the validation of 3-62

calculational methods used for nuclear criticality safety, as indicated in Paragraph 4.3.6 of the standard, to demonstrate the adequacy of the safety mar-gins relative to the bias and to the criticality parameters. The details of

! the validation should also demonstrate that the calculations embrace the range i of variables to which the method will be applied.

The Conceptual Design Report (Parsons, MRS-11, September 1985) states that the MRS would include a neutron-sensitive criticality alarm system to meet the in-tent of ANSI /ANS 8.3, " Criticality Accident Alarm System," and ANSI N16.2,

" Criticality Accident Alarm System." (ANSI /ANS 8.3 is a revision of ANSI N16.2.)

If DOE submits an MRS license application, the staff will evaluate the adequacy of the criticality alarm system in accordance with RG 8.12. " Criticality Acci-dent Alarm Systems." 10 CFR 70.24(a)(1) states that each area in which an alarm is required must be covered by two detectors, in contrast to Paragraph 4.5.1 of the standard, which allows coverage by a single detector.

3.2.12.4 Conclusion On the basis of its review to date, the staff finds that DOE has given adequate attention to nuclear criticality safety. If DOE submits an MRS license appli-cation, the staff will review the final design, safety analyses (including the staff's independent verifications), the operating organization, administrative controls, and engineered features that would ensure the nuclear criticality safety of the operation.

3.2.13 Radiological Protection 3.2.13.1 Requirements l

10 CFR 72.74(a) gives the requirements for radiological protection as follows:

(a) Exposure control. Radiation protection systems shall be pro-vided for all areas and operations where onsite personnel may be exposed to radiation or airborne radioactive materials. Struc-tures, systems and components for which operation, maintenance, 3-63 i

~

l and required inspections may involve such exposure shall be designed, fabricated, located, shielded, controlled and tested so as to control external and internal radiation exposures to personnel. The design shall include means to:

1 (1) prevent the accumulation of radioactive material in those j systems requiring access; (2) decontaminate those systems to which access is required; (3) control access to areas of potential contamination or high radiation within the ISFSI (MRS];

(4) measure and control contamination of areas requiring access; (5) minimize the time required to perform work in the vicinity of radioactive components: for example, by providing suffi-cient space for ease of operation and designing equipment for ease of repair and replacement; and j (6) shield personnel from radiation exposure.

t (b) Radiological alarm systems. Radiological alarm systems shall be provided in accessible work areas to warn operating personnel of radiation and airborne radioactivity levels above a given set-point and of concentrations of radioactive material in effluents above control limits. Such systems shall be designed with pro-visions for calibration and testing their operability.

(c) Ef fluent and direct radiation monitorina.

(1) Effluent systems shall be provided with means for measuring the amount of radionuclides in effluents during normal opera- ,

tions and under accident conditions. A means of measuring the flow of the diluting medium, either air or water, shall also be provided.

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(2) Areas containing radioactive materials shall be provided with systems for measuring the direct radiation levels in 4

and around these areas.

(d) Effluent Control. The ISFSI [MRS] shall be designed to provide means to limit to levels as low as is reasonably achievable the release of radio-active materials in effluents during normal operations; and control the release of radioactive materials under accident conditions. Analyses shall be made to show that releases to the general environment during nor-mal operations and anticipated occurrences will t,e within the exposure limits given in 72.67. Analyses of design basis accidents shall be made to show that releases to the general environment will be within the ex-posure limits given in 72.68. Systems designed to monitor the release of radioactive materials shall have means for calibration and testing their operability.

3.2.13.2 Description 3.2.13.2.1 General 1

i l The MRS design would provide radiation protection for operating personnul in all areas where such onsite personnel might be expected to be exposed to radiation or airborne radioactive materials. DOE has achieved this protection by using as low as reasonably achievable (ALARA) radiation exposure principles in the facility design and equipment layout for operation, maintenance, and re-placement. These factors, combined with sufficient radiation shielding, would

! permit operation without exposing personnel to unacceptable radiation levels.

Also, remote maintenance would be used to reduce personnel exposures.

The following summary of MRS facility radiological protection features is based on the MRS Facility Regulatory Assessment Document and the Conceptual Design Report (Parsons, MRS-11, September 1985).

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3.2.13.2.1.1 Radiation Sources The primary radiation sources at the MRS would be spent fuel assemblies and HLW.

1 These materials would arrive at the MRS in shielded transport casks, which might l have both external and internal radioactive contamination. The exterior contami-nation would be removed at the cask decontamination area, producir"; liquid radwaste. Significant contamination would originate in the process cells of the R&H building as a result of spent fuel disassembly and consolidation. These radiation sources would include activated crud from the fuel assemblies, metal

{ particles from the laser cutting operations, possibly fuel pellet particles from 4

broken fuel rods, and fission product gases. Radiation from these sources would

) be deposited on the in-cell equipment and on the ventilation ducts and HEPA filters.

l 1

l Other radiation sources would include contaminated fluids from the decontamina-l tion of numerous low level radiation areas and their equipment (cask handling j and decontamination area, low level radwaste area, laundry room, analyzer room, j and analytical laboratory), and the following high activity radiation areas and their equipment: process cells, maintenance cells, high activity waste cell, j remote equipment decontamination rooms, remote handling air filtration and filter compaction cells, crane maintenance rooms, process cell ventilation l ducts, internal surfaces between non-testable and testable HEPA filters, drum l decontamination cells, and shielded canyons.

3.2.13.2.1.2 Shielding Shielding requirements for the R&H building process cells would be based on j estimates of (1) spent fuel source terms for fuel aged 10 years from reactor

! discharge with up to 55,000 MWD /MTU burnup; (2) the quantities and radioactive I

source terms of crud abraded from fuel assembly surfaces during handling and consolidation operations and shredding of non-fuel-bearing components; and l (3) the particles produced by laser cutting. In addition, 1% of the spent fuel l in the cells would be assumed to be failed upon arrival or to break during i in-cell handling and consolidation, with 0.1% of this failed fuel assumed to j deposit within the cell.

l.

Shielding requirements for the liquid and solid radwaste areas would be based on the quantitles of decontamination liquid expected to be generated, along 4

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with the expected contamination burdens from crud, laser cut particles, and spent fuel fission product depositions from failed fuel. Liquid radwaste system dilution and evaporator concentration also would be considered.

For the storage casks and drywells, shielding requirements would be based on the contained radiation sources--either spent fuel, HLW, HAW, or RHTRU. Shield-ing would also be provided for the transfer shield for the transport of spent fuel canisters or waste containers and their emplacement into the drywells in the storage facility, and for drywell shield plug emplacement.

Both gamma and neutron radiation sources were considered for the shielding calculations. The greatest uncertainties for the shielding analyses would be associated with the quantities and distribution of radioactive contamination sources in the HEPA filters, on the process cell equipment, and in the liquid radwaste system.

3.2.13.2.1.3 Dose Rate Limits and ALARA In determining annual direct radiation dose equivalents for MRS operating per-sonnel, 00E used 10 CFR 20 limits, with target values of one-fifth of these limits. 00E used ALARA principles (1) in system and component design to mini-mize equipment operation and maintenance exposure time and (2) in determina-i tions of remote versus contact operations and maintenance. A cost-benefit l evaluation of potential dose rate reduction has been recommended in the final design phase.

3.2.13.2.1.4 Plant Layout and Ventilation Zoning In addition to the shielding, the plant layout would restrict and contain radio-active contamination by regulating air flow in the direction of increasing con-tamination and by maintaining the air pressure in contaminated regions below

! ambient. 1 Contingent areas between contaminated areas and at exits from the operating areas would be provided with radiation monitors and decontamination facilities.

Area radiation monitors would be used to track radiation levels and trends in 3-67

the facility and to alert operations personnel to unusual radiation levels by audible and visual alarms, while process samples would provide information on radiation sources and system integrity. Swipes would be used to determine area contamination, and radiation badges would monitor and record personnel radiation exposures. Potentially contaminated work areas would be provided with monitor-ing equipment for checking personnel contamination when they leave these areas.

1 The ventilation systems for the areas containing radioactive materials would be equipped with HEPA filters (see Section 3.2.8 above). The R&H building would be divided into ventilation control zones based on the relative potential for personnel exposure to airborne contamination. Contamination spread would be limited by air-control vestibules, backdraft dampers, and other barriers to separate ventilation control zones from one another. Within the process cells,

airflow would be directed downward around the fuel disassembly / consolidation station to help draw loose contamination (crud and laser cut metal particles) away from the disassembly station. Contaminants deposited on process cells surfaces and equipment would be periodically removed by a vacuum cleaner system.

3.2.13.2.1.5 Access and Contamination Control Contamination barriers would be used to prevent the spread of contamination from 1

the process cells during cask unloading, from the shielded canyons during stor-age cask or transfer shield loading, and from the process cells to the shielded canyon at the canister access port.

All piping systems in operating areas that contain contaminated fluids would be corrosien resistant, isolated from non-contaminated systems, and accessible.

i Personnel access to potentially highly contaminated areas would be controlled by locked and alarmed doors at the air locks leading to contaminated areas. Health physics storage rooms would be located adjacent to the contaminated areas and would be equipped with portable radiation monitoring and survey instrumentation and communications systems. )

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3.2.13.2.1.6 Equipment Repair and Replacement In designing equipment that would be located in potentially contaminated areas, ,

DOE has considered maintenance requirements and effects on personnel. Equipment likely to become highly contaminated or to be located in a highly contaminated

area would be designed to allow remote removal, maintenance, and/or replacement, with provisions for decontamination and repair within a shielded area. Equip-ment likely to be repaired by actual contact would have in place decontamination capability for maintenance.

The removal of contaminated equipment would be facilitated by floor or roof panels and hatches and/or labyrinth and door openings sized to accommodato the [

, largest equipment within the potentially contaminated areas.

1 3.2.13.2.1.7 Housekeeping and Decontaminatiori i Housekeeping requirements to prevent accumulation of radioactive mattrials would include (1) survey and decontamination of all spent fuel and HLW canisters or waste drums before their placement in the storage facility, (2) scheduled decon- ,

tamination of contaminated areas and equipment during operation, and (3) moni-toring of storage units in the storage facility. If a leak in a storage unit were detected, the storage unit and the leaking canister or drum would be decon-taminated, and the canister or container would be overpacked before they were 1

returned to storage.

Decontamination activities would be facilitated by the use of (1) expendable vacuum cleaners, (2) portable decontamination carts with high pressure spray wands, (3) stainless steel liners or decontaminable protective coatings on all cells and areas that could be contaminated, (4) curbs to limit spread of con-taminated liquids, and (5) decontamination foam agents and high pressure water to decontaminate hot cells before personnel are allowed entry.

3.2.13.2.2 Radiological Alarm Systems l Il The MRS would have an automatic monitoring system. Releases of radioactive particulate material and gases would be limited so that exposures to operating 3-69

personnel and the public would be within the limits in 10 CFR 20 and 10 CFR 72.

Health physics instrumentation would be used to limit occupationally related radiation exposures in controlled work areas.

The radiation monitoring and alarm system would transmit signals to the primary distribution control system (DCS) and to the secondary backup DCS terminals located in the control rooms of the R&H building and the site services building, respectively (Parsons, MRS-11, September 1985, Vol III, Dwg H-356786). Failure of one terminal in the control rooms (in the R&H building and the site services building) or signal transmission line would not result in loss of protective functions of the radiation monitoring system. The control terminals would pro-vide information on dose rates, alarm conditions, and operational status or criticality conditions from criticality monitors. The system would provide audible and visible alarms, both locally and remotely, if any radiation detector  ;

indicated a trend, alert, or high alarm condition. DCS terminals transmit radiation monitoring alarm signals to the DCS interface terminals located in 4 the health physics office, the analyzer room, and the administrative building.

The radiation monitors that would be connected to the DCS terminals are (1) air-borne continuous alpha and beta monitors, (2) the continuous gamma area monitors, (3) the shipping cask off gas line monitors, and (4) the R&H building process cell criticality alarm system. The radioactive airborne monitors would be operated continuously to sample the air of R&H building rooms occupied by per-sonnel and where the airborne radionuclide concentration might exceed maximum permissible concentrations.

Area radiation monitors would be provided in the service areas, with both audi-ble and visual alarms for radiation intensities exceeding preset limits. Addi-tional area monitors would be located in the storage facility, with audible and visual alarms at the storage site and in the R&H building control room. Alarm systems would be designed for periodic inspection, calibrations, testing, and routine maintenance.

Portal monitors, fixed filter particulate air sampling systems, emergency moni-toring kits, portable survey instruments, floor monitors, hand and foot monitors, and laundry monitors would have local displays and alarms, and they would be not monitored by the DCS terminals.

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3.2.13.2.3 Effluent und Direct Radiation Monitoring The R&H building liquid and gaseous effluent monitors would transmit signals to the OCS terminals in a way similar to the way the radiological alarms described above are transmitted.

Effluent and survey monitors and alarms would be designed to measure releases of radioactivity from which off-site exposures in uncontrolled areas could be calculated for comparison to 10 CFR 20 limits. Air sampling capability would be provided in the exhaust systems for Zones 1 and 2 by test ports. Continuous j area radiation monitors would be provided on each leg of the exhaust system before the air enters the exhaust stack (see also Section 3.2.8 above).

Contaminated liquid streams generated in the RAH building would be routed to l and treated by the 11guld radwaste system, as described in Section 3.2.15. Non-

! contaminated liquid streams used to support the radwaste treatment process--

such as cooling water and steam and condensate returns from evaporators and heat exchangers--would be monitored for potential contamination. Other normally j noncontaminated liquid effluents from the R&H building, including its sewer systems, would be monitored with alarm instruments to alert if there were a

! release to the environment.

I 3.2.13.2.4 Effluent Control

< As previously noted, the MRS design would limit and control the release of radioactive materials during normal and accident conditions to levels consistent

! with ALARA principles. Releases of radioactive particulate materials and gases

! would be limited so that exposures to operating personnel and the public would be within the limits of 10 CFR 20 and 10 CFR 72. The results of the analyses of the effects of estimated releases on the general environment during normal I

operations, inticipated occurrences, and design-basis accidents indicate that the releases would be a small fraction of the applicable exposure limits of 10 CFR 72. The effluent control monitors would be designed for periodic inspec-I tion, calibv. tion, testing, and routine maintenance.

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3.2.13.3 Discussion The MRS design would meet the radiological protection requirements of 10 CFR 72.74 (a) through (d). The following radiation protection features would be included in the MRS design, ensuring that the criteria of this regulation would be met:

(1) shielding and facility orientation that would reduce radiation dose rates 4

to ALARA levels and conform to the requirements of 10 CFR 20, 10 CFR 72, and other applicable Federal regulations and NRC regulatory guides l

(2) ventilation zoning that would ensure flow of air from areas of lower contamination potential to areas of higher potential i

< (3) maintenance of ventilation flow rates into areas of high contamination potential at levels that would preclude air escape into lower potential areas through openings (4) filtration of gaseous and airborne particulate effluents from contaminated

' or potentially contaminated areas before they would be released to the 1

general environment (5) no planned releases of contaminated liquids; all liquid effluent streams would be monitored to ensure releases to the gener al environment were not t contaminated I

(6) monitoring of gamma dose rates in areas where there would be a potential for exposure rates above preset thresholds (7) monitoring and sampling gaseous effluents and storage area atmosphere to ensure releases to the general environment would be below 10 CFR 20 limits '

l and ALARA (8) monitoring and sampling of the air in potentially contaminated work areas to ensure operating personnel exposures would be below the maximum pre-scribed limits for occupied areas 3-72

(9) personnel survey stations at exits of potentially contaminated areas to detect personnel contamination l (10) personnel decontamination facilities 4

(11) criticality alarms I In general, the MRS design would include the items important to a good health physics program. However, the following are examples of concerns that DOE must I

address in the final design:

i (1) ensure that all reasonable design measures for dose reduction have been incorporated l

(2) ensure that all operations would be in keeping with ALARA principles i

(3) ensure that assumed source terms would be valid or conservative, especially for process cell operations and handling of decontamination solutions Concerns regarding the radiation monitoring for the ventilation system are discussed in Section 3.2.8 above.

3.2.13.4 Conclusions and Recommendations The MRS design appears to meet 10 CFR 72.74(a) through (d). As 00E finalizes the design, DOE should ensure that all ALARA concerns are resolved.

3.2.14 Spent Fuel and Radioactive Waste Storage and Handling 3.2.14.1 Requirements '

l 10 CFR 72.75(a) gives the requirements for spent fuel and radioactive waste  !

l storage and handling as follows: l Spent fuel storage, radioactive waste storage, and other systems that might contain or handle radioactive materials associated with spent 1

3-73

i l fuel, shall be designed to ensure adequate safety under normal and 1

accident conditions. The systems shall be designed with

4 4

(1) a capability to test and monitor components important to safety,

(2) suitable shielding for radiation protection under normal and accident conditions, (3) confinement structures and systems, (4) a heat-removal capability having testability and reliability consistent with its importance to safety, and i

! (5) means to minimize the quantity of radioactive wastes generated

• i 3.2.14.2 Description 3.2.14.2.1 General 1

) The systems and equipment that would be employed for spent fuel and waste handling and storage at the MRS can be categorized as handling systems and i

storage facilities. These can be further divided as follows:

Handlina Systems

)

lifting equipment (overhead cranes, monorails, etc.)

transport equipment (vehicles, carts) remote handling equipment (manipulators, robots, etc.)

cutting equipment (laser cutting, automatic pipe lathe) welding equipment (resistance welding, electron beam welder) l -

mechanical specialties (rod extraction and reconfiguring equipment) other mechanical equipment (chutes, grapples, etc.)

I i

! *8ecause of the interrelation of treatment and handling of transuranic (TRU) l wastes and high activity wastes with the low level wastes, requirements for these radioactive wastes are discussed in Section 3.2.15. Other wastes (e.g.,

non-fuel-bearing components) are discussed in this section.

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decontamination equipment (storage tanks, pumps, piping, etc.)

viewing equipment (windows, closed circuit television)

Storage Systems in process lag storage pits (incoming and discharge) canyon cell lag storage vaults storage units (casks and drywells) 1 To meet 10 CFR 72.75(a), these systems must be designed with capabilities for (1) testing and monitoring, (2) shielding, (3) confinement, (4) heat removal, and (5) minimizing radwaste quantities. Design features of the MRS responding to these requirements include l

(1) testing and monitoring systems testing equipment (ultrasonic testing, helium testing) sampling systems radiation monitoring equipment instrumentation and control equipment (2) shielding systems shielded cells transfer shields stortge unit shielding (3) confinement barriers t

confinement structures l -

ventilation systems waste canisterization canister overpacking liners 1

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(4) heat-removal systems forced cooling equipment I passive cooling system ,

(5) systems for minimizing radioactive waste volumes l liquid waste evaporation equipment shredding equipment compaction equipment recycling systems The description below* covers those systems that have not been addressed in other sections of this report.

3.2.14.2.2 Handling Systems The preparation for storage of spent fuel, solidified HLW and HAW, and RHTRU would take place in the R&H building. The preparation would differ with the type of radioactive materials being received in terms of processing needs, packaging mode, and handling and storage requirements.

)

Figure 3.6 is a simplified schematic drawing

• of the relevant handling and pro-cessing areas of the R&H building. In the following, description, reference is made to this schematic drawing and to letters and numerals that indicate various i operating areas in the building. Figure 3.7 is an isometric rendering of the R&H building. ,

The R&H building contains four processing cells (1, 2, 3, and 4), each having identical equipment and layout. There are also two shielded canyons (5 and 6)

. with nearly identical equipment and layout; two nearly identical lag storage areas and two storage cask discharge areas, each associated with a shielded canyon; and one decontamination cell for the non-fuel-bearing ccmponents drum introduction and decontamination associated with each pair of processing cells.

• Based on Parsons, MRS-11, September 1985, Vols I and III. N 4

3-76

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Pass-throughs between the two shielded canyons permit exchange of canisters, if needed, for welding, decontamination, etc. , or for accessing the two lag storages interchangeably.

The equipment in the process cells can be set up with modules and tooling for processing either PWR or BWR spent fuel assemblies, as needed.

3.2.14.2.2.1 Spent Fuel Spent fuel would be received in the R&H building in shielded rail or truck shipping casks (A). Each cask would move onto a cart after being off-loaded from a truck or rail car. The cask and cart would pass through a handling /

decontamination room into the unloading room (B). A contamination control adaptor, in the form of a circular disc, would be placed on top of the cask in the handling / decontamination room. In the unloading room, the shipping cask would be positioned below an entry port that leads into the main process cell.

The cask would be positioned directly beneath the entry port, and a telescoping device would be lowered into contact with the contamination control adaptor ring. This would form a barrier to transport of radioactive contamination between the shipping cask unloading room and the main process cell above. Nor-mal air flow from the ventilation system would be from the loading room into the process cell. Heavy shield plugs could then be removed from the floor of the main process cell, exposing the top of the cask. The cask would be opened and its contents li)'ted out and placed into the incoming lag storage areas (pits built into the floor of the main process cell (C)). Ir. side the incoming lag storage areas in the process cells, the spent fuel would be passively cooled by natural convection. Although the ventilation system would supply air to the process cells, it would not be relied on for forced cooling of the spent fuel.

To clean the exterior of the shipping cask and the contamination control adap-tor and release the empty cask from the facility, the steps above would be per-formed in reverse order.

Depending upon the rate of handling of spent fuel within the process cell and the amount of fuel in lag storage, the spent fuel assemblies could remain in incoming lag storage for a few days or weeks before they were removed for further handling and packaging.

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L l

4 .

The spent fuel assemblies would be disassembled (D) and the fuel rods reconfig-ured into a close array (" rod consolidation"). Disassembly would be performed with the fuel in both vertical and horizontal positions. Figure 9 of Design l Study 9 in the Conceptual Design Report (Parsons, MRS-11, September 1985, Vol IV) shows the two major fuel consolidation process subsystems: the disassembly subsystem and the fuel rod extraction and reconfiguring equipment.

The assemblies would be placed vertically in a downender. A laser system would cut through the support pieces at the upper end of the assembly to allow access to the fuel rods and to permit the rods to be extracted free of the spacer grids.

The assembly would then be lowered to a horizontal position by the downender and the end pieces removed by the laser cutter for those assemblies requiring this additional operation.

Fuel rods would be extracted by a multiple rod " gripper" that would simul-taneously extract fuel rods from either three PWR or seven BWR assemblies.

The gripper carriage would pull the fuel rods out of their spacer grids struc-ture and translate them over the adjacent reconfiguration structure; the dis-lodged crud would be collected by an air sweep in a receptacle below the structure. During this movement, the fuel rods would be supported along their length by horizontal and vertical " combs." Once the fuel rods were in position over the semicircular reconfiguration slot, the horizontal combs would be ro- j tated out of the way, allowing the fuel rods to be released onto a lowering, l spring-tensioned " strap." After the fuel rods were lowered into the recon-figuration slot, an upper die would then be lowered to form the array into a circular cross-section.

Next, the reconfigured fuel rods would be pushed by the " push mechanism," mounted on the rear of the gripper carriage, into an empty consolidation canister that abuts the shielded canyon wall adjoining the process cell through the access port (E) equipped with a contamination barrier. The loaded canister would then be transferred via an electrically powered cart to the canister welding station. l The welding station (F) consists of three major pieces of equipment: the resis-tance welding system, the canister lid transfer and positioning system, and the 4

canister air-lock tube. The canister would be inserted in the air-lock tube, which also serves as the helium test station and decontamination station. The air in the air-lock tube would be evacuated and replaced with an argon-helium 3-80

mixture; hence the canister interior also would be inerted. Then the canister would be inserted into the welding station, the lid welded, and the canister retracted into the air-lock tube. The argon-helium mixture would be evacuated and the canister sprayed and flooded with freon for decontamination. After decontamination, the air-lock would be evacuated and a helium leak test would be performed. If no leaks were detected, the canister would then be removed from the air-lock tube and, after swiping, would be moved to the ultrasonic test station located next to the welder for the final weld test. If the canis-ter passed this test, it would be ready for transfer either to the lag storage vault (H) or to the storage facility (J).

Should the canister weld be unacceptable, the canister lid would be cut from the canister with an automatic pipe lathe at the cutting station (G). After that, the canister would be re-welded and re-tested at the welding station.

Some fuel assemblies that might be defective or could not undergo rod consolida-tion would be placed in canisters intact. Those fuel rods which may be damaged during fuel disassembly or may not be separated from the spacer grids, will be packaged in separate canisters.

The consolidated spent fuel canisters ready to be accepted at the storage facility--either removed from lag storage or directly from the last packaging step--would be loaded (I) into the storage cask (or transporter for drywells),

which would be transported to the MRS storage facility (J).

The spent fuel storage in casks or drywells at the storage facility is discussed in Section 3.2.8.

3.2.14.2.2.2 HLW and RHTRU Processing Canisters of solidified HLW would be received and processed at the east end (I) of the shielded canyons 5 and 6. The onsite generated RHTRU drums also would be processed in this area. The shipping cask unloading would follow a procedure similar to that described for spent fuel unloading in Section 3.2.14.2.2.1 above.

This area would be equipped for inspection of received canisters and of onsite RHTRU drums for their decontamination, as necessary.

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After the above preparation, the HLW canisters and the RHTRU drums would either be placed in discharge lag storage (L) (distinct from canyon storage for spent fuel) in the shielded canyon, or placed into a storage cask (transfer shield for drywells) for loadout and discharge (I) to the storage facility (J).

3.2.14.2.2.3 Non-Fuel-Bearing Components Non-fuel-bearing components (nozzles or end fittings, grids and spacers) resulting from the fuel disassembly operations would be fed to a shredder (a) located in each of the process cells. A clean 55 gallon drum would be brought to the process cell, on a cart from the clean drum storage (b) via the decon-tamination cell (c) and contamination barrier (d). The drum would be located under the shredder to collect the shredded material; the filled drum would be returned to the non-fuel-bearing decontamination cell (c). When five filled drums accumulated on the pallet in the non-fuel-bearing components decontamina-tion cell (c), they would be individually decontaminated. The pallet with clean drums would be moved to the shielded canyon via a shielded door (e) and trans-ferred to discharge lag storage. When enough drums would accumulate in lag storagetofillastkragecask,theywouldbetransferredtothestoragefacil- ,

ity (J); for drywelli storage, a group of five drums would be aligible for trans-fer to the storage facility.

3.2.14.2.2.4 Overpacking Operations If a waste package (spent fuel, HLW, or RHTRU) were to be overpacked at the MRS it would be performed in shielded canyons 5 and 6 at location K. The welding and subsequent ultrasonic testing of the overpacks would be performed in shielded

, canyon 6 only; overpacks generated in shielded canyon 5 would be transferred ftocanyou6fortheseoperations.

I Before waste packages (spent fuel, HLW, and RHTRU canisters and drums overpacked as above or not) would be shipped from the MRS to the geologic repository, they might be placed in a repository overpack (RO). The RO would be received, loaded, and decontaminated at the east end (K) of shielded canyons 5 and 6.

3-82 f m ._

If overpacking were to be undertaken at the MRS, the R0 would arrive at the MRS in a shipping cask and would be unloaded into the shielded canyon by a procedure similar to that for spent fuel unloading in the process cell (de-scribed in Section 3.2.14.2.2.1 above). The R0 would be placed in a welding /

decontamination cell that accepts waste packages from lag storage. The RO would be designed to hold four 12.75-inch-diameter canisters or two 12.75-inch-diameter canisters and one 14-inch-diameter canister.

The R0 lid would be attached by a jib-crane-mounted electron beam welder that would rotate around the weld. After welding, the R0 would be decontaminated I with freon and dried with air. The decontamination pit first would be covered with a shell to prevent splashing and then the pit's spray nozzles would be activated.

3.2.14.2.3 Storage Systems The storage systems in the R&H building would consist of in process lag storage pits and canyon cell lag storage vaults.

The incoming in process storage pits located in process cells 1, 2, 3, and 4 would provide a steady flow of fuel assemblies for processing to accommodate shipping cask unloading operations. The fuel assemblies would be stored in the pits in vertical racks; the racks would be interchangeable for accommodating PWR and BWR assemblies according to cell operation needs. The fuel assemblies would be cooled by natural convection to the cell atmosphere.

The discharge in process lag storage pits located at the east end of shielded canyons 5 and 6 would provide a buffer for the receipt of HLW canisters and for the overpacked canisters as necessary, before their transfer to the storage facility. The canisters would be stored in the pits in vertical racks; the racks would be fixed for a set combination of canister sizes. The canisters would be cooled by natural convection to the shielded canyon atmosphere.

The lag storage vaults and storage units (casks and drywells) are described in Section 3.2.8.

l 3-83

3.2.14.2.4 Testing and Monitoring

• Canister Testing Canisters containing spent fuel rods or intact spent fuel assemblies would be tested before their transfer to the lag storage vaults or to the storage units.

Testing operations would include helium and ultrasonic methods, as described in Section 3.2.14.2.2.1.

Storage Facility Monitoring Fixed-filter air samplers would be installed in the alarm zone around the pe-rimeter of the storage facility, and a selected number of radiation detectors would be provided for personnel protection. They would be installed among groups of casks or drywells, and the signals from these detectors would be transmitted to the redundant radiation monitoring system terminals in the con-trol rooms of the R&H building and site services buildings.

Storage Units Monitoring A statistically significant fraction of storage casks or drywells would be continuously and remotely monitored for temperature. Thermocouples would be installed in the low, intermediate, and upper levels of the casks or the inter-mediate point of the drywells. The thermocouples would be hard wired to multi-plexers centrally located to serve groups of casks (165) or drywells (500).

The signals would be digitalized and transmitted by shielded cable to the dis-tributed control system (DCS). Temperature data in the UCS would be logged and stored for trending.

Gas samples to determine the presence of helium and Kr-85 would be obtained from quick-disconnect sample ports at the bottom of the casks or in the top of dry-wells. The samples would be analyzed in the R&H building analyzer.

• Described in Parsons, MRS-11, September 1985, Vols I and II.

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3.2.14.2.5 Shielding and Radiation Protection Shielding is addressed in Section 3.2.13, above.

3.2.14.2.6 Confinement Structures and Systems Confinement structures and systems are addressed in Section 3.2.8, above.

3.2.14.2.7 Heat-Removal Capability Heat-removal capability is addressed in Section 3.2.8, above.

3.2.14.2.8 Minimization of Radioactive Wastes Section 3.2.15, below, describes how volume of radioactive wastes would be minimized. In addition, the use of all-dry systems for lag storage, spent fuel consolidation, and packaging, as described in Section 3.2.15, would generate less radioactive waste than would the use of wet systems.

3.2.14.3 Discussion 3.2.14.3.1 General In the following discussion the two major categories--handling systems and storage systems--are evaluated first. Then features needed to meet requirements (1) through (5) of 10 CFR 72.75(a) (listed above) are evaluated.

Because some of these systems have been addressed elsewhere in this report as part of other 10 CFR 72 requirements, reference is made to those sections as applicable. This section discusses only those systems or aspects of systems that have not been covered in other sections. The discussion covers both normal l and accident conditions.

3.2.14.3.2 Handling Systems The large majority of handling systems to be used in the MRS would be concen-trated in the R&H building, which would be the site of re-configuring spent 3-85

fuel for storage, inspection and overpacking received HLW canisters and waste drums, possible overpacking for disposal in a geologic repository, loading the storage casks, and the resulting radioactive waste treatment. The handling systems for the storage facility would be limited to placing waste packages in the storage units (drywells) or to locating the storage units (casks) in the storage field.

Evaluation of the proposed handling systems and equipment indicates that, al-though numerous operations would take place in the R&H building, the large majority of this equipment is commonly used in nuclear and non-nuclear applica-tions with few, if any, problems of general public health and safety concern.

For example, a review of fuel handling during significant operational occur-rences (NRC, February 1985) indicates that radiological consequences were minimal in all cases.

Other fuel-handling system components proposed for the MRS are not in common use. These items would be in the fuel rod consolidation system and include the fuel cutting device and controls and the dry rod extraction and reconfiguring equipment.

There is limited experience in disassembling irradiated nuclear fuel and recon-figuring the fuel rods for storage. The primary example of such experience to date is the 1982 Duke Power Company / Westinghouse Electric Corporation underwater demonstration (Duke Power, 1982) for consolidation of four PWR irradiated fuel assemblies from the reactors at Oconee Units 1 and 2 in their shared spent fuel pool. The demonstration achieved its objectives, including remote operation of portable fuel disassembly and consolidation equipment, to safely extract all l fuel rods in a fuel assembly simultaneously and consolidate _them in a rectangu-lar canister containing the rods from two assemblies. Remote compaction of the fuel assembly skeletons (the non-fuel-bearing components) was also demonstrated, although the compaction ratio was lower than expected and the measured dose rates from the compacted canisters were higher than expected as a result of reactor-induced activation of cobalt impurities in the non-fuel-bearing compo-nents. Another problem experienced was crud release and dispersion in the pool water during the disassembly operations, slowing operations until visual obser-vation capability was restored.

3-86

l The applicability of this demonstration to such operations at the MRS is limited '

because it was done underwater (not dry), vertically (not horizontally), and ,

with mechanical cutters (not a laser system). Therefore, the staff recommends l that DOE conduct a comprehensive test program using equipment and spent nuclear fuel prototypical (including Zircaloy parts) of that planned for the MRS. This test should demonstrate cutting of fuel assemblies with the laser system, con-trol of the laser remotely using computer-controlled instrumentation, removal of the fuel rods, reconfiguring the rods into circular arrays, and overall sys-tem reliability for different fuel assembly types. The welding system also should be demonstrated using canisters and welding equipment prototypical of those expected to be used at the MRS. Such demonstrations could provide infor-mation that would be useful for safety evaluations, including the types and amounts of radioactive contamination generated the amount of Zircaloy fines produced from cutting and rcd removal and their size distribution and oxidation / combustion characteristics (as discussed in Section 3.2.3 of this report) the effects of Zircaloy fines on the shredding operations of non-fuel-bearing components the extent of fuel rod deformation or damage the quality of the welds on canisters the reliability of the laser control system the overall safe performance of the integrated system These demonstrations should be completed in a timely manner so they could be in-corporated in design decisions on the MRS fuel-handling systems, if necessary.

i These demonstration programs would still leave uncertainties as to the viability l l

of disassembly for all of the types and conditions (irradiation history, age, handling and storage history, etc.) of fuel to be received at the MRS, espe-cially considering clad embrittlement behavior of fuel rods upon irradiation 1

, 3-87 l l

l

(NUREG/CR-1729). However, confidence in this regard is provided by the large

. number of irradiated fuel rods that have been routinely and satisfactorily re-moved from fuel assemblies as part of fuel reconstitution, non-destructive test-ing, and repair operations.*

Additional confidence is provided by DOE's plan to place in canisters intact any fuel assemblies suspected of posing a problem for consolidation, which would help ensure that excessive fuel failures during consolidation operations would be prevented. This would be especially true if data from utilities on the failure of fuel in the reactor are considered in the decision to consolidate or not. According to Johnson and Gilbert (1983), the incidence of defective rods is estimated to be less than 0.01%; theretore, a decision not to consolidate any fuel assemblies with previously known defective rods should no+ have a major impact on MRS throughputs or storage space.

The DOE proposed design would place non-fuel bearing components into 55 gallon drums. The use of 55 gallon drums might not be compatible with current repusitory waste package designs if the non-fuel bearing components are disposed of in a geologic repository.

3.2.14.3.3 Storage Systems The in process lag storage of spent fuel and HLW canisters and of RHTRU drums in the R&H building would take place for relatively short periods of time.

Although DOE has assumed that passive cooling by natural convection would be adequate for in process lag storage, this must be further assessed.

The lag storage vaults in shielded canyons 5 and 6 would store waste canister 3 before their transfer to the storage facility, or possibly after their retrieval for shipment off site to a geologic repository. The lag storage vault layout, storage mode, and cooling system, as well as the potential accident of loss of the cooling system, are discussed in Section 3.2.8 above.

More than 51,000 in the U.S. alone, according to Bailey and Johnson, 1984.

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3.2.14.3.4 Testing and Monitoring The leak testing of canisters containing spent fuel rods or assemblies, as de-scribed in Section 3.2.14.2.4, appears generally adequate, if DOE can demon-strate that it would be sufficiently sensitive. The continuous temperature monitoring of storage units (casks and drywells) appears adequate.

During operation, gas would be sampled from a statistically significant number l of casks or drywells. The questions arise as to how many units would be sampled and how they would be chosen. Furthermore, the grab sample method seems to be -

inadequate for detecting possible drywell liner breach. As currently planned, such sampling would indicate a canister breach only; at the time of a canister breach, an undetected liner breach could already exist, and thus both barriers would already be lost.

3.2.14.3.5 Decontamination DOE has proposed to use a freon decontamination method for spent fuel canisters.

This method, however, will generate a waste sludge which is listed by EPA as a hazardous chemical waste (40 CFR 261.31).

3.2.14.4 Conclusions and Recommendations  ;

l l

The large majority of the equipment and systems used in spent fuel handling is l commonly used in nuclear and non-nuclear applications. This experience and other l

spent fuel consolidation and reconstitution experience provides confidence that MRS handling operations could be conducted safely.

The staff recommends, however, that 00E conduct a comprehensive test program for the equipment intended to be used for dry, irradiated nuclear fuel disassem-bly and reconfiguration, and that DOE emphasize operation of the laser system, control of the laser remotely (using computer-controlled instrumentation),

reliability of operation, removal of the fuel rods, and reconfiguring the rods in circular arrays. Other types of information needed for the evaluation of an integrated system include the amount and size of Zircaloy fines produced from cutting and their oxidation rate, and the effect of the shredding operations of the non-fuel-bearing components on the formation of Zircaloy particles.

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DOE should consider the efficacy of maintaining an inert gas atmosphere in a l l

seal welded container for proposed storage periods. 1 DOE also should evaluate the concept of the gas sampling system for drywells, in terms of its effectiveness to detect drywell liner failure. The range of poten-tial accidents should be extended to analyze effects of vaporization of fuel pellets from an array of fuel rods, and the potential for canistered fuel rods overheating during packaging operations. DOE plans to undertake all of these tests.

The staff recommends that DOE evaluate alternative decontamination methods which do not utilize freon or result in a waste which would be designated as a hazardous chemical waste.

The staff recommends that proposed non-fuel bearing component package designs be evaluated to ensure that the designs are compatible with repository waste packages.

Currently, the repository programs have not selected repository overpack fabrication and closure methods. The electron beam welding process and weld inspection techniques should be evaluated to ensure they are compatible with repository waste package designs.

3.2.15 Waste Treatment 3.2.15.1 Requirements 10 CFR 72.75(b) gives the requirements for waste treatment as follows:

Radioactive waste treatment facilities shall be provided. Provisions shall be made for the packaging of site generated low level wastes in l a form suitable for transfer to disposal sites.

3.2.15.2 Description Althcugh 10 CFR 72.75(b) addresses the low level wastes (LLW) and Sec- j tion 72.75(a) covers other types of wastes, for clarity, the evaluation below covers not only LLW, but CHTRU and HAW as well. The reasons for combining 1

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i l

I this discussion are (1) the similarity of some of the chemical processing and (2) the joining of various waste streams in process, depending on their result-ing radioactivity levels.

1 MRS operations would generate low level wastes (LLW), transuranic wastes (TRU), l and high activity wastes (HAW). The generation of these wastes would mainly be confined to the R&H building where consolidation and packaging of spent nuclear

fuel would take place. The spent fuel, HLW, and RHTRU storage facility (casks or drywells) would be a passive, non-operations facility and would not generate any radioactive wastes. Although the CHTRU building would normally not be  !

expected to generate radioactive wastes, it could require cleanup in case of an off-normal event (e.g. , leaking drum, etc. ).  !

i I

The MRS would generate three types of waste forms: liquid, solid, and gaseous.

This section will address liquid and solid wastes only; gaseous wastes are ad-dressed in Sections 3.2.8.2.3 and 3.2.8.3.4. The description of the liquid and solid radwaste treatment is based on material in Parsons, MRS-11, September 1985.

3.2.15.2.1 Liquid Radwaste The liquid radwaste generated in the MRS would consist primarily of the decon-tamination solutions collected in drains and sumps. In addition, liquid .

radwastes would be generated in the analytical laboratory and in the laundry I operations. Figure 3.8 is a simplified block diagram that illustrates the sources of the two liquid waste streams, LLW and HAW, and the processing for radioactive and other particulates removal.

The LLW and HAW processing would be essentially similar. Processing would start with collection of the respective wastes in dedicated tanks, followed by treat-ment in thin-film evaporators for volume reduction by separating the suspended solids. The vapors generated by the evaporators (mostly water) would be con-densed in overhead condensers, and the resulting liquid would be collected in dedicated condensate tanks. This liquid would be further cleaned and deionized by traveling through the resin beds of dedicated ion exchangers. The resulting clean water would then be monitored and, if it were not contaminated, it would be returned to the deionized water recycle for use in decontamination solutions, 3-91

SOURCE MATERIAL TREATMENT LLW

? Condenser C Cond.

Tank Radweste Drains Collection Sumps 1 f

- LLW Laundry Drains  ? Coll.  ? Evaporator C Tank --

10 HAW Solidif. Exc anger Analysis Lab. m Syst.

1 f To LLW Solidif. Systs.

(or to HAW Solidif. Syst. G 88""Y Tank for HAW Syst. Below) l I Rad. Check Water To Users . ,

Storage ta (Decon. Syst.. etc.) , For Recycle g'

LLW LIQUID RADWASTE' HAW ( System Similar In-Cell Decon. Drains C Coll. ?J to LLW as Above Except Noted.

Tank l

4 HAW LIQUID RADWASTE Figure 3.8 Liquid radwaste streams and processing Source: Parsons, October 1985

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etc. The spent resins from both LLW and HAW ion exchangers would be sent to the HAW solidification system.

The other stream leaving the evaporators would contain the non-volatile compo-nents of the wastes. They would be discharged from the bottom of the evapora-  :

tors in the form of slurry. The slurry from LLW liquid radwaste would be col-lected in a dedicated slurry tank and later transferred to the LLW solidifica-tion system; the slurry from the HAW would be collected in another dedi-cated slurry tank and later transferred to the HAW solidification system.

The LLW and CHTRU waste treatment would be located in a dedicated area of the R&H building. Operations and maintenance would be done through contact. The HAW waste treatment area also would be in the R&H building; however, it would be shielded and remotely operated, with most maintenance done remotely.

3.2.15.2.2 Solid Radwaste The solid radwastes that would be generated in the MRS can be divided into four broad categories combustibles non-combustibles sludges and slurries HEPA filters These wastes--which could be LLW, CHTRU, or HAW--would be treated, according to their form and composition, in dedicated areas of the R&H building. Figure 3.9 shows the source materials, the treatment provided for waste solidification  !

and/or packaging, and the intended storage / disposal. In the figure, the source materials are keyed A through H.

Combustibles (shown as A)--which include latex, vinyl, paper, etc.--would be

! collected in plastic bags throughout the R&H building. At the LLW solidifica-tion station, the materials would be placed in a cage in the center of a 55 gallon drum. Grout (cement and sand) would be added to the drum and, when the grout sets, it would encase the cage containing the waste. Drums containing combustible materials would be limited to 25% combustibles per volume (PNL, 1985).

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SOURCE MATERIAL TREATMENT STORAGE / DISPOSAL C C Grout (Coment b Sandl Solidificerqui 6n Drum LLW to 4* fambustibles (Lates. Vinyt. Peper. etc I l 4 OffDesposef Sete C*s, M M Drumel B LLW Eveo . sonoms <

Interrogator numme g Freon Studge. Laundev Fitters. Spent Rosens. (erl Uo

$sudge fr. $wmps tintiermettents j g _

4 CMTRU Structure g* Non Combustibles (Contam.. Teoss. Metal Pieces. etc I (56 Gat. Drumst L Pockeging Compactingi d '

E. "" 'd'ers (tuent en. Cow and F.est St. e -

Packaging iest.bi i 4

LLW AND CHTRU SOLID RADWASTE F,.m.

Shearing C y an Cell HEPA Fi: tees i

O A F.

First Stage Testab4e Hf PA FA g,,,.

Fitt.er/

g, m Media _

Compaction -

HAW HAW to le 200 mR/hrt On-Site CA5K or G'***C******b'** -

De CHTRU i >10 mrthe) I hw C -

temmation HAW (10 200 mR/hri cro.: 4C.m .r n, ..ck.g g l M AW twop. Bottoms. F6fters Backwesh. _

HAW tr LLW Room $surry HAW SOLID RADWASTE Figure 3.9 Solid radwaste sources and processing i Source: Information provided by DOE (Parsons, MRS-11, September 1985, Vol III)

Evaporator bottoms from the LLW thin-film evaporator (B) would be added to the grout used for combustibles.

If freon sludge, spent resins, etc. (shown as C), were of low level activity, they would be intermittently added to the grout used for the combustible. If these wastes were found to be of the HAW type, they would be sent for solidifi-cation in HAW stream H.

Non-combustibles (D)--consisting of contaminated tools, metal pieces, etc.--

would be packaged in 55 gallon drums without further treatment.

HEPA filters (E)--except the in-cell non-testable and the first stage testable filters (see Figure 3.4)--also would be compacted and packaged in 55 gallon drums.

All drums used in treating and packaging LLW and CHTRU wastes would pass through a drum interrogator that would determine the presence of TRU material by gamma pulse height analysis. Drums containing TRU material would be sent to the on-site CHTRU structure for storage; drums containing only LLW would be shipped off the site for disposal.

In-cell HEPA filters and first-stage testable HEPA filters (F) would be removed by remote handling and transferred to the HAW solid radwaste area. This treat-1 ment would consist of (1) separation of filter media from the metal frames,  !

(2) frame shearing and packaging in 55 gallon drums (six frames per drum), and (3) media compaction and packaging in 55 gallon drums (six filters per drum). l The gaseous effluent from the HAW filters compactor would first be filtered ,

1 through a HEPA filter and then vented through the R&H building exhaust. '

Parsons (MRS-11, September 1985, Vol I) states that 4000 HAW HEPA filters per year would be generated. The total number of drums of consolidated frames and filter media would be 1350 per year.

l l

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Non-combustibles (G) that would result from operations involving high activity levels would be transferred to the remote maintenance cell for decontamination.

Wastes decontaminated to below 10 mR/hr would be sent for packaging to the LLW/CHTRU radwaste area; those with 200 mR/hr or more would be remotely trans-ferred to the HAW for packaging together with the HEPA filter media; and those with between 10 and 200 mR/hr would be remotely transferred and packaged in the HAW radwaste area.

HAW evaporator bottoms, resin slurry, etc. (H) would be solidified in 55 gallon drums by mixing with grout (cement and sand).

All drums resulting from the treatment and packaging of HAW would be stored on the site in the spent fuel, HLW, and RHTRU storage facility in dedicated casks or drywells. In the cask storage concept, 45 drums could be stored in one cask.

In the drywell storage concept, five drums could be stored in each drywell.

3.2.15.3 Discussion The staff finds the design provisions for treatment of waste generated during operations at the MRS adequate in general. The design provisions address the following relevant issues related to safe collection, treatment, storage and disposal of radioactive wastes:

1 (1) collecting all liquid radwaste streams in dedicated LLW and HAW tanks.

(2) reducing the volume of liquid radwaste by evaporation and recycle of water after it has been decontaminated by ion exchange and monitored

for radioactivity. The pretreatment of the feed solutions to prevent foaming in the evaporator should be studied, particularly, if organic decontaminating agents would be used.

(3) collecting, treating, and packaging solid radwaste in dedicated facilities for LLW/CHTRU and HAW radwastes. j l

(4) remote handling of all HAW materials from their source to treatment and storage.

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(5) reducing the volume of low level waste HEPA filters through compaction, and reducing the volume of high-activity HEPA filters through frame shear-ing and compaction and media compaction.

(6) solidifying semi-solids (slurries, evaporator bottoms, etc.) in drums, with grout.

(7) solidifying combustible materials in drums with grout. -

(8) limiting combustibles to 25% by volume in drums containing combustibles.

(9) using a drum interrogator to detect presence of TRU in solid radwaste drums for determining either onsite storage of TRO wastes or offsite disposal of non-TRU wastes.

Although these provisions appear adequate, DOE must address additional issues in tt.e final design.

Depending on the results of analyses and demonstrations recommended in Sec-tion 3.2.8, DOE must address the treatment, packaging, and storage / disposal of unoxidized Zircaloy fines resulting from spent fuel consolidation operations.

Also, the potential accummulation of Zircaloy fines or dust on the in-cell non-testable HEPA filters and its effects on treatment and storage must be addressed. If unoxidized Zircaloy fines were captured by the HEPA filters, they could pose the danger of fire or explosion during filter handling, particu-larly during media compaction.

3.2.15.4 Conclusions and Recommendations The waste treatment provisions included in the MRS design appear adequate in general. The staff recommends, however, that DOE address the unoxidized Zircaloy fines in terms of hazards during waste treatment and storage (in con-junction with the recommendations made in Section 3.2.14 above).

The low-level waste program should consider the waste form and waste classification requirements in 10 CFR Part 61. In addition, the assay methods 3-97

for detecting radionuclides should be qualified to obtain the necessary sensitivity considering the actual waste properties and the expected high beta gamma background.

3.2.16 Deconnissioning 3.2.16.1 Requirements 10 CFR 72.76 gives the requirements for decommissioning the MRS as follows:

The ISFSI [MRS] shall be designed for decommissioning. Provisions shall be made to facilitate decontamination of structures and equipment, mini-mize the quantity of radioactive wastes and contaminated equipment, and facilitate the removal of radioactive wastes and contaminated materials at the time the ISFSI [MRS] is permanently decommissioned.

i 3.2.16.2 Description and Discussion The R&H building would De the only likely location in the MRS requiring signi-ticant decontamination. The four process cells in the R&H building would normally t e contaminated as would, to a lesser extent, the filter rooms and non-fuel-bearing waste treatment locations. The R&H building would be designed to limit most of the contamination to those four cells. Some contamination would be expected in incoming casks and the area and solution transfer lines would require infrequent decontamination. Other areas, such as the normally clean canyon cell, would be decontaminated on an infrequent basis, as dictated by circumstance. DOE plans to maintain the canyon cell as a " clean" (non-

contaminated) area insofar as it is practical to do so. The staff endorses that philosophy. The storage areas would not be likely to be contaminated because of

the preventative measures that would be taken.

The staff has not evaluated the difficulty that would be involved in a final decontamination and decommissioning effort for the four process cells. Because of the dry operations involved, the task would probably be less dif ficult than decontaminating and decommissioning cells in reprocessing plants. If DOE sub-mits a license application, the staff will review this issue in detail.

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l 3.2.16.3 Conclusion On the basis of the above discussion, the staff concludes that, even at this early stage of uesign, it appears that the MRS would be designed to minimize the fu al decontamination and decommissioning effort.

3..~ O cident Analysis As part cf its safety review of the receipt, handling, and storage of spent nuclear ftel and solidified HLW at the MRS, the staff has considered design-basis accidents (DBAs) that would have a potential impact on public health and safety. A DBA study provides insights concerning the impact of worst-case accidents that can occur, based on present conceptual designs. Impacts may be acceptable or may require mitigating design features.

For the purpose of evaluating the acceptability of the radiological effect of any postulated DBA, the staff compared the calculated DBA doses with the appro-priate requirements of 10 CFR 72.

3.3.1 Requirements 10 CFR 72.68(b) gives the requirements for the public health and safety of a j spent fuel handling facility similar to the MRS as follows: l Any individual located on or beyond the nearest boundary of the con-trolled area shall not receive a dose greater than 5 rem to the whole '

i body or any orgaa from any design basis accident. The minimum dis-tance from the ft:cilities to the nearest boundary of the controlled area shall be at least 100 meters.

I In addition, 10 CFR 72.74(d) states: ,

The ISFSI [MRS) shall be designed to... control the release of radio-active materials under accident conditions.... Analyses of design basis accidents shall be made to show that releases to the general environment will be within the exposure limits given in S 72.68....

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3.3.2 DBA Description Most of the radioactive material handled at the MRS would be PWR or BWR spent fuel that had been removed from a reactor and stored for 5 or more years. The fission product of primary concern under conditions of long-term spent fuel storage is krypton-85 (an inert neble gas present in fuel rod and void spaces).

For spent fuel to be received at the MRS, PWR reactor fuel exposed to 55,000 MWD /

MTU has the highest inventory of Kr-85, about 5000 curies per tonne.

l Spent fuel at the MRS would be received, unloaded, consolidated, and placed in

} canisters by mechanical operations. A major potential accident associated with the mechanical handling of spent fuel would be damaged fuel resulting from the drop of an assembly, cask, or canister drop. The accident scenario with the most severe ramification would be an open canister drop during the fuel consolidation operation that results in the damage of 100% of the fuel elements of three assemblies.

Evidence to date indicates that Kr-85 leaks rather quickly from damaged fuel elements; it would not be retained by filters. Therefore, an MRS accident analysis must include an analysis of the fuel assembly drop DBA and its poten-tial impact on public health and safety. The impact can be evaluated by calcu-lating off-site whole-body doses that result from this type accident. This dose calculation would entail (1) determining the amount of Kr-85 released as a

result of the accident; (2) considering the atmospheric conditions that exist and transport the released Kr-85; and (3) evaluating the individual whole-body i

dose obtained from Kr-85 exposure.

1 3.3.3 Description of Accident Analysis '

i An accident analysis was performed using conceptual level design information.

} Credible events that might result in an off normal exposure to radioactivity by i

operating personnel and/or the general public were identified (Parsons, MRS-11, September 1985, Vol VI). Examples of these events include: (1) events that

may occur during an operating year (e.g. , pipe leak); (2) events that may occur during the design lifetime (e.g., fuel assembly drop); and (3) events that may

! result in the maximum potential impact on the immediate environment (e.g.,

l 3-100 l

impact of severe natural phenomena). The NRC staff reviewed the events identi-fled (Parsons, MRS-11, September 1985 Vol VI). The description included detec-tion methods, mitigating design features, consequences, and post-event actions.

Also included are offsite maximum integrated individual dose calculations for DBAs. The NRC staff has reviewed these dose calculations and has independently identified and analyzed a severe case accident. The NRC staff assessed this accident's potential radiological impact on public health and safety.

3.3.4 Discussion As rtated above, the staff has performed an offsite dose calculation of the accident that had the potential for a severe impact. The type of accident considered was an open canister drop that results in the breach of 100% of the fuel rods in three fuel assemblies. In all, 30% of the krypton-85,10% of the iodine-129, and 10% of the tritium was assumed to be released as a result of this accident. Breached fuel assemblies were assumed to be from a 10 year stored PWR assembly with a burnup of 55,000 MWD /MTU. Windspeeds were assumed to be 1 meter per second, and Pasquill's F stability (moderately stable) conditions j were assumed. The point of exposure used for dose calculations was assumed to be at the controlled area boundary nearest the facility's stack (approximately )

420 m, which is estimated from Drawing H-3-56726 (Parsons, MRS-11, September 1985, l Vol III) for the Clinch River site). Dose calculations using these assumptions were performed using (1) Equation 2 of RG 1.145 to obtain values for X/Q; (2) immersion dose conversion factors were obtained from NUREG/CR-1918; and (3) internal dose equivalent conversion factors were obtained from NUREG/CR-0150, Vol 2, and the radioactive source (Ci per tonne) for 10 year cooled 55,000 MWD /

MTil fuel was obtained from Table 3.3 of the MRS Environmental Assessment (DOE /

RW 0035, Vol II). These dose calculations indicate that the maximum integrated dose received by an individual at the controlled area boundary from an upper limit accident is approximately 3 mrems. This includes whole-body dose contri-butions from krypton-85 and tritium exposure, which are 1 mrem and 2 mrems, i respectively. Thyroid dose (due to I-129 releases) and skin dose were also calculated; they are 7 mrems and 30 mrems, respectively.

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3.3.5 Conclusion On the basis of the above discussion, the type of accident having the potential for the greatest offsite dose is an open canister drop. The whole-body dose to an individual member of the public located at the controlled area boundary for this type of accident would be about 3 mrems. This is a small fraction of the limit of 5 rems accident exposure guideline in 10 CFR 72.68(b).

On the basis of the above, the accident potential of the conceptual design of the MRS would be limited because of the nature of the facility's operations, which include passive storage and relatively simple spent fuel handling opera- l tions. On the basis of the staff's review of MRS postulated accident events, NRC independent review of a potential upper limit accident, and the nature of MRS operations, the staff further concludes that any accident that could credi-bly occur at an MRS is not likely to have an offsite consequence more severe than that set forth in 10 CFR 72.68(b).  ;

1 f

3.4 Safeguards 3.4.1 Objectives and Review Approach Safeguards comprises the components of physical protection and material account-ability. The objective of the physical protection component is to protect

against sabotage of spent fuel or high level waste at the MRS or enroute to it.

In this context, the term " sabotage" refers to a deliberate, malevolent act that could result in the release of radioactivity in spent fuel or high level waste from its intended containment ard into the environment. The objective of the material accountability component is to protect against undetected theft or diversion of the special nuclear material in spent fuel or high level waste by maintaining vigilance cver the material; governing its internal movement and ,

1 location; monitoring its inventory status; assigning responsibility for it; l maintaining records of all transactions and movements; and issuing reports of j its status at the time of physical inventory. l The radioactive materials to be handled and stored and the structures, equip-ment, operations, and procedures to be used have been described in earlier i

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sections. The activities to be carried out at an MRS are comparable to activi-ties that have, in the past, been subject to the NRC safeguards licensing pro-cess. The safeguards review revealed no proposed activity that would require any new safeguards regulations. Accordingly, current safeguards requirements apply and serve as a guide for the safeguards evaluation of the DOE MRS propo-sal. Applicable NRC regulations address fixed site protection, transportation protection for spent fuel, safeguards contingency planning, and material accountability.

The approach used for this review lists all significant NRC safeguards require-ments that would apply in connection with an operating MRS facility and notes the DOE provision in each instance. Some detailed safeguards requirements have no place as review criteria during this, the conceptual design phase. Such requirements are identified for later consideration as the MRS design process progresses.

This NRC safeguards evaluation is based on information from Parsons, MRS-11, September 1985, Vols I and II.

3.4.2 MRS Site Protection NRC requirements for fixed site protection are found in 10 CFR 72, Subpart H.

They call for a physical protection plan detailing how 10 CFR 73.50 would be met. 10 CFR 73.50 requires a combination of structure, equipment, trained per-sonnel, management organization, and procedures. At this, the conceptual design phase of the MRS, safeguards structures, and to some extent safeguards equipment, must be considered, because these factors could influence MRS layout. Other factors can be deferred until a later stage in the design and licensing process.

A summary of the applicable fixed site requirements and the DOE provisions for each is set forth below.

l l

Physical security organization Requirements: The site must be protected by a security organization espe-cially established and trained for that purpose. The organization must include guards (who are armed) and may include watchmen who need not be

armed. A supervisor must be on duty at all times.

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- Response: Parsons (MRS-11, September 1985, Vols I and II) acknowledges the need for a trained security organization, including armed guards.

Barriers Requirements: The protection system must include barriers arranged so that access to credible targets for sabotage would entail passage through at least two barriers. The outer barrier must be illuminated and the space between the barriers must be monitored to protect against undetected intru-sion. No staff or visitor vehicles may be parked within the space enclosed by the outer barrier.

Response: Parsons (MRS-11, September 1985, Vols I and II) describes a suitable barrier system with illumination and intrusion alarms. Unautho-rized vehicles may not enter or park within the area enclosed by the bar-riers. Areas and activities with no radiological sabotage potential--such as administration offices, site services, and cask manufacturing--would be

~

properly located outside of the area, thereby avoiding unnecessary applica-tion of access constraints to those operations.

Access Requirements: Access to the MRS must be controlled on the basis of need for entry. Persons entering must be identified and their need for access confirmed. Depending on their purpose, some visitors will be treated as staff, while others must be escorted. Packages and vehic,les must be checked for sabotage materials and devices. Locks and keys used in support of the security system must be controlled.

N Response: Parsons (MRS-11, September 1985, Vols I and II) describes access points through the barrier system. The distinction between staff, escorted visitors, and unescorted visitors would be recognized. Appropriate pro-visions for entry of large transport vehicles and the Jarge storage casks (which are fabricated on site but not in the barrier area) without compro-mise of security are described.

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Detection Aids Requirements:

Intrusion alarms used to ensure against unauthorized entry must terminate in a continuously staffed alarm station located within the area enclosed by the outer barrier. Alarms must meet rigorous performance requirements.

Response

Parsons (MRS-11, September 1985, Vols I and II) describes vari-ous detection aids to be used, such as intrusion alarms and television monitors. A hardened alarm station to be appropriately located within the barrier area is also described.

Communications Requirements:

Each guard or watchman on duty must have a two-way radio to enable communication with the central alarm station. The alarm station must have both radio and telephone links with nearby law enforcement authorities who can provide assistance if requested. Communication equip-ment must continue operable if commercial electric power fails.

Response

Parsons (MRS-11, September 1985, Vols I and II) Indicates that a capability for communicating between the alarm station and the guards would be provided.

Security equipment would be among the equipment sup-plied by an onsite power system during any commercial power outage. A commitment for radio communication with local police is implied but is not made explicit.

Testing and Maintenance Requirements:

The protection system must include a test and maintenance program which ensures that all safeguards structures and equipment are maintained in operable condition.

Response

The staff agrees that this is a procedural requirement that would be inappropriate to address in a conceptual design.

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Security Response Requirements: The security organization must respond to and assess any indication of intrusion or introduction of unauthorized articles and must defend credible targets in the facility against sabotage.

Response: The staff agrees that this is a procedural requirement that would be inappropriate to address in a conceptual design.

i 3.4.3 Shipment Protection Requirements applicable to the protection of spent fuel en route to the MRS by road or rail are in 10 CFR 73.37. Although these requirements do not expressly address high level waste, they would apply because the waste contains many of the same radioactive materials as spent fuel. The scope of the current rule may be expanded to include high level waste. This action, if taken, will be carried out well before the rule would be applied to MRS shipments.

The applicable NRC requirements for shipment protection are listed below.

Holter and Braitman (1985) and Parsons (MRS-11, September 1985, Vol II) do not address spent fuel shipment protection, but the NRC shipment protection require-ments do not affect structures or equipment at the MRS site. Accordingly, DOE response to the requirements can be deferred to a later stage of the design and review process.

Route Approval A description of the routes to be used must be submitted for NRC approval prior

to their use for shipments. Prior to approval a survey of the routes would be made. The objective of the survey is to ensure there is adequate planning for protection against sabotage. The advice of local law enforcement agencies along the routes is sought and considered. In addition, each state is given an opportunity to designate preferred routes.

Armed Escorts in Cities l

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Shipments passing through cities must be escorted by armed guards.

Coordination with Police The staff coordinates with police (usually state police) along the proposed route as a part of the route survey process. The objectives of this coordina-tion are (1) to ensure that the police understand the nature of the shipments and (2) to learn of and consider the capabilities of the police units that could 1

be called upon for assistance in the event of a safeguards need.

Communication The shipment vehicle must be provided with onboard radiotelephone or equivalent, capable of communication with a control point that is continuously staffed during the period the shipment is en route. A call must be made from the shipment vehicle to the control point every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The control point staff acts if a call is not received.

Immobilization Device A road vehicle used for shipment must be equipped with a device that will immobilize the shipment vehicle on command from the shipment escort.

No Casual Stops j

A shipment vehicle may make no casual stops during the course of the shipment.

Schedule Information Protection Schedule information pertaining to a shipment must be protected against unautho-rized disclosure until the shipment is completed.

l Advance Notification to States

! The governor (or designee) of each state along the shipment route must be noti-fied prior to transport within the state. The object of this requirement is to 3-107

enable states to contribute to the safety, security, and ease of transport of the shipment.

3.4.4 Safeguards Contingency Planning Detailed requirements for safeguards contingency plans are in Appendix C of 10 CFR 73. A safeguards contingency plan is a documented plan to give guidance to a facility staff in the event of sabotage or a threat of sabotage of the facility.

Parsons (MRS-11, September 1985, Vols I and II) does not address safeguards con-tingency planning. However, contingency planning does not affect structures i and equipment at the MRS site. The required content of a safeguards contingency plan is summarized below, but the DOE response to these requirements can be deferred to a later stage of the design and review process.

Background

The category of information would identify and define the perceived dangers and incidents with which the plan would deal and the general way it would handle these.

Generic Planning Base 4

This category of information defines the criteria for initiation and termination of responses to safeguards contingencies, together with the specific decisions, actions, and supporting information needed to bring about such responses.

Licensee Planning Base This category of information includes the factors affecting contingency planning that are specific to the MRS operation, including transportation.

Responsibility Matrix 3-108

This category of information consists of detailed identification of the organi-zational entities responsible for each decision or action associated with spe-cific responses to safeguards contingencies.

Procedures This category of information details the actions to be taken and decisions to be made by each member o. units of the organization as planned in the responsi-bility matrix.

3.4.5 Material Accountability Record Keeping 10 CFR 72, Subpart 0, gives the applicable requirements for material accountabi-lity and related activities. In general, these requirements do not have a strong effect on MRS structures and equipment and therefore do not influence the con-ceptual design stage of the MRS. DOE is aware of the requirements and has pro-vided information for the more important ones. The requirements and the DOE responses are summarized below.

Records Requirements:

Records must be kept showing receipt, inventory (including locations), disposal, acquisition, and transfer of all spent fuel in storage.

Response: The RAD addresses the receipt of spent fuel and the verifica-tion of the shipping document, DOE /NRC-741, but it does not address records of the history of the' spent fuel. Those records include fuel manufacturer's loading data and utility data on uranium burnup and plutonium production.

These data are necessary to calculate a material balance across the fuel cycle and to calculate the yield of special nuclear material (SNM) at time of recovery. However, comment on these matters can be deferred by DOE until a later design phase of the MRS.

Inventories Requirements: A physical inventory of all spent fuel must be conducted at intervals not to exceed 12 months. NRC guidance calls for a licenseee to 3-109

use item control for inventory purposes. The SNM content of each item may be calculated from 10 CFR 72.15(a) records.

Response: DOE has committed to periodic inventories. Item 25.3.2 of the RAD does not have enough detail to determine how DOE would establish or measure the SNM value to be used for records and inventory. However, the staff anticipates that DOE can satisfactorily control SNM values for record purposes at a subsequent design stage.

l Procedures l

Requirements: Written material control and accounting procedures to enable the licensee to account for the spent fuel in storage must be established, maintained, and followed. Related NRC guidance proposes strict item con-trol, including piece count of fuel rods and assemblies by three knowledge- l able MRS employees who attest to the canister content; the immediate seal- '

ing of each canister in some permanent manner such as by welding; and iden-tifying each canister with permanent markings as necessary. The inventory is verified by item control.

Response

Parsons (MRS-11, September 1985, Vol I) outlines a satisfactory set of procedures. The details of the outline for verifying the inventory must be developed at later design stages.

Duplicate Records Requirements: Records of spent fuel in storage must be kept in duplicate at two separate locations.

Response

The RAD states that two separate records would be kept in two locations at the MRS site. '

Miscellaneous Provisions Requirements: Any loss of SNM must be reported promptly to the NRC regional office. Material status reports must be completed and submitted to the 3-110

NRC in a prescribed format. Each transfer of spent fuel must be documented in a prescribed format.

Response: DOE has not submitted material addressing these requirements.

However, these requirements do not influence the MRS conceptual design and hence can be addressed at a later design stage.

3.4.6 NRC Experience in Applying Requirements All of the preceding saftguards requirements have been in effect for a number of years, and the staff has had considerable experience in their application.

Fixed site protection requirements have been applied at three licensed sites.

Most potential problems are discovered when a licensee's physical protection plan is reviewed and system weaknesses are thereby avoided.

The staff has been applying transportation protection requirements since 1979.

The number of shipments varies each year, but ranges from about 75 to 200 annu-ally. For the first 2 years all shipments were inspected at the point of origin (or destination). The requirements were readily understood, as indicated by the  :

, fact that the inspections revealed no significant safeguards deficiencies.

Thousands of miles of routes have been successfully surveyed (NUREG-0725).

Arrangements for advance notification to states have been made and the notifi-cations have been carried out. State-designated preferred routes have been accepted in instances where states have officially expressed a preference.

Railroad routes also were surveyed initially by the NRC and later by both NRC and DOT representatives. The MRS shipments appear to pose no new problems in i safeguard transportation regulation.

The material control and accounting requirements have been applied to power reactors, non power reactors, and certain fuel cycle facilities. Although the regulations do not require applicants to submit a plan to document how com-pliance will be achieved, the requirements have been found to be sufficiently detailed to enable applicants to gain a thorough understanding of what it re-quired. Inspections have identified no significant deficiencies in the appli-cation of these requirements.

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1 3.4.7 Reassessment of Certain NRC Requirements NRC regulations are organic documents subject to continuing review, change, and updating to reflect the current state of nuclear knowledge, consequence predic-tion, developing technology, and other factors. NRC regulations that would ap-ply to the MRS and related transport are being reviewed and may be amended. In the case of transportation protection requirements for spent fuel shipments, the need to expand the scope of the requirements to include shipments of high level waste has been noted. In addition, the NRC recently completed a lengthy research program to determine the effects of successful sabotage. On the basis of these findings, a proposed rule that would moderate the existing requirements was issued in June 1984. Public comment on that proposed rule is now being considered. In the case of fixed site protection, the NRC is sponsoring research to determine the effect of sabotage on dry casks to be used for long-term stor-age, including casks similar to those that would be used at the MRS. The staff does not expect the results of the rule reviews to result in any problem or delay in the licensing of an MRS.

3.4.8 Safeguards Conclusions The NRC reached the following conclusions concerning the safeguards aspects of the MRS conceptual design.

(1) Safeguards requirements suitable for use at the MRS currently exist.

Special or new requirements beyond those currently in progress need not be developed, but NRC safeguards regulations are subject to continuing evaluation, which could result in changes.

(2) The NRC is experienced in applying the requirements to a variety of fixed sites and to the protection of spent fuel in transit.

(3) DOE is aware of the NRC requirements and has evaluated their influence on the MRS conceptual design.

(4) Requirements for structures, equipment, and access controls have been ade-quately taken into account in the MRS conceptual design.

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(5) Other, more detailed NRC requirements have only a remote influence on MRS conceptual design and are better and more efficiently addressed at a later state.

(6) The NRC safeguards review revealed no reason why NRC safeguards require-ments cannot be met at the MRS.

3.5 Repository Interface DOE intends that the MRS perform certain functions that would have to be under-taken at each individual repository if the MRS were not built. These MRS func-tions involve the preparation of the spent fuel or high level waste for final disposal in packaging compatible with the physical requirements of the repository.

It is the responsibility of the MRS to ensure that MRS operations do not adversely affect the performance of the high level waste package so that repository performance requirements can not be met.

DOE's schedule indicates that DOE would submit a license application for an MRS facility in 1989. The schedule also indicates that the first repository would be selected in 1991 from the three current candidates, 2 years after the MRS application was submitted and, after the staff had completed its safety and environmental reviews for the MRS. Therefore, DOE should show that the MRS waste packages prepared for disposal could accommodate any of the requirements of the three candidate repositories so that when a repository is selected, the packages would be compatible with the requirements of the repository that was selected.

One of the principal aspects of the repository requirements with respect to the packaging of waste would be the selection of the waste package design. In this regard it is possible that each repository would require different, perhaps unique, designs with different dimensional tolerances. The MRS application should demonstrate, through its design and operational information, the technique whereby each of the three candidate repositories would be accommodated.

A second aspect, closely related to design selection, is the close coordina-tion required between the repository and MRS organizations to provide a consis-tent and adequate assurance of quality in the disposal package. The quality 3-113

l l

l 1

l assurance of design selection, fabrication techniques, and seal testing of I disposal packages should be consistent with repository requirements and not foreclose any options available to each repository for packages received for safe disposal.

On the basis of its review and evaluation of the conceptual design of the MRS, )

considering its capability to accommodate a variety of canister sizes and ma-terials, the staff foresees no impediment at the MRS that would foreclose re-1 pository options for package requirements. If the Congress approves the DOE I request to authorize an MRS, the staff would recommend that DOE establish a quality assurance and waste acceptance coordinating committee that would be made up of representatives of the repository and MRS organizations, and would include NRC staff participation.

l 1

3.6 Transportation The transportation of sperit fuel and high level waste to and from the MRS is an important consideration in an evaluation of the environmental impacts that would be associated with such a facility. Regulating safety in the transporta-tion of commercial spent fuel and high level waste is a responsibility shared by NRC and 00T. These responsibilities are described in a memorandum of under-standing (MOU) between NRC and DOT (44 FR 28690). Briefly, the MOU describes the NRC role (for materials of interest here) as including the development of safety standards for packaging and approving package designs. (In addition, NRC regulations require the physical protection of spent fuel shipments against acts of sabotage (see also Section 3.4.3 above).) DOT is primarily responsible for regulating the conditions of carriage and routing. The existing regulatory system for transporting materials such as spent fuel has been in place for many years. The safety record associated with this system has been excellent; NRC is not aware of any injuries caused by the release of radioactive materials in transport. In considering the transportation aspect of the MRS, NRC's primary objective is to ensure that the safety record is maintained as the number of shipments of spent fuel and high level waste increases.

l The Nuclear Waste Policy Act (Act) assigns to DOE the responsibility to design, operate, and maintain the transportation system to ship spent fuel and high 3-114 l

level waste facilities authorized by the Act. The Act also contains sections that pertain to the regulation of these shipments. For example, Section 9 states that nothing in the Act shall be construed to affect Federal, state, or local laws pertt *qing to the transportation of spent nuclear fuel or high level radioactive waste.

Formal agreements have been reached between Federal agencies to deal with specific aspects that concern the regulation of shipments under the Act. In connection with-the development of transportation packaging under the Act, DOE and NRC have executed a procedural agreement (48 FR 51875, November 14, 1983).

The procedural agreement establishes common planning assumptions and outlines procedures that DOE and NRC will observe. As stated in the agreement, DOE plans to use packaging that has been approved by NRC in accordar,ce with 10 CFR 71 for DOE shipments to and from NRC-licensed facilities (e.g., shipments from utili-ties to an MRS or from an MRS to a repository).

DOE has also expressed its intentions regarding transportation activities under the Act in several recent DOE publications. In its mission plan, DOE states:

Commercial spent fuel and high-level waste will be transported in accordance with all applicable 00T and NRC regulations.

In its draft transportation institutional plan, DOE states:

When engaged in future activities related to the transportation of commercial waste to NWPA facilities, the OCRWM (Office of Civilian Radioactive Waste Management) will meet all existing DOT and NRC safety and security standards.

Finally, in its draft transportation business plan, DOE states:

Each design for a new cask used to transport civilian radiocctive l waste must be certified by NRC prior to use. l There is a certain degree of ambiguity and inconsistency among these statements. l i

For example, in one statement reference is made to meeting " existing" standards, 3-115

]

)

whereas in another it is made to " applicable" regulations. One statement refers to meeting " security" standards, but none of the others refer to security require-ments.

Furthermore, all the statements are silent on a related matter, pre-notification for shipments.

Although it is clear that DOE intends to use NRC-certified casks, DOE should clarify its intent with respect to NRC physical protection and prenotification requirements for transport of commercial spent fuel and high-level waste to the MRS and the repository.

3.6.1 Conclusion In support of its proposal to Congress to construct and operate an MRS at the Clinch River site, DOE has prepared an environmental assessment for a monitored retrievable storage facility (DOE /RW 0035, Vol II). In the executive summary of that document, DOE states that the radiological risk to the public from the shipment of spent nuclear fuel is low, and that the proposed MRS could reduce exposures, as opposed to not constructing an MRS. The NRC staff review of the MRS proposal and its supporting environmental assessment has not identified any transportation activities that would result in unacceptable risks for the public.

In 1977, NRC published its " Final Environmental Statement on the Transportation of Radioactive Materials by Air and Other Modes" (NUREG-0170). In 1981, after considering the information developed in NUREG-0170 and the safety record asso-ciated with transportation of radioactive material, the Commission concluded that the present regulations provide a reasonable degree of safety and that no immediate changes in the regulations were needed to improve safety. Because the MRS transportation program would be conducted by DOE in accordance with transportation regulations found to provide adequate public safety, the NRC concludes that MRS transportation can be conducted safely.

The Commission regularly reviews its regulations and makes changes as necessary to take into account new technology or new information. Such a review of the transportation regulations is now in progress. The staff also is planning to update NUREG-0170, which may provide information applicable to the DOE transpor-tation program.

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i 4 REFERENCES Bailey, W. J., and A. B. Johnson, " Surveillance of LWR Spent Fuel in Wet Storage," Electric Power Research Institute report EPRI NP-3765, October 1984.

Bickford, D. F. , and C. M. Jantzen, " Devitrification Behavior of SRL Defense Waste Glass, DP-MS-83-62, paper presented at Materials Research Society Annual Meeting, November 14-17, 1983.

l l

Chick, L. A. , et al. , " West Valley High-Level Nuclear Waste Glass Development:

A Statistically Designed Mixture Study," Pacific Northwest Laboratory report i

PNL-4992, October 1984. '

Departement of Energy (DOE), DOE /RW 0035, " Monitored Retrievable Storage, Submission to Congress," Vol. 1, " Proposal"; Vol. 2. " Environmental Assessment"; Vol. 3 " Program Plan."

t l Duke Power Company / Westinghouse Electric Corporation, " Spent Fuel Consolidation Demonstration, Oconee Nuclear Station," 1982.

Gertz, C. P., "The INEL Demonstration Program on Dry Storage Casks," paper presented at the Atomic Industrial Forum Meeting on Solutions to Nuclear Transportation Issues, June 19, 1985. '

Hawley, G. G., The Cnndensed Chemical Dictionary, 10th edition, Van Nostrand Reinhold Company, 1977.

Holter, G. M. , and J. L. Braitman, " Siting of an MRS Facility: Identification of a Geographic Region that Reduces Transportation Requirements," Pacific Northwest Laboratory report PNL-5424, April 1985.

Johnson, A. B. , and E. R. Gilbert, " Technical Basis for Storage of Zircaloy-

, Clad Spent Fuel in Inert Gases," Pacific Northwest Laboratory report PNL 4835, September 1983.

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Pacific Northwest Laboratory, " Functional Design Criteria for an Integral Monitored Retrievable Storage (MRS) Facility," MRS-3-85, PNL-5673, Revision 4, August 1985.

R. M. Parsons Company (Parsons), MRS-11, " Integral Monitored Retrievable 4 Storage Facility: Conceptual Design Report," Vol I, " Design Description";

Vol II, " Regulatory Assessment Document"; Vol III, " Conceptual Design Draw-ings"; Vol VI, " Design Studies," September 1985.

-- , MRS-11. " Integral Monitored Retrievable Storage Facility: Conceptual Basis for Design," October 1985.

Rockwell Hanford Operations (Rockwell), " Review of Zirconium-Zircaloy Pyrophoricity," SD-WM-TI-104, January 1984.

Rockwell International, " Draft Site Characterization Report for the Basalt Waste Isolation Project," June 1982.

United Kingdom Atomic Energy Authority, IGR-TN/S-578, "A Code of Practice for the Handling of Zirconium and Its Alloys," July 1957.

U.S. Nuclear Regulatory Commission, NUREG-75/039, " Final Environmental State-( ment Related t' o the Construction of Tennessee Valley Authority's Hartsville Nuclear Plants, Plant A, Units 1 and 2, Plant B Units 1 and 2," June 1975.

-- , NUREG-0014, " Safety Evaluation on the Hartsville Nuclear Plant, Units A1, I A2, B1, and B2," April 1976.

-- , NUREG-0139, " Final Environmental Statement Related to the Construction and Operation of Clinch River Breeder Reactor Plant," 1977; Supplement No.1, 1982.

-- , NUREG-0170, Vol. 1,"The Transportation of Radioactive Material by Air and Other Modes," November 1977.

-- , NUREG-0725, Rev. 4," Pub 11c Information Circular for Shipments of Irradiated Reactor Fuel," June 1984.

1 4-2

-- , NUREG-0786, " Site Suitability Report in the Matter of the Clinch River Breeder Reactor Plant," March 1977; revised, June 1982.

j -- , NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plar.ts, LWR Edition," July 1981.

-- , NUREG-1092, " Environmental Assessment for 10 CFR Part 72, ' Licensing Requirement for the Independent Storage of Spent Fuel and High-Level Radioactive Waste,'" August 1984.

/ -- , NUREG/CP-0057, "The Performance of Defected Spent LWR Fuel Rods in Inert and Dry Air Storage Atmosphere," in Transactions of the 12th Water Reactor Safety Research Information Meeting, October 22-26, Gaithersburg, Maryland, 1984.

-- , NUREG/CR-0150, " Estimates of Internal Dose Equivalent to 22 Target Organs for Radionuclides Occurring in Routine Releases from Nuclear Fuel Cycle Facilities," Vol. 2, November 1979.

/ -- , HUREG/CR-1729, Vols.1 and 2," Evaluating Strength and Ductility of Irradi-ated Zircaloy, Task 5," L. M. Lowry et al., May 1981.

1

-- , NUREG/CR-1918, " Dose-Rate Conversion Factors for External Exposure to Photons and Electrons," August 1981.

-- , Office of Inspection and Enforcement (IE) Information Notice 85-12 "Recent Fuel Handling Events," February 11, 1985.

1 Westinghouse Electric Corporation, " Integral Monitored Retrievable Storage Facility Licensing Design Guide," MRS:RJS:84-024, draf t report, May 1984.

Proceedings of the International Workshop on Irradiated Fuel Storage Operating Experience and Development Programs, October 17 and 19, 1984.

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