ML20151A155

Final Generic EIS on Decommissioning of Nuclear Facilities

ML20151A155
Person / Time
Issue date:	10/31/1987
From:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:
Shared Package
ML20151A086	List: IA-88-274, Partially Withheld Version of Commission Paper Obtaining Commission Approval to Publish Final Rule Amending 10CFR30, 40,50,51,70 & 72 by Setting Forth Technical & Financial Requirements for Decommissioning Nuclear Facilities ML20151A083 ML20151A088 ML20151A155 ML20151A178
References
FOIA-88-274 ENVS-871031-01, ENVS-871031-1, NUDOCS 8807190267
Download: ML20151A155 (386)

Text

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t FIML GENERIC ENVIRONMENTAL IMPACT STATEMENT on decommissioning of nuclear facilities U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research October 1987 i

8807190267 880629 PDR FOIA P H I,L,L,1 S88-274 PDR . I l

Enclosure {$

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FOREWORD BY NUCLEAR REGULATORY COMMISSION STAFF The NRC staff is in the process of reappraising its regulatory position relative to the decomissioning of nuclear facilities. The initial part of this activity consisted of obtaining the information base to support any -

subse . Highly detailed studies weae compi Te regulatorv rk.u d, through

s. 4 cu 4 huent chanaestechnical assists,4% eJak Lklossain'ea.n s1 These studies were,Asas w%

ance4arb referenc'ed in this accument). in turn, utilized along with other information., to prepare a Oraf t Generic Envi'r onmental St'atement;on dec6mmi'ssioning Nuclear Fac'ilities, draf t GEIS,', '

NUREG-0586, January 1981. O'n February 11, 1985, the Comission published a notice of proposed rulemaking on decommissioning criteria for nuclear facilities (50 FR 5600).

This Final Generic Environmental Impact Statement on Deccamissioning Nuclear Facilities is being published based on public comment on the draft GEIS and on the proposed rule as well as on updated information in the technical informa-tion base. This statement is required because the regulatory changes that might result from the reevaluation of decommissioning policy may be a major NRC action affecting the quality of the human environment.

The information provided in this Statement, including any coments, will be -

included in the record for consideration by the Commission in establishing criteria and new standards for decommissioning.

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OVERVIEV At the end of a commercial nuclear facility's useful life, termination of its license by the Nuclear Regulatory Commission (NRC) is a desired objective. Such termination requires that the facility be decommissioned. Decommissioning means the removal of a nuclear facility safely from service and reduction of residual radioactivity to a level that permits release of the property for unrestricted g use and termination of the license. It is the objective of NRC regulatory activities in protecting public health and safety to provide to the applicant or licensee appropriate regulations and guidance to accomplish nuclear facility decommissioning.

Although decommissioning is not an imminent health and safety problem, the nuclear industry is maturing. Nuclear facilities have been operating for a ntaber of years, and the number and complexity of facilities that will require decommissioning is expected to increase in the near future. Accordingly, the NRC is reevaluating its regulatory requirements concerning decommissioning.

This final generic environmental impact statement is part of this reevaluation. ,

PAST ACTIVITIES In support of this reevaluation, a data base on the technology, safety, and cost of decommissioning various nuclear facilities and on other matters related -

to decommissioning, including financial assurance, is being completed for the NRC by Battelle Pacific Northwest Laboratory (PNL), by Oak Ridge National Laboratory and by other contractors. Based on this data base and on input from other State and Federal government agencies and the public, NRC has modified -

and amplified its policy considerations and data base requirements in a manner responsive to comments received. Another area addressed is the generic appli-cability of the data base for specific facility types. This has been addressed through expansion of the PNL facility reports to include sensitivity analyses for a variety of parameters potentially affecting safety and cost considerations.

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A draf t generic ereironmental impact s catement was issue: in January,1981 and coenments received have been considereu in the developme" of this final state-sent. On February 11, 1985, the NRC published a notice :f proposed rulemaking on Decommissioning Criteria for Nuclear Facilities (50 Fi 5600).(15) The proposed amendments covered a number of topics related to decommissioning that would be applicable to 10 CFR Parts 30, 40, 50, 51, 70, and 72 applicants and licensees. These topics included decommissioning alterratives, planning, assurance of funds for decommissioning, c,vironmental re,iew requirements, and residual radioactivity.

SCOPE OF THE EIS Regulatory changes are being considered for both fuel cy:le and non-fuel-cycle nuclear facilities. The fuel cycle facilities are pressurized (PWR) and boiling water (BWR) light water reactors (LWRs) for both single and multiple reactor sites, research and test reactors, fuel reprocessing plants (FRPs)

(currently, use of FRPs in the commercial sector is not being considered),

small mixed oxide (M0X) fuel fabrication plants, uranius fuel fabrication plants (U-fab), uranium hexafluoride conversion plants (tF s), and independent spent fuel storage installations (ISFSI). Under non-fuel-cycle facilities, consideration is given to major types such as radiopharsaceutical or industrial radioisotope supplier facilities, various research radioisotope laboratories, and rare metal ore processing plants where uranium and trorium are concentrated in the tailings.

This EIS addresses only those issues involved in the activities carried out at the end of a nuclear facility's useful life which permit the facility to be removed safely from service and the property to be released for unrestricted use. It does not address the considerations involved in extending the life of a nuclear facility. If a licensee makes an application for extending a facility license, an application for license renewal or amendment or for a new license would be submitted and reviewed according to appropriate existing regulations.

This is not considered to be decommissioning and therefore is outside the scope of this EIS.

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High-level waste repositories, low-level waste burial facilities, and urar.ium tills and their associated mill tailings piles are covered in separate rulemakings and are not included here. The first two items are covered in Title 10 of the Code of Federal Reg.,1ations (10 CFR) Parts 60 and 61. The last item is covered in amendments to 10 CFR Part 40.

REGULATORY OBJECTIVE It is the responsibility of the NRC to ensure, through regulations and other guidance, that apprcpriate procedures are followed in decommissioning to protect the health and safety of the public. Presert regulatory requirements and guidance cover the requirements and criteria for decommissioning in a limited way and are not adequate to regulate decommissioning actions effectively.

Areas needing further criteria include decommissioning alternatives, financial assurance, planning and residual radioactivity levels as discussed below:

Decommissioning Alternatives. It is the responsibility of the NRC, in protecting public health and safety, to ensure that af ter a nuclear facility ceases opera-tion its license is terminated in a timely manner. License termination requires decommissioning. Analysis of the technical data base. establishes that decoe-missioning can be accomplished and the facility released for unrestricted use shortly af ter cessation of operations or, in certain situations for certain facilities, delayed and completed af ter a period of storage. These situations would include considerations where the potential exists for occupational expo-sure and waste volume reduction, resulting from radioactive decay, or the -

inability to dispose of waste due to lack of disposal capacity, or other site specific factors which may affect safety. Completing decommissioning and releasing the site for unrestricted use eliminates the potential problems that may result from an increasing number of sites contaminated with radioactive material, as well as eliminating potential health, safety, regulatory, and economic problems associated with maintaining the nuclear facility.

Based on the technical data base, it appears that completing decommissioning shortly af ter cessation of facility operations or delaying completion of decom-missioning for a 30 to 50 year period are reasonable options for decommir,sion-ing light water power reactors. Delay beyond that period may be acceptable if l

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there is an inability to dispose of waste due to lack of disposal capacity or if there are site specific factors af fecting safety such as if the safety of an adjacent reactor might be af fected by dismantlement procedures.

For research and test reactors and for nuclear facilities licensed under 10 CFR Part:, 30, 40, 70, and 72, occupational doses would be in most cases much less significant thn power reactors. Thus, completing cecommissioning shortly af ter cessatic , of operations is considered the most reasonable option. De-laying completion of decommissioning to allow short lived nuclides to decay may be justified in some cases, however, any extended delay would rarely be justi-fiable.

Financial Assurance. Consistent with the regulatory objective of decommis-sioning as described above, reasonable assurance is required from the nuclear facility licensee that adequate funds are available to decommission the facility. The funding mechanisms considered reasonable for providing the necessary assurance include prepayment of funds into a segregated account, insurance, surety bonds, letters of credit, and certain other guarantee methods, and a sinking fund deposited into a segregated account. Based on the information base developed in part of the NRC's reevaluation, another funding mechanism, internal reserve, is considered to provide reasonable assurance of the availability of funds for decommissioning for electric utility licensees owning more than one generating facility.

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Planning. Planning for decommissioning is a critical item for ensuring that the decommissioning activities can be accomplished in a safe and timely manner.

Development of detailed plans at the application stage is not possible because many factors (e.g., technology, regulatory requirements, economics) will change before the license period ends. Thus, most of the planning for the actual decommissioning will occur near final shutdown. However, a certain amount of preliminary planning should be done at the application stage.

Information on decoar'ssioning funding provisions must be submitted with an application for a license for a nuclear facility. This information should include the method of assuring funds for decommissic,;ng (as discussed above under Financial Assurance) and an indication of the amount being set aside.

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• revisions should also be made to adjust cost levels and associated funding levels over the life of the facility.

Facilitation of decommissioning in the design of a facility or during its operation can be beneficial in reducing operational exposures and waste volumes requiring disposal at the time of decommissioning. Although many aspects of facilitation can be covered under existing regulations, s;ecific requirements that records of relevant operational and design information important to decommissioning be maintained should be added.

A final detailed decommissioning plan is required for review and approval by the NRC prior to cessation of facility operation or shortly thereafter. Besides the description of the decommissioning alternative which will be used, the final plan should include a description of the plans to ensure occupational and public safety and to protect the environment during decommissioning; a description of the final radiation survey to ensure that remaining residual radioactivity is within levels permitted for releasing the property for unrestricted use; an updated cost estimate; and for certain facilities as appropriate a description of quality assurance and saf 2 guards provisions. The plan should include an estimate of the cost required to accomplish the decommissior.ing.

Residual Radioactivity levels. The selection of an acceptable level is outside the scope of rulemaking supported by this EIS. The Commission is participating in an EPA organized interagency working group which is developing Federal guid-ance on acceptable residual radioactivity for unrestricted use. Proposed ~

Federal guidance is anticipated to be published by EPA. NRC is planning to implement this guidance through rulemaking as soon as possible, as well as by issuing regulatory guides and standard review plan sections. Currently, criteria for residual contamination levels do exist and research and test reactors are being decommissioned using present guidance contained in Regula-tory Guide 1.86 for surface contamination plus 5 pr/hr above background measured at 1 meter from the surface for direct radiation. The cost estimate for decommissioning can be based on current criteria and guidance regarding residual radioactivity levels for unrestricted use. The information in the studies performed as part of the reevaluation on decommissioning have indicated that in any reasonable range of residual radioactivity limits, the cost of viii

de:ommissioning is relatively insensitive to the radioactivity level and use of cest data based on current criteria should provide a reasonable estimate. Even in situations where the residual radioactivity level might have an effect on decomissioning cost, by use of update provisions in the rulemaking, it is e.xpected that the decomissioning fund available at the end of facility life vill appr:ximate closely the actual cost of deccerissioning.

EWIRONMENTAL IMPACT STATEMENT Generally, the major environmental impact from decomissioning, especially for power reactors, occurs when the decision is made to operate the reactor.

Provided decommissioning rules are in place and based on the conclusions of Chapters 4 and 5 regarding impacts from reactor decommissioning alternatives, it is not expected that any significant environmental impacts will result from decommissioning. Therefore current 10 CFR Part 51 needs to be amended to delete the manditory EIS requirement for decomissioning of power reactors. An EIS may still be needed but this should be based on site specific factors.

Consequently a licensee should submit a supplemental environmental report and safety analysis and, based on these submittals, the NRC should consider prepara-rlo tion and issuance of an environk. ital assessment and a finding of environmental 4

impact. This is expected to be reasonable for most situations.

It is imperative that decommissioning rule amendments in 10 CFR Parts 30, 40, 50, 51, 70, and 72 be issued at this time because it is important to establish -

financial assurance provisions, as well as other decommissioning planning provisions, as soon as possible so that funds will be available to carry out decomissioning in a manner which protects public health and safety. Based on this need for the decomissioning provisions currently existing as well as those contained in the proposed rule amendments, the Comission believes that the rule can and should be issued now.

CONCLUSIONS ON DEcomlSSIONING IMPACTS Consideration of the decommissioning data base including coments on the Draft Generic Environmental Statement and on the proposed rule and of the need for ix

regulatory activity has led to the following conclusions in the 'inal Generic Environmental Impact State ent:

(1) The technology for decommissioning nuclear facilities is weil in hand and, while technical improvements in decomissioning techniques are to be ex-pected, decomissioning at the present time can be performed safely and at' reasonable cost. Radiation dose to the public due to de:: rissioning activities should be very small and be primarily due to tra sportation of decommissioning waste to waste burial facilities. Radiatit9 dose to de-comissioning workers should be a small fraction of their exposure experi-enced over the operating lifetime of the facility and be well within the occupational exposure limits imposed by regulatory requiresents. Decom-missioning costs are reasonable and are, at least for the larger facilities such as reactors, a small fraction of the present worth commissioning costs (i.e., less than 10%).

(2) Decommissioning of nuclear facilities is not an iminent health and safety problem. However, planning for decommissioning as an integ al activity prior to comissioning as well as during facility life is a critical item that can have an impact on health and safe' . .s well as cost. Essential to such planning activity is reasonable assurance that funds will be avail-able for performing required decommissioning activities at the cessation of facility operation.

(3) Decomissioning of a nuclear facility generally has a positive environ- -

mental impact. At the end of facility life, termination of a nuclear license is the goal. Termination requires decontamination of the facility so that the level of any residual radioactivity remaining in the facility or on the site is low enough to allow unrestricted use of t.9e facility and site. Commitment of resources, compared to operational aspects, is generally small. The major environmental impact of decommissioning is the commitment of small amounts of land for waste burial in excnange for reuse of the facility and site for other purposes. Since in many instances, such as at a reactor facility, the land is a valuable resou ce, return of this land to the commercial or public sector is highly desi able.

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INCCDDORATION OF EIS CONCLUSIONS IN REGULATIONS l l

• t is recommended that specific implementation of regulatory activities be per- l f orced by rulemaking as amendments to existing regulations (i.e. ,10 CFR Parts 3",

I 40, 50, 51, 70 and 72) rather than as a separate regulation solely covering decommissioning. Because decommissioning overlaps so many areas covered by c esent regulations, such incorporation we.lc be more efficient.

ORGANIZATION OF THE EIS Sections 1 to 3 of the main text of the EIS contain material common to all the facilities considered and should be read for discussion of generic issues.

Sections 4 to 14 contain specific facility considerations. These separate f acility sections were kept as self-contained as possible (recognizing that some redundancy would be inevitable for such an organizational approach), so that a user interested in a particular facility type need primarily read only that section, as well as introductory, generic, and policy sections. Section 15 contains details on how the conclusions of the EIS will affect regulatory policy considerations. The last section of the EIS is a glossary which pro-vides the reader definitions of terms used in this report, including those used in a special sense in this report. Finally, in the Appendices, discussion and resolution of comments on the DGEIS is presented in Appendix A along with the original comments presented in Appendix B.

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TABLE OF CONTENTS f.ag STUDY CONTRIBUTORS . ............................................ i F O R EWO R D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii AB S T R A C T . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii 0VERVIEW .............. ................................... ...... iv T AB L E O F C O N T E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . xii LIST OF TABLES ................................................... xix LIST OF FIGURES .................................................. xxiii

1.0 INTRODUCTION

................................................ 1-1 1.1 PurposeofEIS......................................... 1-1 1.1.1 NEPA Requirements ............................... 1-2 1.2 O rgani zation o f the EIS . . . . . . . . . . . . . . . . . . . . . . . . ...... 1-3 1.3 Purpose of Decommi s s' oni ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-4 1.4 Responsibility for Decommi ssioning . . . . . . . . . . . . . . . . . . . . . 1-4 1.4.1 Existing Criteria and Regulations for Decommissioning ................................. 1-4 1.4.2 Current Rulemaking Activities . . . . . . . . . . . . . . . . . . . 1-5 1.5 History, Background, and Experience With D e c o mm i s s i o n i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-5 RE F E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-9 2 GENERIC NUCLEAR FACILITY DECOMMISSIONING CONSIDERATION $ . . . . . . . . 2-1 2.1 Nuclear Facilities Operational Description . . . . . . . . . . . . . 2-1 2.1.1 The Nucl ear Fuel Cycl e . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 2.1.2 Research and Test Reactors . . . . . . . . . . . . . . . . . . . . . . 2-6 2.1.3 Non-Fuel-Cycl e Nucl ear Facili ties . . . . . . . . . . . . . . . 2-6 2.2 Facilities Considered in EIS . . . . . . . . . . . . . . . . . . . . . . . . . . 2-7 2.3 Definition of Decommissioning . . . . . . . . . . . . . . . . . . . . . . . . . . 2-7 2.4 Decommissioning Alternatives ........................... 2-7 2.4.1 No Ac t i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-11 2.4.2 DECOM ........................................... 2-11 2.4.3 SAFSTOR......................................... 2-12 2.4.4 ENTOMB .......................................... 2-15 xii

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TABLE OF CONTENTS (Continued)

P. age 2.5 Residual Radioactivity Levels for Unrestricted Use of a Facility ............................................... 2-16 2.6 Financial Assurance .................................... 2-19 2.6.1 Present Regulatory Guidance ..................... 2-20 2.6.2 Implementation of Financial Assurance Requirements ................................... 2-20 2.7 Management of Radioactive Wastes and Interim Storage ... 2-26 2.8 Safeguards ............................................. 2-28 RE F E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-30 3 AFFECTED ENVIRONMENT - GENERIC SITE DESCRIPTION . . . . . . . . . . . . . . . 3-1 3.1 Fuel Cycle Facility Site ............................... 3-1 2EFERENCES ....................................................... 3-34 4 PRESSURIZED WATER REACTOR ..................................... 4-1 4.1 PYR Dese.ription ........................................ 4-2 4.2 Reactor Decommissioning Experience . . . . . . . . . . . . . . . . . . . . . 4-4 4.3 Decommissioning Alternatives ........................... 4-5 4.3.1 DECON ........................................... 4-5 4.3.2 SAFSTOR ......................................... 4-11 4.3.3 ENTOMB .......................................... 4-12 4.3.4 Sensitivity Analyses ............................ 4-15 4.4 Environmental Consequences ............................. 4-19 4.5 Comparison of Decommissioning Alternatives . . . . . . . . . . . . . 4-24 .

REFERENCES .................................. .................... 4-26 5 BO I LI NG WAT E R REACTO R . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 Boiling Water Reactor Description . . . . . . . . . . . . . . . . . . . . . . 5-2 5.2 BWR Cecommissioning Experience ......................... 5-4 5.3 Decommissioning Alternatives .............. ............ 5-4 5.3.1 DECON ...........................................' 5-5 5.3.2 SAFSTOR ......................................... 5-11 5.3.3 ENTOMB .......................................... 5-12 5.3.4 Sensitivity Analyses ............................ 5-15 5.4 E nv i ro nme n tal C o n s e qu e n c e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-19 5.5 Comparison of Decommis sioning Alternatives . . . . . . . . . . . . . 5-24 REFERENCES ....................................................... 5-26 xiii

l TABLE OF CONTENTS (Coritinued)

Page 6 MULTIPLE REACTOR STATION ...................................... 6-1

- 6.1 Mul tiple-Reactor Station Description . . . . . . . . . . . . . . . . . . . 6-2 6.1.1 Multiple-Reactor Station Concepts . . . . . . . . . . . . . . . 6-2 6.1.2 Multiple-Reactor Station Scenarios .............. 6-4 '

6.1.3 Reference Light Water Reactors .................. 6-6 6.2 Multiple Reactor Station Decommissioning Experience . . . . 6-7 6.3 Multiple Reactor Station Decommissioning Scenarios . . . . . 6-7 6.3.1 DECON ........................................... 6-23 6.3.2 SAFSTOR ......................................... 6-25 6.3.3 ENTOMB .......................................... 6-27 6.4 Envi ronmental Cons equence s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-28 6.5 Comparisons of Reactor Decommissioning at Multiple-Reactor Stations and at Single-Reactor Stations . . . . . . . . 6-29 REFERENCES ....................................................... 6-32 7 RESEARCH AND TEST REACTORS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.1 Description of R&T Reactors . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-2 7.1.1 Reference Research Reactor ...................... 7-2 7.1.2 Reference Test Reactor .........,................ 7-2

7. 2 Research and Test Reactor Decommissioning Experience ... 7-3

7. 3 Decommi s s i oni ng Al te rnati ves . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-3 7.3.1 DECON ............. .......................... 7-3 7.3.2 SAFSTOR ........... ......................... 7-8 -

7.3.3 ENTOMB ............. ......................... 7-9 7.3.4 Sensitivity Analysis ......................... 7-10 7.4 Envi ronmental Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-12

7. 5 Comparison of Decommi ssioning Alternatives . . . . . . . . . . . . . 7-15 REFERENCES ....................................................... 7-16 8 DECOPHISSIONING OF REACTORS THAT HAVE BEEN INVOLVED IN A C C I D E NT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-1 8.1 Reference Facility Description and Reference Accident Scenarios .............................................. 8-3 8.2 Post Accident Decommissioning Experience . . . . . . . . . . . . . . . 8-4 8.3 Deconti s s i oni ng Al te rna ti ves . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-6 xiv

TABLE OF C')NTENTS (Continued) 1

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8.3.1 DECON ........................................... 8-6 l 8.3.2 SAFSTOR ......................................... 8-7 8.3.3 ENTOMB .......................................... 8-11 8.4 Environmental Consequences ............................. 8-12 8.5 Comparison of Decommissioning Alternatives ............. E-14 REFERENCES ....................................................... 8-17 9 FU E L RE PROC E SSING P LANT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.1 Description of Fuel Reprocessing Process and Facility .. 9- 1 9.1.1 Process Description ............................. 9-1 9.1.2 Plant Description .............................. 9-3 9.1.3 Estimates of Radioactivity Levels at FRP Shutdown. 9-5 9.2 Fuel Reprocessing Plant Decommissioning Experience ..... 9-5 9.3 Decommissioning Alternatives ........................... 9- 6 9.3.1 DECON ........................................... 9-6 9.3.2 SAFSTOR ......................................... 9- 9 9.3.3 S i te Decommi s s ioni ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-13 9.3.4 S umma ry o f R a d i a t i o n S a f e ty . . . . . . . . . . . . . . . . . . . . . 9-13 9.3.5 Missing 9.3.6 Decommissioning Costs ........................... 9-14 9.4 Envi ronmental Consequences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-17 9.4.1 Wastes .......................................... 9-17

., 9.4.2 Nonradiological Safety Impacts .................. 9-19 9.4.3 Socioconomic Impacts ........................... 9-19 .

9.E Comparison of Decommissioning Alternatives ............. 9-21 REFERENCES ....................................................... 9-22 10 SMALL MIXED OXIDE FUEL FABRICATION PLANT . . . . . . . . . . . . . . . . . . . . . 10-1 10.1 Description of the Reference MOX Fuel Fabrication Plant. 10-1 10.2 M0X Decommissioning Experience ......................... 10-2 10.3 Decommissioning Alternatives ........................... 10-2 10.3.1 DECON .......................................... 10-3 10.3.2 SAFSTOR ........................................ 10-5 10.3.3 ENTOM8 ......................................... 10-6 10.3.4 Summary of Radiation Safety and Decommissioning Costs ............................. ............ 10-7 xv

i TABLE OF CONTENTS (Continued)

P_gte 10.4 Environmental Consequences ........... ................. 10-15 10.4.1 Waste .......................................... 10-15

.10.4.2 Nonradiological Safety ......................... 10-16 10.4.3 Socioeconomic Effects .......................... 10-17 10.4.4 Noise and Aesthetics ......... . ............... 10-18

10. 5 Compari son of Decommi ssioning Al ternative s . . . . . . . . . . . . . 10-19 REFERENCE ........................................................ 10-20 11 URANIUM HEXAFLUORIDE CONVERSION PLANT . . . . . . . . . . . . . . . . . . . . . . . . 11-1 11.1 Uranium Hexafluoride Conversion Plant Description . . . . . . 11-2 11.1.1 Plant and Process Description . . . . . . . . . . . . . . . . . . 11-2 11.1.2 Estimates of Radioactivity Levels at UFs Plant Shutdowr. ....................................... 11-3 11.2 Uranium Hexafluoride Conversion Plant Decommissioning Experience ............................................. 11-5

11. 3 Cecomi ssioning Al ternatives . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-5 11.3.1 DECON .......................................... 11-6 11.3.2 SAFSTOR ........................................ 11-8 11.3.3 ENTOMB ......................................... 11-9 11.3.4 Si te Decomi s s i oni ng . . . . . . . . . . . . . . . . . . . . . . . . . . . 11-9 11.4 Environmental Consequences ............................. 11-9 11.4.1 Wast Disposal .................................. 11-10 11.4.2 Additional Effects of Decommissioning .......... 11-10

11. 5 Comparison of Decommissioning Al ternatives . . . . . . . . . . . . . 11-10 REFERENCE........................................................ 11-12 12 URAN I UM F U E L FAB RI C AT ION P LANT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-1 12,1 U- F ab P l a nt D e s c ri p t i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-1

12. 2 U-Fab Plant Decommissioning Experience . . . . . . . . . . . . . . . . . 12-2

12. 3 Decomi ssioni ng Al te rnative s . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-4 12.3.1 DECON .......................................... 12-4 12.3.2 SAFSTOR (Custodial) ............................ 12-5 12.3.3 ENTOMB ......................................... 12-7 12.3.4 Sumary of Radiation Safety and Decommissioning Costs .......................................... 12-8 xvi

TABLE OF CONTENTS (Continued)

P_ age 12.4 Environmental Consequences ............................. 12-13 12.4.1 Nonradi ol ogical Sa fe ty . . . . . . . . . . . . . . . . . . . . . . . . . 12-13 12.4.2 Commitment of Resources ........................ 12-13 12.4.3 Socioeconomic E f fects . . . . . . . . . . . . . . . . . . . . . . . . . . 12-14 12.5 Comparison of Deccmmissioning Alternatives ...... ...... 12-14 R EF E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-15 13 INDEPENDENT SPENT FUEL STORAGE INSTALLATION . . . . . . . . . . . . . . . . . . 13-1 13.1 Description of an Independent Spent Fuel Storage Installation (ISFSI) ................................... 13-1 13.1.1 Wet Storage ISFSI .............................. 13-1 13.1.2 D ry S to r a ge I S F S I . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-2 13.2 ISFSI Decommissioning Experience ....................... 13-4 13.3 Decommissioning Alternatives ......................... . 13-4 13.3.1 DECON .......................................... 13-5 13.3.2 SAFSTOR ........................................ 13-5 13.3.3 ENTOMB ......................................... 13-6 13.3.4 Summary of Radiation Safety and Oecommissioning Costs .......................................... 13-7

13. 4 Envi ronmental Cons equence s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-9 13.4.1 Waste Disposal ................................. 13-9 13.4.2 Soc i oeconomi c E f f ects . . . . . . . . . . . . . . . . . . . . . . . . . . 13-10

13. 5 Comparison of Decommi ssioning Al ternatives . . . . . . . . . . . . .

13-10 RE F E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13-12 14 NON-FUEL-CYCLE NUCLEAR FACILITIES ............................ 14-1 14.1 Facilities Descriptions ................................ 14-3 14.1.1 Selected Types of Materials Facilities . . . . . . . . . 14-3 14.1.2 Reference Facilities and Sites ................. 14-7 14.2 Non-Fuel-Cycle Materials Facilities Decommissioning l Experience ............................................. 14-9 l

14. 3 Decocmiss ioning Alternatives . . . . . . . . . . . . . . . . . . . . . . . . . . . 14-11 )

l l

1 xvii

TABLE OF CONTENTS (Continued)

P_ag 14.3.1 Decommissioning Alternatives for Non-Fuel-Cycle Facilities ..................................... 14-11 14.3.2 Decommissioning Alternatives for Sealed Source and Radiochemical Manufacturers ................ 14-14 14.3.3 Decommissioning Alternatives for Broad Research and Development Program Facilities.............. 14-18 14.3.4 Decommissioning Alternatives for Processors of Radioactive Ore ................................ 24-20 14.4 Environmental Consequences ............................. 14-21 14.5 Comparison of Decommissioning Alternatives ............. 14-22 REFERENCES ....................................................... 14-24 15 NRC POLICY CONSIDERATIONS .................................... 15-1 15.1 Major Regulatory Particulars . . . . . . . . . . . . . . . . . . . . . . . . . . . 15-3 15.1.1 Decommissioning Alternatives ................... 15-3 15.1.2 Planning ....................................... 15-9 15.1.3 Financial Assurance ............................ 15-12 15.1.4 Residual Radioactivity Levels for Unrestricted Use of a Facility .............................. 15-15 15.1.5 Environmental Impact Statement ................. 1E-16 15.2 Regulations ............................................ 15-16 REFERENCES ....................................................... 15-18 G-1 G L O S S A wu R Y .w.

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APPENDI X A . .f.'.d.</J .c. . .R 0il'.*A*dl. .T-dfMt. .M.'M4.C. . . . . . . . A-1 -

APPENDIX B . b :W d ...k d f.G................................... B-1 xviii

LIST OF TABLES Pag 1.5-1 Summary of Nuclear Reactor Decommissionings. . . . . . . . . . . . . 1-7 1.5-2 Nonreactor Nuclear Facility Decommissioning Information. 1-8 2.4-1 Summary of the Elements of the Decommissioning Alternatives ........................................... 2-6 2.7-1 Classification of Low-Level Decommissioning Wastes From Power Reactors.......................................... 2-27 4.3-1 Summary of Estimated Costs for Decommissioning the Re fe renc e PWR i n $ Mi l l i on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-7 4.3-2 Summary of Radiation Dose Analyses for Decommissioning the Reference PWR ...................................... 4-9 4.3-3 Estimated Ismediate Dismantlement Costs for Plants Smaller Than the Reference PWR, Based on Previously-Derived Overall Scaling Factors . . . . . . . . . . . . . . . . . . . . . . . . 4-16 4.3-4 Estimated Costs and Occupational Radiation Doses for Decommissioning Di f ferent-Si zed PWR Plants . . . . . . . . . . . . . 4-17 4.4-1 Estimated Burial Volume of Low-Level Radioactive Waste and Rubble for the Reference PWR ....................... 4-20 4.4-2 Summary of Radiation Doses to the Maximally-Exposed Individual from Accidental Airborne Radionuclide Releases During Decommi s sioni ng Operations . . . . . . . . . . . . . . . . . . . . . . 4-22

<4.4-3 Estimated frequencies and Radioactivity Releases for Selected Truck Transport Accidents . . . . . . . . . . . . . . . . . . . . . 4-23 ,

5.3-1 Summaar of Reevaluated Decommissioning Costs for the Reference BWR in $ Millions ............................ 5-6 5.3-2 Summary of Radiation Dose Analyses for Decommissioaing the Reference BWR ....................................... 5-9 5.3-3 Estimated Immediate Dismantlement Costs (in millions) for Plants Smaller Than the Reference BWR, Based on Previously-Derived Overall Scaling Factors . . . . . . . . . . . . . ' 5-15 5.3-4 Estimated Costs and Occupational Radiation Doses for Decommissioning Di f ferent-Sized BWR Plants . . . . . . . . . . . . . 5-17 5.4-1 Estimated Burial Volume of Low-Level Radioactive Waste and Rubble for the Ref erence BWR . . . . . . . . . . . . . . . . . . . . . . . 5-20 xix

LIST OF TABLES (Continued)

Page 5.4-2 Summary of Radiation Doses to the Maximally-Exposed Individual from Accidental Airborne Radionuclide Releases during BWR Decommissioning and Trans of Wastes ...................................portation ........... 5-22 6.1-1 Mul tiple-Reactor Station Scenarios. . . . . . . . . . . . . . . . . . . . . . 6-5 6.3-1 Summary of Estimated Cost Reductions When Decomm- sioning Each Reference PWR at a Multiple-Reactor Station ....... 6-8 6.3-2 Summary of Estimated Cost Reductions When Decommissioning Sach Reference BWR at a Multiple-Reactor Station ....... 6-10 6.3-3 Sumary of Estimated Occupational Dose Reductions When Decommissioning Each Reference PWR at a Multiple-Reactor Station ................................................. 6-12 6.3-4 Summary of Estimated Occupational Dose Reduction When Decommissioning Each Reference BWR at a Multiple-Reactor Station ................................................ 6-14 6.3-5 Summary of Estimated Public Dose Reduction When Decommissioning Each Reference PWR at a Multiple-Reactor Station ................................................ 6-16 6.3-6 Summary of Estimated Public Dose Reduction When Decommissioning Each Reference BWR at a Multiple-Reactor Station ................................................ 6-18 7.2-1 Experience With Research and Test Reactor Decommissionings ....................................... 7-4 7.3-1 Summary of Estimated Costs for Oecommissioning the Reference Research Reactor in $ Millions. . . . . . . . . . . . . . . . 7-6 7.3-2 Summary of Estimated Costs for Decommissioning the Reference Test Reactor in $ Millions ................... 7-6 i 7.3-3 Summary of Radiation Safety Analyses for Decommissioning the Reference Research Reactor ......................... 7-7 7.3-4 Summary of Radiation Safety Analyses for Decommissioning the Reference Test Reactor ............................. 7-7 7.3-5 Comparison of Data From Selected Cases of Research Reactor Decommissioning ............. .................. 7-11 xx

LIST OF TABLES (Continued)

Page I 7.4-1 Summary of Radiation Doses to the Maximum-Exp; sed Indivic.al From Accidental Radionuclide Releases During i Decomissioning at the Reference Research Reactor . . . . . . 7-13 i I

7.4-2 Sumary of Radiation Doses to the Maximum-Expesed i Individ.al From Accidental Radionuclide Releases During Decommissioning at the Reference Test Reactor ......... 7-14 8.2-1 Summary of Nuclear Reactor Post-Accident Clear ~p and Decommissioning Experience ............................. 8-5 8.3-1 Summary of Estimated Costs for Decommissioning of the Reference PWR Following Accident Cleanup . . . . . . . . . . . . . . . 8-8 8.3-2 Sumary of Radiation Safety Analysis for Deconmissioning of the Reference PWR Following Accident Cleanuo ........ 8-9 8.4-1 Burial Volume of Radioactive Waste and Rubble for the Reference PWR Following the Accident Cleanup at a Reactor Involved in an Accident . . . . . . . . . . . . . . . . . . . . . . . . 8-13 8.4.2 Sumary of Radiation Doses to the Maximum-Exposed Individaal from Postulated Releases Due to Ind;strial Accidents During Post-Accident Decomissioning . . . . . . . . . 8-15 8.4.3 Summary of Estimated Radiation doses to the Maximum-Exposed Individual from Postulated Transportation Accidents During Decommiasioning . . . . . . . . . . . . . . . . . . . . . . 8-16 9.3-1 Packaging and Shipping Information for Wastes Generated From DECON ............................................. 9-8 9.3-2 Sumary of Radiation Safety Analysis for Decomnissioning the Reference FRP ...................................... 9-10 9.3-4 Summary of Estimated Cos'.s for Decomissioning a Fuel Reprocessing Plant ..................................... 9-16 9.4-1 Radioactive Waste Resulting from Decommissionirg a R e f e r e nc e F R P . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-18 ,

9.4-2 Summary of Nonradiological Safety Impacts . . . .. . . . . . . . . . 9-20 l l

10.3-1 Sumary of Radiation Safety Analyses for Routi e l Decomissioning of the Reference H0X Plant . . . . . . . . . . . . 10-8 10.3-2 Sumary of Radiation Doses to the Maximum-Expesed ,

Individ.al From Accidental Airborne Radionucli:e Releases 1 During tecomissioning Activities . . . . . . . . . . . . . . . . . . . . . 10-9 '

l l

xxi l

_. .- - - . .. ----l

_ LIST OF TABLES (Continued)

Page 10.3-3 Estimated Frequencies, Radioactivity Releases and Doses for Selected Truck Transport Accidents ................. 10-13 10.3-4 Summary of Estimtted Costs for Decommissioning the Reference Small N0X Fuel Fabrication Plant ............. 10-14 10.4-1 Burial Volume of Radioac*.ive Waste and Rubble Resulting from Decommissioning a Reference MOX Plant ............. 10-16 11.5-1 Summary of Decommissioning Alternatives . . . . . . . . . . . . . . . . 11-11 12.3-1 Summary of Radiation Safety Analyses for Routine Decommissioning of the Reference U-fab Plant ........... 12-6 12.3-2 Summary of Radiation Doses to the Maximum-Exposed Individual From Accidental Airborne Radionuclide Releases During Decommissioning Activities for Either Decommissioning Alternative ............................ 12-10 12.3-3 Estimated Frequencies and Radioactivity Releases for Selected transportation accidents ...................... 12-11 L2.3-4 Summary of Estimated Costs for Decommissioning the Reference U-fab Plant .................................. 12-12 13.3-1 Summary of Occupational Radiation Safety Analyses for Routine Decommissioning of the Reference ISFSIS ........ 13-8 13.3-2 Summary of Estimated Costs for Decommissioning the Reference ISFSIS .. .................................... 13-9 14.0-1 NRC Material Licenses as of June 1978 . . . . . . . . . . . . . . . . . . 14-2 .

14.0-2 Agreement State Licenses (June 1978) . . . . . . . . . . . . . . . . . . 14-2 14.3-1 Summary of Estimated Requirements and Costs for DECON of Six Reference Laboratories That Process or Use Radioisotopes .......................................... 14-15 14.3 2 Summary of Estimated Manpower Requirements, Costs, arid Radiation Doses for Decommissioning Three Reference Sites ........................................ 14-17 15.0.1 Summary of Estimated Radiation Doses from Decommissioning Nuclear Fuel Cycle Facilities . . . . . . . . . . 15-4 15.0.2 Summary of Estimated Costs for Decommissioning Nuclear Fuel Cycle Facili ties . . . . . . . . . . . . . . . . . . . . . . . . . . 15-5 ,

xxii

d 4

i I

LIST OF FIGURES P. ag.e 4

2.1-1 Diagram of the Steps in the Nuclear Fuel Cycle. . . . . . . . . . 2-3 4.1-1 Pressurized' Water Reactor............................... 4-3 5.1-1 Boiling Water Reactor................... ............... 5-3 9.1-1 Sirplified Process Flow Diagram of a Fue' Reprocessing Plant .................................................. 9-2 11.1-1 UFs Production - West Solvent Extraction Fluorination Process Simpli fied Block Flow Diagram . . . . . . . . . . . . . . . . . . 11-4 11.1-2 UFs Production - Dry Hydrofluorinaticn Process Simplified Block Flow Diagiam ..................................... 11-4 i

k i

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i j'

i xxiii

1 INTRODUCTION Commercial nuclear facilities that come under the Nuclear Regulatory Commis-  !

sion's (NRC) regulatory authority include those dealing with fuel cycle and i non-fuel-cycle operation. The generation of electric power from steam sup-plied by nuclest reactors requires a series of processes collectively known as the nuclear fuel cycle. This cycle begins with the mining and milling ol ura- I nium ore, includes the operation of power reactors, and ends with the disposi-tion of radioactive wastes. Each step in the cycle requires the handling of radioactive materials, which are specifically designated as cource saterials, byproduct materials, or special nuclear materials. Non-fuel-cycle facilities can also use byproduct, source, and special nuclear materials. Non-fuel-cycle facilities include those involved in academic, pharmaceutical and industrial radioisotopic use and in rare metal ore processing. The handling of these materials and the processes involved have given rise to several issues of funda-sental importance to the American public. These issues include the safe opera-tion of all steps in the nuclear fuel cycle and of other nuclear facilities, especially the safe operation of power reactors; the safe disposition of radio-active wastes; and the safe decommissioning of all nuclear facilities. The first two issues have received much attention from Congress and from federal regulatory agencies, beginning in 1954 with the passage of the Atomic Energy Act. The third issue, decommissioning, is now receiving an increasing amount .

of attention because the nuclear field is maturing, in that nuclear facilities have been operating for a number of years, and the number and complexity of facilities that will require decommissioning is expected to increase in the future. It is this third issue which is the subject of this document.

1.1 Purpose of EIS The purpose of this environmental impact statement (EIS) is to assist the Nuclear Regulatory Commission (NRC) in developing policies and in promulgating amended regulations with respect to the decommissioning of licensed nuclear facilities. It is prepared pursuant to the requirements of the National Environmental Policy Act (NEPA). The decommissioning of uranium mills and mill 10/07/87 1-1 NUO586 CH 1 '

_ . _ . . _ _. _ ~_ _ .__ __. _

l l

tailings, (this includes all fac-lities assochted with extracting uranium from

<,.s

+ aces', such as in situs, heap leach, and milling facilities) low-level waste burial f acilities and high-level waste repositories has been treated in 10 CFR Parts 40, 60 and 61. In addition, also excluded from this action are uranium l mines which come under the jurisdiction of the states and other Federal agencies.

The generic analyses of this EIS are applicable to specific facilities based on  !

the deconmissioning information base studies which included sensitivity analyses )

of such parameters as the size of the facility, contamination level, waste )

disposal costs, labor costs, etc. (See References of Section 1) I 1.1.1 N:PA Requirements Section 102(1) of the National Environmental Policy Act (42 U.S.C. 4321 et seq.) requires that "the policies, regulations, and public laws of the United States shall be interpreted and administered in accordance with the policies set forth in this Act." Section 102(2)(C) requires all agencies of the Federal Government to "include in every recommendation or report on proposals for legis-lation and other major Fede.ral actions significantly affecting the quality of the human environment, a detailed statement by the responsible official on:

(i) the environmental impact of the proposed action, (ii) any adverse environmental effects which cannot be avoided should the proposal be implemented, (iii) alternatives to the proposed action, (iv) the relationship between local short-term uses of man's environment and the maintenance and enhancement of long-term productivity, and (v) any irreversible and irretrievable commitments of resources which would be involved in the proposed action should it be implemented."

10/07/87 1-2 NU0586 CH 1  ;

1

1. 2 Organization of the EIS The first three sections of this EIS contain material comon to all of the facilities discussed in the statement. Regulatory matters are discussed in Section 1. Section 2 discusses in a generic manner the following: nuclear facilities; decommissioning alternatives; acceptable residual radioactivity levels for permitting release of the site for unrestricted use; financial assurance that sufficient funds are available for decomissioning; the manage-ment of radioactive wastes; and safeguards. Facility sites (i.e., the affected environment) are discussed generically in Section 3. Reactor facilities are discussed in Sections 4 through 8. Fuel cycle facilities are discussed in Sec-tions 9 through 13 and non-fuel-cycle facilities in Section 14. These sections include descriptions el each facility, discussions of decomissioning alterna-tives, and sumaries of radiation exposures and decomissioning costs. Other environmental consequences are also discussed. Regulatory policy considerations are discussed in Section 15.

It is intended in this report to provide a document sufficient in detail to be useful to the NRC in establishing policies and in promulgating amended regula-tions, yet not so lengthy or detailed as to be overwhelming to the general public and to others who have a valid interest in the subject. Detailed reports have been prepared which constitute information bases on the technology, safety and costs of decornissioning of the nuclear facilities discussed in this report,l'10 These facilities are pressurized water reactors, boiling water

~

reactors, multiple reactor power stations, research and test reactors, fuel reprocessing plants, small mixed oxide fuel fabrication plants, uranium hexafluoride conversion plants, uranium fuel fabrication plants, independent tpent fuel storage installations, and non-fuelcycle materials facilities. Many of those reports have been available for critical coment for some time, have been found to be useful as a data base, and have been used in preparation of decommissioning studies. The decommissioning of uranius sills and tsilings piles is discussed in a separate EIS.11 The decomissioning o' low-level waste burial facilities is also discussed in a separate EIS.22 1-3 71U0586 CH 1 10/07/87

This E'S represents a compendium of what would otherwise have been many sepa-rate EIS's on the nuclear facilities considered in this report. To make the report more useful to the user, the separate facility sections (Section 4 through 14) were kept as self-contained as possible, so that a user interested in a particular facility type need primarily read only that section, as well as the introduction, the section on generic issues and the section on policy.

Such an approach causes some unavoidable redundancy in presentation of informa-tion contained in the various facility sections. In addition, an overview of this report is presented to enable a user to gain a perspective of the objectives and conclusions reached in this report.

1.3 Purpose of Decommissioning The purpose of decommissioning nuclear facilities is to take tne facility '

safely from service and to reduce residual radioactivity to a level that per-mits release of the property for unrestricted use and termination of license.

Alternative methods of accomplishing this purpose, and the environmental impacts of each alternative are discussed in this EIS.

1.4 Responsibility for Decommissioning The responsibility for decommissioning a commercial nuclear facility belongs to the licensee. Regulatory and policy guidance for decommissioning is the responsibility of the NRC and is implemented either by the NRC or Agreement State as applicable.

1.4.1 Existing Criteria and Regulations for Decommissioning Statutory authority for the regulation of activities related to the commercial nuclear fuel cycle is contained in the Atomic Energy Act of 1954 (42 U.S.C. 2011 et seq.) and the Energy Reorganization Act of 1974 (42 U.S.C. 5841 et seq.) and in subsequent amendments. Pursuant to these acts, the NRC has promulgated regulations which appear in Title 10 of the Code of Federal Regulations. The NRC has also published Regulatory Guides for the purpose of assisting applicants and licensees in carrying out their regulatory obligations.

l 10/07/87 1-4 NUO586 CH 1

Present hegulations specifically pertaining to decommissi ning are contained in 10 CFR Parts 40, 61, and 72 and in Section 50.33(f), Section 50.82, and Appendix F of 10 CFR Part 50. General guidance is contai ed in NRC Regulatory Guides 1.86 and 3.5 (Rev.1) and in NRC staf f guidelines.

1.4.2 Current Rulemaking Activities The NRC is currently developing an explicit overell policy for decommissioning commercial nuclear facilities and acending its regulations in 10 CFR Chapter I to include more specific decommissioning guidance for production and utiliza-tion facility licensees and byproduct, source, and special nuclear material licensees.13 On February 11, 1985, the NRC published a notice of proposed rulemaking on Deconmissioning Criteria for Nuclear Facilities (50 FR 5600).

The proposed amendments covered a number of topics related to decomissioning that would be applicable to 10 CFR Parts 30, 40, 50, 70', and 72 applicants and licensees. These topics included decomissioning alternatives, planning, assurance of funds for decomissioning, environmental review requirements, and residual radioactivity.

1.5 History, Background, and Experience With Decomissionino Facilities identified with the portion of the nuclear fuel cycle between mining and reactor operation, uranium hexafluoride conversion planta and uranium fuel fabrication plants, call for relatively routine decomissioning procedures.

These facilities usually contain low-level radioactivity dich is well confined ~

to the facility. Mixed oxide fuel fabrication plants involve plutonium and thus call for special procedures. Pressurized water reactors, boiling water reactors, fuel reprocessing plants, and spent fuel storage facilities contain high levels of radioactivity that require special precautions and procedures. The differences among research and test reactors that have a variety of functions and the complexity of non-fuel-cycle facilities that handle byproduct, source, or special nuclear materials depend on the activities carried out and the materials handled. However, their problems in decomissioning these facilities are more from the great number and variety, than in any technical difficulties.

10/07/87 1-5 NUO586 CH 1

Since 1960, five licensed power reactors, four demoristration reactors, six licensed test reactors, one licensed ship reactor, and 52 licensed research reactors and critical facilities have been or are being decommissioned by the methods discussed in this EIS. Forty-two research reactors and critical f acil-ities have been dismantled. Only one power reactor, the Elk River demonstra-tion reactor, has been completely dismantled. Three other deiaonstration power reactors of small size have been entombed. The decommissioning status of the more important reactors is listed in Table 1.5-1. Some military reactors are included, while licensed research reactors cnd critical facilities have been omitted.

Decommissioning experience with come of the specific types of facilities is limited, but a broad base of experience with various facilities exists which is generally relevant to the decommissioning of any type of nuclear facility.

A sampling of non-reactor facilities which have been decommissioned is pre-sented in Table 1.5-2.

1-6 NU0$86 CH 1 10/07/87 o

Table 1.5-1 Sunvnary of nuclear reactor de:e vnissionings

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Table 1.5-2 Nonreactor nuclear facility decommissioning information Year Type of Facility Location Oecommissioned Decommissioning 1

Polonium-210 Miamisburg, Ohio 1950 Partial dissan-Facilities (Units tienent; decon-III & IV) taminated to un-restricted re-lease 1e.els Cave Facility Miamisburg, Ohio 1967 Partial e., tomb-(Radium-226 ment, remainder and Actinium- decontaminated 227 Processing to unrestricted Facility) release levels SM Facility (Space Miamisburg, Ohio 1972 Decontaminated Programs Pluto- and placed in nium-238 passive safe Facility) storage (noth-balled) avait-ing final dis-position by DOF Plutonium Filter Los Alamos, NM 1973 Dissantled Facility (Building 12)

Laboratory for Richland, WA 1974 Dismantled Plutonium Criticality Studies (P-11)

Plutonium Physics Los Alamos, NM 1975 Dismantled Study Building -

No 21 10/07/87 1-8 NUC$S6 CH 1 J

REFERENCES

1. R.1. Smith, G.. J. Konzek and W. E. Kennecy, Jr. , Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station, NUREG/CR-0130,. Prepared by Pacific Northwest Laboratory for U.S.

Nuclear Regulatory Commis'ston, June 1978, Addendum, August 1979, Addendum 2, July 1983, and Addendum.3, September.1984.

^

2. H. D. Oak et al. , Technology, Safety and Costs of 0ecoinmissioning a Reference Boiling Water Reactor Power Station, NUREG/CR-0672, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, Jur.e 1980, Addendum 1, July 1983, and Addendum 2, September 1984.

3. K. J. Schneider and C. E. .Jenkins, Technology, Safety and Costs of .

Decossissioning a Reference Nuclear fuel Reprocessing Plant, NUREG-0278, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, October 1977.

~

4. C. E. Jenkins, l .E . S;; Murphy and' K. J. Schneider, Technology, Safety and Costs of Decommissioning a Reference Small Mixed Oxide Fuel Fabrication Plant, NUREG/CR-0129, Prepared by Pacific Northwest Laboratory for U.S.

Nuclear Regulatory Comission, February 1979.

5. h. K. Elder, Technology, Safety and Costs of Decommissioning Reference UFs Conversion Plant, NUREG/CR-1288, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, October 1981.

6. H. K. Elder and O. E. Blahnin, Technology, Safety and Costs of Decommis-sioning a Reference Uranium Fuel Fabrication Plant, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Comission, NUREG/CR-1266, October 1980.

7. N. G. Wittenbrock, Technology, Safety and Costs of Decomissioning Nuclear Reactors at Multiple-Reactor Stations, NUREG/CR-1755, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Comission, January ,

1982.

8. E. S. Murphy, Technology, Safety and Costs of Decommissioning Reference Non-Fuel-Cycle Nuclear Facilities, Prepared by Pacific Northwest Labo ory for U.S. Nuclear Regulatory Commission, NUREG/CR-1754, February 1981. *

9. G. J. Konzek, Technology, Safety and Costs of Decommissioning Reference Nuclear Research and Test Reactors, NUREG/CR-1756, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, March 1982.

10. E. 5. Murphy, Technology, Safety and Cost of Cleanup and Decommissioning at a Reference Pressurized Water Reactor involved in an Accident, NUREG/CR-2601, Prepared by Pacific Northwest Laboratory for the U.S.

Nuclear Regulatory Commission, to be published.

11. Final Generic Environmental Impact Statement on Uranium Milling, U.S.

Nuclear Regulatory Comm'ssion, NUREG-0706, September 1980.

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12. Draf t Environmental Impact Statement on 10 CFR Part 61, "Licensing Require-ments for Land Disposal of Radioactive Waste," NUREG-0782, U.S. Nuclear Regulatory Commission, September 1981.

13. Plar, for Reeval'uation of NRC Policy on Decommissioning of No: lear Facilities, U.S. Nuclear Regulatory Commission, NUREG-0436, Rev.1, December 1978 and iN. Supplemental F<. A< <A.

1, August 1980,e.,gi<W ,yes.r. <o-il,19*C>

c foF4 5' ooint of all referencesstdocuments may be purchased through the U.S.

overnment Printing Office by calling (202) 275-2060 or by citing to the U.S. Government Printing Of fice, P.O. Box 37082, Washington, OC 20013-7082.

Copies may also be purchased from the National Technical Information Service, U.S. Department of Commerce, 5285 Port Royal Road, Springfield, VA 22161. A copy is available for inspection or copying for a fee in the NRC Publig Document Room,1717 H Street HW. , Washington, DC 20555. -

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10/07/87 1-10 NUO586 CH 1

2 GENERIC NUCLEAR FACILITY DECOMMISSIONING CONSIDERATIONS In this section consideration is given to generic items required for implement-ing a decommissioning program for the facilities considered in this EIS.

First, for an overview, a brief discussion is presented of the nuclear fuel ,

cycle for light-water-reactors. Research and test reactors and non-fuel-cycle nuclear facilities are also briefly discussed. Consideration is then given to:

(1) decomissioning alternatives and their advantages and disadvantages, (2) acceptable residual radioactivity levels for permitting release of a decom-missioned nuclear facility for unrestricted access, (3) assurance that funds to pay for decommissioning will be available, (4) waste management for radioactive waste needing to be disposed of during nuclear facility decommissioning, and (5) safeguarding requirements during decossissioning.

2.1 Nuclear Facilities Operational Description 2.1.1 The Nuclear Fuel Cycle A nuclear power plant is a facility designed to generate electricity by utiliz-ing the heat produced by controlled nuclear fission of uranium and plutonium.

This is the desired production step in the fuel cycle. It is preceded by several steps in the fuel cycle in which uranium ore is processed into fuel elements, and is followed by several steps in which fuel removed from the reactor is stored and then either reprocessed to recover usable fuel or disposed of in some manner. The basic steps in the nuclear fuel cycle are shown in Figure 2.1-1. Each box in the diagram represents a separate facility and each arrow represents the transportation of the product between facilities. Spent 10/07/87 2-1 NUO586 CH 2

1 i

fuel is being stored at the reactor sites pending eventual disposal at spent fuel storage facilities or high-level waste repositorie,. j The steps in Figure 2.1-1 for the typical fuel cycle for power plants are described more fully below.

J M lling The uranium ores that are mined and milled in the United States are sedimen-tary deposits in which the uranium occurs as a coating on sand grains. Small quantities of radium and thorium are also found in the ore. The uranium con-tent is only about 1 to 3 kg per tonne (2 to 6 lb per ton). The milling pro-cess dissolves the uranium and separates it from the sand. This involves crushing arid grinding the ore, dissolving the uranium by acid or alkaline leach, and precipitating a semi-refined product, called yellowcake. The tail-ings from this process are r.;ostly sand, but they also include the original quantities of radium, thorium, and other decay products that do not extract with the uranium. The tailings are carried as a slurry to impoundment areas where the water is allowed to evaporate. The tailings are then stabilized to reduce future potential contamination problems.

Conversion The yellowcake is shipped to a conversion plant where it is converted to UFs by one of two processes. One is the "dry" or hydrofluor pro:ess in which the -

yellowcake goes through a series of reduction, hydrofluorination, and fluorina-tion steps in fluidized bed reactors. The cther is a "wet" process in which the yellowcake is first processed to produce a high purity uranium dioxide feed that undergoes reduction, hydrofluorination, and fiuorination.

Enrichment The UFs produced by the conversion process contains about 0.7% 23s0, which must be increased to 2 to 4% prior to fabrication into LWR fuel assemblies.

Enrichment is accomplished by a gaseous diffusion process in wtiich assVFs me'. e-23: UFs molecules, cules pass more readily through a porous membrane than do 10/07/87 2-2 NU0586 CH 2

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thus producing a product strea a that is enriched in 23sVFs. This process is repeated through many such stages until the desired degree of enrichment is attained. The er.riched UFs is then shipped to a fuel fabrication plant.

Fuel Fabrication In the preparatien of LWR fue' . the enriched UFs first undergoes chemical treatment to convert it to UO2 . The U0 2 is mechanically and thermally treated to produce high-density ceramic fuel pellets that are placed in metal fuel tubes. These tubes or rods are then clustered into fuel assemblies for reactor cores.

Reactors A light water reactor (LWR) as used in a power plant utilizes the heat pro-duced by controlled nuclear fission within the fuel assemblies in the reactor core to heat water and generate steam which drives a turbirie generator. There

, are two basic LWR types: the pressurized water reactor (PWR) and the boiling water reactor (BWR). In a PWR the water in the reactor core is kept under pressure to allow heat build-up without boiling. This heated water is circu-lated through a heat exchanger where water in a second circulating system is converted to steam to drive the turbines. In a BWR the water S the reactor core is allowed to boil, directly producing the steam to drive the turbines.

Spent Fuel Storage Facilities The partially depleted LWR spent fuel assemblies are removed from the reactor and stored in spent fuel pools at the reactor for a minimum of 90 days.

This cooling period allows the short-lived radionuclides to decay and reduce the radioactivity and thermal heat emission of the fuel assemblies.

Spent fuel is currently being stored at reactor spent fuel pools for extended time periods as plans for further disposition of the spent fuel are being developed. Storage of spent fuel at away-from-reactor independent spent fuel storage installations (ISFSI) is being considered as an interim measure. One ISFSI design is similar to that of the reactor storage pools except that the 10/07/87 2-4 NUO586 CH 2

storage capacity is significantly greater. An alternative ISFSI desig* is to store the spent fuel in a dry stcrage environment such as an air-coolec vault.

Fuel Reprocessing LWR spent fuel assemblies can be chemically reprocessed to separate the remain-ing uranium and the generated plutonium from the radioactive wastes prc:.:ed during reactor operation. The chemical separation is accomplished by c c: ping the fuel rods into short sections, dissolving the pellets with nitric a:id, extracting uranium and plutonium nitrates from the fission products, and then separating the uranium from the plutonium. The uranyl nitrate is converted to UFs and the plutonium nitrate is oxidized to plutonium dioxide. Both can then be inserted into the fuel cycle for reuse. At the present time no comnercial spent fuel is being reprocessed in the United States.

Mixed Oxide Fuel Fabrication A mixed oxide fuel fabrication plant produces fuel elements that contai, a mix-ture of UO2 and Pu02 . For example, 002 and Pu02 powders are mixed and the mixture is formed into pellets by mechanical and thermal treatment. These pel-lets are sealed in metal cladding to form fuel elements. Only small mixed oxide plants are currently in use commercially and are used tc fabricata experimental fuel elements.

Low-Level Waste Burial Facilities Low-Level radioactive wastes which do not contain transuranic elements above certain concentrations are disposed of in shallow-land burial facilities.

These kinds of materials may be generated at reactors or at any of the facili-ties where fuel is processed, and consist of contaminated trash, filters, and equipment. These wastes are placed in boxas or drums to facilitate handling and are buried at sites that are monitored and are restricted froe public access.

10/07/87 2-5 NUOM6 CH 2

I I

1

@ Level Vaste Repositories High-level wastes are either intact fuel assemblies that are being discarded l af ter serving their useful life in a reactor core (spent fuel) or certain fission product and actinide wastes generated during fuel reprocessing. High-level waste burial at deep geologic repositories is currently under consideration. There are currently no facilities of this type.

2.1.2 Research and Test Reactors A research reactor is defined in 10 CFR 170.3(h) as a nuclear reactor licensed for. operation at a thermal power level of 10 megawatts or less, and which is not a testing facility. A testing facility (i.e., a test reactor) is defined in 10 CFR 50.2 as a nuclear reactor licensed for operation at: (1) a thermal power level in excess of 10 megawatts, or (2) a thermal power level in excess of 1 megawatt if the reactor is to contain: a circulating loop through the core in which the applicant proposes to conduct fuel experiments, or a liquid fuel loading, or an experimental facility in the core in excess of 16 square inches in cross-section. There are 84 nonpower research and test (R&T) reactors in the U.S. that are licensed by the NRC. Of these 76 are research reactors, and 8 are test reactors. The level of activity of these facilities ranges from no longer operational, to occasional use, to intermittent use, to steady and scheduled use.

2.1.3 Non-Fuel-Cycle Nuclear Facili'.ies Non-fuel-cycis facilities are .. facilities which handle by product, source and/or special nuclear materials, but which are not involved in the production of power as outlined in Figure 2.1-1. Non-fuel-cycle facilities must be licensed by the NRC, Precise definitions and licensing requirements for the materials listed above are published in 10 CFR Parts 30, 40, and 70, respectively. Broadly speaking, source materials consist of uranium and thorium, special nuclear materials consist of plutonium or enriched uranium, and byproduct materials consist of saterials made radioactive by special nuclear material. These facil-ities include a wide range of applications in industry, medicine and research such as manufacture of packaged products containing small sealed sources and of 10/07/87 2-6 NV0586 CH 2

radiochemicals, research and development institutions, and processcrs of ores in which the tailings contain licensable quantities of radionuclides.

2. 2 Facilities Considered in EIS The facilities considered in this EIS are: (1) pressurized water reactors, (2) boiling water reactors, (3) multiple reactor stations, (4) research and test rectors, (5) fuel reprocessing plants, (6) small mixed oxide fuel fabrication plants, (7) uranium hexafluoride conversion plants, 8) uranium fuel fabrication plants, (9) independent spent fuel storage installations, and (10) non-fuel-cycle nuclear facilities. The facilities not considered include uranium mills and mill tailings, low-level waste burial facilities and high-level waste repositories because they are covered by separate rulemaking; and uranium mines and the existing government owned uranium enrichment plants

%cau< e they are not under NRC jurisdiction.

2. 3 Definition of Decommissioning Jecommissioning means to remove a nuclear facility safely from service and to reduce residual radioactivity to a level that permits release of the property for unrestricted use and termination of the license. Decommissioning activities do not include the removal and disposal of spent fuel which is considered to be an operational activity or the removal and disposal of nonradioactive structures and materials beyond that necessary to terminate the NRC license. Disposal of nonradioactive hazardous waste not necessary for NRC license termination is not '

covered in detail by this EIS but would be treated by other agencies having responsibility over these wastes as appropriate. 1

2. 4 Decommissioning Alternatives Once a nuclear facility has reached the end of its useful life, it'aust be decommissioned according to the definition contained in Section 2.3. Several .

alternatives are possible, although not all may be satisfactory for all nuclear l

facilities. These alternatives are: no action, DECON, SAFSTOR, and ENTOMB. l The terms DECON, SAFSTOR, and ENTOMB are relatively new in use. in the past, the nomenclature for describing these alternatives has not been consistent.

10/07/87 2-7 N'J0586 CH 2 l

Different documents have of ten used dif ferent terminology when referring to the 1

same decommissioning alternative, thus causing some confusion. In the interest I of ending the confusion, this section lists the following definitions of the major aecommissioning alternatives and the following pseudoacronyms to clearly I delineate each alternative:

DECON is the alternative in which the equipoent, structures, and portions of the f acility and site containing radioactivv contaminants are removed or decon-taminated to a level that permits the property to be released for unrestricted use shortly after cessation of operations.

SAFSTOR is the alternative in which the nuclear facility is placed and maintained in a condition that allows the nuclear facility to be safely stored and subse-quently decontaminated (deferred decontamination) to levels that permit release for unrestricted use.

ENTOMB is the alternative in which radioactive contaminants are encased in a structurally long-lived material, such as concrete; the entombed structure is appropriately maintained and continued surveillance is carried out until the radioactivity decays to a level permitting release of the property for unre-stricted use.

Table 2,4-1 presents a summary of the various activities that will be in effect during DECON, SAFSTOR and ENTOMB.

Conversion to a new or modified use is also considered. Conversion, however, is not considered to be a decommissioning alternative whether the new use involves radioactivity or not. If the intended new use involved radioactive material and, thus was under NRC licensing authority, an application for license renewal or amendment or for a new license would be submitted and reviewed according to appropriate existing regulations. If the intended new use does not involve radioactive materials, i.e. , unrestricted public use, then such new use would be contingent on prior decommissioning and termination of license. As such, it would have to use one of the decommissioning alternatives indicated above, namely DECON, SAFSTOR, or ENTOMB. In this case, the new use 10/07/87 2-8 NV0586 CH 2

Table 2.4-1 Summary of the elements of the decommissioning alternativea Comments, Facility / Site Use ElementsI *) Facility Status Equipment - removed if radioactive Fact;ity - Unrestricted use reaching Decontamination [to permissible levels levels permitting Continuing Care Staff - none unrestricted use Security - none Site - Unrestricted use after of the facility] Environmental Monitoring - none reaching permissible levels Radiosctivity - removed Surveillance - none Structures - removal optional Safe Storage Custodial Equipment - some operating Safe storage alone is not an (Layaway) Continuing Care Staff - some required acceptable decommissioning mode; Security - continuous it must be followed by decon-Environmental Monitoring - continuous tamination to unrestricted use.

Radioactivity - confined Surveillance - continuous Facility - Nuclear Only Structures - intact Site 'liuclear Only Passive Equipment - none operating Facility - Nuclear Only Continuing Care Staff - optional (onsite) - Site - Conditional Non-nuclear routine inspections Security - remote alarms Environmental Monitoring - routine periodic Radioactivity - immobilized /sometimes sealed Surveillance - periodic Structures - intact

+

Table 2.4-1 (Continued)

Elements (*) Facility Status Comments, Facility / Site Use Hardened Equipment - none operating Facility - Conditional Non-nuclear Continuing Care Staff - none on site Site - Conditional Non-nuclear Security - hardened barriers, fencing and posting Environmental Monitoring - infrequent Radioactivity - hardened sealing Surveillance - infrequent Structures partial removal optional Entombment Equipment - some removed, the rest encased in Facility - Unusable for an extended concrete time period Site - unrestricted Site - Unrestricted use Continuing Care Staff - none Security - hardened barriers Environmental Monitoring - infrequent Radioactivity - encased in concrete Surveillance - infrequent Structures - intact

• Elements are the specific activities involved in each of the decommissioning alternatives, e.g., SAFSTOR is made up of the following elements: preparation for safe storage, safe storage and decontamination.

e

except as it affects the decommissioning alternative chosen. For these reasons, conversion to a new or modified facility is not considered further in this EIS.

2.4.1 No action The objective of decommissioning is to restore a radioactive facility to a condition such that there is no unreasonable rit,k from t?e decommissioned facility to the public health and safety. In order to e sure that at the end of its life the risk from a facility is within acceptable bounds, some action is required, even if it is as minimal as making a terminal radiation survey to verify the radioactivity levels and notifying the NRC of the results of the survey. Thus, independent of the type of facility and its level of contami-nation, No Action, implying that a licensee would simply abandon or leave a facility after ceasing operations, is not a viable decommissioning alternative.

Therefore, because no action is not considered viable for any facility discussed in this EIS, this alternative is not considered further in this report.

2.4.2 DECON DECON is the alternative in which the equipment, structures, and portions of a facility and site containing radioactive contaminants are removed or decontami-nated to a level that permits the property to be released for unrestricted use shortly af ter cessation of operations. DECON is the only one of the decommis-sioning alternatives presented here which leads to termir.ation of the facility license and release of the facility and site for unrestricted use shortly after '

cessation of facility operations. DECON is estimated to take from fairly short time periods for small facilities to up to approximately 6 years for a large LWR.

Because all of the DECON work is completed within a few months or years following shutdown, personnel radiation exposures are generally higher than for other decommissioning alternatives which spread the decommissioning work over longer time periods thus allowing for radioactive decay. Similarly, larger commitments of money and waste disposal site space are also required for DECON in a relatively short time f rame compared to the other alternatives.

2-11 NUO586 CH 2 10/07/87

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1 i

Thus, the primary advantage of DECON, which is termi1ating the facility license l and making the facility and site available for some other beneficial use, is accomplished at the expense of larger initial commitments of money, personel radiation exposure, and waste disposal site space than for the other alter-natives. Other advantages of DECON include the availability of a work force highly knowledgeable about the facility and the elimination of the need for long-term security, maintenance and surveillance of the facility which weald be required for the other decommissioning alternatives.

In DECON, nonradioactive equipment and structures need not be torn dowr, or removed as part of a decontamination procedure for termination of the NRC license and release for unrestricted use. Once the radioactive facility structures are decontaminated to radioactivity levels permitting unrestricted use of the facility, they may either be put to some other use or demolished at the owner's option.

2.4.3 SAFSTOR SASTOR is the alternative in which the nuclear facility is placed (preparation for safe storage) and maintained in a condition that allows the nuclear facility to be safely stered (safe storage) and subsequently decontaminated to levels that permit release for unrestricted use (deferred contamination). SAFSTOR consists of a short period of preparation for safe storage (up to 2 years af ter final reactor shutdown), a variable safe storage period of continuing care consisting of security, surveillance, and maintenance (up to 60 years af ter final shutdown depending on the type of facility), and including a short period of deferred decontamination. Several subcategories of SAFSTOR are possible:

1. Custodial SAFSTOR require: a minimum cleanup and decontamination effort initially, followed by a period of continuing care with the active protec-tion systems (principally the ventilation system) kept in service through-out the storage period. Full-time onsite surveillance by operating and security forces is required to carry out radiation monitoring, to maintain the equiptent, and to prevent accidental or deliberate intrusion into the 10/07/87 2-12 NUO586 CH 2

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facility and the subsequent exposure to radiation or the dispersal of radioactivity beyond the confines of the facility.

2. Passive SAFSTOR requires a more comprehensive cleanup and decontamination effort initially, sufficient to permit deactivation of the active protec-tive (ventilation) system during the continuing care period. The struc-tures are strongly secured and electronic surveillance is provided to detect accidental or deliberate intrusion. Periodic monitoring and maintenance of the integrity of the structures is required.

3. Hardened SAFSTOR requires comprehensive cleanup and decontamination and the construction of barriers around areas containing significant quantities of radioactivity. These barriers are of suf ficient strength to make acci-dental intrusion impossible and deliberate intrusion extremely difficult.

Surveillance requirements are limited to detection of attack upon the barriers, to maintenance of the integrity of the structures, and to infrequent monitoring.

All categories of safe storage require some positive action at the conclusion of the period of continuing care to release the property for unrestricted use -

and terminate the license for radioactive materials. Depending on the nature ci *be nuclear facility and its operating history, the necessary action can range from a radiation survey that shows that the radioactivity has decayed and the property is releasable, to dismantlement and removal of residual radio-active materials. These latter actions, whatever their scale, are generically '

identified as deferred decontamination.

SAFSTOR is used as a means to satisfy the requirements for protection of the public while minimizing the initial commitments of time, money, occupational radiation exposure, and waste dispesal space. In addition, SAFSTOR may have some advantage where there are other operational nuclear facilitiek at the same site, and may also become necessary in other situations if there is a shortage of radioactive waste disposal space offsite. Modifications to the facilities are limited to those which ensure the security of the buildings against intruders, and to those required to ensure containment of radicactive or toxic material.

It is not intended that che facilities will ever be reactivated. In highly 10/07/87 2-13 NUO586 CH 2

contaminated facilities ano/or facilities with large amounts of activation pro-dets, there is the potential for incurring larger occupational radiation expo-s res if complete decontamination is performed immediately af ter shutcown (DECON).

bever, as a result of radioactive decay of this contamination, reductions in personnel exposure and simplifications in the complexity of operations can be achieved by deferring major decontamination efforts for a number of years. Also, because many of the contamination and activation products present in the f acility will have decayed to background levels af ter a lengthy storage period, the volume of material that must be packaged for disposal will be redaced.

The reduced initial effort (and cost) of the preparation of safe storage is tempered somewhat by the need for continuing surveillance and physical security to ensure the protection of the public. Electronic surveillance devices, which are presently available, could be in service fulltime, with offshift readouts in a local law enforcement office or private security agency. These devices which monitor for intruders, increases in radiation levels, and detection of fires will require periodic checks and maintenance.

Maintenance of the facility's structures and an ongoing program of environmen-tal surveillance are also necessary. The duration of the storage and surveil-lance and dismantlement period can vary from a few years to up to 60 years depending on the type of facility. If SAFSTGR is used, the decision on the length of the safe storage period will be made by the facility owner, with the approval of the NRC, based on consideration of factors including desirability of terminating the license, radiation dose and waste volume reductions, availabilit'y of waste disposal capacity, and other site specific factors affecting safety, such as presence of other nuclear facilities at the site. Similarly, the decision on the extent of decontamination during the period cf preparation for safe storage, and the resultant subcategory of SAFSTOR to be used, depends upon safety considerations and the planned length of the storage and surveillance pe riod. If for example, 80Co is the controlling source of occuoational exposure, a chemical decontamination campaign achieving a decontamination factor (DF) of 10 (i.e. , radinactivity levels reduced to 1/10 of original) will result in approximately the same dose reduction as a decay period of 17 years.

10/07/87 2-14 NUO586 CH 2

At the end of t le period of safe storage, several things will remain to be done before the facility can be released for unrestricted use. In most cases, radio-activity in some areas within the facility will be significantly above levels acceptable for unrestricted release of the facility, necessitating the removal, packaging and disposal of selected m&terials at a regulated disposal site. If the safe storage period is sufficiently long, radioactive materials in the facility ray have decayed to levels low enough to permit the facility to be released for unrestricted use without additional decontamination. This would not apply in the case of a reactor, if the reactor had been operated long enough to produce significant amounts of the long-lived isotnpes 58 Ni and SiNb.

Deferred decontamination, even for a major facility such as a LWR, is a relatively straight-forward disassembly job com;licated by whatever radio-activity remains. Removal and transport of the materials containing the radio-activity to a disposal site are the principal task.s that must be completed.

Further action following termination of the NRC license and release for unre-stricted use, such as disassembly of the various non-radioactive systems and use or demolition of the buildings, would be at the owner's discretion.

A disadvantage of SAFSTOR is the potential lack of personnel familiar with the facility at the time of deferred decontamination. More time and training would be needed. One potential solution to this probies would be the establishment of companies specializing in the decommissioning of nuclear reactor power station and other nuclear facilities. Other disadvantages include the fact '

that the site is tied up in a non-useful purpose for extended time period, regulatory uncertainties in the future, and the continuing need for maintenante, security and surveillance.

2.4.4 ENTOMB ENTOMB is the alternative in which radioactive contaminants are encased in a structurally long-lived material, such as concrete; the entombed structure is appropriately maintained and continued surveillance is carried out until the radioactivity decays to a level permitting release of the property for l

l l

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s. restricted use. ENTOMB is intended for use where the residual radioactivity will decay to levels permitting unrestricted release of the f acility within reasonable time periods (i.e. , within the time period of continued structural integrity of the entombing structure as well as confidence in the reliability of continued radioactivity containment and access restriction, perhaps the order of 100 years). However, a few radioactive isotopes found in fuel reprocessing plants, nuclear reactors, fuel storage facilities, and mixed oxide facilities have half-lives in excess of 100 years and the radioactivity will not decay to levels permitting release of the facilities for unrestricted use within the foreseeable lifetime of any man-made structure. Thus, the basic requirement of continued structural integrity of the entombment cannot be en-sured for these facilities, and ENTOMB would not be a viable alternative in these circumstances. On the other hand, if the entombing structure can be expected to last many half-lives of the most objectionable long-lived isotope, then ENTOMB becomes a viable alternative because of the reduced occupational i and public exposure to radiation. However, even in these circumstances, one of the difficulties with ENTOMB for any complex structure such as a reactor is that the radioactive materials remaining in the entombed structure would need to be characterized well enough to be sure that they will have decayed to acceptable levels at the end of the surveillance period. If this cannot be done adequately, deferred decontamination would become necessary, which would make ENTOMB more difficult and costly than DECON or SAFSTOR. Some method would have to be provided to demonstrate that the entombed radioactivity will decay to levels permitting release of the property for unrestricted use within the order of 100 years, which would be difficult. ENTOMB does, of course, contri-bute to the problems associated with increased numbers of sites dedicated for very long periods to the containment of radioactive materials.

2.5 Residual Radioactivity Levels for Unrestricted Use of a Facility Decommissioning requires reduction of the radioactivity remaining in the facil-ity to residual levels that permit release of the facility for unrestricted use and NRC license termination.

The Commission is participating in an EPA organized interagency working group which is developing Federal guidance on acceptable re:,idual radioactivity levels 2-16 NUO586 CH 2 10/07/87

for unrestricted use. Proposed Federal guidance is anticipated to be published

, by EPA. NRC is planning to implement this guidance through rulemaking as soon  !

as possible. The selection of an acceptable level is outside the scope of r alesaking supported by this EIS. Currently, criteria for residual contamina-tion levels do exist and research and test reactors are being decommissioned using present guidance contained in Regulatory Guide 1.865 for surface con-tamination plus 5 pr/hr bove background as reasured at 1 meter direct radia-tion. The NRC provided such criteria in letters to Stanford University, dated 3/17/81 and 4/21/82 providing "Radiation criteria for release of the dismantled Stanford Research Reactor to unrestricted access." The cost estimate for decommissioning can be based on current criteria and guidance regarding residual radioactivity levels for unrestricted use. The information in the studies by Battelle Northwest Laboratory and Oak Ridge National Laboratory on decommission-ing have indicated that in any reasonable range of residual radioactivity limits,

'the cost of decommissioning is relatively insensitive to the radioactivity level and use of cost data based on current criteria should provide a reasonable estimate.

For example, in ORNL studies t .2 for a PWR, certification surveys at realistic dose values 10 and 25 mrem / year were considered. It was indicated that a survey for the 10 mrem / year value was considered to be well within technical capability and could be done for a cost of approximately $250,000 (i.e. , less thar, about 0.6% of estimated PWR decommissioning costs); and a survey for the 25 mrem / year value is estimated to cost not much less than that for 10 mrem / year (about

$225,000). '

There should be no significant additional decontamination effort required as a result of the termination survey, perhaps only cleanup of a few hot spots indicated by the survey. This is because the extensive efforts required to de-contaminate the highly contaminated facility to low radioactivity level will result in residual radioactivity levels well below the limits which permit unre-stricted release of the facility. It is also the case because spot surveys will be carried out periodically during the decommissioning period so that at the time of the termination survey the licensee is confident that decontamination efforts have achieved the acceptable residual radioactivity levels in most instances. Thus, because there should not be significant additional 10/07/87 2-17 NUO586 CH 2

decontamination necessary af ter completion of the terminatic survey, the major cost and ef fort expected for verifying the required resid.a' radioactivity levels for unrestricted facility use should come from the ce-tification survey.

As indicated above for the PWR example, these survey costs are expected to be a small fraction of the total decommissioning cost, and thus the effort to certify that the facility is available for unrestricted use should not add significantly to the overall decomissioning cost.

In addition, cost-benefit considerations are involved in tre evaluation of the extent of facility decontamination necessary to reduce radioactive contamina- >

tion to levels considered acceptable for releasing the facility for unrestricted use. As is discussed by PNL in NUREG/CR-0130,8 and in NUREG/CR-0278,4 and as is also inherent in the reports prepared by PNL for the other nuclear facilities discussed in this EIS, the cost of decontamination of a facility and thus its decommissioning cost, is essentially independent of the level to which it must be decontaminated as long as that level is in the range of 10 to 25 mrea/yr to an exposed individual. This is because, as indicated above, it is expected that the extensive efforts required to decontaminate the highly contaminated facility to low radioactivity levels will result in residal radioactivity level well below the limits to permit release of the facility fer unrestricted use.  :

An additional cost-benefit consideration relates to decontamination of rooms which are mildly contaminated with radioactivity. Most rooms should not be  !

mildly contaminated with radioactivity in excess of levels which are acceptable for unrestricted facility use since it is assumed that good housekeeping and ALARA practices will be used during facility operations te control the spread of contamination. In areas where there is mild contamination, techniques such as having previously painted surfaces should make decontamination easier and less costly. A source of data for the evaluation of cost for decontamination of mildly contaminated rooms is in NUREG/CR-17546 which evaluates decontamina-tion of a number of specific components. As an example, for a hot cell contami-nated with Cs-137, the manpower needed for decontamination would be approximately 5 man-days and the associated costs would be approximately $5,000. Costs for  :

decontamination of other specific components would be about the same order.

These costs for decontamination of specific mildly contaminated components are small in comparison to the overall decommissioning costs. T.erefore, based on the above discussions, while cost-benefit is a consideration, it is not expected 10/07/87 2-18 NU0586 CH 2

have a major impact on the GEIS ressits concerning reactor or most nonrea: tor a:omissionings.

E.en in situations where the residual radioactivity level night have an effect o- decomissioning cost, by use of update provision in the rulemaking it is expected that the decommissioning fund available at the end of facility life v il approximate closely the actual ccst of decommissioning.

It is imperative that these decommissioning rule amendments in 10 CFR Parts 30, 40, 50, 70, and 72 be issued at this time because it is important to establish financial assurance provisions, as well as other decommissioning planning pro-visions, as soon as possible so that funds will be available to carry out decommissioning in a manner which protects public health and safety. Based on this need for the decommissioning rule and provisions currently existing and those contained in the rule amendments, the Comission believes that the rule can and should be issued now.

2.6 Financial Assurance The primary objective of the NRC with respect to decommissioning is to protect the health cad safety of the public. An important aspect of this objective is to have reasonable assurance that, at the time of termination of facility oper-ations, adequate funds are available to decommission the facility in a safe and timely manner resulting in its release for unrestricted use, and that lack of "

funds does not result in delays in decoamissioning that may cause potential health and safety problems for the public. The need to provide this assurance arises from the fact that there are uncertainties concerning the avai' ability of funds at the time of decomissioning. The nuclear facility licensee has the responsibility for completing decommissioning ir, a manner which protects public health and safety. Satisfaction of this objective requires that the licensee provide reasonable assurance that adequate funds for performing decommissioning vill be available at the cessation of facility operation.

I 2-19 NUO586 CH 2 10/07/87

2.6.1 Present Regulatory Guidance Fresent regulatory requirements concerning the degree cf financial assurance required of a licensee are not specific enough. 10 CFR 50.33(f) requires that, except for an electric utility applicant for a license to operate a utilization facility, an applicant for a production or utilization facility operating license demonstrate financial capability both to operate the facility and to shut it down and maintain it safely. 10 CFR 50, Appen:ix F, requires the applicant for a fuel reprocessing plant operating license to demonstrate his financial qualifications "to provide for removal and disposal of radioactive wastes during operation and upon decommissioning." 10 CFR 72 requires an appli-cant for a license for an independent spent fuel storage installation to provide information on funding for decommissioning. These reg;1ations do not contain sufficient criteria for assuring funds for decommissioning the facilities covered by this EIS.

2.6.2 Implementation of Financial Assurance Requirements In providing reasonable assurance that funds will be available for decommis-sioning, there are several possible financing mechanisas, outlined below, which are available to applicants and licensees. The many different types of nuclear facilities present a wide diversity in the cost of decome t .ioning, in the risk that decommissioning funds might be unavailable, and in the licensees' finan-cial situations. This diversity necessitates that the NRC allow latitude in the implementation of these financing mechanisms. For example, the situation '

for a large power reactor can be significantly different from that for a small research or testing facility or for a materials license. Generally, for a power reactor, state utility commissions regulate retail rates and the Federal Energy Regulatory Commission regulates wholesale rates, thus permitting utilities to recover the cost of providing electricity from their customers, the decommis-sioning costs are higher than for small facilities, and the licensees are required by 50 CFR 10.54(w) to carry substantial levels of insurance for post-accident decontamination and cleanup. This is significantly different than the situation for a small non-fuel-cycle facility which is not rate regulated and has low decommissioning costs.

10/07/87 2-20 NUO586 CH 2

In analyzing funding methods, the NRC has developed the follow ng major classification of funding alternatives.

(1) Prepayment - The deposit prior to the start of operation into in account segregated from licensee assets and outside the licensee's administrative control of cash or liquid assets such that the amount of funds would be sufficient to pay decommissi:ning costs. Prepayment could be in '.*e form of a trust, escrow account, government fund, certificate of deposit, or deposit of government securities.

(2) Surety bonds, letters of credit, lines of credit, insurance, or other guarantee methods - These mechanisms guarantee that the decommissioning costs will be paid should the licensee default. The licensee still must provide funding for decomissioning through some other method. It appears questionable that surety methods of the size necessary and for the time involved with power reactors will be available. However, they appear to be available for facilities that involve smaller costs and periods. The contractual arrangement guaranteeing the surety methods, insurance, or guarantee must include provisions for insuring that these methods will in fact result in funds being available for decomissioning. It should be kept in mind that sureties would only be called if at the time of cessa-tion of facility operation or impending discontiauance of surety by the guarantor, licensee decommissioning funds were inadequate or unavailable.

(3) External sinking funds - A fund established and maintained by setting funds' aside periodically in an account segregated from licenset assets and out-side the licensee's administrative control in which the total amount of funds would be sufficient to pay decomissioning costs at the time termina-tion of operation is expected. An external sinking furd could be in the form of a trust, escrow account, government fund, certificate of depc. sit, or deposit of government securities. The weakness of the sinking fund approach is that M the event of premature closure of a f acility the decourissioning fund wed 1,e insufficient. Therefore, the sinking fund would have to be supplemented 'uy insurance or surety bonds, or letters or line, of credit or other guarantee methods of ites (2).

2-21 NUO536 CH 2 10/07/87

i (4) Internal reserve or unsegregated sinking fund - A fund established and maintained by the periodic deposit or credit.ng of a prescribed amount into an account or reserve which is not segregated from licensee assets and is withir tha licensee's administrative control in which the total amount of the periodic deposits or funds reserved plus accumulated earnings would be i sufficient to pay for decomissioning at the time termination of operation is expected. In this mechanism, the funds are not segregated from the )

utility's assets, rather they may be invested in utility assets and, at the end of facility life, internal funds are used to pay for decommission-ing by, for example, issuance of bonds against licensee assets and the funds raised are used to pay for decommissioning. An internal reserve may also be in the form of an internal sinking fund which is similar to an external sinking fund except that the fund is held and invested by the licensee.

Such a mechanism is generally considered to be less expensive in terms of.

net present value than the options listed above, although, as discussed below, whichever funding mechanism is used should not have a significant impact on the revenue requirements. The problem with the internal or unsegregated funding method is the lesser level of assurance that funds will be available to pay for decommissioning than the other mechanisms because this method depends on financing internal to the licensee, and therefore, is vulnerable to events that undennine the financial solvency of a utility.

The NRC has considered the use of all of these methods, and in particular internal reserve, in several documents. These include.NUREG-0584, Revs. 1-3, -

"Assuring the Availability of Funds for Decommissioning Nuclear Facilities,"7 ,

NUREG/CR-1481, "Financing Strategies for Nuclear Power Plant Decommissioning,"8 and NUREG/CR-3899, "Utility Financial Stability and the Availability of Funds for Decomissioning"9 In addition, the Comission held a meeting soliciting pelic and industry views of decomissioning on September 18, 1984 and the NRC staff has reviewed coments in the area of financial assurance submitted on NUREG-0586, "Draft Generic Environmental Impact Statement on Decomissioning Nuclear Facilities" and submitted in response to the preposed rule on decom-missioning (50 FR 5600)20 10/07/87 2-22 NUO586 CH 2

4' i I

4 The NRC has evaluated the acequacy of various funding methods in light of financial problems encountered by some utilities which, faced witt lower growth in electricity demand than tney projected and rapidly increasing costs of con-struction, had been forced to cancel nuclear plants in advanced stages of con-struction and the ramifications these cor.ditions could have on a utility's ultimate ability to pay for decommissicning. Details of this evaluation are

. 3 9

~

rontained in NUREG/CR-3899 , prepared by an NRC consultant, Dr. J. Siegel of the Wharton School, University of Pennsylvania. Dr. Siegal evaluated one ptblicity owned and four investor-owned utilities then experiencirg severe financial stress and also addressed the issue of what would occur if these or future utilities were to face bankruptcy.

The analysis in NUREG/CR-3859 indicated that the tnarket value of the utilities, even those involved in the most extreme financial crisis is still far in excess of decommissioning costs. Therefore, NUREG/CR-3899 concluded that even in the most severe instances, the value of remaining assets, both tangible and intangi-ble, are more than adequate to cover future projected decomissioning costs.

Based on this analysis, NUREG/CR-38S9 concluded that from an econceic and finan-cial standpoint, any method of funding decomissioning, i.e. , external reserves or internal reserves, is acceptable and provides excellent assurance of the availability of funds.

In addition, NUREG/CR-3899, Supp.1,9 concludes that, even with the threat of non-recoverable or cancelled plants, investors believe that the ongoing value of the utility industry based on future prospects is substantial and increasing,'

that these utilities can, if necessary, attract outside funds, and that because of the substantial margin between the value of the securities of the utilities and the cost of decomissioning, owners and investors in the utility cannot walk away from the financial responsibility of decomissioning without for-feiting the values of their securities.

With regard to concerns regarding insolvency, and liquidity of assets, NUREG/

CR-3899 and Supplement 1 indicate that it is not necessarily true that bank-ruptcy of a utility is tantamount to default on decomissioning ebligations.

In a bankruptcy involving liquidation of assets, there may be sone question as to the priority of carrying out decommissioning under current ba-kruptcy law.

2-23 NUO586 CH 2 10/07/87

Hcwever, the more likely situation with regard to a utility facing bankruptcy would be some form of reorganization of the company. Because electric service is considered an essential service, there will, of necessity, be a successor to an insolvent utility. A successor will retain the obligation to decommission.

The market value of the utility apportioned to its successor would be greater than the cost to complete the decommissioning.

Based on the analysis, NUREG/CR-3899, Supp.19 reaches the following con-clusions:

1. The financial health of utilities, especially those involved in substantial nuclear construction, has substantially improved over the past eighteen months. Recent rulings of the public utility commissions indicate that even after substantial write-offs of nuclear plants are made, investors perceive substantial value in the remaining assets of the utility and can obtain funds without difficulty for decommissioning.

2. Therefore, from a financial standpoint, internal reserves currently provide sufficient assurance of the availability of funds for decommissioning and should be permitted, as proposed by the NRC on February 11, 1985.

In summary, NRC has considered the analysis of NUREG/CR-3899, Supp.1, as well as the documents discussed above. NRC has also considered pertinent fa". ors affecting funding of decommissioning by electric utilities such as the feo

: that they are regulated entities providing a basic necessity of modern life, their long history of stability, and the situation which may occur in an actual bankruptcy, and the requirements that utilities maintain over one billion ciollars of property in.curance which reduces one of the major threats to utility solvency. Based on th.se considerations, it is the Commission's conclusion that the internal reserve method currently allowed by the proposed rule provides a reasonable level of a:,su snce of the availability of funds and that even in the unlikely event of utility bankruptcy, there is reasonable assurance that a reactor will r.ot become a risk to public health and safety.

I I

Whatever funding mechanism is used, its use requires establishing the cost required for decommissioning a facility. This cost should be included as part l

10/07/87 2-24 NUO586 CH 2 i

- - - - _ . _ . . _. _ . , , , _ . . _ , __ _ _ _ .__ . _ , . , _l

of financit.1 provisions submitted by an applicant prior to facility commission-ing. To minimize administrative ef fort while still maintair.ing reasonable assurance of funding, for certain facilities the financial provisions may be based on setting aside an amount which is at least equal to amounts prescribed in the NRC regulations. These amounts vary for the different facilities covered by the regulations.

As information on decornissioning costs become more definitive in time, due to technology improvements, enhanced decommissioning experience, and inflation /

deflation cost factors, a licensee's funding provisions should be updated. In this way, it is expected that the decommissioning fund available at the time of facility shutdown will not differ significantly from actual costs of decommissioning.

It is difficult to accurately estimate what the projected costs for the various funding mechanisms will be at the time of decommissioning. Based on Battelle cost analyses 3'11 presented in this EIS, for the generic PWR and BWR 1175 MWe reactors, decommissioning costs have been estimated at approximately $105 and

$135 million respectively. These estimates do not include the costs of demolition of nonradioactive systems or structures beyond that necessary to ter-minate the NRC license or the cost of site restoration. This results in a cost of a few tenths of a sill (0.1 cent) per kilowatt-hour when averaged over the expected 30 yect reactor operating life. The $105 million cost, while not insignificant, is only a small amount compared to PWR operating capital, perhaps comparable to the cost of a full core reload. Furthermore, whichever funding ~

mechanism used should not have a significant impact on the cost to consumers.

One studys has estimated that the difference in cost between the various funding mechanisms would result in less than a 1% difference in the total bill of a representative utility customer.

In summary, the NRC objective of protecting the public health and hafety requires that there be reasonable assurance of funds for decommissioning.

There should not be any significant financial burden on the applicant in pro-viding a funding mechanism for decommissioning costs either through prepayment, surety bonds, a sinking fund, insurance, or some combination thereof.

10/07/87 2-25 NU0586 CH 2

2.7 Management of Radioactive Wastes and Interim Storage During the decommissioning of a nuclear facility radioactive waste which was generated during the faci.lity operating lifetime must be disposed of at waste disposal sites. These wastes include equipment and structures made radioactive both by neutron activat',n and by radicactive contaminants, incluoe radioactive wastes resulting from chemical decontarination of the facility, and include miscellaneous cleaning equipment.

Disposal of these wastes is covered by existing NRC and other applied Federal and State regulations and is beyond the scope of the rulemaking action supported by the EIS. Disposal of spent fuel will be via geologic repository pursuant to requirements set forth in NRC's regulation 10 CFR Part 60. Disposal of low-level wastes is covered under NRC's regulation 10 CFR Part 61. Because low-level wastes cover a wide range in radionuclide types and activities,10 CFR Part 61 includes a waste classification system that establishes three classes of waste generally suitable for near-surface disposal: Class A, Class 8, and

. Class C. This classification system provides for successively stricter disposal requirements so that the potential risks from disposal of each class of waste are essentially eq'rivalent to one another. In particular, the classi-fication system limits to safe levels the concentrations of both short- and long-lived radionuclides of concern to low-level waste disposal. The radio-nuclides considered in the waste classification system of 10 CFR Part 61 include long-lived activation products such as Ni-59 or Nb-94, as well as .

"intense emitters" such as Co-60.

Wastes exceeding Class C limits are considered to be not generally suitable for near-surface disposal, and those small quantities currently being generated are being safely stored pending development of disposal capacity. The recently enacted Low-Level Rac'ioactive Waste Policy Amendments Act of 1985 (Pub. L. ,99-240, approved January 15,1986, 99 Stat.1842) provides that disposal of l wastes exceeding Class C concentrations is the responsibility of the Federal government. The Act also requires a report by DOE t o s with recom- l mendations for safe disposal of these wastes. DOE p M!'c M d this report, l

"Recommendations for Management of Greater than Class C Low-Level Radioactive Waste," DOE /NE-0077, in February 1987.

10/07/87 2-26 NUO586 CH 2

}

As far as decommissioning wastes are concerned, technical studies coupled with practical experie ce from decommissioning of small reactor units indicate that wastes from future decommissionings of large power reactors will have very similar physical and radiological characteristics to these currently being generated from reactor operations. Two of the studies performed by NRC include NUREG/CR-0130, Addendum 3,3 and NUREG/CR-0672, Addendum 2,11 which specifically address classification of wastes from decommissioning large pressurized water reactor (PWR) and large boiling water reactor (BWR) nuclear power stations.

These studies indicate that the classification of low-level decommissioning wastes from power reactors will be roughly as shown in Table 2.7-1.

Table 2.7-1 Classification of low-level decommissioning wastes from power reactors Waste Class PWR (Vol. %) BWR (Vol. %)

A 98.0 97.5 B 1. 2 2.0 0 0.1 0.3 Above C 0.7 0.2 As shown, the great majority of the waste volume from decommissioning will be classified as Class A waste. Only a small fraction of the wastes will exceed Class C limits.

Transportation of decommissioning wastes will involve no additional technical considerations beyond those for transportation of existing radioactive material.

Existing regulaticos covering transportation of radioactive material are covered under NRC regulations in 10 CFR Parts 20, 71, and 73, and Department of Trans-portation regulations in 49 CFR Parts 170-189.

An operating 1000 MWe reactor will generate approximately 25.4 MTHH (metric tons of heavy metal) (9.4 u3 ) of spent fuel each year and l300 m 8 of low-level waste each year. When multiplied over the 40 year operating lifetime of the plant, these values can be compared to the 11 3m of activated material 3

(greater than Class C) and 17,900 m of Tow-level waste resulting from DECON of 10/07/87 2-27 NUO586 CH 2

a PWR of similar size (see Secticn 4.4), and it can be seer that decom'ssioning will generate an appreciable fraction of the low-level waste generated :j a PWR over its lifetime. However, in any given year, the quantity of waste f*cm all operating reactors will considerably exceed that generated from those f acilities being decommissioned. The low-level wastes generated in 1980 from commercial nuclear fuel cycle activities totaled 81,000 m3 and low-level wastes frcm commercial non-fuel-cycle activities totaled 28,000 m3. Hence, any pro:lems in waste disposal capacity will be the result primarily of operating nuclea-facility waste inputs rather than decommissioning waste inputs. The following is a discussion of the current situation in this area.

Disposal capacity for Class A, Class B, and Class C wastes currently exists.

Development of new disposal capacity under the State compacting process is covered under the low-Level Radioactive Waste Policy Amendments Act referenced above. This Act provides for incentives for development of such capacity, as well as penalties for failure to develop such capacity. For wastes exceeding Class C concentrations, DOE has offered to accept such waste for storage pending development of disposal criteria and capacity. For spent fuel which as noted in Section 2.4 could impact the decommissioning schedule, a detailed schedule for development of monitored retrievable storage and geologic disposal capacity is provided in the Nuclear Waste Policy Act of 1982.

Hence, based on the above discussion, before decommissioning of a nuclear facility occurs, licensees should assess current waste disposal conditions and their potential impact on decommissioning. Although the DECON decommissioning ~

alternative assumes availability of capacity to dispose of waste, alternative methods of decomissioning are available (e.g. , SAFSTOR) including delay in completion of decommissioning during which time there can be temporary stcrage of wastes. Delay in decomcissioning can result in a reduction of occupational dose and waste volume due to radioactive decay.

2.8 Safeguards Just prior to decommissioning, the same safeguards rneasures may be required that are required while the facility is operating. During the actual dec:m-missioning, levels of special nuclear material in '.he facility should be 10/07/87 P-28 NUO586 CH 2

l decreas :d as a result of cleanout of the facility. In the case of DECON, decreased levels of safeguards measures should be continued until the quantity ,

of special nuclear material is reduced below safeguards levels, at which time safeguards measures can be discontinued. Regulations defining required pro-cedures and safeguard levels are found in 10 CFR Part 70 Special Nuclear Materials and 10 CFR Part 73 Physical Protection of Plant and Materials. In the case of SAFSTOR, depending on the quantity of special nuclear material as compared to the safeguards levels, continuous manned security may be required or may be replaced by continuous remote monitoring of intrusion, fire, and radiation' alarms during the continuinq care period. Immediate response is, of course, required in case any alarm is activated. Engineered barriers, such as fences and high-security locks, are maintained and inspected regularly.

Deferred decontamination requires similar safeguards provisions as are required during DECON depending on the quantity of special nuclear material remaining at that-time. The long-term care period of ENTOMB requires remote monitoring of intrusion, fire, and radiation alarms and engineered barriers if special nuclear material quantities are above safeguard levels.

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l REFERENCES l

1. C. F. Holoway and Others, Monitoring for Compliance with Decommissioning Termination Survey Criteria, NUREG/CR-2082, Oak Ridge National Laboratory i for U.S. Nuclear Regulatory Commission, June 1981. l

2. J. P. Witherspoon, Technology and Cost of Termination Surveys Associated with Decommissioning of Nuclear Facilities, NUREG/CR-2241, Prepared by Oak Ridge National Laboratory for U.S. Nuclear Regulatory Commission, February 1982.

3. R. I. Smith, G. J. Konzek, and W. E. Kennedy, Jr. , Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Station, NUREG-0130, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, June 1978, Addendum 1, August 1979, Addendum 2, July 1983, and Addendum 3, September 1984.

4. K. J. Schneider and C. E. Jenkins, Technology, Safety and Costs of Decommissioning a Reference Nuclear Fuel Reprocessing Plant, NUREG-0278, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, October 1977.

5. Termination of Operating Licenses for Nuclear Reactors, U.S. Nuclear Regulatory Commission, Regulatory Guide 1.86, June 1974.

6. E. S. Murphy, Technology, Safety and Costs of Decommissioning Reference Non-Fuel-Cycle Nuclear Facilities, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, NUREG/CR-1754, February 1981.

7. Robert S. Wood, Assuring the Availability of Funds for Decommissioning Nuclear Facilities, Draft Report, NUREG-0584, Revision 2, U.S. Nuclear Regulatory Commission, October 1980.

8. Financing Strategies for Nuclear Power Plant Decommissioning, NUREG/CR-1481, Prepared by Now England Conference of Public Utilities Comm:ssioners, Ir.c.,

in conjunction with Temple, Barker and Sloane, Inc. for U.S. Nuclear Regulatory Commission, July 1980.

9. J. J. Siegel, Utility Financial Stability and the Availability of Funds for Decommissioning, NUREG/CR-3899, September 1984, and Supplement 1, To be Published, Prepared by Engineering and Economics Research, Inc., for the U.S. Nuclear Regulatory Commission.

10. 50 FR 5600, February 11, 1985.

11. H. D. Oak, et al., Technology, Safety, and Costs of Decommissioning a Reference Boiling Water Reactor Power Station, NUREG/CR-0672, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, June 19C0, Addendum 1, July 1983, and Addendum 2, September 1984.

.y kE M ** M @A V /" 9- L4 10/07/87 2-30 NUO586 CH 2

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3 AFFECTED ENVIRONMENT - GENERIC SITE DESCRIPTION This section describes the characteristics of the sites used as bases for the decommissioning studies of the nuclear facilities discussed in this document.

Each fccility, with the exception of non-fuel-cycle nuclear facilities, is con-sidered to be located on a reference site. The site described is considered to be representative of the site of a large nuclear installation. Based on the analyses done in Sections 4 through 14 of this EIS, it was found that, while some details may vary from installation to installation, these differences are not expected to have any major impact on the results of the study. The generic fuel cycle facility site is described in Section 3.1.

3.1 Fuel Cycle Facility Site A reference environment was developed to aid in assessing the public safety and potential environmental effects of decommissioning nuclear facilities by various alternative methods. The meteorology parameters and population distributions were taken from the ALAP Study 1 for a river site in the year 2000. The ecologi-cal information was derived from the environment of one operating nuclear re-actor.2 The remainder of the information was obtained from a variety of sources or developed specifically for these studies, and is felt to be representative of potential sites for fuel cycle facilities.

Individual features of any specific nuclear fuel cycle facility will vary slightly from those of a generic site. Howeve", it is believed that use of a generic site will result in a more meaningful overall analysis of potential impacts associated with decommissioning nuclear fuel cycle facilities. Site-specific assessments will be required for the safety analysis and the environ-mental report submitted with the application for license modification prior to decommissioning a specific facility.

The generic fuel cycle facility site occupies 470 hectares (1160 acres) in a 1 rectangular shape of 2 km (1.24 miles) by 2.35 km (1.46 miles). A moderate sized i

10/07/87 3-1 NU0586 CH 3 )

l

river runs through one corner of the site. The site is located in a rural area that has.relatively low population density. Higher population densities are located at distances of 16 to 64 km (10 to 40 miles), and gradually reducing

. population densities are encountered out to 177 km (110 miles). The closest moderately large city,l population 40,000, is about 32 km (20 miles) distant.

The closest large city, population 1,800,000 is about 48 km (30 miles) away.

The total population in a radius of 80 km (50 miles) is 3.52 million.

The plant facilities are located inside a 12-hectare (30-acre) fenced portion of'the site. The minimum distance from the point of plant airborne releases to the outer site boundary is 1 km. Of the area surrounding the site, about 80%

of the land is used for farming.

-The relatively clean river flowing through the site has an average flow rate of 1,420 m3 /sec (50,000 ft3/sec). The river. is ased for irrigation, fishing, boat-ing'and other aquatic recreational activities, ad is a source of drinking water for the larger communities. Large supplies of flo<ing ground water exist at modest depths around the site. This water is widely used for drinking and irrigation.

-The reference site occupies a relatively flat terrace that has a low bluff form-ing one bank of the river. Young soils cover the old basement rocks in the area.

This site is in a relatively passive seismic area and ic lecated at an elevation above the estimated maximum probable flood-level.

The climate at the site is typical for internal continental areas. It has wide temperature variations 3nd moderate precipitation. Meteorology used in this study is an average taken from 16 nuclear reactor sites.

Less' than 20% of the land around the site is covered with pristine vegetation.

The original vegetation was primarily a climax deciducus forest A number of species of migratory birds are present in the area, as well as some annual birds.

A number of mammals occupy the general area.

The site is slightly contaminated with radioactive material as a result of depo-sition from the release of normal operating effluents over the operating lifetime 10/07/87 3-2 NUO586 CH 3

I \

1 of the-facility, it is expected that any accidental releases of radioactive material will be cleaned up immediately following the event. The individual site. contamination estimates are based on the predicted normal operating releases of gaseous effluents from the specific type of facility.

I s

- 10/07/87 3-3 NUO586 CH 3

REFERENCES ng N

As Practicable" for Radioactiveeactor Eff u Ac io for 0 eration to Meet6C-8, aterial n L19 Criteria page 60-12.

" As Lowt Water-Cooled Nucle I icello

/ birectorate oY RegFigure 68-1, t pageto68-43 Related O and5 Figure eration of Mont

'No. 50-263, November 1972, pp. 11-1 U.S. AEC, Final Environmental Statemen -

2, Nuclear Generating Plant, Docket h sed th ough the U.S. Governmen through 11-26. ing to the U.S.

l wr n, DC 20013-7082. i Copies of eferenced documents may Washing P Techni btInformationad, al Mrc ace by calling j

Springfield, (202)

-Printing Of ,

ting Office, P.O. Box 37082, for a purchased from the Natio l

Government Pr fee in the  ;

Copies may also ent of Commerce, spection or copyin C 20555.

Service, U.S.A Depa copy is ilable fortreet NW. , Washington, VA 22161. i RC Public Document Room,1717 e

0* m I

f lWh k t b 4f .

I 1

l I

NUO586 CH 3 3-4

?-

.4 PRESSURIZED WATER REACTOR ility for converting the thermal A pressurized water reactor (PWR) is a facto drivetoaa high turbine-generator a energy of a nuclear reaction into steam i ed duce electricity. The conversion is accomplished by heating

. temperature and pressure in the rea erator, and driving the turbine-het water to produce steam in the steam gen generator with the steam.

PWR is described in Section 3.1 l The generic site for the referencethe1175-MWe basis of operational and regu- i to decommissioning as well as to f The soecific site for a reactor it is chosen on latory criteria, some of which are appropr a eFor example, transportat l reactor construction and operation. uired for construction and operation, l supply, and a skilled labor supply ig Usually, are reqhowever, the most suit- l i

and are also necessary for decommiss on n .t depend upon the g able decommissioning alternative Rather it will depend on such

' tions.will no tion or upon specific siting consideralicense, land use considerations The at factors as desirability of terminating thel radiation exposures, and cost the time of decomissioning, occupationa also depend upon whether or not the choice of decommissioning alternative l retirement may age because of pre-will depend facility must be decommissioned before normaIn h sature closure.

almost entirely upon circumstances at t e upon earlier siting considerations.

nsored Pacific Northwest Laboratory C

Much of what follows is based on the NR -spod cost of dec d General Electric Company's (PNL) studies on the technology,l safety an In the parent study,1 PNL selected the Port anOregon, PNL as the re 1175-MWe Trojan Nuclear Plant at Rainier,ite typical of reactor locat assumed it to be located on i a generic s on the available technolo then developed and reported informat missioning on the reference facility 2 to this study, considerations, and probable costs fer decom at the end of its operating life.

Also, i as part of an addendumthe ef PNL did a sensitivity analysis to determ ne NUO586 CH 4 4-1 m /n? /87 ~- - - - ~ _ _ , , w~

l parameters might have on the conclusions in the original study regarding doses and costs of decommissioning. The parameters that were varied in the addendum included reactor size, degree of radioactive contamination, decommissioning alternatives, etc. The incremental costs of utilizing an external contractor for decommissioning and of additional staff needed to assure that the decom-missioning staff do not exceed radiation dose limits have been evaluated in a related follow-on analysis.3 In anothar related follow-on study,4 the estimated decommissioning cost and dose impacts of post-TMI backfit require-ments on the reference PWR have been examined and assessed. The results of all of these recent studies are included in the estimated decommissioning cost and dose estimates presented in this chapter for the reference PWR.

4.1 PWR Description The major components of a PWR are a reactor core and pressure vessel, steam generators, steam turbines, an electric generator, and a steam condenser system (Figure 4.1-1). Water is heated to a high temperature under pressure inside the reactor and is then pumped in the primary circulation loop to the steam generator. Within the steam generator, water in the secondary circulation loop is converted to steam that drives the turbines. The turbines turn the generator to produce electricity. The steam leaving the turbines is condensed by water in the tertiary loop and returned to the steam generator. The tertiary loop water then flows to cooling towers where it is, in turn, cooled by evaporation.

The tertiary loop is open to the atmosphere, but the primary and secondary cool-ing loops are not. -

Buildings or structures associated with the reference PWR include (1) the heavily reinforced concrete containment building, which houses the pressure vessel, the steam generators, and the pressurizer system, (2) the turbine building, which contains the turbines and the generator, (3) the cooling towers, (4) the fuel building, which contains fresh and spent fuel handling facilities, the spent fuel storage pool and its cooling system, and the solid radioactive waste system, (5) the auxiliary building, which contains the liquid radioactive waste treatment systems, the filter and ion exchanger vaults, the gaseous radioactive waste treatment systen, and the ventilation systems for the containment, fuel, and auxiliary buildings, (6) the control building, which houses the reactor control 10/07/87 4-2 NU0586 CH 4

t a mes etwis stattes re<mxts sete m teiftali $#ti4G .

Os M a tystint eueCa fJui as tLIC? tit gelat OR if sesuodt ht Sw v4 Poe a suCias WC** *EE 85 atissors i ema tyst a stat on ndi edas to o'rtohti ci m -tuau gsiCdt f%eitet weAepw iseaMe ig sos e g *lD el pup waevum 14W heerest Coat 4M@s Tsangleeste it t estamine auct fd(1 e f 4 CCW( f.e & Cathpe entle el art west rett %#t iD ta'.e

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NUO586 CH 4 I 4-3 l

10/07/87

room and personnel facilities, (7) water intake structures, (8) the administra-tion building, and (9) perhaps other structures such as warehouses and nonradio-active shops.

In a PWR, the reactor core and its pressure vessel are highly radioactive. So are the steam generators and the piping between the reactor and steam generators.

Because the turbines are not directly connected to the primary loop, they are usually not radioactive unless there has been tube leakage in the steam generators. The cooling towers and associated piping are normally not radio-active. Much equipment in the auxiliary building is radioactive, as is the spent fuel storage pool and its associated equipment.

The major radiation problems in decommissioning are associated with the reactor itself, the primary loop, the steam generators, the radioactive waste handling systems, and the concrete biological shield that surrounds the pressure vessel.

4. 2 Reactor Decommissioning Experience At the present time, the Elk River, Minnesota, demonstration reactor is the only power reactor that has been completely dismantled. This was a 58.2-MWt BWR that was dismantled between 1971 and 1974. Though this reactor was quite small compared to present day commercial power reactors, one lesson stands out:

reactors can be decontaminated with reasonable occupational radiation exposure and with virtually no public radiation exposure. At Elk River the containment l building was kept intact until the pressure vessel and the biological shield -

were removed. Only after all of the radioactive metal components and concrete areas were removed, was the concrete containment building demolished. Of par-ticular interest was the development of a remotely operated plasma arc torch that was used for cutting 1\-inch-thick stainless steel under water and 3\-inch-thick carbon steel in air.5 For large reactors,1,000-MWe, the cutting of 23/4 -inch-thick stainless steel under water and 9-inch-thick carbon steel in air will be7 8 required.s Based on current technology, this shoulo easily be accomplished. '

Other power reactors, all of them relatively small, have been placed in safe storage or entombed (see Table 1.5-1). These methods of decommissioning re-quire some sort of surveillance as mentioned in Section 2.3, and also require 10/07/87 4-4 NUO586 CH 4 t

1 l

l l

retencion of a possession-only license. In the case of the Elk River reactor, its licenses were terminated.

4.3 Decommissioning Alternatives The decommissioning alternatives considered in this section are DECON, SAFSTOR, and ENTOMB.

f 4.3.1 DECON DECON is defined as the immediate removal and disposal of all radioactivity in excess of levels which would permit release of the facility for unrestricted use. Nonradioactive equipment and structures need not be torn down or removed as part of a DECON procedure. The end result is the release of the site and any remaining structures for unrestricted use as early as the 6 years estimated for decommissioning af ter the end of reactor operation.

DECON is advantageous b(cause it allows termination of the NRC license shortly after cessation of facility operations and eliminates a radioactive site. DECON is advantageous if the site is required for other purposes, if the site is extremely valuable, or, if for some reason the site must be immediately released for unrestricted use. It is also advantageous in that the reactor operating staff is available to assist with decommissioning and that continued surveillance and maintenance is not required. A disadvantage is the higher occupational radiation dose which occurs during DECON compared to the other alternatives. -

The basic estimates in the original PNL studies have Leen adjusted by PNL analysts to reflect January 1986 costs. The revised estimate for the reference PWR shows that DECON would require 6 years to complete, including 2 years of planning prior to reactor shutdown, and would cost $88.7 million in 1986 dollars (Table 4.3-1). In addition to the values escalated from the PNL reports (NUREG/CR-0130 and NUREG/CR-0130, Addendum 1), the table also includes the cost additions--for pre-decommissioning engineering, additional staff to assure meet-ing the 5 rem / year dose limit for personnel, extra supplies for the additional staff, and the additional costs associated with the option of utilizing an 10/07/87 4-5 NUO586 CH 4

l l

l external contractor to conduct the decommissioning ef fort- which were devcloped in the PNL cost update done for the Electric Power Research Institute.3 The estimated decommissioning cost impacts of post-TMI-2 requirements on the refer-ence PWR4 are included in the table as well. It can be seen from the table that the total cost of DECON is about $103.5 million under the utility plus-contractor option. For comparison purposes, the time required to plan and build a large power reactor is presently about 12 years and the cost is well over two billica dollars.

Three important radiation exposure pathways need to be considered in the evalu-ation of the radiation safety of normal reactor decommissioning operations:

inhalation, ingestion, and external exposure to radioactive materials. For decommissioning workers, external exposure to radioactive materials is the domi-nant exposure pathway during decommissioning since inhalation and ingestion can be minimized or eliminated as pathways by protective techniques, clothing and breathing apparatus. Inhalation is considered to be the dominant pathway of public radiation exposure, since exposure to radioactive surfaces and ingestion can be minimized or eliminated as radiation pathways to the public during decom-missioning. During the transport of radioactive wastes, inhalation and inges-tion can be minimized or eliminated as radiation pathways to workers and to the public by techniques similar to those used during decommissioning. Therefore, exposure to radioactive materials is considered to be the dominant mode of radiation exposure to the public and to workers during waste transport. PNL calculated radiation doses for only the dominant pathways, and assumed the radiation doses from other pathways to be essentially zero. A summary of these doses is presented in Table 4.3-2.

The aggregate occupational radiation dose from external exposure to surface contamination and activated material, not including transportation of radio-active waste, is estimated to be about 1115 man-rem over 4 years (Table 4.3-2) or an average of about 279 man rem oer year. The aggregate occupational radia-tion dose from the transportation of radioactive wastes is estimated to be about 100.2 man-rem to truck transportation workers from DECON waste shipments. For comparison purposes, the average aggregate annual occupational radiation dose from operation, maintenance, and refueling of PWRs from 1974 through 1978 was 10/07/87 4-6 NU0586 CH 4

• ~

i the referenc3 PWR la $ M111(ons Summary of estleeted costs for decommission ng ENT0fe N _

Tobie 4.3-1 100 years of Internals Survelliance (h)

I _ Internals Removed

. SAF5 TOR 100 ') Years Included (g)

Prep. for 307eFs I Safe Storage (d) W Years Decommissioning DECON ') -

(3.9)

(24.7) 6.4 Element (39.9) (21.0) 46.6 (39.2) (40.8) 38.2 Base C:se Estimated (31.0) (9.5)

Decommissioning Costs:

(1978 dollars) 73.5 NA NA 1986 dollars 21.8Id) (h) 21.8(d) 12.6 (h) 17.1 21.8Id) 3.7 NA Safe Storage (d) NA' 1.I NA NA Preparation NA 69.4 40.4 Continuing Care . 69.4 NA Deferred MA Decontamination 3.9 P4ssible Additional Costs (S} 3.1

- Additional Staff Needed to Reduce Average Annual 7.5 1.1 11.4 Radiation Oose to: 10.5 5 ree per year Use of External Decgs- 12.9 4.6 sioning Contractor 5.6 4.5 5.6 4.5 4.5 7.5 Pre-Decommission;ng 7.4 5.6 3.4 E

I"9 ""I"9; Internal (utility)(j) 7.4 4.5 0.7 or 0.6 External (contractor) >0.1 *1.0 1.2 0.1

~0.3I ") -1.0I} +0.1 Supplies for Extra staff (5 rem /yr4.1 +0.1 I) average cost)III

-0.3 -0.3 I3) >0.1 so.3 NRC Licensing Activities $0.8 negligible negifgible "I s0.8

- Post-TNI-2 Impacts: s0.8 Internal (utility) or e

4-7 . ..

m ,.

Tabl2 4.3-1 (Contfr d ) II EllT0le }  ;

hmh i m m E 1 m years afSurveillanc2 (h)

M -

~

wm 16iri included (g)

Removed drep.fer I,} 3D Vears 4.3 16 Tears 4.3 Safe Storage Decommissioning DECON(C) -

7.4 Element negligible 57.2 4.9 80.3 47.9 7.4 External (contractor) "I 97.7 100.5 60.2 70.5 88.7 21.8 64.6 Subtr.tsi (<5 rea/yr): Utility (Internal) Staffing 77.9 27.5 80.3 103.5 100.5 Contractor (external) Staffing 97.7 88.7 f T!TAL Estimated Cost: ,, 103.5 f Utility Staffing or t of deep geologic Contractor Staffing radioactive structures, and exclude cos t nt 1986 dollars.

/

(c) Values include a 25% contingency and are in cons at core, exclude cost of demolition of non ponents. indicated.

II Values exclude cost1of disposal 1 and Table H.5-2, of lasdisposal unless otherwise of dismantled, i b 3 care, and deferred decontamination.

highly activated com IC Adapted f rom Reference 1. Table 10. - l H.5-2, unless otherwise indicated.arations aterials, for safe storage, cont nu w the demolition (d) Adapted from Reference l s otherwise indicated. 1. Table d2.9-3 disposal and of the entombed Tab eI')The radioactive values m shown for I Adapted from Reference 2, 4.5-1, un i ted es with the removal, packaging, an 064 million.

59 Dose g include the eventual costs assoc ademolition of the Reactor Building. cture is estimated to be ab t of the entombment structure, or and maintenance for the entombed s ru - s anticipated to be i

I The annual cost of surveillance ll as the Costs associated with inspect on U}NA-not appilcable. less otherwise indicated.

UI Adapted from Reference 3. Table 1.1, un sts of NRC Ifcensing activities as we te agencies.

The values shown include required by other Federal and sta the estimated co i 2.5-4. Reference 2, Section 6.3.

Adapted from Reference 4. Table (a) Adapted from Reference 4, Table 2.5-4 and from

" Negilgible means less than 50.025 million.

O 4 *

,m G

- - - - - _ _ . _ _ ~ ' - - - . _ . _ _

i i ng the rafcr:nco PWR '

Summaryofradiationdog*galysesfordecommisson -

Table 4.3-2 ENTOMB (values are in man-rem) )

SAFSTOR Internals Internals Included Removed 100 Years 10 Years 30 Years DECON NA I) NA 282.4 282.4( ) neg.

Occur,ational Exposure NA II) 282.4(k) 14 neg.

14 NA Safe Storage Preparation (C) NA 10 f) 1 NA Id) 337.5(k) 24.6 1,000 Continuing Care 1,114.5(k) 900 _

NA

) NA NA NA NA Decontamination ( NA 10.2 10.2 NA l I9) (h) 10.2 neg. NA Entombment NA 1.7 i

24.2 21 Safe Stor. Prep. Truck Shipments 16 (h) 100.2(k) NA NA 1,021 Decontamination Truck Shipments I9) NA NA 664(k) 333(k) Qkl 916 Entombment Truck Shipments 1,215(k)

NA NA Total neg.

neg. neg. neg. neg.

Public Exposure NA neg.

neg. NA Safe Storage Preparation ( ) neg. NA NA neg.

fi) neg. neg. neg. neg.

Continuing Care neg.

NA NA NA NA NA Decontamination ( ) NA 2.1 2.1 2.1 NA NA Entombment (9} ) NA neg.

I) 0.4 4 Safe Stor. Prep. Truck(h) Shipments 20.6( ) (h S NA NA 4

4 NA 4 Decontamination Truck Shipments NA _ 2 3

Entombment Truck Shipments (9I 21( )

7(")

Total S

.( -.e= =wG 9

e= %em.eete*e hM h-

so . ,e. .s, . .. -

4. .

Table 4.3-2 (Continued) i ted.

(0)A11 references are from Reference 1, unless othemise ind (b) Values exclude radiation dose from disposal of the last core .

IC) Table 11.3-2.

(d) Table 11.3-4.

(*) Table 11.3-1.

(I) Table H.6-1. for radioactive j

(9) Tables 3.5-1 and 4.6-1 from Reference 2, with no allowanc f, decay (see text for discussion).

(h) Table 11.4-2, with allowances for radioactive decay, l (I) Table 11.2.2.

t TMI-2 (3)NA-not applicable. ~

l II) Values affected by the C4As 'rd A estimated additional CeteefsE For a radia  !

impacts on decommissioning operations,f rom post-TMI-2 impacts to detailed explanation of the minor contributions 2.4-1 of Reference the total estimates given, consult Table i

NUO586 CH 1 i 4-10 y m' ro7 - - . _ _ - . '"~ - - - - . _ . _ _ _ _ . - .

I

550 man-rem per reactor.' In 1979 it was 924 man-rems,10 and in 1980 it was 1,101 man-rems.

This increase is considered to be due to build-up of radioactive contaminants with increasing reactor agell and to increasing reactor size 12 and special man rem intensive maintenance tasks.

The inhalation radiation dose to the public from airborne radionuclide releases during DECON is estimated to be negligible. The radiation dose to the public is calculated to be about 20.6 man-rem from the truck transport of radioactive wastes from DECON.

4.3.2 SAFSTOR Generally, the purpose of SAFSTOR is to permit SOCo to decay to levels that will reduce occupational radiation exposure during decontamination. As indi-cated in Table 4.3-2, most of the occupational dose reduction due to decay occurs during the first 30 years after shutdown with considerably less dose reduction thereafter. The public dose, which will always be small, will also experiencebost of its reduction during the first 30 years. Nonradioactive equipment and structures need not be removed, but eventually all radioactivity in excess of that allowed for unrestricted use of the facility must be removed.

Hence, in contrast to DECON, to take advantage of the dose reduction, SAFSTOR could be as long as 60 years including final decontamination. The end result -

is the same: release of the site and any remaining structures for unrestricted use.

SAFSTOR is advantageous in that it results in reduced occupational radiation exposure in situations where urgent land use considerations do not exist.

Disadvantages are that the licensee is required to maintain a possession-only license under 10 CFR Part 50 and to meet its requirements at all times, thus contributing to the number of sites dedicated to radioactive confinement for an extended time period. Other disadvantages are that surveillance is required, the dollar costs are higher than for DECON, and the experienced operating staff may not be available at the end of the safe storage period to assist in the decontamination.

.-m,m, . ,, ,,,,n e c e c., ,

The PNL study shows that the costs of SATSTOR for a 30 year period are greater tnan those of DECON and vary with the number of years of safe storage. For example, the total cost of 30 year SAFSTOR is estimated to be $100.5 million in 1986 dollars compared with the total cost of $88.7 million for DECON. However,  !

the total cost of 100 year SAFSTOR is estimated to $80.3 million in 1986 dollars. l The lower cost for 100 year SAFSTOR compared to 30 year SAFSTOR is the result  !

of lower costs for deferred decontamination due to the radioactivity having decayed. PNL's cost estimates for the decommissioning alternatives are pre-sented in Table 4.3-1.

SAFSTOR results in lower radiation doses to both the work force and to the public than DECON. The PNL study (Table 4.3-2) shows the aggregate occupa-l l tional radiation dose to be approximately 321 man-rem for a 30 year SAFSTOR (282.4 man-rem from safe storage preparation, 14 man-rem for continuing care and surveillance, and 24.6 man-rem from deferred decontamination), not includ-

! ing transportation. The occupational radiation dose from the truck transport of radioactive wastes is calculated to be about 12 man-rem. 100 year SAFSTOR results in little additional reduction in the aggregate occupational radiation dose compared to 30 year SAFSTOR.

Radiation doses to the public from airborne radionuclide releases during prepa-ration for safe storage are estimated to be negligible. The radiation dose to the public from the truck transport of radioactive wastes during preparation for safe storage is estimated to be about 2.1 man rem, and that from the truck trunsport of radioactive wastes during deferred decontamination after 30 years -

of safe storage is estimated to be about 0.4 man-rem.

4.3.3 ENTOM8 i

ENTOMB means the complete isolation of radioactivity from the environment by means of massive concrete and metal barriers until the radioactivity has decayed to levels which permit unrestricted release of the facility. These barriers must prevent the escape of radioactivity and prevent deliberate or inadvertent intrusion. The length of time the integrity of the entombing structure must be maintained depends on the inventory of radioactive nuclides present. A PWR 10/07/87 4-12 NUO586 CH 4

that has been operated only a short time will contain 60Co as the largest con-tributor to radiation dose and smaller amounts of dominant fission products such as 187Cs with about 30 year half-life. In this case, the integrity of the entombing structure need only be maintained for a few hundred years, as the disappearance of radioactivity is initially controlled by the 5.27 year half-life of 60Co and later by 30 year half-life fission products. If, on the other hand, the reactor has been operated for 30 or 40 years, substantial amounts of 59Ni and **Nb (80,000 year and 20,000 year half-lives, respectively,) will have been accumulated as activation products in the reactor vessel internals. The dose rate from the 94Nb present in the reactor vessel internals has been esti-mated to be approximately 2 rem / hour while the dose from the 59Ni in the inter-nals is 0.1 rem / hour. These dose levels are substantially above acceptable residual radioactivity levels and, because of the long half-lifes of 9'Nb and SSNi, would not decrease by an appreciable amount, due to radioactive decay, for thousands of years. In addition, there are an estimated 1,300 curies of SSNi in the reactor vessel internals which could result in potential internal exposures in the event of a breach of the entombed structure and subsequent introduction of the SSNi in an exposure pathway during the long half-life of SSNi. Thus, the long-lived isotopes will have to be removed or the integrity of the entombing structure will have to be maintained for many thousands of years.

ENTOMB of a PWR is limited to the containment building because its unique structure lends itself to entombment and because it contains most of the radio-activity in the facility. The other radioactive buildings associated with a -

reactor must be <1ecommissioned by another method such as DECON. It is possible, however, to move some radioactive components from the fuel building or auxiliary building to the containment building and entomb them there, rather than ship them of f site.

ENTOMB is advantageous because of reduced occupational and public exposure to radiation compared to DECON, because little surveillance is required, and because little land is required. It is disadvantageous because the integrity of the entombing structure must be assured in some cases for hundreds of thousands of years, because a possession-only license under 10 CFR Part 50 would be required, and because entombing contributes to the number of sites permanently dedicated to radioactive materials containment.

10/07/87 4-13 NUO586 CH 4

PNL considered two approaches to entombment in an addendum2 to its earlier PVR study.1 In both approaches, as much solid radioactive material from the entire facility as can be accommodated is sealed in the containment building beneath the operating floor by means of a continuous concrete slab. All openings to the exterior beneath the operating floor are sealed. Above the operating floor, radioactive - c, n'.1 are removed to sufficiently permit release of that portion of the fac. 'vr unrestricted use.

In the first approach, the pressure vessel internals and their long-lived SSNi and 94Nb isotopes are entombed, along with other radioactive material. This results in less cost and radiation exposure because the pressure vessel and its internals will not have to be removed, dismantled, and transported to a deep geologic waste repository. It will also, however, result in the requirement for a possession-only license and surveillance in perpetuity because of the presence of the long-lived isotopes. Because of the many variables involved, PNL made no firm estimate of the costs for possible deferred dismantlement of the entombment structure. However, these costs are anticipated to be at least of the same order of magnitude as those for deferred dismantlement of the reference PWR af ter a period of safe storage (see Table 4.3-1).

In the second approach, the pressure vessel internals and their long-lived SSNi and 84Nb isotopes are removed, dismantled, and transported to a radioactive waste repository (a careful inventory of radioactivity would need to be made to ensure that only relatively short-lived isotopes remained). This approach results in more cost and radiation dose, but offers the possibility that sur-veillance and the possession-only license could be terminated at some time within several hundred years, thereby releasing the entire facility for unre-stricted use.

Radioactive materials not entombed would have to be packaged and transported to a disposal site. Costs and radiation doses for this portion of the entombment procedure would be the same as for DECON. Cost savings and radiation dose re-ductions result from a lesser volume of radioactive equipment and material having to be dismantled, packaged, and transported. In all cases, spent fuel would bc removed.

10/07/87 4-14 NUO586 CH 4

ECOMB for the reference PWR, including the pressure vessel and its internals, is estimated to cost $47.9 mfilion, with an annual maintenance cost of 564,000.

It results in an aggregate radiation dose of 900 man-rem to decommissioning workers,16 man-rem to transportation wnrkers, and 4 man-rem to the general public. ENTOMB for the reference PWR, with the pressure vessel internals removed, is estimated to cost $57.2 million with an annual maintenance cost of

$64,000, and to result in an aggregate radiation dose of 1000 man-rem to de-commissioning workers, 21 man-rem to transportation workers, and 4 man-rem to the general public. These estimates are listed in Tables 4.3-1 and 4.3-2.

Although task-wise schedules were developed for DECON,1 no comparable schedules were developed for the ENTOMB analysis.2 As a result, the estimated occupa-tional exposures shown in Table 4.3-2 are not decay-corrected; thus, they represent conservative, upper-bound estimates.

4.3.4 Sensitivity Analyses An addendum to the initial PNL study was developed 2 to analyze a variety of realistic decommissioning situations that might significantly impact on the original conclusions regarding doses and costs for the various decommissioning alternatives. While there were some differences in results, the conclusion of the sensitivity analysis is that these differences do not substantially affect the original cost and dose conclusions. Of the various situations analyzed by PNL in the addendum, the most important with regard to their potential effect on dose and cost estimates are reactor size and degree of contamination.

Based on an analysie l similar to that for the reference PWR (NUREG/CR-0130 l

Addendum 1) and incorporating selected cost adders (described in References 3 and 4 and escalated to constant 1986 dollars as shown in Table 4.3-1), upper-bound estimates were made of the costs for immediate dismantlement of reactor plants smaller than the reference plant. The analysis was limited to plants with thermal power ratings greater than 1200 MWt and was based on the assump-tion that all costs (staff labor, equipment, supplies, etc.) except radioactive waste disposal are independent of plant size. The results are shown in Table 4.3-3.

10/07/87 4-15 NUO586 CH 4

1 l

l T:ible 4.3-3 Estimated immediate dismantlement costs for plants smaller than the reference PWR, l derived overall scaling factors,bgsed on previously-(millions of 1 dollars)

Waste Scaling Remaining Escalated Reactor MWt. Disposal Factor Costs Adders Total Costs (c)

Trojan 3500 40.223 1.000 24.174 14.385 88.782 Turkey Pt.' 2550 40.223 0.789 34.174 14.385 80.295 R. E. Giona 1300 40.223 0.518 34.174 14.385 69.395 (a)All costs are in constant 1986 dollars and include a 25% contingency.

(b) Derivation of previously-dm J rall scaling factors can be found in Reference P..

(c) Total costs shown above si for the utility-only cost option.

Using the results from Table 4.3-3, a linear equation can be derived for the scaling of the immediate dismantlement costs for plants in the 1200 to 3500 MWt range:

Cost = 57.911 + (8.808 x 10 8)(Wt)

Revised overall scaling factors for the Turkey Point and Ginna plants were obtained by dividing the results of the linear equation by the cost of the reference plant. Based on this formula, a list of variations in dose and cost for these PWRs is presented in Table 4.3-4. .

The addendum 2 also analyzed the sensitivity of decomissioning costs and radiation doses related to a postulated tripling of radiation dose rates from radionuclides deposited in PWR coolant system piping during reactor operation over a period of 30 to 40 years. This tripling of dose rate is postulated as an upper limit on the basis of recent trends for operating reactors. If no corrective action is taken to reduce the radiation dose rates, the accumulated radiation dose to decommissioning workers for DECON would be increased about 1,250 man-rem (*),

and the total decommis:ionirg costs could be increased by about $5.2 million (a)This number excl des remo'al of last core and allows for radioactive decay.

l 10/07/87 4-16 NUO586 CH 4 L

(

Table 4.3-4 Estimated costs and occupational radiation doses for decommissioning different-sized PWR plants (a,b)

Station R. E. Ginna Turkey Point Trojan Power Rating (thermal Overall Scaling megawatts) 1.300 2.550 3.500 Factor (OSF[MWt]) 0.781 0.905 1.000 DECON ($ millions) 69.3 80.3 88.7 (man-rem) 1097. 1.271 1.404 ENTOM8(d) w/ internals ($ millions)(d) 37.4 43.3 47.9 (man-rem) 703 815 900 w/o internals ($ millions) 44.7 51.8 57.2 (man-rem) 781 905 1.000 SAFSTOR Preparations for Safe Storage ($, millions) 17.0 19.7 21.8 (man-rem) 333 386 426 Safe Storage for 30 years ($ millions) 3.7 3.7 3.7 (man-rem) 14 14 14 for 50 years ($ millions) 6.2 6.2 6.2 (man-rem) 14 14 14 for 100 years ($ millions) 12.6 12.6 12.6 (man-rem) 14 14 14 Deferred Dismantlement:

• after 30 years ($ million) 54.2 62.8 69.4 (man-rem) 23.4 27.2 30 after 50 years ($ million) 31.6 36.7 40.5 (man-rem) 1.9 2.2 2.4 after 100 years ($ million) 31.6 36.6 40.4 (man-rem) 0.9 1.1 1. 2

(*) Values include a 25% contingency and are in 1986 dollars.

(b) Costs do not include spent-fuel disposal or demolition of nonradioactive structures.

(c) Doses are taken from Ref. 2 and do not include transportation doses and do not take credit for radioactive decay during decommissioning.

(d) Entombment costs do not include continuing care cost (50.064 M/yr.).

10/07/87 4-17 NUO586 CP 4

for DECON. For ENTOMB the radiation dose would be nearly doubled and the total cost could be increased about $3.6 million. For preparations for safe storage, the radiation dose would be increased about 130 man-rem, and there would be no significant change in the cost. If corrective action is taken, such as an ex-tended chemical decontamination cycle, the total additional cost could be abe;t

$170,000.

In order to handle these postulated higher initial radiation levels, it appears that additional chemical decontamination during decommissioning would be the most cost-effective approach. For example, it is estimated that increasing the circulation time of the chemical solution about 50% would reduce the postulated increased radiation levels by a factor of 3, thus reducing these levels to approximately the same dose rate conditions assumed in the reference case analysis. This approach would also be more consistent with the principles of ALARA, since the occupational radiation dose associated with a chemical decon-tamination cycle is relatively small, compared with the radiation dose associated with installing temporary shielding, or with attempting to perform the dismantlement withnut additional shielding. In addition, it appears likely that the large buildups of radionuclides prevalent today on piping systems will be prevented as periodic decontamination tiuring normal operation of the reat. tor coolant system and related fluid-handling sy>tems become standard procedures when the present technology development for decoetamination solutions has been completed.

One of the circumstances that has changed since the origi. Tai PWR decommissioning ~

reports t2 were prepared whici could influence the development of the cost and dose estimates presented in his GEIS is an assessment of post-TMI-2 require-ments on the decommissioni") of the reference PWR. Actions judged necessary by the NRC to correct or improve the regulation and operation of nuclear power plants based on the experience from the accident at THI-2 resulted in a number of recommendations that were subsequently issued to the utilities as requirements. Some of those requirements resulted in equipment and hardware changes and/or additions to the reference PWR that could eventually expand the scope of decommissioning activities, since those materials could reasonably be expected to become contaminated or radioacti,e during the remaining operational lifetime of the 0. 1ant. For the reference PWR, it was concluded by PNL in a 10/07/87 4-18 NUO586 CH 4

recent study' that the original immeciate dismantlement decommissioning cost estimates could be expected to increase only slightly overall (less than 1% in constant 1986 dollars)r due to a slightly expanded scope of decommissioning activities associated with changes in the reference plants characteristics.

The radiation dose would be increased by about 32 man-rem, due largely to the dismantling operations associated with the removal of a significantly greater mass of spent fuel pool storage racks.

There are many areas where various planned design and operational features could facilitate decommissioaing. Exploration of such areas was considered by PNL1 in their initial decommissioning study. It was concluded that appropriate mea-sures could not only significantly reduce decommissioning occupational dose and radioactively contaminated waste volume but could also reduce occupational dose during reactor operation. Preliminary considerations of various design and operational features that could further facilitate decommissioning and their impacts on doses and costs are discussed in NUREG/CR-0569. H 4.4 invironmental Consequences Radiation doses and costs associated with possible decommisi,ioning alternatives are discussed in Section 4.3. It is noted for perspective that in the cases of DECON and SAFSTOR, the environmental effects of greatest concern (i.e., radia-tion dose and radioactivity released to the environment) are substantially less than the same effects resulting from reactor operation and maintenance. It should also be noted that while the dollar costs of ENTOMB are les: than those of DECON, the environmental impacts could be quite high should large acounts of rad'oactivity escape from a breached structure during the entombment period.

Other environaiental consequences are rather different from the environmental consequences usually discussed in environmental impact statements. This is because, usually, an environmental impact statement is addressed to the consequences of building a facility that will require land, labor, capital investment, materials, continuing use of air, water, and fuel; a socio-economic infrastructure; and so on. Decommissioning, on the other hand, is an attempt to restore things to their original condition, which requires a much smaller commitment of resources than did building and operating the facility.

i 10/01/87 4-19 NUO586 CH 4

i A major environmental consequence of decommissioning, other than radiation d and dollar cost, is the commitment of land area to the disposal of radioactive waste.

PNL made estimates (shown in Table 4.4-1) of the low-level wast Table 4.4-1 Estimated burial volume of low-level radioactive waste and PWR rubble for the reference Decommissioning ilternative Volume (m3 )

DECON 18,340 SAFSTOR Deferred DecontaminationID) following Safe Storage for: 10 Years 18,340(3) 30 Years 18,340(a,c) 50 Years 1,830 100 Years 1,780 ENTOMB (d) 1,740 (a) Includes about 4403 mof radioactive waste attributable to removal of back-fitted material adapted from Table 5.1-9, Reference 4).

(b) Radioactive wastes from preparation for safe storage and during safe storage are small in comparison to those of deferred decontamination.

(C)Although, in actuality, there is a '

gradual decrease in waste volume over time, it is not indicated here for clarity of presentation.

(d)Does not include the volume of the entombing structure or of the wastes within, disposal volume required to accommodate radioactive weste and rubble removed from the facility and transported to a licensed site for disposal. Reduction in waste volume for SAFSTOR occurs as many of the contamination and activation '

products present in the facility will have decayed to background levels. The volume for ENTOMB does not include the volume of the entombing structure or of the wastes entombed within it, only the wastes shipped off-site. The entombing structure is, in effect, a new radioactive waste burial ground, separate and 10/07/87 4-20 NUO586 CH 4

Tabla 4.4-2 Summary cf radiation doses ta the maximally-exposed individual from accidente,1 airborne radionucilde releases during decommissioning operations DECON Preparations for Safe Storage Airborne First-Year Dose Fifty-Year Dose Airborne First-Year Dose Fif ty-Year Dose Release (arem) Commitment (ares) Release (area) Commitment (aree)

Incident (pCl)

Total Body (a) Lung Total Body Lung (mCl) Total Body (a) Lung Total Body Lung Explosion of LPG Leased from e Front End Loader 3.6 x 108 3.6 x 10 2 4.7 x 10 2 4.4 x 10 8 5.4"x 10 8 ---(c)

Explosion of Oxyacetylene During Segmenting of the ,

Reactor vessel Shell 3.6 x 102 4.3 x 10 5 6.1 x 10 8 6.9 x 10 8 6.9 x 10 8 ---

Explosion and/or Fire in the Ion Exchange Resin -

3.8 x 108 3.8 x 10

• 5.0 x 10

• 3.6 x 10

• 5.7 x 10 * ---

Gross Leak during In Situ Decontamination 2.1 x 108 2.1 x 10

• 2.8 x 10

• 2.5 x 10

• 3.2 x 10

• 2.1 x 108 2.1 x 10

• 2.8 x 10

• 2.5 m 10

• Segmentation of RCS Piping With Unremoved Contamination 1.1 x 108 4.6 x 10 ' 7.3 x 10

• 4.8 x 10 8 7.9 x 10 * ---

Less of Contamination Control Envelope During 0xyacetylene Cstting of the Reactor V2ssel Shell 2.3 x 108 --- --- ---

4.4 x 10 * ---

Vacuum Bag Rupture ---

1.0 x 108 1.1 x 10 ' 1.3 x 10 5 1.2 x 10 5 Accidental Cutting of Contaminated Piping ---

1.8 x 10 8 ---

1.2 x 10 5 ---

Accidental Spraying of Concentrated Contamination With the High Pressure Spray ---

1.2 x 10 8 i

1.6 x 10 ' 1. 5 x 10

• I The average annual total body dose 'to an individual in the U.S. from natural sources range. from 80 to 170 mrem. United Nations Scientific Committee on the Effects of Atomic Radiation, Ionizing Radiation: Levels and Effects. Volume 1. United Nations, pp. 29-63, 1972.

Frequency of occurrence: high >1.0 x 10 2; medium 1.0 x 10 2 to 1.0 x 10 5; low <1.0 x 10.s per year.

I'}A dash indicates a dose less than 1.0 x 10 ' area or that this action does not apply to the decommissioning mode shown.

m I

4- % .

Table 4.4-3 Estimated frequencies and radioactivity releases for selected truck transport accidents Radiation Dose for Maximally Exposed Individual, (rem)(a)

Frequency of Frequency of 50 Yr Dose Accidents per Accidents per Release, 1st Year Oose Commitment Accident Description DECON SAFSTOR Curies Bone Lung Bone Lung TruckTransportg)Dgemis-sioning Wastes Minor Accident with Closed Van 8.8 x 10 8 9.0 x 10 2 No Release -- -- -- --

Moderate Accidents with Closed Van 2.1 x 10 1 2.1 x 10 2 1 x 10 4 0.01 0.2 0.01 0.2 Severe Accieent with Closed Van 5.6 x 10 3 5.7 x 10 4 1 x 10 2 1.1 21 1.1 24 (a) Maximally-Exposed individual is assumed at 100 m from the site of the accident.

(b) Based on an inventory of 100 Ci per truck shipment.

(C) Release fractions for respirable material for moderate and severe accidents are assumed to be 10 6 and 10 4, respectively.

greater than when the PWR was in operatir.n and will be much less than when the PWF was under construction. The transportation network is already in place, but will require some maintenance if the SAFSTOR alternative is selected.

Disturbance of the ground cover need not take place to any appreciable extent except for filling holes and leveling the ground following removal of underground structures, unless extended operation of the plant has resulted in contamination of the ground around the plant. Plowing of the ground would generally result in lowering average soil contamination levels to those acceptable for releasing the site for unrestricted use, except for a few more highly contaminated areas where material would have to be removed. In this case, soil to a depth of several centimeters and some paving may have to be removed, packaged, and shipped to a disposal facility before the site can be released for unrestricted use.

The biggest socioeconomic impact will have occurred before decommissioning started, at the time the plant ceased operation and the tax income created by the plant was reduced. No additional public services will be required becausa the decommissioning staff will be somewhat smaller than the operating staff. In the case of deferred decontamination, the decontamination staff will be larger than the surveillance staff.

4. 5 Comparison of Decommissioning Alternatives from careful examination of Tables 4.3-1 and 4.3-2 it appears that DECON or 30 year SAFSTOR are reasonable options for decommissioning a PWR. 100 year SAFSTOR is not considered a reasonable option since it results in the continued presence of a site dedicated to radioactivity containment for an extended time period with little benefit in aggregate dose reduction compared to 30 year SAFSTOR. DECON costs less than SAFSTOR and its larger annual occupational radiation dose, which is similar to the routine annual dose from plant operations is considered of marginal significance to health and safety.

Either ENTOMB option requires indefinite dedication of the site as a radioactive waste burial ground. In the ENTOMB option with the reactor internals and its long-lived activation products entombed, tne security of the site could not be assured for thousands of years necessary for radioactive decay, so this option 10/07/87 4-24 NU0586 CH 4

is not considered viable. .n the ENTOMB option with the reactor internals removed, it may be possible to release the site for unrestricted use at some time within the order of a hundred years if calculations demonstrate that the radioactive inventory has decayed to acceptable residual levels. However, even this ENTOMB alternative appears to be less desirable than either DECON or SAFSTOR based on consideration of the fact that ENTOMB results in higher radiation exposure and hiober initial costs than 30 year SAFSTOR, that the overall cost of ENTOMB over the entombment period is approximately the same as DECON, and the fact that regulatory changes occurring during the long entombment period might result in additional costly decommissioning activity in order to release the facility for unrestricted use.

Consideration was given to the situation where, at the end of the reactor operational life, it is not possible to disposc of waste offsite for a limited period of time, but not exceeding 100 years (see Section 2.7). Such a constraint needs to be accounted for in the decommissioning alternatives. Based on en analysis by PNL of the technology, safety and cost considerations on telection of decommissioning alternatives,H it was concluded that SAFSTOR is an acceptably viable alternative. While DECON and conversion of the spent N el pool to an independent spent fuel storage pool is certainly a possibility for the case where all other radioactive wastes can be removed offsite, there ones not appear to be any significant safety difference between this alternative and SAFSTOR and the choice should be a licensee decision. The active phase of maintaining the spent fuel in the pool is not considered to be part of the regulatory require-ments for decommissioning, but would be considered under the usual operating licensing aspects regarding health and safety with consideration given to facil-itation for decommissioning. Aside from the expenses incurred from storing spent fuel, other costs for keeping radioactive wastes onsite for the reactor in a safe storage mode were estimated to have minimal effect on the SAFSTOR alternative compared to this alternative for radioactive wastes being sent offsite. Site security for storage of spent fuel (which is considered as an operational rather than a decommissioning consideration) was estimated at about

$0.94 million per year (in 1986 dollars)(a) In a multireactor site, such security could result in less cost because of a sharing of required overheads.

(a) Adapted from Reference 14.

10/07/87 4-25 NUO586 CH 4

I 1

REFERENCES 1

1. R.1. Smith, G. J. Konzek, and W. E. Kennedy, Jr. , Technology, Safety and l Costs of Decommissioning a Reference Pressurized Water Reactor Power Station, NUREG/CR-0130, Prepared by Pacific No hwest Laboratory for the U.S. Nuclear Regulatory Commission, June 1978

2. R. I. Smith and L. M. Polentz, Technology, Safety and Costs of Decommis-sioning a Reference Pressurized Water Reactor Power Station. Addendum to NUREG/CR-0130, Prepared by Pacific Northwest Laboratory for the U.S.

Nuclear Regulatory Commission, August 1979.*

3. R. I. Smith et al. , Updated Costs for Decommissioning Nuclear Power Facilities, EPRI NP-4012, Prepared by Battelle, Pacific Northwest Laboratories for the Electric Power Research Institute, Final Report, May 1985.

4. G. J. Konzek, Estimated Impacts of Post-TMI-2 Requirements and Other Selected Regulatory Changes on Oecommissioning of Two Reference Light Water Reactors, NUREG/CR , Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, , 1987.

5. AEC-Elk River Reactor Final Program Report, CD0-651-93, United Power Association, Elk River MN, Revised November 1974.

6. W. J. Manion and T. S. LaGuardia, An Engineering Evaluation of Nuclear Power Reactor Decommissioning Alternatives, Atomic Industrial Forum, November 1376.

7. Ninth Annual Occupational Radiation Exposure Report,1976, U.S. Nuclear Regulatory Dommission, NUREG-0322, p. 10, October 1977.

8. B. G. Brooks, Occupational Radiation Exposure at Commercial Nuclear Power '

Reactors 1980, U.S. Nuclear Regulatory Commission, NUREG-0713, December 198 *

9. R. O. Pohl, "Radiation Exposure in LWRs Higher than Predicted," Nuclear ,

Engineering International, p. 36, January 1977.

10. A. Martin, "Occupational Radiation Exposure in LWRs Increasing," Nuclear Engineering International , p. 32, January 1977.

11. Letter from R. I. Smith, Pacific Northwest Laboratories, to Carl Feldman, U.S. Regulatory Commission,

Subject:

Updated Costs for Decommissioning the Reference PWR and BWR as Developed for the GEIS, November 12, 1986 (available in USNRC Public Document Room).

12. E. B. Moore, Jr. , Facilitation of Decommissioning Light Water Reactors, NUREG/CR-0569, Prepared by Pacific Northwest Laboratory for the U.S.

Nuclear Regulatory Commission, December 1979.

• See footnote to reference in Chapter 1 for document purchasing availability.

10/07/87 4-26 NUO586 CH 4

0

13. R. L. Engel et al., 1986. 1986 Annual Review of the Adequacy of the 1.0 mill per kWh Waste Disposal Fee, PNL-5877, Pacific Northwest Laborato ry. Richland, Washington.

14. G. M. Holter, Technology Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station, Addendum 2 to NUREG/CR-0130, Effects on Oecommissioning of an Interim Inability to Dispose of Wastes Offsite, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, July 1983.

D l

l 10/07/87 4-27 NUO586 CH 4 l

5 BOILING WATER REACTOR A boiling water reactor (BWR), like a pressurized water reactor (PWR), is a f acility for converting the thermal energy of a nuclear reaction into the kinetic energy of steam to drive a turbine generator and produce electricity.

In a BWR, the conversion is accomplished by heating water to boiling in the reactor pressure vessel and using the resulting steam to drive the turbines.

The intermediate step, present in a PWR, of converting pressurized hot water into steam through a heat exchanger in a steam generator is not used in a BWR.

Elimination of this step also eliminates one cooling loop.

The generic site for the reference 1155-MWe BWR is assumed to be typical of reactor locations and is described in Section 3.1. As in the case of a PWR, the specific site for a BWR is chosen on the basis of operational and regula-tory criteria, usually with little regard for decommissioning. Fortunately, factors that are appropriate for siting, such as transportation access, water supply, and skilled labor supply, are also appropriate for decommissioning.

Thui,, the decommissioning alternative chosen will not usually depend on siting considerations, but rather on safety, costs, and land use options at the time of decommission;.ig. These considerations are discussed in Section 4 for a PWR, and apply equally to a BWR.

In this section, we have used information prapared for the study on the tech-nology, safety and costs of decommissioning a reference BWR, which was con-ducted by Pacific Northwest Laboratory (PNL) for the NRC.1 In the BWR study, PNL selected the Washington Public Power Supply System's WNP-21155-MWe reactor at Hanford, Washington, as the reference BWR and assumed it to be located on the generic site. PNL then developed and reported infonnation on the available technology, safety considerations, and probable costs for decommissioning the reference facility at the er.d of its operating life. As part of this study, PNL did a sensitivity study to analvte the effect that variation of certain 1

10/07/87 5-1 NU0586 CH 5 l

- ~ _ __ - . _-

parameters might have on radiation doses and costs associated with decom-missioning. The parameters which were varied included reactor size, degree of radioactive contamination, dif ferent contract arrangements, type of containment structure, etc.

The incremental costs of utilizing an external contractor for decommissioning were updated in a related follow-on analysis. In another related follow-on study,3 the estimated decommissioning cost and dose impacts of post-THI-2 requirements on the reference BWR have been examined and assessed. The results of these two recent studies are included in the estimated decommissioning cost and dose estimates presented in this chapter for the reference BWR, 5.1 Boiling Water Reactor Description Ine major components of a BWR are a reactor core and pressure vessel, steam turbines, an electric generator, and a steam condenser system (Figure 5.1-1).

Vater is boiled in the reactor pressure vessel to create steam at high tempera-ture and pressure, which then passes through the primary circulation loop to drive the turbines. The turbines turn the generator, which produces electricity.

The steam leaving the turbines is condensed by water in the secondary loop and flows back to the reactor. The water in the secondary loop flows to the cooling towers where it is in turn cooled by evaporation. The secondary cooling loop is open to the atmosphere, but the primary loop is not.

Buildings or st jctur6s associated with the reference BWR include 1) the reactor building which houses the reactor pressure vessel, the containment structure, l the biological shield, new and spent fuel pools, and fuel handling equipment;

2) the turbine generator building which houses the turbines and electric gen-erator; 3) the radwaste and control building which houses the solid, liquid, and gaseous radioactive waste treatment systems, and the main control room; 4) the cooling towers; 5) the diesel generator building which houses auxiliary diesel generators; 6) water intake structures and pump houses; 7) the service building which houses the makeup water treatment system, machine shops, and offices; and

8) other minor structures.

10/07/87 5-2 NUO586 CH 5

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In reference BWR, the reactor building, the turbine generator building, and the radwaste building are the only buildings containing radioactive materia s.

Tra reactor core and its pressure vessel are highly radioactive, as is the piping to the turbines. The turbines are also radioactive, but the cooling tcwers and associated piping are not, since the design of the system is such that any leakage would be from the nonradioactive secondary loop to the primary loop. Much equipment in the radwaste building is radioactively contaminated, as is the spent fuel pool in the reactor building.

The major sources of radiation in decommissioning a BWR are associated with the reactor itself, the containment structure, the concrete biological shield, the primary loop, the turbines, and the radwaste handling systems.

5. 2 BWR Decommissioning Experience At the present time, the Elk River, Minnesota, demonstration reactor is the only power reactor that has been completely dismantled.' This was a 58.2-MWt BVR that was dismantled between 1971 and 1974. While this reactor was quite small compared to present-day power reactors, its decommissioning served to desnonstrate that reactors can be decontaminated safely with little occupational or public risk. At Elk River, the containment building was kept intact until the pressure vessel and biological shield were removed. Only after all of the radioactive metal components and concrete areas were removed was the concrete containment structure demolished.

Other reactors, all of them relatively small, have been placed in safe storage or entombed (Table 1.5-1). Safe storage and entombment require surveillance and retention of a possession-only license. At Elk River, the license was te rminated.

5. 3 Decommissioning Alternatives The decommissioning alternatives considered in this section are DECON, SAFSTOR, and ENTOMB.

10/07/87 5-4 NU0586 CH 5

5.3.1 DECON DECON meansym the N C@ removal and disposal of all radioactivity in excess of levels which would permit r01 ease of the facility for unrestricted use.

Nonradioactive equipment and structures need not be torn down or removed as part of a DECON procedure. The end result is the release of the site and any remaining structures for unrestricted use as early as 6 years after the end of reactor operation.

DECON is advantageous because it allows termination of the NRC license shortly after cessation of facility operations and eliminates a radioactive site.

DECON is advantageous if the site is required for other purposes, if the site has become extremely valuable, or if the site for some reason must be ime-diately released for unrestricted use. It is al.so advantageous in that the ,

reactor operating staff is available to assist with decommissioning and that continued surveillance and maintenance is not required. A disadvantage is the higher occupational radiation dose which occurs during DECON compared to the other alternatives.

The basic estimates in the original PNL studies have been adjusted by PNL analysts to reflect January 1986 costs. The revised estimate for the reference BWR shows that DECON would require 6 years to complete, including 2 years of planning prior to reactor shutdown, and would cost $108.9 million in 1986  ;

dollars (Table 5.3-1). In addition to the values escalated from the PNL report, (NUREG/CR-0672),1 the table also includes the cost additions--for pre-decommissioning engineering, additional staff to assure meeting the 5 ren/ year dose limit for personnel, extra supplies for the additional staff, and the addi-tional costs associated with the option of utilizing an external contractor to conduct the decommissioning effort--which were developed in the PNL cost update done for the Electric Power Research Institute.2 The estimated decommissioning cost impacts of post-TMI-2 requirements on the reference BWR8 are included in the table as well. It can be seen from the table that the total cost of DECON is about $131.8 million under the utility plus-contractor option. For com-parison purposes, the time required to plan and build a large power reactor is presently about 12 years and the cost is well over two billion dollars.

10/07/87 5-5 NUO586 CH 5

Taole 5.3-1 Summary of reevaluated decommissioning costs for the reference ShNt in 5 Millions I** I Decommi ssioning EN1088I 'E Prep. for SAFSTOR I' Internals Element DECON ICI Safe Storage IdI 10 Years Internals 100 years of 30 Years 100 Years Included (g) Removed Servelliance (h)

Base Case Estimated Decommissioning Costs: -

(1978 dollars) (43.6) (21.3) 1986 dollars 98.5 (57.4) -

(58.9) (55.0) (35.0) (40.6) (3.9)

E 8.7 81.4 6.4 Safe storage Preparation NA8 37,5 Continuing Care,, 41.0 41.0 NA M MA M4 0.9 Il 41.0(3 Deferred 3.3 Il} 11.6 (h) (h)

DecontaminationId) NA NA 82.2 82.2 48.0 NA NA Possible Additional CostsI3)

Additional Staff Needed to

- duce Average Annual Radiation Dose to:

5 rem per year 4.4 1.1 2.7 2.3 UseofExternalDecgs-sioning Contractor 21.1 8.8

) 17.8 21.3 Pre-Decommissioning Engineering:

Internal (utility) 5.6 3.4 4.5 4.5 or 4.5 5.6 5. 6 Enternal (contractor) 7. 4 4.5 7. 4 7.5 Supplies for Entra Staf f 4

($ rea/yr average dose)III .02 0.1 s0.1 +0.1 NRC Licensing Activities (" > 0.1 s0.1 so.I II so.3 II s1.0 II sO.1 +0.1 s1. 0 Post-TMI-2 Impacts:

Internal (utility)I"I *0.1 negligible IP} +0.1 or so.1 negilgible s0.1 +0.1 +0.3 I

i E-G w

Table 5.3-1 (Continued)

• ENTOIS II)

Decommissioning Prep. for $ I """"* I"

• I #*" *

[lement DECONIg) Safe Storage (d) 10 Yeers 30 Years 100 Years Included (g) Removed Surveillance (h)

[aternal (contractor)I*) < 0.1 negligible *

. <0.1 <0.1 Subtotal (<5 ree/yr):

Utility (Internal) 108.9 41.0 128.3 - 130.4 106.1 77.3 89.6 7.4 Contractor ( ternal) Staffing 131.8  %.2 112.8 7.4 10TAL [stimated Cost: ..

Utility Staffing 108.9 128.3 131.4 106.1 84.7 97.0 or Contractor Staffing 131.8 104.3 120.2 j

I'T

TABLE 5.3-1 Footnctes (a) Values include a 25% contingency and are in constant 1986 dollars.

(b) Values exclude cost of disposal of last core, exclude cost nf demolition of norradioactive structures, and exclude cost of deep geologic disposal of dismantled, highly activated components.

(c) Adapted from Reference 1, Table 10.1-1, unless otherwise indicated.

(d) Adapted from Reference 1, Table 10.2-1, unless otherwise indicated.

(e) The values shown for SAFSTOR include the costs of the preparations for safe storage, continuing care, and deferred decontamination.

(f) Adapted from Reference 1, Table 10.3-1 and Appendix K.

(g) Does not include the eventual costs associated with the removal, packaging, and disposal of the entombed radioactive materials, the demolition of the entombment structure, or demolition of the Reactor Building.

(h) The annual cost of surveillance and maintenance for the entombed structure is estimated to be about $0.064 million.

(i) NA-not applicable.

I (j) Adapted from Reference 1 Table 2.10-4.

(k) Adapted from Reference 1, Table J.7-2.

(1) Adapted from Reference 2, Table 1.1, unless otherwise indicated.

(a) The values shown include the estimated costs of NRC licensing activities as well as the costs associated with inspections anticipated to be required by other Federal and state agencies. -

(n) Adapted from Reference 3, Table 2.5-7.

(o) Adapted from Reference 3, Table 2.5-7 and from Reference 1 Appendix 0.

(p) Negligible means less than $0.025 million.

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le 5.3-2 Summary of radiation dose analyses for decommissioning the reference BWR (values are in man-rem)(*)

ENTOMB with SAFSTOR After Internals Interi.:A DECON 10 Years 30 Years 100 Years Included Removed Gccupational Exposure sah Storage Preparation NA(b) 294 294 294 NA NA Concinuing Care NA 1 7 10 neg neg Decontar.ination 1764 495 36 neg NA NA

'l Entombment NA NA NA NA 1492 1603 Safe Stor. Prep. Truck Ship e ts NA 22 22 22 NA NA Decontamination Trc k inipe nts 110 22 2 neg NA NA Entombment Truck SMpcents NA NA NA NA 51 69 Total 1874 834 361 326 1543 1672 Public Exposure Safe Storage Preparation NA neg neg neg NA NA Continuing Care NA neg neg neg neg neg Decontamination neg neg neg neg NA NA Entombment NA NA NA NA neg neg Safe Stor. Prep. Truck Shipments NA 2 2 2 NA NA Decontamination Truck Shipments 10 2 neg neg NA NA Entombment Truck Shipments NA NA NA NA 5 7 Total 10 4 2 2 5 7

(*)All entries are,from Reference 1. Values exclude radiation dose from disposal of last core.

(b)MA means not applicable and neg means negligible.

i Three important radiation exposure pathways need to be considered in the evalu-ation of the radiation safety of normal reactor decommissioning operations:

inhalation, ingestion, and external exposure to radioactive materials. For l reasons similar to that discussed for PWRs in Section 4.3.1, during decommis- I sioning the dominant exposure pathway to workers is external exposure while for the public the dominant exposure pathway is inhalation. During the trans-port of radioactive waste, the dominant exposure pathway is external exposure for both transportation workers and the public. A summary of the radiation doses resulting from these pathways is presented in Table 5.3-2.

The aggregate occupational radiation dose from external exposure to surface contamination and activated material, not including transportation of radio-active waste, is estimated to be about 1764 man-rem over 4 years, or an average of 440 man-rem per year. (Table 5.3-2). The occupational radiation dose to truck transportation workers from DECON waste shipments is estimated to be about 110 man-rem.(a) In comparison, the average annual occupational radiation dose from operation, maintenance, and refueling of BWRs from 1974 through 1979 was approximately 670 man-rem per reactor 5 and 1,136 man-rem in 1980.

The inhalation radiation dose to the public from airborne radionuclide releases during DECON is estimated to be negligible. The radiation dose to the public from the truck transportation of radioactive wastes from DECON is estimated to be about 10 man-rem.

A major reason for the difference in cost and radiation dose between DECON of a BWR and a PWR is the requirement to dismantle, remove, and dispose of the radioactive turbine, condenser, and main steam piping of a BWR. A PWR turbine is not significantly contaminated with radioactivity since the major portion f

l of the radioactivity is confined to the primary coolant systems.

(*)For a detailed explanation of the minor contributions (e.g. , less than 0.08 man-rem for DECON) from post-THI-2 impacts to the total estimates shown in Table 5.3-2, consult Table 2.4-2 of Reference 3.

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t 5.3.2 SAFSTOR Generally, the purpose of SAFSTOR is to pennit residual radioactivity to decay tc levels that will reduce occupational radiation exposure during subsequent, final decontamination. As indicated in Table 5.3-2, most of the occupational dese reduction due to decay occurs during the first 30 years after shutdown with considerably less dose reduction thereaf ter. The public dose will always be small and will also experiences most of its reduction during decommissioning within the first 30 years. Nonradioactive equipment and structures need not be removed, but eventually all radioactivity in excess of that allowed for un-restricted use of the facility must be removed. Hence, in contrast to DECON, to take advantage of the dose reduction, the safe storage period could be as long as 60 years including final decontamination. The end result is the same: release of the site and any remaining structures for unrestricted use.

SAFSTOR is advantageous in that it can result in reduced occupational radiation egosure in situations where urgent land use considerations do not exist. Dis-advantages are that the owner is required to maintain a possession-only license <

under 10 CFR Part 50 during the safe storage phase and to meet its requirements at all times, thus contributing to the number of sites dedicated to radioactive materials storage for an extended time period. Other disadvantages are that surveillance and monitoring are required, the cumulative dollar costs are higher than for DECON, and the original operating staff will not be availabla at the end of the safe storage period to assist in the decontamination.

The PNL study shows that the costs of SAFSTOR for a 30 year period are greater than those of DECON and vary with the number of years of safe storage. For example, the total cost of 30 year SAFSTOR is estimated to be $131.4 million in 1986 dollars compared with the total cost of $108.9 million for DECON.

However, the total cost of 100 year SAFSTOR is estimated to $106.1 million in 1986 dollars. The lower cost of 100 year SAFSTOR compared to 30 year SAFSTOR is the result of lower costs for deferred decontamination due to the radio-activity having decayed. PNL's cost estimates for the decomissioning alter-natives are presented in Table 5.3-1.

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SAFSTOR results in lower radiation doses to both the work force and the public than DECON or ENTOMB. The aggregate occupational radiation dose is estimated to be approximately 337 man rem for 30 year SAFSTOR (294 man-rem from safe storage preparation, 7 man rem from continuing care, and 36 man-rem from deferred decontamination), not including transportation (Table 5.3-2). The occupational radiation dose from the truck transport of radioactive wastes is estimated to be about 24 man-rem. For 100 year SAFSTOR the estimated occupational radiation dose is estimated to be approximately 326 man-rem (294 man-rem from safe storage preparation,10 man-rem f rom continuing care, and a negligible dose from deferred decontamination). The occupational radiation dose from the truck transpot t of radioactive wastes is estimated to be about 22 man-rem. Thus,100 year SAFSTOR results in little additional reduction in the aggregate occupational radiation dose compared to 30 year SAFSTOR.

Radiation doses to the public from airborne radionuclide releases resulting from SAFSTOR are estimated to be negligible. The radiation dose to the public from the truck transport of radioactive wastes during the preparation for safe storage is estimated to be about 2 man-rem, and that from the truck transport of radioactive wastes during deferred decontamination af ter 30 and 100 years of safe storage is estimated to be negligible.

5.3.3 ENTOMB ENTOMB means the complete isolation of radioactivity from the environment by means of massive concrete and metal barriers until the radioactivity has decayed to levels which permit unrestricted release of the facility. These barriers must prevent the escape of radioactivity and prevent deliberate or inadvertent intrusion. The length of time the integrity of the entombing structure must be maintained depends on the inventory of radioactive nuclides present. A BWR will contain 80Co as the largest contributor to radiation dose. If it has been operated only a short time the integrity of the entombing structure need only be maintained for a few hundred years, as the disappearance of radioactivity is controlled by the 5.27 year half-life of 80Co and the 30 year half-life fission products such as 137Cs. If, on the other hand, the reactor has been operated for 30 or 40 years, substantial amounts of SSNi and H Nb (80,000 year 10/07/87 5-12 NUO586 CH 5 l

ad 20,000 year half-lives, respectively) will have been accumulated as ccti-vation products in the reactor vessel internals. The dose rate from the 94Nb present in the reactor vessel internals has been estimated to be approximately C.7 rem / hour while the dose from the 5SNi in the internals is 0.07 rem / hour.

These dose levels are substantially above acceptable residual radioactivity levels and, because of the long half-lives of 84Nb and 59Ni, would not decrease by an appreciable amount, due to radicactive decay, for thousands of years. In addition, there are an estimated 1,000 curies of 59Ni in the reactor vessel internals which could result in potential internal exposures in the event of a breach of the entombed structure and subsequent introduction of the 59Ni in an exposure pathway during the long half-life of 59Ni. Thus, the long-lived isotopes will have to be removed or the integrity of the entombing structure will have to be maintained for many thousands of years.

ENTOMB for a BWR is limited to the containment vessel because its unique struc-ture lends itself to entombment and because it contains most of the radioactiv-ity in the facility. Other buildings associated with a reactor must be decom-missioned by another method such as DECON. It is possible, however, to move some radioactive components from other buildings to the containment vessel and ENTOMB them there, rather than shipping them offsite.

ENTOMB is advantageous because of reduced occupational and public exposure to radiation compared to DECON, because little surveillance is required, and because little land is required. It is disadvantageous because the integrity of the entombing structure must be assured in some cases for hundreds of thou- ~

sands of years, because a possession-only license under 10 CFR Part 50 would be required which in turn requires some surveillance, monitoring, and main-tenance, and because entombing contributes to the number of sites dedicated to radioactive materials containment for very long time periods.

Two approaches to the ENTOMB alternative for a BWR are possible. In the first approach, the pressure vessel internals and their long-lived ssNi and 84Nb isotopes are entombed, along with other radioactive material. This results in less cost and radiation dose because the pressure vessel and its internals will not have to be removed, dismantled, and transported to a deep geologic waste 10/07/87 5-13 NUO586 CH 5

reposi tory. It will also, however, result in the requirement for a posse ssion-only license and indefinite surveillance because of the presence of the long-lived isotopes.

In the second approach, the pressure vessel internals, with their long-lived t 5'Ni and S4Nb isotopes, are removed, dismantled, and transported to a radio-active waste repository. This results in more cost and radiation dese, but offers the possibility that surveillance and the possession unly license could be terminated at some time within several hundred years, thereby releasing the entire facility for unrestricted use. At the outset, a careful inventory of radioactivity would need to be made to ensure that only relatively short-lived isotopes were present.

In both approaches, as much solid radioactive material from the entire facility as can be accommodated is sealed within the containment vessel. All openings to the exterior of the containment vessel are sealed. Radioactive material outside the containment vessel is removed down to levels which permit release of the remainder of the facility for unrestricted use.

Radioactive materials not entombed would have to be packaged and transported to a disposal site. Cost savings and radiation dose reductions would result from the lesser volume of radioactive equipment and material having to be dis-mantled, packaged, and transported. In any case, all spent fuel would be removed.

ENTOMB for the reference BWR, including the pressure vessel and its internals, '

is estimated to cost $77.3 million, with an annual surveillance and maintenance cost of $64,000. It results in an aggregate radiation dose of 1492 man-rem to decommissioning workers, 51 man-rem to transportation workers, and 5 man-rem to the general public. ENTOMB for the reference BWR, with the pressure vessel internals removed, is estimated to cost $89.6 million, with an annual surveil-lance and maintenance cost of $64,000, and to result in an aggregate radiation dose of 1603 man-rem to decommissioning workers, 69 man-rem to transportation workers, and 7 man-rem to the general public. These estimates are listed in Tables 5.3-1 and 5.3-2, 10/07/87 5-14 NU0586 CH 5

5. 3. 4 Sensitivity Anaiyses In addition to the reference BWR, PNL also analyzer! a variety of realistic decommissioning situations.1 These variations were studied to determine if they might have significant impact on the conclusions reached for the reference BWR regarding doses and costs for the decomissioning alternatives. While there were some differences in results, the conclusion of the sensitivity analysic is that these differences do not substantially affect the original cost and radiation dose conclusions. Of the various situations analyzed by PNL, the most important with regard to their potential effect on dose and cost estimates are reactor size, degree of contarination and type of containment structure.

Based on an analysis 8 similar to that for the reference BWR (NUREG/CR-0672) and incorporating selected cost adders (described in References 2 and 3 and escalated to constant 1986 dollars as shown in Table 5.3-1), upper-bound estimates were made of the costs for immediate dismantlement of reactor plants smaller than the reference plant. The analysis was limited to plants with thermal power ratings greater than 1200 MWt and was based on the assumption that all costs (staff labor, equipment, supplies, etc.) except radioactive waste disposal are independent of plant size. The results are shown in Table 5.3-3.

Table 5.3-3 Estimated imediate dismantlement costs (in millions) for plants smaller than the referegegR, based on previously-derived .

overall scaling factors Waste Scaling Remaining Escalated Costs Total (c)

Reactor MWt Disposal Factor Costs Adders WNP-2 3320 44.201 1.000 54.464 10.230 108.894 Cooper 2381 44.201 0.809 54.464 10.230 100.453 Vermont 1593 44.201 0.648 54.465 10.230' 93.336 Yankee

(*)All costs are in constant 1986 dollars and include a 25% contingency.

(b) Derivation of previously-derived overall scaling factors can be found in Reference 1.

(C) Total costs shown above are for the utility-only cost option.

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Using the results from Table 5.3-3, a linear ecJation can be derived for the scaling of the immediate dismantlement costs of plants in the 1200 to 3500 W t range:

Cost = 78.993 + (9.008 x 10 8) (Wt)

Revised overall scaling factors fer the Cooper and Vermont Yankee pla ts were obtained by dividing the results of the linear equation by the cost cd the reference plant. Based on this formula, a list of variations in dose and cost of these BWRs is presented in Table 5.3-4.

Also analyzed was the sensitivity of decommissioning costs and radiation doses to a postulated tripling of radiation dose rates from radionuclides deposited in BWR coolant system piping during reactor operation over a period of 30 to 40 years. This tripling of dose rate is postulated as an upper limit on the basis of recent trends for operating reaci. ors. If no corrective action is taken to reduce the radiation dose rates, the accumulated radiation dose to decom-missioning workers for DECON would be increased from 1764 man-rem to 4573 man-rem,1 and the total decommissioning costs could be increased by about 12 million for DECON. For ENTOMB the radiation dose would be increased from 1604 man-rem to 4154 man-rem and the total cost could be increased about 12 million.

For preparation for safe storage, the radiation dose would be increased from 294 man-rem to 759 man-rem, and there would t'e no significant change in the cost.

In order to handle these postulated higher initial radiation levels, it appears that additional chemical decontamination during decommissioning would be the most cost-effective approach. For example, it is estimated that increasing the circulation time of the chemical solution about 50% would reduce the postulated increased radiation levels by a factor of 3, thus reducing these levels to approximately the same dose rate conditions assumed in the reference case analysis. This approach would also be more consistent with the principles of ALARA, since the occupational radiation dose associated with a chemical decon-tamination cycle is relatively small, compared with the radiation dose asso-ciated with installing temporary shielding, or with attempting to perform the dismantlement without additional shielding. In addition, it appears likely that the large buildups of radionuclides prevalent, today on piping systems will 10/07/87 5-16 NUC586 CH 5

Table 5.3-4 Estimated costs and occupations 1 radiation doses for decommissioning dif ferent-sized BWR plants (a, b, c)

Station Vermont Yankee Cooper WNP-2 Power Rating (thermal megawatts) 1,593 2,381 3,320 Overall Scaling Factor (OSF) 0.857 0.922 1.000 DECON ($ millions) 93.3 100.4 108.9 (man-rem) 1.5 1 1,701 1,845(c)

ENTOMB (d) w/ internals ($ millions)(C) 66.2 71.3 77.3 (man res) 1,348 1,450 1,573 w/o internals ($ millions) 76.8 82.6 89.6 (man-rem) 1,443 1,553 1,684 SAFSTOR Preparations for Safe Storage ($ millions) 35.1 37.8 41.0 (man-res) 321 346 375 Safe Storage:

for 30 g 3rs ($ millions) 3.3 3.3 3.3 (man-rem) 6.5 6.5 6.5 for 50 yer:1, ($ millions) 5.6 5.6 5.6 (man-ren) 10 10 10 for 100 years ($ millions) 11.7 11.7 11.7 (man-rem) 10 10 10 Deferred Disuantlement:

after 30 years ($ millions) 70.4 75.8 82.2 (man-rem) 31 33 36 after 50 years ($ millions) 41.4 44.5 48.3 (man-rem) 2.6 2.8 3 after 100 years ($ millions) 41.1 44.3 48 (man-rem) >1 >l >l Facility Demolitian ($ millions) 16.4 18.0 19.9

(*) Values include a 25% contingency and are in 1986 dollars.

(b) Costs do not include spent-fuel disposal or dersettien of nonradioactive structures.

Doses are taken from Reference 1 and do not include those due to transportation of wastes.

fC d

ENTOMB costs do not include continuing care costs (0.064 M/yr).

l l

l be prevented as periodic decontamination during normal operation of the reactor l coolant system and related fluid-handling systems becomes standard procedure.

Analysis was also done to determine if variation in design of the BhR containment structure would have significant impact on doses or costs of dei esissioning.

There are three principal designs of BWR containments and pressu suppression systems, namely Mark 1, Mars II, and Mark III and these were an' .ed by P:ll.

The conclusion reached by thi analysis was that for BWR plants of equivalent power rating, olfferences in containment design have very little effect on the total cost of decommissioning of a BWR, One of the circumstances that has changed since the original BWR decommissioning report 1was prepared which could influence the development of the cost of dose estimates presented in this GEIS is an assessment of post-TMI-2 requirements on the decommissioning of the reference BWR. Actions judged necessary by the NRC to correct or improve the regulation and operation of nuclear power plants based on the experience from the accident at TMI-2 result,d in a number of recommerida-tions that were subsequently issued to the utilities as requirements. Some of those requirements resulted in equipment and hardware changes andor additions to the reference BWR that could eventually expand the scope of decommissioning activities, since those materials could reasonably be expected to become con-taminated or radioactive during the remaining operational lifetime of the plant.

For the reference BWR, it was concluded by PNL in a recent study 3 that the original immediate disn. o 'ement decommissioning cost estimates could be expected to increase very ightly overall (considerably less than 1% in -

constant 1986 dollars), due to a slightly expanded scope of decommissioning activities a3sociated with changes in the reference plant's characteristics.

The radiation dose would be increased by about 3 man-rem, due entirely to decommissioning operations associaced with the removal and packaging of a small additioral quantity of contasinated materials.

6 Other methods of facilitating decommissioning, in addition to additional chemical decontamination, are discussed in NUREG/CR-0569.7 These include improved documentation, reduction of radwaste volume by incineration, electro-polishing of pioing and components as a decontamination technique, remote main-tenance and decommissioning equipment (robots), improved access to piping and components, and improved concrete protection.

10/07/87 5-18 NUO586 CH 5

l l

l 5.4 Environmental Consequences l

Radiation doses and costs associated with possible decommissioning alternatives I are discussed in Section 5.3. It is to be emphasized for perspective that for any viable decommissioning alternative, the environmental effects of greatest concern, i.e. , radiation dose and radioactivity released to the environment, am substantially less than the same ef fects resulting from reactor operation and maintenance. It should also be noted that while the dollar costs of ENTOMB are less than those of DECON, the environmental impacts could be quite high should large amounts of radioactivity escape from a breached structure during the entombment period.

Other environmental consequences are rather different from the environmental consequences usually discussed in environmental impact statements. This is f because, usually, an environmental impact statement is addressed to the conse-quences of building a facility that will require land, labor, capital investment, saterials, cuntinuing use of air, water and fuel, a socioeconomic infra-structure, etc. Decommissioning, on the other hand, is an attempt to restore things to their original condition, which requires a much smaller commitment of resources than did building and operating the facility.

A major environmental consequence of decomnissioning, other than radiation dose and dollar cost, is the commitment of land area to the disposal of radioactive waste. Estimates are shown in Table 5.4-1 of the low-level waste disposal .

volume required to accomtrodate radioactive waste and rubble removed from the facility and transported to a licensed site for disposal. The volume for ENT0M8 does not include the volume of the entombing structure or of the wastes entombed within it, only the wastes shipped off-site. The entombing structure is, in effect, a new radioactive waste burial ground, separate and distinct from the ones in which the wastes in Table 5.4-1 are buried, and may necessitate licens-ing consideration such as those for a low-level waste burial ground under I (10 CFR 61).

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Table 5.4-1 Estimated burial volume of low-level radioactive waste and rubble for the reference BWR Decommissioning Alternative Volume (m3 )

DECON 18,975(a)

SAFSTOR Deferred Decontamination (b) following Safe Storage f or: 10 Years 18,975(a,c) 30 Years 18,975 50 Years 1,783 100 Years 1,673 ENTOMB (d)

Internals Included 8,042 Internals Removed 8,420 (a) Includes about 36m 3 of radioactive waste attributable to removal of oackfitted material (adapted from Table 5.2-8, Reference 3).

(b) Radioactive wastes from preparations for safe storage are small in compari- ,

son to those from deferred decontaminetion.

(C)Although, in actuality, there is a gradual decrease in waste volume over time, it is not indicated here for clarity of presentation.

~

(d) Volume of entombing structure and the wastes within are not included.

If shallow-land burial of radioactive wastes in standard trenches is assumed, then a burial volume of about 18,975 m 3 of radioactive waste can be accommo-dated in less than 2 acres. The two acres is small in comparison with the 1,160 acres used as the site of the reference BWR.

Certain highly activated corponents of the reactor and its internals may require disposal in a deep geologic disposal facility rather than in a shallow-land burial ground because of the large initial level of radioactivity and the very long half-lives of 5SNi and 94Nb. Only about 11.5 m3 of material would be involved and would require approximately 89 m3of waste, disposal space.

10/07/87 5-20 NUO586 CH 5

The cost for disposing of these materials in deep geologic disposals was esti-mated by PNL to be about $2.9 milliun (in 1978 dollars).1 Based on recent estimates of deep geologic disposal costs,s it is currently estimated by PNL that deep geologic disposal of the highly activated materials would cost abot,t l 16.2 million (in 1986 dollars). This cost has not been included in the costs of decommissioning shown in Table 5.3-1.

PNL considered accidental releases of radioactivity both during de:ommissioning during transport of wastes and the results are presented in Table 5.4-2. Racia-tion doses to the maximally-exposed individual from accidental airborne radio-activity releases during decommissioning opera; ions were calculated to be quite low. Radiation doses to the maximally-exposed individual from accidental radioactivity releases resulting from transportation accidents wera calculated to be low for the most severe accident.

Other environmental consequences of decoarnissioning are minor cor. pared to the environmental consequences of building and operating a BWR. Water use and evaporation at the rate of as much as 27 x 108 m 3/yr ceased when the reactor ceased operation. The total water use for decommissioning is estimated to be about 18 x 103 m. 3 The number of workers on site at any time will be no greater than when the BWR was in operation and will be much less than when the BWR was under construction. The transportation network is already in place, but will require some maintenance if the SAFSTOR mode is selected.

Disturbance of the ground cover need not take place to any appreciable extent except for filling holes and leveling the ground following removal of under-ground structures, unless operation of the plant has resulted in contamination of the ground around the plant. Plowing of the ground would generally result in lowering average soil contamination levels to those acceptible for releasing the site for unrestricted use, except for a few more nighly contaminated areas where materials would have to be removed. In this case, soil to depth of several centimeters and some paving may have to be removed, packaged, and shipped to a disposal facility before the site can be released for unrestricted use.

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Table 5.4-2. Summary of radiation doses to the maximally-exposed fr.dividual from accidental alrborne radionuclide releases during BW decommissioning and transportation of wastes Total Atmospheric Radiation Dose to Lung (in ree) from:

Release DECON SAF5 TOR ENTOMB Incident (Ci/hr)(b) First-Year Fifty-Year First-Year Fifty-Year First-Year fifty-Year Occurrence (a)

Severe Transportition -

Accident 2.0 x 10 2 9.0 x 10 8 2.0 x 10 8 9.0 x 10 2 2.0 x 10 8 9.0 x 10 2 2.0 x 10 8 Low Explosion of LPG Leaked -

from a Front end Loader 8.6 x 10 8 7.9 x 10

• 1.5 x 10

• N/Ac N/A N/A N/A low vacuum Filter-Bag Rupture 8.5 x 10

• 8. 3 x lo

• 1.8 x 10 4 8.3 x 10 S 1.8 x 10 4 8.3 x 10 5 1.8 x 10

• Medium Minor Transportation Accident 5.0 x 10

• 2.2 x 10 8 4.5 x 10 8 2.2 x 10 8 4.5 x 10 8 2.2 x 10 8 4.5 x 10 8 Low Contamination Control Envelope Rupture 1.4 x 10

• 1.0 x 10 8 1.9 x 10 8 N/A N/A N/A N/A High Oxyacetylene f 71osion 1.2 x 10

• 8.7 x 10 7 1.6 x 10 8 N/A N/A N/A N/A Medium Contaminated Sweeping Compound Fire 1.1 x 10

• 1.1 x 10 7 2.3 x 10 7 1.1 x 10 7 2.3 x 10 7 1.1 x 10 7 2.3 x 10 7 Medium

, Cross Leak During Loop Cheefcal Decontamination 1.0 x 10

• 9.8 x 10

• 2.1 x 10 7 9.8 x 10 7 2.1 x 10 7 9.8 x 10.a 2.1 x 10 7 Low Filter Damage from Blast-ing Surges 1.3 x 10 7 1.2 x 10 ' N/A N/A N/A N/A N/A Medium 4

e I

Table 5.4-2 (Continued)

Total Atmospheric Radiatinn Dose to Luna (in ree) from:

Release

• DECON SAF5 TOR ENT0Pe Incident (C1/hr)(b) . First-Year Fifty-Year First-Year Fifty-Year First-Year fifty-Year Occurrence (a)

Combustible waste Fire 6.0 x 10 ' 5.9 x 10 8' 1.2 x 10 ' 5.9 x 10 ,8' 1.2 x 10 8 5.9 x 10 8' 1.2 x 10

• High Detonation of Unused Emplosives 4.8 x 10 8' 4.4 x 10 82 8.6 x 10 82 N/A N/A N/A N/A Medium The frequency of occurrence considers not only the probability of the accident, but also the probability of an atmospheric release of the ~

calculated magnitude. The frequency of occurrence is listed as "high" if the occurrence of a release of stellar or greater magnitude per year is >10-2, as "medium" if between 10-2 and 10 5, and as "low" If <10 5 (b)All atmospheric releases 4re assumed to occur during a 1-hr period for comparison purposes.

IC N/A = Not applicable.

The biggest socioeconomic impact will have occurred before decommissii ning started, at the time the plant ceased operation and the tax income created by the plant was reduced. No additional public services will be required because the decommissioning staff will be somewhat smaller than the operating staff.

In the case of deferred decontamination, the decontamination staff will be larger than the surveillance staff.

5. 5 Comparison of Decommissioning Alternatives From careful examination of Tables 5.3-1 and 5.3-2 it appears that DECON or 30 year SAFSTOR are reasonable options for decommissioning a BWR. 100 year SAFSTOR is not considered a reasonable option since it results in the continued presence of a site dedicated to radioactivity containment for an extended time period with little benefit in aggregate dose reduction compared to 30 year

-SAFSTOR. DECON costs less than SAFSTOR and its larger on an annual basis occupational radiation dose, which is consistent with routine annual operational dose for plant operatiens is considered of marginal significance to health and safety.

Either ENTOMB option requires indefinite dedication of the site as a radioactive waste burial ground. In the ENTOMB option with the reactor internals and its long-lived activation products entombed, the security of the site could not be assured for thousands of years necessary for radioactive decay, so this option is not considered viable. In the ENTOMB option with the reactor internals removed, it say be possible to release the site for unrestricted use at some time within the order of a hundred years if calculations demonstrate that the radioactive inventory has decayed to acceptable residual levels. However, even this ENTOMB alternative appears to be less desirable than either DECON or SAFSTOR based on consideration of the fact that ENTOMB results in higher radiation exposure and higher initial costs than 30 year SAFSTOR, that the overall cost of ENTOMB over the entombment period is approximately'the same as DECON, and the fact that regulatory changes occurring during the long entomb-ment period eight result in additional costly deccanissioning activity in order to release the facility for unrestricted use.

10/07/87 5-24 NU0586 CH 5

. 1 1

l l

l Cc . sideration was given to the situation where, at the er.d of the reactor opera-tional life, it is not possible to dispose of waste of fsite for a lir.ited period of time, but not exceeding 100 years (see Section 2.7). Such a cor.straint needs to be accounted for in selecting the decommissioning alternative. Based on an analysis by PNL of the technology, safety and cost considerations on selection of decommissioning elternatives,9 it was concluded that SAFSTOR I is an acceptably viable alternative. Unlike the PWR case, DECON ard conversion of the spent fuel pool to an independent spent fuel storage pool for a BWR is an unlikely possibility for the case where all other radioactive wastes can be removed offsite. The active phase of maintaining the spent fuel in the pool is not considered to be part of the regulatory requirements for decor.nissioning, but would be considered under the usual operating licensing aspects regarding health and safety with consideration given to facilitation for decommissioning.

Aside from the expenses incurred from storing spent fuel, other costs for keeping radioactive wastes onsite for the reactor in a safe storage mode were estimated to have minimal effect on the SAFSTOR alternative compared to this alternative for radioactive wastes being sent offsite. Site security for stor-age of spent fuel (which is considered as an operational rather than a decom-missioning consideration) was estimated at about $0.94 million per year (in 1986 dollars)(a) For a multi-reactor site, such security could result in a l lesser cost because of a sharing of required overheads.

l (a) Adapted from Reference 9.

10/07/87 5-25 hUO586 CH 5

l l

i

' $~

REFERENCEb

1. H. D. Oak et al. , Technology, Safety and Costs of Decommissioning a Reference Boiling Water Reactor Power Station, NUREG/CR-0672, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, June 1980.*

2. R. I. Smith et al. , Updated Costs for Oe:cmmissioning Nuclear Power Facilities, EPRI NP-4012, Prepared by Battelle, Pacific Northwest Laboratories for the Electric Power Research Institute Final Report, May 1985.

3. G. J. Konzek, Estimated Impacts of Post-TMI-2 Requirement. ad Other Selected Regulatory Change on Decommissioning of Two References Light Water Reactors, NUREG/CR (to be published), Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission.

4. AEC-Eik River Reactor Final Program Report, C00-65193. United Power Association, Elk River, Minnesota, revised November 1974.

5. B. G. Brooks, Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1982, U.S. Nuclear Regulatory Commission, NUREG-0713, December 1981.

6. Letter from R.I. Smith, Pacific Northwest Laboratories, to Carl Feldman, U.S. Nuclear Regulatory Commission,

Subject:

Updated Costs for I

Decomissioning the Reference PWR and BW as Developed for the GEIS, November 12,1986 (available in USNRC Public Document Room).

7. E. B. Moore, Jr. , Facilitation of Decommissioning Light Water Reactors, NUREG/CR-05G9, Prepared by Pacific N west Laboratory for U.S. Nuclear Regulatory Commission, December 197 *

8. R. L. Engel et al. ,1986. 1986 Annual Review of the Adequacy of the 1.0 mill per kWh Waste Disposal Fee, FNL-5877 Pacific Northwest Laboratory.,

Richland, Washington.

9. G. M. Holter, Technology, Safety and Costs of Decommissioning a Reference Boiling Water Reactor Power Station, Addendum 1 to NUREG/CR-0672, Ef fects on Oecommissioning of Interim Inability to Dispose of Wastes Offsite, Prepared by Pacific No hwest Laboratory for U.S. Nuclear Regulatory Commission, July 198

• Avnt&DtB for purchase from the NRC/GP0 Sales Program, U.S. Nu%

Regulatory Consnission, Washington, DC 20555, and the National Technican Information Service, Springfield, VA 22161.

M

• • Available for purchase from the National Technical Information Service.

f Y 10/07/87 5-26 NUO586 CH 5

6 MULTIPLE REACTOR STATION i Mest of the operating or planned nuclear power reactors in the United States are located at sites with two or more reactors. Twenty-six 2-reactor sites are in operation and an additional nine 2-reactor sites are being constructed.

Five 3-reactor sites are in operation. The possibility of locating mu'tiple facilities at a single site is discussed in References 1 through 4. Possibili-ties range from a small site containing two or three reactors to a very large site with up to 40 reactors and cther fuel cycle facilities as well. The 1974 AEC study! contemplated up to 40,000 We of generating capacity at a single site, together with reprocessing plants, fuel enrichment plants, and waste handling and storage facilities. The 1975 NRC study 2 contemplated power plant centers, fuel cycle centers, and combined centers. The power plant centers would con-sist of 10 to 401200 MWe-capacity nuclear reactors; the fuel cycle centers would include fuel reprocessing plants, mixed oxide fuel fabrication plants, and radioactive waste management facilities; and the combined centers would contain both nuclear power reactors and other fuel cycle facilities. The Hanford 3

Nuclear Energy Center study assumed that 20 to 40 nuclear power plants would be waste management facilities. A Science magazine article 4 examines some of these alternatives and argues for a small number of large sites each containing several reactors, as opposed to a large number of small sites with only one or two reactors at each site. .

It is the purpose of this section to investigate whether significant differences in the costs, safety, and other environmental consequences of decommissioning might exist between a reactor at a single-reactor site and one at a multiple-reactor station and whether these differences could have an effect on regulatory considerations. Most of this section is based on a PNL study of the technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations.5 In the PNL study, consideration was given to interim storage of waste, permanent onsite disposal of low-level waste, the dedication of the site to nuclear power 10/07/87 6-1 NUO586 CH 6 n ,-. - - - - . , ,- -. +- - . - ~ . -

1 generation, the avt.ilability of centralized services,(a) and the type and number of reactors present at the station. In addition, major facilitation aspects such as modular construction concepts which would allow for intact removal of the reactor pressure vessel during decommissioning were examined.

6.1 Multiple-Reactor Station Description Although most of the operating or planned nuclear power reactors in the United Statt:s are located at stations with two or three reactors, no commercial site presently exists with more than three reactors, and no multiple-reactor sites have been decommissioned. Therefore, it is necessary to develop a model that permits the identification of factors which could affect the cost and safety of decommissioning a nuclear reactor at a site where other reactors are operatina, being built, or being deconnissioned.

6.1.1 Multiple-Reactor Station Concepts The PNL studys identified several variables that could result in differences between the costs and radiation doses anticipated for decommissioning a reactor at a multiple-reactor station and decommissioning an identical reactor at a single-reactor station. These variables include the number of reactors at the multiple-reactor station, the type of reactor, the nuclear waste disposal option, dedication of the site to nuclear power generation, and the provision of central services.

In the PNL study, sites with 4 and 10 reactors were considered. It is more likely that the reactors at a multiple-reactor station with a small number of reactors (i.e. , four reactors) will be of the same type and design than it i:

for a station with a larger number of reactors. However, even at a station with a large number of reactors including both PWRs and BWRs it is probable that there will be several reactors of a given type and design. Standardization of design results in several advantages which can reduce costs and improve safety (a) Central services include health physics services, security forces, solid waste processing, and equipment decontamination services.

10/07/87 6-2 NUO586 CH 6

darino the decommissioning of identical reactors at a mu'.4ple-reactor station.

These advantages include:

minimizing the planning ef fort for decommissioni.7g t e second and later reactors of an identical or similar design improving the productivity of decommissioning worke i due to experience gained on the first reactor improving the planning of decommissioning techniques and effectively im-plementing the lessons of past experience.

Nuclear waste disposal is the major contributor to the puclic radiation dose from decomissioning a nuclear reactor and is a significant item in the decom-missioning cost. Decomissioning a reactor at a multiple-reactor station results in the same quantity of nuclear waste for disposal as dec:mnissioning an identi-cal reactor at a single-reactor station. However, options for the management of this waste which may be available at the multiple reac*.or station can result in significant cost and radiation dose reductions. To pe sit release of a site for unrestricted use, the radioactive waste from decommissioning an LWR at a single-reactor station would require disposal at an offsi*.e, licensed nuclear waste disposal facility. However, at a dedicated nuclear site (which remains restricted during dedication), options for the disposal o' decommissioning wastes include:

1. disposal at an offsite licensed low-level waste disp: sal facility

2. interim onsite storage with transfer to an offsite license low-level waste disposal facility at a later date

3. disposal at a permanent onsite low-level nuclear waste disposal facility l

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s Options 2 and 3 generally result ir .vwer osts and smaller occupational and public radiation doses for waste d'sposal than option 1.(a)

Other cost and safety benefits may result from the location of multiple electric generating facilities at nuclear erergy centers. Dedication of a site to nuclear power generation results in replacement reactors being constructed on a s:5edule to achieve startup of a replacement reactor as an old reactor is shut dow and decommissioned. At such dedicated sites, improvements in efficiency as the labor force gains experience and reduction in the planning effort required for decommissioning the second and subsequent reactors of the same or similar types could result in lower decommissioning manpower costs in reduced occupational radiation doses.

A number of onsite, centralized services may be available during decommissioning of a reactor at a multiple-reactor station. The major i.mpact of having cen-tralized services available would be reduction in the cost of decommissioning each reactor.

6.1.2 Multiple-Reactor Station Scenarios Three multiple-reactor station scenarios are chosen for illustration of the estimated ef fects of the variables described in Section 6.1.1. Details of the three scenarios are shown in Table 6.1-1. Summaries of estimated cost and occupational and public radiation dese reductions for decommissioning a reactor at a multiple-reactor station relative to decommissioning a reactor at a single-~

reactor station are given in Section 6.3.

(a)However, option 3 would necessitate licensing as a low-level waste burial ground under 10 CFR 61 in additien to a possession-only license under 10 CFR 50 for the retired reactor (s).

10/07/87 6-4 NV0586 CH 6

Table 6.1-1 Multiple-reactor station scenarios Number of Waste Disposa' Reactors of Onsite Onsite Dedicated Central Scenario Same Type Immediate Interim Permanent Site Services Number Decommissioned Offsite Storage Disposal No Yes No Yes Single-reactor 1 X X X station I 2 X X X II 4 X X X III 4 X X X The three scenarios evaluated for multiple-reactor station decommissioning a re:

Scenario I

- the station is small (e.g. , four reactors onsite) the two reactors being decommissioned are of the same type ,

- nuclear waste is temporarily stored onsite and moved later to an offsite licensed disposal facility

- the site is not dedicated to nuclear power generation (i.e. , a replacement reactor is not started up as each old reactor is shut down) ,

- central facilities are not provided onsite Scenario II

- the station is large (e.g. , ten reactors onsite)

- the four reactors being decommissioned are of the same type

- nuclear waste is temporarily stored onsite and moved later to an offsite licensed disposal facility 10/07/87 6-5 NUO586 CH 6

I l

the site is dedicated to nuclear power generation (i.e. , a replacement reactor is started up as each old reactor is shut down) central facilities are provided onsite Scenario III the station is large (e.g., ten reactors onsite) the four reactors being decommissioned are of the same type low-level nuclear waste is permanently disposed of onsite the site is dedicated to nuclear power generation (i.e., a replacement reactor is started up as each old reactor is shut down) central facilities are provided onsite.

6.1.3 Reference Light Water Reactors The reference reactors for this analysis of reactor decommissioning at multiple-reactor stations are the same as those described in PNL studiess 7.s of the l

decornmissioning of light water reactors at single-reactor power stations. The reference PWR plant is an 1175 We (3500-Wt) Westinghouse pressurized water reactor, specifically the Trojan Nuclear Plant at Rainier, Oregon, operated oy ~

the Portland General Electric Company. The reference BWR plant is an 1155-We (3220-Wt) General Electric boiling water reactor operated by the Washington Public Power Supply System; it is designated as the WPPSS Nuclear Project No. 2 and is located near Richland, Washington. These reactors are also used as bases for the decommissioning cost and safety information presented in Chapters 4 and 5 of this GEIS. A brief description of the reference PWR is given in Section 4.1; a brief description of the reference BWR is given in Section 5.1.

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)

6.2 Multiple-Reactor Station Decommissioning Experience ho multiple-reactor stations containing more than three reactors have been built in the United States, and no multiple-reactor statiois have been decommissioned.

Therefore, there is no decommissioning history to report. Brief histories of decommissioning individual commercial nuclear power reactors are given in Sections 4.2 and 5.2 of this EIS.

6. 3 Multiple-Reactor Station Decommissioning Scenarios In this section, the costs and radiation doses for decomrrissioning a reactor at a multiple-reactor station are compared with those for decommissioning an identical reactor at a single-reactor station. The decommissioning alternatives considered are DECON, SAFSTOR, and ENTOMB. Decommissioning costs are summarized in Table 6.3-1 for the reference PWR and in Table 6.3-2 for the reference BWR.

Costs are in 1986 dollars and include a 25% contingency. Occupational dose information is summarized in Table 6.3-3 for the reference PWR and in Table 6.3-4 for the reference BWR. Public dose information is summarized in Table 6.3-5 for the reference PWR and in Table 6.3-6 for the reference BWR.

The data in these tables are derived from the PNL study 5 on the technology, safety and costs of decommissioning nuclear reactors at multiple reactor sta-tions and include, where applicable, the costs and doses attributable to nuclear wastes associated with post-THI-2 backfit requirements (safety upgrades).9 The bases and assumptions used to estimate decommissioning costs and safety are given in the PNL reports. -

Waste disposal options evaluated include: (1) interim onsite storage of waste with later permanent disposal offsite, and (2) permanent onsite disposal. In-teria onsite storage would be designed to remotely place the conta:ners of waste in storage cells and remotely remove the containers at the end of the storage period. Onsite storage involves the following tasks:

packaging transporting to interim onsite storage placing in interim storage 10/07/87 6-7 NUO586 CH 6

Table 6.3-1 Summary of estimated cost reductions when decommissioning each reference PWR at a multiple-reactor station

• DECON SAFSTOR ENTOMB Cost,C . Cost Reduction Cost,C Cost Reduction Cast factor $ millions Cost.C Cost Reduction 5 millions percent $ allifons 5 millions percent 5 millions 5 millions percent Waste Disposai d

• Immediate Offsite Disposal 40.112 -- '

40.827 -- --

12.609 -- --

Onsite Intgrim Storage for 30 Years 59.770 (19.658)g (49.0)7 37.042 3.785 9.3 13.436 (0.827) (6.6)

Onsite Intgrim Storage for 50 Years 37.339 2.773 6. 9 36.742 4.085 10.0 10.514 2.095 16.6 Onsite Integio StoragF for 100 Years 37.259 2.853 7.1 36.567 4.260 10.4 to 418 7. t in

!=nediate Onsite Olerneal 17.19'. I.911 19.1 Jt. ins 8.b42 21.2 ti t ti. 6 t t 5. *>9tt 4 / . e.

Decommissioning Staff Labor 8 No. of Reactors of Same Type:

1 29.183 --

2 31.473 -- --

24.802 --

26.750 2.433 8.3 28.024 3.449 4 25.009 11.0 22.478 2.324 9.4 4.174 14.3 25.798 5.675 18.0 20.891 3.911 15.8 C ntral Services i

Without Central Services 9.384 -- --

11.489 -- --

I With Central Services 4.998 4.386 9.384 -- --

46.7 5.866 5.623 48.9 4.998 4.386 46.7 Tstalsforgecommissioning Scenarios Single-Reactor Station 88.7 --

Scenario I 96.8 -- --

57.2 -- --

Interim Storige for 30 Years 105.9 (17.2)

Interim Storage for 50 Years (19) 89.6 7.2 7.4 55.7 1. 5 83.5 5.2 6 89.3 3 Interim Storage for 100 Years 83.4 5.3

7. 5 7.8 57.8 4.4 8 6 89.1 7. 7 8.0 52.7 4.5 8 l

1 6T -

5ummertred from Chapter 8, Append 1m A, and Appendia B of Reference 5.

b For 30 years safe storage. Values are the sum of the cost of preparations for safe storage plus deferred decontaelnation.

"Casts are in 1986 dollars and include a 251 contingency.

Values exclude the costs of disposal of the last fuel core, exclude cost of demolition of nonradioactive structures and exclude costs of deep geologic disposal of activated components.

' Includes the cost of placement in interim storage plus the cost of removal at a later date to periaanent offsite disposal.

Parentheses indicate a cost increase.

balues include labor costs for both planning and preparation'and decommissioning operations. Security force labor costs are not included.

"Central services include health physics services, security services, solid waste processing, and equipment decontamination services.

I Nittple-reactor station scenarios are described in detail in Section 6.1.2 1

l

Table 6.3-2 Summary of estimated cost reductions when decommissioning each reference 8 Wit at a multiple reactor station

• DECON SAFSTOR 9 ENTOMB Cost,C . Cost Reduction Cost," Cost Reduction Cost C Cost Reduction

$ millions percent 5 elllions 5 millions percent 5 milliois I millions percent Cast factor 5 millions Weite Disposai d .

Immediate Offsite Disposal 44.159 -- -- 40.159 -- --

25.814 -- --

Onsite Intgrim Storage for (30.7), 34.778 9.381 21.2 27.630 (1.816) (7.0) 30 Years 57.703 (13.544)7 Onsite Interim Storage for 50 Years 33.69T 10.467 23.7 32.74R 11.411 25.8 17.110 8.444 37. 7 Onsite IntegIn Storage for 26.7 11.030 8.184 34.0 100 Years 33.335 10.824 24.5 32.359 11.800 Immediate Onsite Storage 29.633 14.526 32.9 29.500 14.659 33.2 14.063 11.751 45.5 Decommissioning Staff Labor 9 No. of Reactors of Same Type:

1 40.195 -- --

56.443 -- --

38.844 -- --

2 37.216 2.979 7.4 51.940 4.501 8.0 35.906 2.938 7.6 4 34.974 5.221 13.0 48.641 7.800 13.8 33.715 5.129 13.2 h

Central Services without Central Services 14.512 -- --

20.020 -- --

14.976 --

With Central Services 8.986 5.526 38.1 12.403 7.617 38.1 9.213 5.763 38.5 T4talsforgecommissioning Scenarios Single-Reactor Station 108.9 -- --

128.1 -- --

89.6 -- --

9 G - i<,

tabla 6.3-2 (Continued)

DECON SAFSTOR(d) gg Cost.IC) Cost Reduction Cost IC) Cost Reduction Cost.IC) Cest Reduction Cost factor $ millions 5 millions percent $ mil 11oas 5 millions percent 5 millions G llions percent Scenario I -

Interim Storage for 30 Years 119.5 (10.6) (10) 114.2 13.9 11 88.5 1.1 1 Interim Storage for 50 Years 95.5 13.4 12 112.2 15.9 12 78.2 11.4 13 Interim Storage for 100 Years 95.1 13.8 13 111.8 16.3 13 77.9 11.7 13 <

Scenario II Interim storage for 30 Years 111.7 (2.8) (3) 103.3 24.8 19 80.5 9.1 10 Interim Storage for 50 Years 87.7 21.2 19 101.1 26.8 Interim Storage for 100 Years 21 10.3 19.3 22 87.3 21.6 20 100.9 27.2 21 69.9 19.7 22 Scenario III 83.6 25.3 23 98.0 30.1 24 67.0 22.6  ?*.

(a) Scusarized from Chapter 8, Appendix A, and Appendix B of Reference 5.

(b) For 30 years safe storage. Values are the sum of the cost of preparations for safe storage plus deferred decontamination.

(c) Costs are in 1986 dollars and include a 251 contingency.

(d) Values exclude the costs of disposal of the last fuel core, exclude cost of demolition of nonradioactive structures and exclude cost of deep geological disposal of activated components.

(e) Includes the cost of placement in interim storage plus the cost of removal at a later date to permanent offsite disposal.

(f) Parentheses indicate a cost increase.

(g) Values include labor costs for both planning and preparation and decommissioning operations. Security force labor costs are not included.

(h) Central services include health physics services, security forces, solid waste processing, and equipment decontamination services.

(i) Multiple reactor station scenarios are described in detail in Section 6.1.2.

I J

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Table 6.3-3 Swamery of estimated dose reductions when decommissioning each reference PWit at a evittple reactor station

• b DECON SAFSTOR ENTONB Occupational Occupational Occupational Occupational Occupational Occupational Dose, Dose Reduction Dose, Dose Red.sction Dose, Dose Reduction Dose factor man-ree man-rem pe rcent man-ren man-res percent man rem man rem percent d

Weste Disposai Immediate Ortsite Disposal 222.8 -- --

113:9 -- --

64.9 -- --

Onsite Intgrim Storage for 30 Years 292.0 (69.2)d (31.1)d 40.0 73.9 64.9 71.0 (6.1) (9. 4 )

Onsite Iteria Storage for 50 Years - 150.2 72.6 32.6 36.5 77.4 68.0 53.3 11.6 17.9 Onsite Inte[in Storage for 100 Years 147.1 75.7 34.0 36.0 77.9 68.4 52.1 12.8 19.7 Imediate Onsite Disposal 132.7 90.1 40.4 27.3 86.6 76.0 45.2 19.7 30.4 Decommissior.ing Staff Labor No. of Reactors of Same Type:

1 1117 -- --

307 -- --

914 -- --

2 1089 28 2.5 299 8 2. 6 891 23 2. 5 4 1050 67 6.0 289 18 5.9 859 55 6.0 Solid Waste Processing Without Solid Waste Processing 4.4 -- --

1. 9 -- --

4.~ -- --

With Solid Waste Processing 0.6 3. 8 86.4 0.4 1.5 80.0 0.6 3.8 86.4 1

4-st

Table 6.3-3 (Continued)

DECON SAF5 TON ENT0Pe Occupational Occupational Occupational Occupational Occupational Dose, Occupational Dose Reduction Dose, Dose Reduction Dose, Dose Reduction Dose factor man-rem man-rem percent man ree man-rem percent man-rem -man rem percent T4tals for Decommissioning Scenarios Single-Reactor Station 1477 -- --

558 --

Scenario I 979 -- --

Interim Storage for 30 Years 1518 (41)

Interim Storage for (3) 476 82 15 %2 17 2 50 Years 1376 101 7 472 86 15 944 Interim Storage for 35 4 100 Years 1373 104 7 472 86 15 943 Scenario II 36 4 Interim Storage for 30 Years 1475 7 <1 465 Interim Storage for 93 17 930 49 S 50 Years 1334 143 10 461 97 17 909 Interim Storage for 70 7 100 Years 1330 147 10 460 98 18 Scenario III 1316 161 11 452 907 72 7 106 19 900 79 8 (a) Summarized from Chapter 9 and Appendix C of Reference 5.

(b) For 30 years safe storage. Values are the sum of doses for preparations for safe storage plus deferred decontamination.

(c) Includes disposal,the sum of doses including from placement in interim storage, retrieval from interim storage, and placement in permanent of fsite transportation.

(d) Parentheses indicate a dose increase.

Table 6.3-4 Summary of estimated dose ptions when decommissioning each reference SWR at a multiple reactor station D

DECOM SAFSTOR ENTOPS Occupational Occupational Occupational Occupational Occupational Occupational Dose. Dose Reduction Dose, Dose Reduction Dose, Dose Reduction Dost factor man ree man-rem parcent man-rem , man-rem percent man-re= man-sem seri eM Weste Disposal d^

^

Immediate Offsite Disposal 274.3 -- --

128*8 -- --

207.1 -- --

Onsite Intgrim Storage for 30 Years 297.1 (22.8)d (8.3)d 60.8 68.0 52.8 216.1 Onsite Interim Storage for (9.0)d (4.4)d 50 Years *-

195.2 79.1 28.8 42.3 86.5 67.2 156.9 50.2 24.2 OositeIntegiaStoragefor 100 Years 190.5 83.8 30.6 41.0 87.8 68.2 153.6 Immediate Onsite Disposal 53.5 25.8 173.7 100.6 36.7 36.3 92.5 71.8 139.7 67.4 32.5 Decommiissioning Staff Labor No. of Reactors of Same Type:

1 1767 -- --

331 -- --

1606 -- --

2 1723 44 2.5 323 8 2.4 4

1566 40 2. 5 1661 106 6.0 311 20 6.0 1510 96 6.0 Solid Waste Processing Without Solid Weste Processing 6.3 -- --

3.3 -- --

6.3 -- --

With Solid Waste Processing 1. 3 5. 0 79.4 0.8 2.5 75.8 1.3 5. 0 M.4

[ . H( ,

s Table 5.3-4 (Continued)

DECON SAFSTOR ENTOMB Occupational Occupational Occupational Occupational Occupational Occupational Dose, Dose Reduction Dose. Dose Reduction Dose, Dose Reduction Dose factor man-rew man-rem percent men-ren man rem percent man-rem man-rem percent Tatals for Decommissioning Scenarios .

51ngle-Reactor Station 2122 -- --

549 -- --

1894 -- --

Scenario I .

Interim Storage for 30 Years 2101 21 <1 473 76 14 1863 31 2 Interim Storage for 50 Years ,. 1999 123 6 454 95 17 1804 90 5 Interim Storage for 100 Years 1994 128 6 453 96 17 1800 94 5 Scenario II Interia starno, for 30 Years 2034 88 4 458 91 -

17 1807- 97 's Interim Storage for

• 50 Years 1932 190 9 440 109 20 1743 151 8 Interim Storage for 100 Years 1927 195 9 439 110 20 1739 155 8 Scenario III 1910 212 10 434 115 21 1726 168 9 (c) Summarized from Chapter 9 and Appendix C of Reference 5.

(b) For 30 years safe storage. Values are the sum of doses for preparations for safe storage plus deferred decontamination.

(c) Includes the sum of doses from placement in interim storage, retrieval from interim storage, and placement in permanent offsite disposal, including transportation.

(d) Parentheses indicate a dose increase.

l *

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i-i STUDY CONTRIBUTORS

The overall responsibility for the preparation of this report was assigned to the' Materials Branch of the Office of Nuclear Regulatory Research, Nuclear )

I Regulatory Commission (NRC). The report was prepared by NRC with input from j Battelle Pacific Northwest Laboratories (PNL). Major contributors to this report were the following:

i Carl Feldman, GEIS Task Leader, NRC Frank Cardile, NRC Keith Steyer, NRC Robert Wood, NRC Peter Erickson, NRC Richard % th. PNL )

. George Konzek, PNL l

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Table 6.3-5 Summary of estimated pubil5 d 5' "d"cta5 "hea decommissioning each reference PWR at a multiple-reactor station DECON b SAFSTOR ENTOMB Puolic Pubtle Public Pub 1(c Public Public Dose, Dose Reduction Dose, Dose Reduction Dose, Dose factor man-rem man rem Dose Reduction _

percent man-ren . mar.-ree percent man-rem man-res per u..t Waste Transportation Transport to Immediate Offsite Disposal 20.6 -- --

19.9 -- --

4.5 -- --

Transport to Offsite DJsposal After Interim Storage for:

30 Years 17.9 2.7 13.1 2.4 17. % 87.9 3.4 1.1 24.4 50 Years 3.0 17.6 85.4 1.8 18.1 91.0 1.5 3. 0 (6.1 100 Years 3.0 17. f. 85.4 1. 8 18.1 91.0 1. 4 31 tJt 9 Transport to Immediate Onsite Disposal 0.0 20.6 100.0 0.0 19.9 100.0 0. 0 4.5 100.0 T*tal Puby Dose for Decommissioning Scenarios

$1ngle-Reactor Station 20.6 -- --

19.9 -- -- -- -- --

Scenario I Interim Storage for 30 Years 17.9 2. 7 13.1 2.4 17.5 87.9 3.4 1.1 24.4 Interim Storage for 50 Years 3.0 17.6 85.4 1. 8 18.1 91.0 1. 5 3.0 66.7 Interim Storage for 100 Years 3.0 17.6 85.4 1.8 18.1 91.0 1.4 3.1 68.9

Table 6.3-5 (Continued) b DECON SAFSTOR . ENTOMB Public

• Pubitc Pubile Public lic Public Dose, Dose Reduction Dose, Dose Reduction Dose, Dose Reduction Dose factor man rem man-rem percent man-rem man rem percent man-rem man-ree percent Scenario II Interim Storage for 30 Years 17.9 2.7 13.1 2.4 17.5 87.9 3.4 1.1 24.4 Interim Storage for 50 Years 3.0 17.6 85.4 1.8 18.1 91.C 1. 5 3.0 66.7 Interim Storage for 100 Years 3.0 17.6 85.4 1.8 18.1 91.0 1.4 3.1 68.9 Scenario III ,,

0.0 20.6 100.0 0.0 19.9 100.0 0.0 4.5 100.0 (a) Summarized from Chapter 9 of Reference 5.

(b) For 30 years of safe storage. Values are the sum of doses for preparations for safe storage plus deferred decontamination.

(c) Doses from routine decommissioning operatloas are estimated te be less than 0.001 m2n-rem for all decommissioning alternatives.

Hence, the dose to the public is estimated to result almost entirely from the transportation of nuclear waste to of fsite disposal.

i l

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Table 6.3-6 Summary of estimated pub i i i f Oldt at a multiple-reactorstation}Icdosereductionswhendecommssonngeachreerence b ENT0pe DECON SAFSTOR Public Public Public Public Public Pubile Dose, Oose Reduction Dose, Dose Reduction Dose, D se Reduction Dese factor man ree man-ree percent man-ree san-ree percent een-ree e n-ree p= . .:.' t.

Waste Transportation .

Transport to Ismediate Offsite Disposal 22.4 -- --

20.8 -- --

9.6 -- --

Iransport to Offsite Disposal Aftee Interia Storage for:

30 Years 17.0 5.4 24.1 4.1 16.7 80.3 9.2 0.4 4.2 50 Years 3.5 IP. 9 84.4 1.2 19.6 94.2 2.9 6. 7 69.8 100 Years 3.4 19.0 84.8 1.0 19.8 95.2 2.8 6.8 70.8 Transpt,rt to Immediate Onsite Disposal 0.0 22.4 100.0 0.0 20.8 100.0 0.0 9.6 100.0 TotalPubig0oseforDecommissioning Scenarios Single-Reactor Station 22.4 -- --

20.8 -- --

9.6 -- --

Scenario 1 Interim Storace for 30 Years 17.0 5.4 24 4.1 16.7 80 9.2 0.4 4 Interin Storage for 50 Years 3. 5 18.9 84 1.2 19.6 94 2.9 6.7 70 Interim storage for 100 Yean 3.4 19.0 85 1.0 19.8 95 2-8 6. 8 71 I

i

(-11 1

'e=u-- __ _ _ j

Table 6.3-6 (Continied) b DECON SAFSTOR g ,yp ,g Pubile Public Public Pubitc Public .

Public <

Dose. -

Dose Reduction Dose, Dose Reduction Dose. Dose Reduction Dose factor aan ree man res percent man-rem man res percent man-ree rean-ree percent Scenario II

• Interim Storage for 30 Years 17.0 5.4 24 4.1 16.7 80 9.2 0.4 4 Interie Storage for 50 Years 3.5 18.9 84 1.2 19.6 94 2.9 6.7 70 Interim Storage for 100 Years 3=4 19.0 85 1. 0 19.8 95 2.8 6.8 71

Scenario III ,

0.0 22.4 160.0 0.0 20.8 100.0 0.0 9.6 100 (a) Sammarized from Chapter 9 of Reference 5.

(b) For 30 year of safe storage. valves are the sum of do<es for preperations for safe storage plus deferred decontamination.

(c) Doses from routs.w decommissioning operations are estimated to be less than 0.05 man ree for all decommissioning alternatives. i Hence, the dose io the public is estimated to result almost entirely from the transportation of nuclear waste to of fsite disposal.

4 i

4 i

1 4

)

e i

(,-i )

- retrieving from interim storage transporting to a permanent disposal facility placing in a permanent disposal f acility Because of the necessity of handling each waste package three times, interim onsite storage could result in increased costs and increased occupational expo-sures unless the waste is stored for a long enough peric: to result in signifi-cant radioactive cecay prior to shipment to offsite disp: sal. The cost and safety of interim onsite storage are evaluated in the PN. study for onsite storage periods of 30, 50, and 100 years.

Sites where large numbers of nuclear power reactors are located conceivably will be large enough to include a permanent onsite low-level nuclear waste disposal facility. Permanent onsite disposal facilities will be operated only at those multiple-reactor stations where the site is not subject to flooding and where the disposal facility can be designed and operated in accordance with the cri-teria of 10 CFR 61.10 Any decommissioning wastes that d: not meet the criteria on waste classification and waste form given in 10 CFR 61 will be sent offsite to a storage or disposal facility for non-low-level wastes.

It is expected that the efficiency of decommissioning the reactors at a multiple-reactor station will improve after the first reactor is cecommissioned due to the learning process. Improved efficiency will result ir reduced manpower requirements for decommissioning subsequent reactors of the same type and in reduced labor costs and occupational radiation doses. C:st and dose reductions result from the following factors:

minimization of the planning effort for decommissioning the second or later reactors of the same type

- standardization and improvement of decommissioning techniques stabilization of N work force, resulting in less time spent in learning or rehearsing (t '.w issioning procedures e

10/07/87 6-20 NU0586 CH 6

improvement of the productivity of decommissioning workers as a resL't of the learning experience on the first reactor The PNL study used the following assumptions as bases for estimating redu:tions in costs and occupational exposures for decommissioning reactors of the same type at a multiple-reactor station:

1. The cost reduction factor for planning and preparation for the secon: and each succeeding reactor of a particular type (PWR or BWR) is 0.50.

2. The cost and occupational dose r3 duction factor for decommissioning operations for the second reactor of a particular type is 0.95.

3. The cost and occupatienal dose raductiun factor for decommissioning opera-tions for the third and each succeeding reactor of a particular type is 0.90.

A number of centralized services may be available at a multiple-reactor station, including:

health physics services security forces

- solid waste processing equipment decontamination services

- maintenance shops and services -

- laundry services transportation services central stores.

The availability of the first four of these services is estimated to result in significant cost savings for decommissioning. Solid waste processing is also j estimated to result in a reduced occupational radiation dose.

Centralized health physics services at a multiple-reactor station could gnatly reduce the costs of health physics activities at each reactor, during bott the 10/07/87 6-21 NU0586 CH 6

reactor operating life and the decommissioning peric:: following operation. The two major f actors postulated to contribute to this' cost reduction are:

. the reduced health physics staff overhead at ea:n reactor, resulting from the sharing of certain staff members between several reactors at the site

. the reduced peak-load staf fing requir?ments per reactor, because the large pool of health physics techniques at the site can be shared between reactors as needed Two factors that account for a reduction in security force costs during decommissioning at a multiple-reactor station are:

. the overhead structure for each reactor can be reduced by sharing certain staf f members between reactors

. the off-shif t coverage at a reactor being decommissioned can be reduced or eliminated after the spent fuel has been shipped (no special nuclear material at the reactor) if provision is made for routine spot-checks by roving security patrolmen, reducing the overall persnnnel requirement.

At a multiple-reactor station, a central waste incinerator to serve the whole station can reduce the volume of combustible radioactive waste by about a factor of 25. Therefore, a central waste incinerator can p ovide significant savings in waste disposal costs and in occupational exposure to transportation workers '

for both the operating and decommissioning phases of reactor life.

Equipment decontamination services can be more fully utilized at a multiple- l reactor station than at a single-reactor station, thereby increasing the economy of these services and the economic incentive to provide improved services and facilities at a multiple-reactor station. Several types of equipment decontami-nation services are considered to be available at a multiple-reactor station, including:

. decontamination of special tools and equipment used for decommissioning, allowing maintenance and reuse of these items 6-22 NUO586 CH 6 10/07/87 1

I

1 I

l

. mobile decontamination systems for in situ chemical decontamina ic of piping and components ]

l

- central electropolishing and chemical decontamination facilities for improved decontaminacion of pipe sections and components In estimating the net reductic- in decommicsioning costs resulting f :- the availability of these services, account is taken of the cost of prov' ding the services as well as the cost savings from reuse of equipment. Savin;s resulting from electropolishing and salvage of stainless steel are two-fold. The material does not require disposal as radioactive waste and the metal can be sohi as scrap, However, these cost savings are partially offset by the cost of con-struction and operation of the central electropolishing facility. At a multiple-reactor station this cost is assumed to be shared by all of the reactors using the facility.

6.3.1 DECON DECON in the prompt removal and disposal of all materials containing or contami-nated with radioactivity in excess of levels permitted for release / the faciif ty for unrestricted use. Under present regulatory requirements, DECON is the only decommissioning alternative that 7.llows termination of the facility license in a short time period. Demolition and removal of decontaminated or un:entaminated structures are not part of DECON but may be performed at the option of the owner and local government agencies. .

The PNL study shows that significant reductions in the cost of DECON say be achievable at a multiple-reactor statun. With the exception of 30 year interim onsite storage of the nuclear waste, waste disposal costs are substantially reduced by using either interia onsite storage or permanent onsite disposal of the nuclear waste, compared with immediate offsite disposal. Interia onsite storage for 30 years results in a higher waste disposal cost for both the refer-ence reactors because the 30 year storage period is too short for the radio-activity in the contaminated caterial to decay to the level at which signifi-cant quantities of material can be released. Savings in staf f labor costs can be achieved if more than one reactor of the same type is decommissio ed due to 10/07/87 6-23 N.0586 CH 6

improvements in the efficiency of decommissioring the seco d and subsequent reactors of the same type. Significant savings in decommissioning costs are achievable by providing centralized health physics services, centralized security forces, solid waste processing, and decontamination services. While the mag-nitudes of the cost reductions for DECON, shown in Tables 6.3-1 and 6.3-2, are different for the reference PWR and the reference BWR, the percentage reductions are comparable in most instances.

The total rosts of DECON for the reference reactors at multiple-reactor stations are also shown in Tables 6.3-1 and 6.3-2. The multiple reactor station scenarios are those described in Section 6.1.2. Changes in decommissioning costs are the sums of cost reductions (or cost increases) for the individual cost factors shown in the tables. With the exception of the scenarios that include interim onsite storage of nuclear waste for 30 years, all of the scenarios result in an estimated reduction in the total cost of decomissioning a reactor at a multiple-reactor station. The greatest cost reduction per reactor occurs for Scenario III, which include: immediate onsite disposal of nuclear waste and the decommissioning of four reactors of the same type. For the reference PWR, changes in the total cost (in 1986 dollars) of DECON at a multiple-reactor sta-tion range from an increase of about $17.2 million for Scenario I with interim onsite storage for 30 years to a reduction of about $16.5 million for Sce-nario III. For the reference BWR, changes in the total cost of DECON at a multiple-reactor station range from an increase of about $10.6 million for Scenario I with interim onsite storage for 30 years to a reduction of about

$25.3 million for Scenario III. -

The same factors examined in the cost analysis are considered in estimating changes in occupational radiation dose from OECON for a reactor at a multiple-reactor station. With the exception of 30 year interim onsite storage of the nuclear waste from DECON, all of the factors considered result in a reduction in occupational dose. The greatest dose reduction results from immediate onsite disposal of the nuclear waste because of the large reduction in dose to trans-portation workers. The largest percentage reduction in occupational dose results froen solid waste processing. However, the absolute value of this dose reduction is small because the total dose from the packaging of contaminated combustible wastes for shipment is small. For each of the dose factors, the percentage 10/07/87 6-24 NUO586 CH 6

reductions in occupational exposure are about the same for both the PWR and the 86.

The changes in total occupational dose shown in Tables 6.3-3 and 6.3-4 are the sums of the dose reductions or dose increases) for the individul dose factors shown in the tables. With the exception of multiple-reactor station scenarios that involve interim onsite storage of nuclear waste for 30 yea's, the total occupational dose from DECON at a reactor at a multiple-reactor station is estimated to be smaller than that from DECON at a single-reactor station. For the reference PWR, changes in the total occupational dose from CfCON for a reactor at a multiple-reactor station range from an increase of about 41 man-rem for Scenario I with interia onsite storage for 30 years to a re&ction of about 161 man rem for Scenario III. For the reference BWR, changes ir, the total occupational dose from DECON at a multiple-reactor station range from a decrease of about 21 man-rem for Scenario I with interim onsite storage for 30 years to a reduction of about 212 san rem for Scenario III.

As shown in the reference PNL studies8 '7.s on the decommissioning of nuclear reactors at single-reactor stations, the public dose from normal decommissioning activities is small and comes principally from the transportation of nuclear wastes to a licensed offsite disposal facility. Interim onsite storage of the nuclear waste from decommissioning can significantly reduce this already small public radiation dose, especially if the onsite storage period is 50 to 100 years.

Permanent onsite disposal of the nuclear waste from decommissior.ing reduces the dose to the public from waste transportation activities to zero.

6.3.2 SAFSTOR SAFSTOR is defined as those activities required to place a reactor in a condi-tion that poses an acceptable risk to the public (preparations for safe storage) and safely stores the property for as long as desired to allow decay of some of the radioactivity, followed by decontamination of the facility to levels which permit release of the facility for unrestricted use (deferred dMontamination).

As shown previously in Chapter 4 and Chapter 5, SAFSTOR results in greatly re-duced occupational radiation doses because decommissioning actisities that must l

be performed imediately af ter reactor shutdown when radiation esposure levels 6-25 NUO586 CH 6 10/07/87

are high are kept to a minimum, a id the rajor decommissioning activities (de-fe-red decontamination activities) take p' ace af ter SOCo has decayed to levels that result in significantly reduced radiation dose rates. SAFSTOR may be used to advantage at a multiple-reactor station where there is less incentive to decontaminate a reactor to unrestricted use levels immediately following shutdown.

One of the principal disadvantages of SAF5 TOR, namely that personnel familiar with the construction and operation of the plant may not be available at the end of the safe storage period to assist in deferred decontamination, may be less of a problem at a multiple-reactor station than at a single-reactor station.

Personnel would normally be available onsite at a multiple-reactor station who have similar construction and operating experience, even though they might not be intimately familiar with the plant currently being decommissioned.

The information in Tables 6.3-1 through 6.3-6 on cost and dose reductions for the SAFSTOR alternative assumes a safe storage period (the period following reactor shutdown until deferred decontamination takes place) of 30 years.

Information on cost and dose reductions for SAFSTOR at multiple-reactor stations with 50- and 100 year safe storage periods is presented in the PNL study 5 on which this chapter is based. In general, the cost and radiation dose reductions for interim onsite storage or onsite disposal of nuclear waste are not as great following safe storage periods of 50 or 100 years as they are following a safe storage period of 30 years. This is because the radioactive decay associated with the 00- and 100 year safe storage periods results in waste management requirements which are already significantly reduced from what would be required for of fsite disposal of the waste immediately following reactor shutdown.

1 The cost and occupational dose values for 30 year SAFSTOR presented in Tables 6.3-1 through 6.3-4 are the sum of values for preparations for safe storage plus deferred decontamination. In general, the estimated percentage decreases in decommissioning costs for multiple-reactor station decommissioning are approximately the same for 30 year SAFSTOR as they are for DECON. The esti-sated percentage decreases in occupational dose for multiple-reactor station decommissioning are approximately twice as large for 30 year SAFSTOR as they are for DECON. An exception is the case of onsite interim storage of nuclear 10/07/87 6-26 NUO586 CH 6

waste for 30 years v ich is estimated to result in cost an: cese increases for DECON but in cost ant dose decreases for 30 year SAFSTOR. The decreases for SAFSTOR result from the fact that a major portion of the et:cmissioning waste from this alternative is generated during deferred decontanination, and the 30 year safe storage period followed by 30 years of onsite storage results in ,

1 significant radioactive decay and in reduced disposal requ'reeents. I i

As in the case for DECON, radiation dose to the public fron S AFSTOR results l

almost entirely from the transportation of nuclear waste t: cn offsite licensed '

disposal facility. Interim onsite storage of the nuclear waste from SAFSTOR results in a significantly reduced public dose from waste transportation activities, and permanent onsite disposal of the waste redLces this dose to zero.

6.3.3 ENTOMB ENTOMB is the encasement and maintenance of the nonreleasatle radioactive materials in a monolithic structure to ensure complete isolation of the radio-nuclides from the environment until the radioactivity has cecayed to levels which permit release of the facility for unrestricted use.

Two approaches to ENTOMB are possible: 1) the reactor vessel internals, which have extremely long-livea radioactivity, are removed and s'ipped to a nuclear waste repository, and 2) the reactor vessel internals are 'ef t in place. In each case, as much of the contaminated equipment from outs'de the entombment structure as can be stored in the entombment structure is noved there. In the first case, because of the relatively short half-lives of the entombed radio-activity, it may be possible, without dismantling the structure, to terminate the amended nuclear license and release the entombment stru-ture for unrestricted use af ter a continuing care period of about 100 years. (H: wever, present regulatory guidance does not allow such action without a comprehen; 'urvey to

( establish that radioactive contamination is within acceptable relea.e .imits.)

l In the second case, existing regulations require the amended nuclear license to l

remain in force for an indefinite period of continuing care, unless the reactor l

vessel internals are removed at a later date.

10/07/87 6-27 NUO586 CH 6

o' Men it becomes desirable to terminate the amended nuclear license for ENTCHB,

ismantling of the entombment structure may be required for the second approach.

~his represents a task that is much Mre dif ficult than dismantling the unen-

• oebed

. facility, since the entombment structure is built to endure for a long period of time. Therefore, the second approach to ENTOMB, and perhaps the first approach also, must be viewed as an almost irreversible commitment to long-term s.aintenance of the nuclear license. (-:=ever, dismantlement of the entomb ent structure is not impossible, only very difficult.) Based on the above consider-ations, the second approach to ENTOMB, and perhaps the first approach also, must be viewed as relatively unattractive decommissioning alternatives for a multiple-reactor station.

The cost and dose information presented in Tables 6.3-1 through 6.3-6 are based on the first approach to ENTOMB (removal of the reactor vessel internals prior to entombment). On a percentage basis, cost and dose reductions from ENT02 for a reactor at a multiple-reactor station are estimated to be comparable to cost and dose reductions from DECON. The radiation dose to the public is significantly

-educed for interim onsite storage of radioactive wastes followed by later dis-posal at a licensed offsite facility, and is reduced to near zero for confinement cf wastes to the site (multiple-reactor station Scenario III).

E.4 Environmental Consequences A.s shown in Sections 4.3 and 5.3, the greatest radiological impact to the public f rom decommissioning of a nuclear pcver reactor is the possible radiation dose -

f rom truck shipment of the nuclear waste to a shallow-land disposal site. At a multiple-reactor site, interim storage of the waste to permit radioactive decay or permanent onsite disposal would reduce or eliminate the already small dose to the public from transportation of the decommissioning vastes. Releases of radioactivity to water during decommissioning will be negligible, as in the case of facilities on single-station sites. Impacts to the public'from releases of radioactivity to the air will be less than in the case of single-reactor sites. This is because the public will be, on the average, farther away from each reactor because of the large area occupied by a multiple-reactor station.

10/07/87 6-28 NUO586 CH 6

i l

Radiological impacts to transportation workers will be less then they would be if the wastes were immediately transported to an of fsite disposal location.

Hcwever, for interim onsite storage of the wastes, the total radiation dose to workers who must handle the wastes during emplacement and retrieval operations would increase. The possibility is excellent that the radiation dose to decom-sissioning workers can be reduced because of the experience gained from the repetition of the decommissioning process.

Waste disposal at a site dedicated for nuclear power generation would require approximnely 1 kan 2 of land to be used for a shallow-land burial ground.

Approximately 10% of the burial ground area is estimated to be required for the storage or disposal of decomtr.issioning wastes. Appropriate control of inventory and site will allow for unrestricted release in several hundred years following shallow-land burial.10'11 Radioactive wastes that would require longer time periods to achieve unrestricted release are assumed to be placed in appropriate intermediate-depth burial grounds as per 10 CFR Part 61 either onsite or in a deep geologic repository offsite.

A major socioeconomic impact will occur at the time construction of the last reactor is completed at a dedicated multiple-reactor station. If decommissioning has proceeded as older reactors are retired from service, decommissioning crews will already be on site and construction crews will be discharged when construc-tion is completed. Decommissioning of the final reactors retired from service will be performed by personnel who have operated these reactors. Following decommissioning of the last reactor, only a minimal crew will be required for -

surveillance of reactors that are being maintained in safe storage and to provide surveillance activities for the radioactive waste buried onsite.

6. 5 Comparisons of Reactor Oecommissioning at Multiple-Reactor Stations and at Single-Reactor Stations Based on the information presented in Section 6.3 and in Tables 6.3-1 through 6.3-6, the following conclusions may be drawn with regard to the cost and safety of decommission;..g a nuclear power reactor at a multiple-reactor station.

6-29 NUO586 CH 6 10/07/87

I r

1. Decommissioning of a iight water reactor at a multiple-reactor station probably will be less costly and result in lower occupational radiation doses than decommissioning of an identical reactor at a single-reactor station. The option of onsite storage or disposal of the nuclear waste at a multiple-reactor station has the potential of reducing the public radiation dose from reactor decommissioning to near zero.

2. Although the magnitudes of the decommissioning costs and occupaticnal radiation doses are less, the relative standing of the costs and doses for the three decommissioning alternatives is not change:1 at the multiple-reactor station compared to a single-reactor station. SAFSTOR results in the lowest occupational radiatica case but generally has the highest costs (in constant dollars). ENTOMB, if the reactor can be released for unre-stricted use after 100 years of surveillance, is estimated to have the lowest cost. DECON is estimated to have the highest radiation dose and an intermediate decommissioning cost.

3. Decommissioning costs and occupational radiation doses for the two types of reactors (PWR and BWR) are affected in about the same way by the factors studied at multiple-reactor stations. In determining if there is a cost or dose advantage for decommissioning nuclear reactors at a multiple-reactor station versus a single-reactor station, the type of reactor (PWR or BWR) has little influence on the result.

4. All the factors investigated in the PNL study5 --interim onsite nuclear waste storage, permanent onsite waste disposal, dedication of the site to nuclear power generation, and provision of centralized services--can con-tribute to reduced decommtssioning costs and occupational doses. The number of reactors at a multiple-reactor station may influence the availability of onsite storage, site dedication, and centralized services.

5. The possibility of onsite interim storage or of permanent onsite disposal of decommissioning wastes at a multiple-reactor station could facilitate reactor decommissioning in the event of the unavailability of facilities for the offsite disposal of low-level radioactive wastes.

10/07/87 6-30 NUO586 CH 6

One of the alternatives for reactor retirement .i conversion to a new nuclear-or fossil-fueled steam supply system. Reuse of the facilities at a nuclear pcwer station that can be refurbished makes good economic sense. Capital cost ,

studies of PWRs1 1 and BWRst2 have shown that the structures and equipment other than the nuclear steam supply system account for about 70% of the initial direct construction cost. At a multiple-reactor station dedicated to nuclear power ge eratior., conversion of a retired reactor to a <ew nuclear-fueled steam supply system may be particularly advantageous.

I Analyses of removing the old reactor vessel intact from a retired PWR or BWR 1 and replacing it with a new vessel indicate that such action is feasible, but difficult. Examples of design features that could be incorporated in a light-water reactor to facilitate the later removal or replacement of the reactor pressure vessel and other large equipment pieces include:

an equipment hatch in the reactor containment building large enough to accomodate the intact reactor pressure vessel an equipment hatch located so that there is sufficient lay-down area, both in the containment building and in any adjoining building, to line up the reactor vessel with the hatch adequate supports in the reactor building to handle the special cranes needed for very heavy loads such as the reactor pressure vessel and steam generators a readily removable roof section in the fuel building of a PWR and in the reactor building of a BWR large enough to accommodate the reactor pressure vessel an inner shield of modular design that can be removed and/or replaced.

10/07/87 6-31 NUO586 CH 6

REFERENCES

1. Evaluation of Nuclear Energy Centers. 1974. WASH-1288, U. S. Atomic Energy Commission, Washington, DC.

2. Nuclear Energy Center Site Survey - 1975. 1976. NUREG-0001, U. S.

Nuclear. Regulatory Commission, Washington, DC.

3. H. Harty. 1978. The Haaford Nuclear Energy Center, A Conce:tual Study.

PNL-2640, Pacific Nortr-est Laboratory, Richiand, Washingtor.

4. C. C. Burwell, M. J. Chanian and A. M. Weinberg. 1979. "A Siting Policy for an Acceptable Nuclear Future." Science, 204:1043, June 8,1979.

5. N. G. Wittenbrock. 1982. Technology, Safety and Costs of Decommissioning Nuclear Reactors at Multiple-Reactor Stations. NUREG/CR-1755, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission.

6. .R. I. Smith, G. J. Konzek, and W. E. Kennedy, Jr. 1978. Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station. NUREG/CR-0130, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission.

7. H. D. Oak et al. 1980. Technology, Safety and Costs of Dec:amissioaing a Reference Boiling Water Reactor Power Station. NUREG/CR-0672, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission.

8. R. I. Smith and L. N. Polentz, 1979. Technology, Safety anc Costs of Decommissioning a Reference Pressurized Water Reactor Power Station.

NUREG/CR-0130 Addendum, Prepared by Pacific Northwest Laboratory for U.S.

Nuclear Regulatory Comission.

9. G. J. Konzek. Estimated Impacts of Post-TMI-2 Requirements and Other Selected Regulatory Changes on Decommissioning of Two Refere ce Light Water Reactors. NUREG/CR- (to be published), Prepared by Pa:ific ,

Nortnwest Laboratory for the U.S. Nuclear Regulatory Commiss'en,1987.

10. "Licensing Requirements for Land Disposal of Radioactive Waste." 1982.

Federal Register, 47(248):57446-57482, December 27, 1982.

11. Capital Cost: Pressurized Water Reactor Plant. 1977. NUREG-0241, United Engineers and Constructors, Inc. for U.S. Nuclear Reg.latory Commission.

12. Capital Cost: Boiling Water Reactor Plant. 1977. NUREG-0242, United Engineers and Constructors, Inc. for U.S. Nuclear Regulatory Commission.

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l l

l 7 RESEARCH AND TEST REACTORS A research reactor is defined in 10 CFR 170.3(h)1 as a nuclear reactor licensed for o;e-ation at a thermal power level of 10 megawatts or less, and which is not a testing facility. A testing facility (i.e. , a test reactor) is defined in 10 CFR 50.2 as a nuclear reactor licensed for operation at: 1) a thermal power level in excess of 10 megawatts, or 2) a thermal power level in excess of 1 megawatt if tne reactor is to contain: a circulating loop through the core in which the applicant proposes to conduct fuel experiments, or a liquid fuel loading, or an experimental facility in the core in excess of 16 square inches in cross-section. There are 84 nonpower research and test (R&T) reactors in the U.S. that are licensed by the NRC. Of these 76 are research reactors, and 8 are test reactors. The level of activity of these facilities ranges from no longer operational, to occasional use, to intermittent use, to steady and scheduled use.

Because of the diversity in types and sizes of R&T reactor facilities and in the operational schedules and lifetimes associated with them, the level of ef-fort required to decommission them varies greatly. Necessary actions can range from simple, relatively inexpensive decommissioning activities and administra-tive procedures to extensive deconta.ination and disposal activities costing millions of dollars. This section presents an assessment of the environmental ef fects that may be expected from the decommissioning of R&T reactors and is based primarily on information from a study 2 of the conceptual decommissioning of a reference research reactor and of a reference test reactor. The study l

focused on one research facility and on one test facility, each representing a

( significant decommissioning task. Because it was not practical to include in I one study examples of the decomissioning of all classes of R&T reactors, by examining selected facilities and some components and operations common to many

f acilities, the study provided data that would be useful in estimating the re-l quirements and costs of decommissioning other facilities not specifically con-

! sidered.

l The reference test reactor is assumed to be located at the generic site described in Section 3.1. The reference site used for the study o' the 10/07/87 7-1 NUO586 CH 7

reference research reactor is the campus of a large uni.arsity and is described in Section 3.2. As part of the study, PNL developed ir. formation on the avail-able technology, safety considerations, and probable costs for decommissioning the reference R&T reactors at the end of their useful operational lives. In addition as part of an addendum 3 to the study,2 PNL analyzed selected cases to consider the sensitivity of decommissioning costs and radiation doses to reactor size.

7.1 Description of R&T Reactors 7.1.1 Reference Research Reactor The reference research reactor is the Oregon State University TRIGA Reactor at Corvallis, Oregon. This reactor is a 1 MWt, above ground, open pool nuclear training and research facility. The reference research reactor is made up of a reactor tank and a core structure and a TRIGA type control system. Major structures comprising the reference research reactor include a reactor building (housing the TRIGA reactor and support area), a cooling tower, an annex (hous-ing a hot laboratory area and hot cell), a heat exchanger building (housing a water purification system, water pumping systems, and air compressor systems), a pump house (housing a liquid waste retention tank), and a radiation center building (housing a waste processing and storage room).

7.1.2 Reference Test Reactor The reference test reactor is the Plum Brook Reactor at Sandusky, Ohio operated by the National Aeronautics and Space Administration. The Plum Brook reactor is a 60 MWt materials test reactor, light water moderated and cooled, used in testing materials for certain applications. Although Plum Brook has been actu-ally shut down since 1973 it is analyzed in the study 2 as if it had recently been shut down. The testing system of the Plum Brook reactor is made up prin-cipally of the test reactor vessel (containing the nuclear core and experimen-tal beam tubes) and the reactor water recirculation system. Majorstructures comprising the reference test reactor include a reactor building (housing the test reactor), a hot laboratory building with seven hot cells, a primary pump house, an office and laboratory building (housing radiochemistry laboratories),

10/07/87 7-2 NUO586 CH 7

a fan house (housing ventilation systems ard waste ion exchangers and filters),

a hot retention area (holding waste tanks), a cold retention area, an emergency retention basin, and a waste handling building.

7. 2 Research and Test Reactor Decommissioning Experience Due to the relatively large number of research and test reactors in the V.S.

and the diversity of their use, a number of research and test reactors have either been decommissioned by the use of DECON or are being decommissioned by placing them into safe storage. A list of experience with decommissioning of research and test reactors is given in Table 7.2-1. These experiences indicate that the basic technologies for decontamination and dismantlement of these types of R&T reactors have been carried out successfully and can be modified as necessary to suit site-specific conditions.

7. 3 Decomissioning Alternatives Once a research or test reactor has reached the need of its useful operating

' life it must be decommissioned. As discussed in Section 2.3, this means safely removing the facility from service and disposing of all radioactive materials in excess of levels which would permit unrestricted use of the property. Several alternatives are considered here as to their potential for satisfying this gen-eral requirement for decommissioning. The decommissioning alternatives consid-ered and discussed here are DECON, SAFSTOR, and ENTOMB. The alternative used depends on such considerations as cost, dose, physical design of the facility, -

types of residual radioactivity present, proposed use of the site, and desir-ability of terminating the license.

Discussion of the decommissioning alternatives follows:

7.3.1 DEC00 j DECON is defined as the imediate removal and disposal of all radioactivity in excess of levels which would permit release of the f acility for unrestricted use.

Nonradioactive equipment and structures need not be torn down or removed as part l of a DECON procedure. To accomplish DECON, all potentially contaminated systems 7-3 NV0586 CH 7 i 10/07/87 l

l

Table 7.2-1 Experience with r! search and test reactor cecommissionings(a)

Thermal End cf Decommissioning Power Operation Method Illinois Inst of Tech. 100 kW 1967 DECON USN Research Lab 1 MW 1970 DECON NC State 100 W 1963 DECON Industrial Reacto Labs 5 MW 1975 DECON US Navy Post Graduate School 0.1 W 1971 DECON North American Aviation 5W 1955 DECON Oklahoma State Univ. 0.1 W 1974 DECON Navy Hospital 5W 1962 DECON University of Akron 0.1 W 1967 DECON Univ. of Calif. 0.1 W 1966 DECON Univ. of Delaware 0.1 W 1977 DECON Gulf United Nuclear 100 W 1971 DECON Oregon State Univ. 0.1 W 1974 DECON Rice Univ. 15 W 1965 DECON Univ, of Wyoming 10 W 1974 DECON Polytechnic Inst, of NY 0.1 W 1973 ECON Walter Reed Medical Ctr. 50 kW 1971 DECON Lockheed 3 MW 1970 DECON Univ. of Nevada 10 W 1974 DECON General Dynamics 500 W 1965 DECON General Atomic Co 1.5 MW 1973 DECON Gulf General Atomic 500 W 1967 DECON Gulf Oil 500 W 1973 DECON NUMEC 1 MW 1966 DECON Battelle Memorial Inst. 2 MW 1974 SAFSTOR Watertown Arsenal 5 MW 197C SAFSTOR Rockwell Inter. Corp 10 W 1974 SAFSTOR Oregon State Univ. 0.1 W 1978 SAFSTOR NC State Univ. 10 kW 1973 SAFSTOR West Virginia Univ. 75 W 1972 SAFSTOR Stanford Univ 10 kW 1974 SAFSTOR NASA Hock-up 100 kW 1973 SAFSTOR Calif. Polytech. Univ. 0.1 W 1978 SAFSTOR Diamond Ordnance Facility 250 kW -

DECON Ames Laboratory 5 MW -

DECON I

Lynchburg Pool Reactor 1 MW -

DECON l Westinghouse Test Reactor 60 MW 1962 SAFSTOR Plum Brook Test Reactor 60 MW 1974 SAFSTOR Saxton Test Reactor 28 MW 1972 SAFSTOR i GE EVESR Test Reactor 17 MW 1967 SAFSTOR I B&W 8AWTR Test Reactor 6 MW 1971 DECON l SEFOR Sodium Test Reactor 20 MW 1972 SAFSTOR l

Information updated through 1987.

(a) Adapted from References 2 and 3.

10/07/87 7-4 NUO586 CH 7 l __

must be disassembled and removed a-d all contaminated material must be re cved from the facility and be transported to a regulated disposal site. The end result is the release of the site and any remaining structures for unrestricted use shortly af ter the end of facility operation. Also DECON assumes the availa-bility of capacity to handle wastes requiring disposal.

DECON is advantageous because it a'lews for termination of the NRC licerse shortly af ter cessation of facilit/ operations and removes a radioactive site.

DECON is advantageous if the site is required for other purposes or if the site is extremely valuable. It is also advantageous in that the facility operating staff is available to assist with decommissioning and that continued surveil-lance is not required. A disadvantage of DECON is the higher occupational dose than for other alternatives for research reactors and than for SAFSTOR fer test reactors, although as discussed below the difference in dose for the reference research reactor is very small and for the reference test reactor it is rot substantial .

The PNL study shows that, for the reference research reactor, DECON would re-quire about 1.7 years to complete, including 1 year for planning and prepara-tion, prior to final reactor shutdewn, and, for the reference test reacter, DECON would require about 4.1 years to complete, including 2 years for planning and preparation. The costs (updated to 1986 dollars) for DECON for the refer-ence R&T reactors are given in Tables 7.3-1 and 7.3-2, respectively.

Three important radiation exposure pathways need to be considered in the evalu ~

ation of the radiation safety of normal reactor decommissioning operatio s:

inhalation, ingestior., and external exposure to radioactive materials. For reasons similar to that discussed for PWRs in Section 4.3.1, during decommis-l sioning the dominant exposure pathway to workers is external exposure while for the public the dominant exposure pathway is inhalation. During the transport of radioactive waste, the dominant exposure pathway is external exposure for l

both transportation workers and the public. A summary of the occupational doses resulting from these pathways for the reference research and test reactors is presented in Tables 7.3-3 and 7.3-4, respectively. The dose to the p.blic from radionuclide releases during l ECON activities and from truck transp:rtation of radioactive waste from DECON at the reference research reactor is esti mated 10/07/87 7-5 NUO585 CH 7

T ible 7. 3-1 I Summary of estimated costs for (gcy.issioning the reference  !

research reactor in $ millions '

Oecommissiong SAFSTOR Element DECON 10 Years 30 Years 100 Years ENTOM8(c)

DECON 1.22 NA(d) NA NA NA Entombment NA NA NA NA 0.74 Safe Storage Preparation NA 0.67 0.67 0.67 NA Continuing Care NA 0.41 1.3 4.3 0. 008/yr Deferred Decontamination NA 1.21 1.08 0.95 NA Total 1.22 2.29 3.05 5.92 0.74 + $8K/yr

(*) Values include a 25% contingency and are in constant 1986 dollars .

(b) Values exclude cost of disposal of last core and cost of demolition of nonradioactive structures.

(c) Adapted from Reference 2.

(d)NA-not applicable.

Table 7.3-2 Summary of estimated costs [ggecommissioning the reference test reactor in $ millions Decommissioning SAFSTOR Element (C) DECON 10 Years 30 Years 100 Years ENTOM8(C)

DECON 24.2 NA(d) NA NA NA Entombment NA NA NA NA 21.3 Safe Storage Preparation NA 10.9 10.9 10.9 NA Continuing Care NA 1. 5 4.6 15.5 0.052/yr Deferred Decontamination M 14.4 14.4 11.2 NA Total 24.2 26.8 29.9 37.6 21.3 + $52K/yr

(*) Values include a 25% contingency and are in constant 1986 dollars.

(b) Values exclude cost of disposal of last core and cost of demolition of nonradioactive structures.

(c) Adapted from Reference 2.

(d)NA-not applicable.

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Table 7.3-3 Summary of ractation safety analyses for decommissging the reference research reactor (values are in man-rem)

SAFSTOR DECON 10 Years 30 Years 100 Years ENTOMB Occupational Exposure Safe Storage Preparation NA 13.1 13.1 13. '. NA Continuing Care NA 0.5 0.8 0.8 neg.

Decontamination 18.3 1.5 0.1 0.1 NA Entombment NA NA NA NA 16.6 Safe Stor. Prep. Truck Shipments NA 0.1 0.1 0.1 NA Decontamination Truck Shipments 0.3 neg neg neg NA Entombment Truck Shipments NA NA NA NA 0.1 Total 18.6 15.2 14.1 14.1 16.7 (a)All entries are from Reference 2. NA means not applicable and neg means negligible.

Table 7.3-4 sioning the Sumary referenceoftest radiation safety analyses reactor (values for decog) are in man-rem SAFSTOR DECON 10 Years 30 Years 100 Years ENTOMB Occupational Exposure Safe Storage Preparation NA 112 112 112 NA .

Continuing Care NA neg neg neg neg Decontamination 322 86 6 1 NA Entombment NA NA NA NA 425 Safe Stor. Prep. Truck Shipments NA 12 12 12 NA Decontamination Truck Shipments 22 2 neg. neg NA Entombment Truck Shipments NA NA NA NA 19 Total 344 212 130 125 444

(*)All entries are from Reference 2. NA means not applicable and neg means negligible.

10/07/87 7-7 NLO586 CH 7

to be negligible (less than 0.1 man-rem). The dose to the public from routine releases during DECON activities at the reference test reactor are estimated to be negligible and the dose to the public from truck transport of wastes from the reference test reactor is estimated to be 2.2 man-rem.

7.3.2 SAFSTOR SAFSTOR is defined as those activities required to place (preparation for safe storage) and maintain (safe storage) a research or test reactor in such condi-tion that the risk to safety is within acceptable bounds, and that the facility can be safely stored and subsequently decontaminated to levels which permit release of the facility for unrestricted use (deferred decontamination).

An advantage of SAFSTOR is that there is reduction in occupational and public dose although as can be seen from Tables 7.3-3 and 7.3-4 the occupational doses from the reference research reactor do not decay by a large amount and the dose reduction from the reference test reactor is marginally significant. In addi-tion as no'ed in Section 7.3.1 the public dose from the reference research re-actor is negligible and from the reference test reactor is very small. Other reasons for use of SAFSTOR include shortage of radioactive waste disposal space of fsite or presence of other nuclear facilities onsite. A disadvantage of SAFSTOR is that the licensee is required to maintain a possession only license and to sect its requirements at all times during safe storage thus contributing to the number of site,s dedicated to radioactive confinement for an extended time period. Other disadvantages are that surveillance is required, the dollar '

costs are higher than for DECON, and the experienced operating staff may not be available at the end of the safe storage period to assist in the deferred decontamination.

The PNL study shows that the costs of SAFSTOR are greater than those of DECON and vary with the number of years of safe storage. Tables 7.3-1 and 7.3-2 pre-sent a summary of estimated costs (updated to 1986 dollars) for decommissioning the reference research and test reactors, respectively, The estir.ated radiation doses due to SAFSTOR at the reference research and )

test reactors are estimated in the PNL study 2 and a summary of the occupational doses for these facilities are contained in Tables 7.3-3 and 7.3-4. The dose  ;

10/07/87 7-8 NUO586 CH 7

to the public during SAFSTOR activities and truck transport of radioactive wastes from SAFSTOR at the reference research reactor are estimated to be negligible (less than 0.1 man-rem). For the reference test reactor, the dose to the public f rom routine releases during SAFSTOR activities are estimated to be negligible and the dose to the public from truck transport of wastes is estimated to be 0.35, 0.14 and 0.11 manrem for storage periods of 10, 30, and 100 years respectively.

7.3.3 ENTOMB ENTOMB of a research or test reactor requires its encasement in concrete to protect the public from radiation exposure until its radic3ctivity has de-cayed to levels permitting release of the facility for unrestricted use. The amount and the half-life of the residual radioactive material in the facility to be entombed determines the time period that the integrity of the structure must be assured. ENTOMB includes the entire process of first entombing and then continuing some surveillance to assure the integrity of the structure until the encased material is confirmed to have decayed enough to allow unretricted release.

ENTOMB also requires a nuclear license to remain in force. The facility and site preparations include comprehensive cleanup and decontamination outside of and confinement of nonreleasable materials within the encasement structure. Contin-uing care activities are minimal.

For much the same reasons as is discussed in Sections 4.3.3 and 5.3.3 ENTOMB with the internals in place would probably net be viable due to the long-lived nuclides contained in the internals. The information presented in Tables 7.3-1 through 7.3-4 are based on entombing the reactor with the reactor internals removed. The postulated entombment structure for the reference research reactor is the entire concrete structure housing the TRIGA reactor, and for the reference test reactor the entombment structure encompasses the below grade portion of the reactor containment vessel. Radioactive materials not entombed would have to be packaged and transported to a burial site. dNTOMB has some advantage because of reduced occupational exposure at the reference research reactor however the amount of reduction is very small (less than 2 manrem). For the reference test reactor ENTOMB results in increased occupational exposure partly due to the exposure received in constructing the entombment structure.

10/07/87 7-9 NUO586 CH 7

As- was noted for SAFSTOR in Section 7.3,? the ef fect of use of ENTOMB on publi; dose is small since public doses are already very small even for DECON. Another advantage of ENTOMB occurs if there is a shortage of disposal capacity although waste volumes for research reactors are small. Disadvantages of ENTOMB include the fact that the integrity of the entombing structure must be assured and sur-veillance and monitoring would be required for an extended time, and that entomo-ing contributes to the number of sites de:fcated to radioactive containment for very long time periods. A difficulty with ENTOMB is that the radioactive mate-rials remaining in the entombed structure would need to be characterized well enough to be sure that they decay to acceptable levels at the end of the surveil-lance period, otherwise deferred decontasination would become necessary which would make ENTOMB more costly and difficult. Also, ENTOMB would seem an unlikely choice for a university research reactor where space is at a premium.

The costs (updated to 1986 dollars) of ENTOMB for the reference rtisearch and test reactors are summarized in Tables 7.3-1 and 7.3-2 respectively. As can be seen, the cost of ENTOMB is higher than the cost of DECON when the costs of surveil-lance for an extended time are added in.

The estimated radiation doses due to ENTCMB are summarized in Tables 7.3-3 and 7.3-4. For the reference research reactor the dose to the public during ENTOMB activities and truck transport are estimated to be negligible (less than 0.1 man-rem). For the reference test reactor, the dose to the public during ENTOMB activities is estimated to be negligible and the dose during truck transport of wastes is estimated to be 1.3 manrem (Table N. 5-2 Vol. 2 of reference 2).

7.3.4 Sensitivity Analysis In an addendue3 to the original PNL study 2 PNL analyzed five selected cases to consider the sensitivity of decommissioning costs and radiation doses to plant size. The five cases are listed in Table 7.3-5. The analysis took the form of obtaining data on the radiation doses and costs from these cases and putting the costs on a common year basis of 1981 dollars. The costs (updated to 1986 dollars) and doses are also summarized in Table 7.3-5. The PNL study noted that quantitative data sufficient to correlate radiation dose to reactor size or 10/07/87 7-10 NUO586 CH 7 1

type in a meaningful way do not exist. Costs of decoernissioning 4 appear to have some relationship to power rating although no scaling factor or correlation was developed. The benefit of this analysis is that it provides information on the type of ranges of dose and costs of decommiss'.oning that may be encountered for various types of research and test reactors. An important item noted in the :

l addendum is that the sensitivity results presented are subject to a large num-ber of variables, each with wide ranges of values, that ca9 possibly impact on costs and radiation exposure estimates for other nuclear R&T facilities. Due to the many variables involved, including facility size, number and type of ancillary facilities, facility design and construction, type of labor utilized, use of subcontractors, and operating practices during the facility lifetime, l the relationship noted is not necessarily a fixed relationship. Hence interpo-lation of the data for different type facilities can be risleading and in par-ticular extrapolation of the data to larger power facilities is not practical.

Table 7.3-5 Comparison of data from selected cases of research reactor decommissioning Occupational Adjusted 1 Dose Cost, Millions Reactor Thermal Power (man-rem) (1986 dollars)

Diamond Ordnance 250 kW <2 0.497(a)

Facility l

Ames Laboratory 5 MW 69 5.931(a)

Lynchburg Pool 200 kW (natural <0.1 0.102(a)

Reactor convection) 1 MW (forced convection)

NC State University 10 kW < 2(b) 0.230(b)

Oregon State University 0.1 W Ney(*) 0.014(a)

(*) Adapted from Reference 3.

(b) Based ois Reference 4.

(c)Neg means negligible.

10/07/87 7-11 NUO586 CH 7

1-l 1

l I

7.4 Environmental Consequences An environmental consequence of decomissioning, other than radiation do>e which is discussed above in Section 7.3, is the commitment of land area to the disposal of radioactive waste. The volume of low-level radioactive waste to be disposed of during DECON is estimated to be 160 m3 and 4930 m3 for the refer-ence research reactor and ref erence test reactor respectively. Waste vol.mes will decrease during SAFSTOR due to reduced quantities of radionuclides a.d corresponding waste quantities as a result of radioactive decay and for the reference research reactor are estimated 2 to be 100, 29, and 29 m3 for 30, 50, and 100 years of storage, respectively, and for the reference test reactor are estimate to be 4930, 2960, and 2940 m 3 for 30, 50, and 100 years of storage, respectively. For ENTOMB, the waste volumes are estimated to be 21 m3 and 2930 m3 for the reference research reactor and test reactor, respectively. The volumes indicated are those required to accommodate radioactive waste and rubble removed from the facility and be transported to a licensed site for disposal.

The volume for ENTOMB does not include the volume cf the entombing structure or of the wastes entombed within it. The waste volumes requiring burial would represent a use of less than 0.1 acre of land for disposal for the reference research reactor and about one-half acre for the reference test reactor. This amount is not large in comparison with the size of the reference research reactor site (approximately 40 acres) and the reference test reactor site (approximately 1200 acres) which could now be released for unrestricted use.

PNL considered accidental releases of radioactivity both during decommissioning during transport of wastes and the results are presented in Tables 7.4-1 and 7.4-2. Radiation doses to the maximum-exposed individual from accidental air-borne radioactivity releases during decommissioning operations were calculated to be quite low. Radiation dnses to the maximum-exposed individual f rom acci-dental radioactivity releases resulting from transportation accidents were cal-culated to be low for the most severe accident.

The socioeconomic impacts are mainly from the shutdown (not decommissioning) of the research or test reactor which would result in the loss of certain jets and income to the community. The overall impact from the reference research reactor is likely to be small since the facility is not a revenue producing f acility.

10/07/87 7-12 N00586 CH 7

Table 7.4-1 Summary of radiation doses to the maximum-exposed individual from accidental g) radionuclide releases during decommissioning at the reference research reactor Radiation Dose to Lung (ree) from Toul First-Year 50-Year Committed Atmosphgc Frequencyg) Dose Equivalent Accident Release (Ci/Hr) Occurrence Dose Oxyacetylene Explosion 5.2 x 10 2 Medium 1.2 x 10 3 1.6 x 10 3 HEPA Filte 2.6 x 10 4 Low 7.3 x 10 7 7.8 x 10 7 Failure d) 1.0 x 10 5 2.4 x 10 7 3.1 x 10 7 SevereTranggogation Accident 5.2 x 10 5 8.3 x 10 4 Low 4.1 x 10

• LPG Explosion Id) 1.4 x 10 5 Low 3.9 x 10 8 4.2 x 10 8 Vacuum Filter-Bag Rupture 1.8 x 10 8 Medium 4.3 x !') s 5.6 x 10 8 MinorTransgtation Accident 1.3 x 10 8 Low 1.0 x 10 5 2.1 x 10 5 Accidental Cutting of (d)

Activated Al in Air 2.9 x 10 7 High 6.9 x 10 9 9.1 x 10 9 ContaminatedSwe{ gig -

Compound Fire -

1.9 x 10 8 Medium 5.3 x 10 12 5.7 x 10 2 Combustglg) Waste Fire

• 9.0 x 10 30 High 1.5 x 10 88 3.2 x 10 88 .

(a) Adapted from Reference 2. [

I) For comparison, all accidental releases are assumed to occur in a 1-hr period. '

IC) The frequency of occurrence considers not only the probability of the accident, but also the probability of {

an atmospheric release of the calculated magnitt:de. The frequency of occurrence is listed as "high" if the ;

occurrence of a release of similar magnitude is > 10 2 per year, as "medium" if between 10 2 and 10 5, and ,

as "low" if <10 5 l (d) The accident shown applies to both DECON and SAFSTOR. '.

I') The accident shown applies to both DECON and ENTOMB.

1,

)

Table 7.4-2 Summary of radiation doses to the maximum-exposed individual ft om accidentg) radionuclide releases durlag decommissioning at the reference test reactor g Radiation Dose to Lung (rem) from:

Relea Frequency g Fi rst-Year 50-Year Committed Accident (Ci/hr)tS) Occurrence Dose Dose Equivalent Oxyacetylene 5.6 x 10 2 Medium 1.6 x 10 4 1.7 x 10

• Explosion LPG Explosion Id} 6.5 x 10 3 Low 1.3 x 10 5 2.0 x 10 5 Severe Transportation Accident 1.0 x 10 3 Low 7.8 x 10 3 1.6 x 10 2 HEPAFiltg) 5.7 x 10 4 1.5 x 10 " 1.2 x 10

• Failure 3.8 x 10

• Low 9.1 x 10 " 1.2 x 10 7 Accidental Cutting of -

Activa [gjStainless Steel 8.8 x 10 5 High 2.5 x 10 4 2.6 x 10 7 -

I VacuumFigegBag Rupture 2.9 x 10 5 Medium 8.1 x 10 s 8.7 x 10 s ,

Minor Transportation Accident 2.5 x 10 5 Low 3.8 x 10 5 8.0 x 10 5  ?

ContaminatedSygpg Compound Fire 3.6 x 10 8 Medium 1.0 x 10 80 1.1 x 10 80 .

Combusybg Waste '

Fire

• 1.8 x 10 8 .91gh 5.0 x 10 88 5.4 x 10 12 1 I*)

Adapted from Reference 2. .

(D)

For comparison, all accidental releases are assumed to occur in a 1-hr period.

IC)

The freqsency of occurrence considers not only the probability of the accident, but al".o the probability of an atmospheric release of the calculated magnitude. The frequency of occurrence is listed as "high" if the 1 occur-rence of a release of similar magnitude is >10 2 per year, as "medium" if between 10 2 and 10 5, and -

as "low" if <10 5 '

(d) The accident shown applies to both DECON and SAFSTOR.

I")

It'e accident shown applies to both DECON and ENiOMB.

5 7.5 Comparisto of Decommissioning Alternatives f rom examination of Tables 7.3-3 and 7.3-4, occupational and public doses I

are much less significant and much easier to manage than for the power reactors

. discussed earlier in the final GEIS. Hence, DECON is probably the most reason-able option. In addition, costs of DECON are less than those for SAFSTOR.

30' year cr 50 year SAFSTOR may be justified in some cases where other factors exist sucn as waste disposal problems or presence of other nuclear facilities on-site, combined with the potential for reduced occupational dose. 100 year SAFSTOR is not considered a reasonable option since it results in the continued presence of a site dedicated to radioactivity containment for an extended time period with Ittle benefit in dose or waste volume reduction comparea to 30 year or 50 year SAFSTOR. ENTOMB is unlikely to be a reasonable option for research and test reactors since it results in the presence of a radioactive site for an extended period of time, and due to the lack of significant benefit in dose or waste volume reduction compared to the other alternatives, and the lack of significant cost reduction compared to the other alternatives. In ad-dition, uncertainties regarding characterizaton of residual radioactiv ity over the entomboent period might result in additional costly decommissioning activ-ity in order to release the facility for unrestricted use.

10/07/87 7-15 NUO586 CH 7

REFERENCES b

1. Code of Federal Regulations, Title 10, Energy, Section 170.3.

2. G.J. Konzek et al. , Technology, Safety, and Costs of Decosusissioning Reference Nuclear Research and Test Reactors, NUREG/CR-1756, prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, February 1982.

3. G. J. Konzek, Technology, Safety, and Costs of Decommissioning Reference Nuclear Research and Test Reactors, NUREG/CR-1756, Addendum, prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, July 1983. .

4. Links, B. W. and R. L. Miller, Summary Report North Carolina State' University Research and Training Reactor, NUREG/CR-3370, prepared by UNC Nuclear Industries for the U.S. Nuclear Regulatory Commissien, August 1983.

1 e

O 6

10/07/87 7-16 NUO586 CH 7

. . . , . ,....._,.....m,...,,,.s.m_. .m..

8 Ceco m ISSIONING OF REACTORS THAT HAVE BEEN INVOLVED IN ACCIDENTS The facilities discussed in the preceding sections are representative of f aci'.ities which would undergo routine deccmissioning at the end of their nomal lifetimes. An additional significant area of consideration is the decommissioning that occurs as a result of the premature closure of a reactor due to an accident. A post-operations activities flow sheet showing both a nomal decommissioning and the situation for a reactor involved in an accident is shown in Figure 8.0-1.

As can be seen from the figtre, the activities following shutdown of a facility involved in an accident are somewhat different from the normal situation. These activities include a stabilization period. The stabilization period is the period during which time the accident is brought under control and the facility is brought to a statilized condition. Once the situation is stabilized, acci-dent cleanup can begin. Accident cleanup is considered to be those activities leading to defueling the reactor and to cleanup of contamination and processing and disposal of wastes generated by the accident. As shown in Figure 8.0-1,

! - the accident cleanup period could either be followed by recovery of the facility for a restart, or by decommissioning. If, as is analyzed in the GEIS, it is decided that the facility is to be retired from service, decommissioning activi-l ties are considered to begin following completion of the accident cleanup. -

l Much of what folicws is based on the NRC-sponsored Pacific Northwest Laboratory (PNL) Study on the technology, safety and costs of decommissioning hference light water reactors following postulated in accidents! dr illustration purposes, only the more detailed PWR results are presented. The study did not analyze the stabilization perlod or the recovery of a facility for' restart.

The study did present an analysis of the accident cleanup period, including a consideration of the sensitivity of the costs of accident cleanup to several l

facters, including delays in the cleanup, alternative processing systems, additional structures, alternative disposal requirements, ai,J storage of waste onsite. The accident cleanup period is postulated to include the following 10/07/87 8-1 hf86 CH 8

.<- .- . ,, , i., .-. -

l o l 1

= 1 1

l l

NUCLEAR F ACILITY l

NORMAL ACTIONS POST ACCIDENT ACTIONS V

STOP OPER ATIONS STABILI2E ACTiv; TIES THE PLANT 9P

• ACCIDENT CLLAN UP ACTivlTIES V

OECISION POIN T V V 9P RECOVERY DEcouMISSIONINC DECOMMI5510NINC FOR REUSE 9P 9P RELE ASE F ACILITY RELI ASE F ACILITY FOA UN AESTFICTED USE FOR UNRESTRICTED USE ITERMIN ATE LICENSE) (TERMIN ATE LICENSE) ,

FIGURE 8.0-1. Post-Operations Activities flow Sheet O

?-2  % n./-N /

. .. a. .. .....s...

.;c_.. .,n.. .;

1 l

l l

l tasks: (1) processing the contaminated water generated by the accident (and by ce:ontamination operations); l2) initial decontamination of building surfaces; (3) removal of spent fuel (undamaged and damaged) from the reactor; (4) cleanup cf the reactor coolant system; and (5) solidification and packaging of wastes from accident cleanup operations.

As discussed in the PNL study,1 these accident cleanup tasks are necessary and w uld be approximately the same whether the reactor is ultimately refurbished or decomiss'oned, and if decommissioned, the same regardless of which decoe-sissioning alternative is chosen. The rationale for this is that decontami-r.ation during the accident cleanup period (whether for eventual restart or decommissioning) cannot be too corrosive since this could compromise the in-tegrity of systems which must remain intact during cleanup and decommissioning, especially if a delayed decomissioning alternative, such as SAFSTOR, is cho s e n.1 ' 2 In addition, major equipment items such as the reactor vessel, reactor coolant pumps, and steam generators could not be dismantled until af ter accident cleanup is completed since they form part of the primary systems.

Thus, even if it were decided to permanently shut down a facility following an a:cident, the seauence of activities would be accident cleanup followed by decommissioning. Because, as discussed in Section 2.6.2, the period of acci-dent cleanup is covered by regulations which require insurance (10 CFR Part 50.54(w)),3 this GEIS does net present further details on the accident cleanup period. This GEIS does include the effects that the accident and the activities during the accident cleanup period would have on the decommissioning '

of the facility.

This GEIS section presents a summary of the cetailed analysis done by PNL on the decommissioning of a reactor following an accident.1 Following the completion of the accident cleanup activities, decommissioning activities begin. As a result of the efforts during accident cleanup, the decommissioning activities are considered to be not greatly affected by the condition of the plant immediately following the accident. In addition, many of the uncertain conditions have been removed during the accident cleanup, specifically the da.maged core has been removed from the reactor, the large volumes of uncontained highly radioactive water have been processed, the large areas of contaminated ,_

building surfaces have been treated, and construction of necessary systems cnd l

10 /07/87 6-2 NUO5% CP Q

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l r

structures has been completed. Hence, decommissioning can be carried out it a more stable environment than the accident cleanup. Nevertheless, there would be certain impacts on the decommissioning from the accident and the accident cleanup activities including increased levels and spread of contamination compared to normal decommissioning still remaining after the cleanup activities, the need to decommission systems and structures built and used during accident cleanup, and the potential need to store wastes generated by the accident, and during the accident cleanup period, onsite on an interim basis for an extended time period.

' 8.1 Reference Facility Description and Reference Accident Scenarios The reactors used as the reference facilities for the post-accident decommis-sioning analysis are the same as those used as the reference PWR and BWR in Chapters 4 and 5, respectively. The choice of these facilities as the reference reactors is made to facilitate comparisons between the requirements and costs of post-accident decommissioning given in this section and the requirements and costs of normal shutdown decommissioning given in the earlier chapters, and is not intended u imply anything about the reliability and/or safety of these reference reactors relative to other PWRs or BWRs in operation or under construction. The reference site used in this section is the same as that indicated in Section 3.

Three reference accident scenarios are analyzed to illustrate a range of tech-nological requirements, public and occupational doses, and costs that are -

greater than those estimated for decommissioning following normal shutdown.

For the purposes of this GEIS, the consequences of an accident (i.e., the radiological and physical condition of the plant following an accident) are auch more important than the sequence of events that occur during the accident.

Therefore, detailed descriptions of accident sequences were not analyzed. The reference accident scenarios provide information about radioactive contamina-tion, radiation exposure rates, and damage to the fuel core and to the contain-ment building. The consequence scenarios chosen for this study are believed to be credible with respect to initiating circumstances and are in agreement with scenarios currently considered as design basis by the NRC in safety

evaluations. The postulated scenarios, listed in increasing orJer of tne dif ficulty of post-accident decorrnissioning, are:

1. A small loss-of-coolant accident (LOCA), e.g. , a small steam line break or the inadvertent opening of a safety or relief valve) in which the emergency core cooling system (ECCS) functions to cool the core and to limit the release of radioactivity. Some fuel claddirs rupture is postulated, but no fuel melting. The consequence scenario includes moderate contamination of the containment building but no significant physical damage to the building and equipment.

2. A small LOCA in which ECCS is delayed, resulting in 50% fuel cladding failure and a small amount of fuel melting. The consequence scenario includes extensive radioactive contamination of the containment building %t only minor physical damage to the building and equipment.

It also includes radioactive contamination of auxiliary and fuel handling buildings.

3. A major LOCA (e.g., the rupture of a main coolant line) in which ECCS is delayed, resulting in 100% fuel cladding failure and significant fuel melting and core damage. The postulated consequences include extensive radioactive conttmination of the containment building and major physical damage to structures and equipment. Some radioactive contamination of the auxiliary and fuel buildings is also postulated.

This GEIS does not consider the advisability or merit of permanently shutting down a facility which has been involved in one of the accident scenarios described above.

8. 2 Post Accident Decommissioning Experience Very few reactor accidents have occ'rred that have necessitated extensive post-accident cleanup operations or i. ave resulted in a requirement to decommis-sion the reactor. Primarily, the accidents that have resulted in significant contamination have occurred at small experimental or test reactors. One large reactor, the Three Mile Island Nuclear Plant, has experienced an accident that

... . . . . . . .........n . . ~ . . . . ... .i.. . , .s

. . . r . ..;

.,. .. .. . . q l

resulted in significant contaminatio. similar to

• iat discussed in Sec-  !

tion 8.1. 2 Information on cleanup and decommissioning experience is con-tained in Table 8.2-1. The experience at these facilities provides useful I information at,out cleanup procedures and decommissioning accident damaged facilities.

Table 8.2-1 Summary of nuclear reactor posyccident cleanup and decommissioning experience Facility name Power Year of Status following and location Reactor type level accident accident cleanup Canadian NRX Research, pool 10W 1952 Returned to service Canadian NUR Research, heavy 200W 1958 Returned to service water SL-1 Reactor Military, BWR 3W 1961 Oc:ommissioned PRTR Research, heavy --

1965 Returned to service water Enrico Fermi Fast breeder --

1966 Returned to service Lucens Experimental, 30W 1969 Decommissioned heavy water Three Mile Island Commercial, PWR 2800W 1979 Still in accident cleanup (2)

(1) Data in table taken f rom Reference 1.

(2) No decision made as to eventual plant status. .

Most of the techniques and procedures used to decontaminate or decommission a reactor following an accident are similar to those used for reactor decommis-sioning following normal shutdown, although considerations must be given to

! the problems of working in higiner radiation environments than normal. Some reactor accidents have resulted in high levels of radioactive contamination on building surfaces and equipment and in high radiation exposure rates to accident cleanup personnel. In all cases where contamination has occurred, methods and procedures have been devised to safely remove the contamination j with only medest total radiation doses to decontasination workers.

,.,n,,e, --

The March 28, 1979, accident at Three Mile Island, Unit 22 (TMI-2) resulted in an accident cleanup effort at that facility which will involve years of work.2 Cleanup of TMI-2 will provide experience in procedures and techniques related to the processing of highly contaminated liquids, the removal of damaged fuel from a reactor, and the handling and disposal of high-activity ra:ioactive waste.

8. 3 Decomissioning Alternatives Under normal circumstances, decommissioning follows the orderly sNtdown of the facility at the end of its planned life. However, as discussed above in Section 8.0, decommissioning at a reactor which has been involved in an accident would take place following stabilization and accident cleanup activities. As defined in Section 2.3 decommissioning means safely disposing of all radio-active materials in excess of levels which would permit unrestricted use of the facility.

The accident and the subsequent accident cleanup activities have a t effect on decommissioning activities, on the decommissioning alternatives, a d on the cost, safety and environmental consequences of those alternatives. These effects include the larger levels and sp ead of contamination than would be the case for normal decommissioning with resultant higher occupati nal exposures; different types of contamination (i.e. , Sr-90 and Cs-137 control occupational exposure for post accident decommissioning, whereas Cc-60 controls for normal decommissioning); the need to decommission accident cleanup systems; '

and the potential for interia onsite storage of wastes generated by the acci-dent and by the accident cleanup activities. The following sections discuss the impact of an accident and the accident cleanup activities on tre alternatives DECON, SAFSTOR and ENTOMB.

8.3.1 DECON DECON is defined as the imediate removal and disposal of all radicactivity in excess of levels which would permit release of the facility for unrestricted use. The end result is the release of the site and any remaining structures l

l 1 . . . . . . .

. . . , , . . . . . . . . . . . . . . .- ...... .. . .. .. ~. ...

for unrestricted use. Tc achieve an unrestricted use condition the following tasks must be performed during OECON: 1) remove activated and contaminated materials from the reactor building; 2) decontaminate the reactor building to unrestricted release levels; 3) dismantle and decontaminate fuel and auxiliary {

buildings, and turbine building and other buildings; 4) package and snip all i contaminated materials; 5) dispose of all fuels, damaged and undamaged; and

6) survey facility and site for acceptable levels of residual radioactivity.

DECON has the same advantages as outlined in Section 4.3.1, such as making the site available for unrestricted use, the availability of a knowledgeable work force, and the elimination of the need for long term security and surveillance.

Dis 0dvantages are also similar to those indicated in Section 4.3.1, including the larger occupational exposure and larger initial requirement for waste dis-posal space compared to the other alternatives. In particular, following an accident the difference in occupational exposure between DECON and SAFSTOR is higher than it is for normal decommissioning (see Table 8.3.2). Also, following an accident, there is a potential that the reactor may be unable to dispose of wastes generated during the accident cleanup which could result in the need for extended onsite storage of wastes. (These wastes could include low-level wastes, as well as high level wastes and fuel assemblies). If this occurs, DECON of the reactor site would not be feasible.

The cost of DECON as estimated by the PNL study following the accident cleanup activities is given in Table 8.3-1 for the three reference accident scenarios.

The occupational and public exposure resulting from the DECON activities, as estimated by the PNL study, is given in Table 8.3-2 for the three reference accident scenarios.

8.3.2 SAFSTOR SAFSTOR is defined as those activitie', required to place (preparation for safe storage) and maintain (safe storage) a reactor in such condition that the risk to safety is within acceptable bounds, and that the facility can be stored and subsequently decontaminated to levels which permit release of the facility for unrestricted use (deferred decontamination).

1

. .. ,.> . . . . . . . ~ . > . . . , .. . . .. . . . . . . .

I l

Table 8.3-1 Sumcary of estimated costs for decornissioning of the (

reference PWR following accident cleanup  ;

I Costs of ($ millions)(a)(b)

Scenario 1 Scenario 2 Scenario 3 Decommissioning alternatives accident accident accident DECON 79.5 104.1 154.8 30 year - SAFSTOR Preparations for safe storage 20.0 22.6 28.8 Continuing care costs 4.0 4.0 4.0 Deferred decontamination 67.6 88.5 131.6 Total 30 year SAFSTOR costs 9T ii 115.1 164.4 100 year - SAFSTOR Preparations for safe storage 20.0 22.6 28.8 Continuing care costs 13.8 13.8 13.8 Deferred decontamination 52.5 68.7 102.2 Total 100 year SAFSTOR costs 86.3 10EI 144.8 ENTOM8(c)

Entombment 57.4 76.1 111.8 Con'.inuing care costs (for 100 years) 7. 2 7. 2 7. 2

(*) Costs are in early 1986 dollars and include a 25% contingency.

(b) Updated frcm Raference 1, Table 2.10-5.

(c)If require.1, deferred decontamination at the end of the continuing care period for ENTOMB is estimated to cost at least as much and perhaps more .

than deferred decontamination at the end of the corresponding continuing care period for SAFSTOR.

Table 8.3-2 Summary of radiation safety analysis for gcommissioning the reference PWR following accident cleanupg Dose (man-rem)

SAFSTOR DECON 30 years 100 years ENTOM3 Occupational exposure (D)(d)

Safe storage preparation NA(I) 429 429 NA Continuing Care NA 120 225 2,518 Decontamination 3,063 1,500 300 NA Entombment NA NA NA 2 Safe stur, prep. truck shipments NA 13 13 NA Decontamination truck shipments 200 100 20 NA Entombment truck shipments NA NA NA 90 Total 3,263 2,162 987 2,608 Ptblic uposure(c)(e)

Safe storage preparation NA neg. neg. NA Continuing care NA neg. neg. neg.

Decontamination neg. neg. neg. NA Entombment NA NA NA neg.

Safe stor. prep. truck shipments NA 1.2 1. 2 NA Decontamination truck shipments 19 9.5 1. 9 NA Entombment truck shipments NA NA NA 8. 4 Total 19 10.7 3.1 8. 4 (a) Values given are for decommissioning following the accident cleanup of the scenario 2 accident. .

(b) Values for occupational exposure for deconaissioning following a scenario 3 accident are estimated to be a factor of 2 to 3 times higher than the scenaric 2; for a scenario 1 a;cident, exposures are estimated to be 2 to 5 times lower.1 (C) Values for public exposure for decommissioning following a scenario 3 accident are estimated to be a factor of 2 to 5 times higher than for the scenario 2; for scenario 1 accident, exposures are estimated to be 2 to 5 times lower.

(d)from Reference 1, Tables 14.3-4, 14.3-5, and 14.3-7.

(*)From Reference 1, Tables 14.3-2 and 14.3-7.

(I)NA not applicable.

(9)Neg - dose is estimated to be less than 0.001 man-rem.

. . . . . . . e : : . ,- . ..

The advantages of SAFSTOR are similar to those indicated in Section 4.3.2, including the red.ction in occupational exposure resulting from the deferral of some decommissioning tasks, and the reduction in waste disposal requirements.

Disadvantages of SAFSTOR are similar to those indicated in Section 4.3.2, including need for continuing security and surveillance, the need to maintain a site, and the neec to use personnel unfamiliar with the facility for the deferred decontami ation.

In particular, following an accident the amount of benefit in dose reduction is not as great wi;h SAFSTOR as it is for normal decommissioning. This is because the occupa;ional exposures during post accident decommissioning are primarily due to S-90 and Cs-137 which are released from the fuel during the accident and contaninate building and piping surfaces. Sr-90 and Cs-137 have half-lives of appr:ximately 30 years. This is different from the normal situation where oc:vpational exposures are primarily due to Co-60 which has a half-life of 5.27 years. Because of the long half-life of the controlling nuclides it would take a longer time perind to reduce occupational exposures.

30 year SAFSTOR recuces exposures by a factor of approximately 1.5 and 100 year SAFSTOR only reduces dose by a factor of 4 (compared to normal decommissioning where 30 year SAFST3R results in a doss reduction of 4). Thus, long SAFSTOR periods would be necessary to accomplish occupational dose reduction.

Use of SAFSTOR migh be likely if it is necessary to provide for interim onsite storage of wastes for an extended period of time into the decommissioning period. This might occur because of political or regulatory constraints -

against disposal of waste, because of inadequate disposal capacity for low level waste, or lact of disposal sites for high level waste, including the spent fuel. It is unlikely that most reactor sites could qualify as permanent waste repositories ucause of such factors as nearby population densities and hydrology. Therefore, storage of wastes onsite would be an interim measure, albeit for an exten:ed time, followed ultimately by decontamination of the f acility and site.

The cost of SAFSTOR of the reactor as estimated by the PNL study following the accident cleanup activities is given in Table 8.3-1 for the three accident scenarios. The occmational and public exposure resulting from the SAFSTOR

activities, as estimated by the PNL study, is given in Table 8.3-2 for the three reference accident scenarios.

8.3.3 ENTOMB l ENTOMB is the compete isolation of radioactivity from the environment by means of massive concrete barriers until the radioactivity has decayed to levels which permit release of the facility for unrestricted use.

ENTOMB is intended for use where the residual radioactivity will decay to levels permitting unrestricted release of the facility within reasonable time periods. Recommended policy on reliance on' institutional control for contain-ment of radioactivity is approximately 100 years.4 Some of the discussion of Section 4.3.3 concerning ENTOMB is pertinent here, including advantages and disadvantages, structures which would be entombed, and certain nuclides which would be involved. However, there are certain important considerations for ENTOMB as a post-accident decommissioning alternative that makes it less attractive as an alternative than it is for normal decommissioning. This is because of the higher levels of the entombed radioactivity resulting from accident generated contamination in the plant, and slower decay of the post-accident radionuclide inventory which is controlled by Sr-90 and Cs-137, with 30 year half-lives. Therefore, use of ENTOMB as the decommissioning alterna-tive following an accident would necessitate eitner a period of retention of the entombed structure for longer than 100 years to allow decay of radioactivity to unrestricted use levels, or an eventual deferred contamination of the en-tombed structure. This decontamination would involve significantly greater j time and manpower commitments and costs expenditures than, for example, deferred decontamination for an unentombed structure, since the entombed structure is built to endure for a long period of time.

The occupational and public exposure resulting from ENTOM8 activities, as esti-l mated by the PNL study, is given in Table 8.3-2 for the three reference accident scenarios. The cost given in Table 8.3 .' includes the cost of entombing the structure and the annual continuing care costs, but does not include the cost of deferred decontamiriation which may likely be necessay after approximately 100 years to reduce radioactivity to unrestricted use levels. The cost of the 10/07/87 8-11 NUO586 CH B

deferred decontamination for ENTOMB is estimated to add at least $33 million,

$45 million, and $70 million to the cost of ENTOMB for the reference accident scenarios 1, 2 and 3, respectively.

8.4 Environmental Consequences This section discusses environmental consequences other than the radiation dose consequences discussed above in Section 8.3. These other consequences include waste disposal, radioactivity released due to industrial accidents during decommissioning, and socioeconomic impacts. .

With regard to waste disposal, the volumes of waste to be disposed of during the decommissioning of a reactor, following the accident cleanup of each of the three reference accident scenarios, are contained in Table 8.4-1. These wastes include disposal of neutron activated steel and concrete, contaminated concrete and equipment, and dry and wet radioactive wastes. In arriving at the data in Table 8.4-1, it is assumed that the wastes generated during the accident cleanuo period are disposed of prior to the decommissioning period. These wastes include low-level radioactive wastes, as well as highly radioactive and/or transuranic wastes, and damaged and undamaged fuel assemblies. Based on the criteria of 10 CFR Part 61, the low level radioactive wastes resulting from accident cleanup are assumed to be disposed of by shallow land burial. Because the criteria of 10 CFR 61 may result in the high level radioactive wastes and transuranic wastes generated during accident cleanup being deemed unsuitable for shallow land burial, they are assumed to be sent to a federal repository. Similarly, because the criteria for disposal of the damaged and undamaged fuel is not yet well de-l fined it is assumed to be sent to a Federal repository.

Because of the potential that a reactor involved in an accident may be unable to dispose of the wastes including spent fuel generated during accident cleanup as assumed in the previous paragraph, either because of lack of disposal capa-city or regulatory or political constraints, there may be onsite storage of both accident cleanup wastes and decomissioning wastes for an extended period of time. This would result in additional surveillance costs and an extension i

of the completion of decommissioning. Details of this storage are discussed in 1

Section 2.7.

10/07/87 8-12 NUO586 CH 8

Table 8.4-1 Burial volume of radioactive waste and rubble for the reference PWR following the accident cleanup at a reactor involved in an accident (a'b)

Decommissioning alternative Volume (m3)(c)

DECON 18800 SAFSTOR Total of preparations for safe storage, continuing care and deferred decontam,ination(b) following safe storage for: 30 years 18800 100 years 18800 ENTOMB (d) 8200 (a) Values given are for decomissioning following the accident cleanup of the scenario 2 accident.

(b) Values of waste volumes for decommissioning following the accident cleanup of a scenario 2 or scenario 3 accident are estimsted to be less than 15% difference from the values in the table.1 (c)from Reference 1, Tables H.1-3, H.2-3, H.2-8, and H.3-3.

(d) Volume of entombing structure and wastes within are not included.

l l

10/07/87 8-13 NUO586 CH 8

PNL considered releases of radioactivity resulting from indJstrial accidents during the decommissioning activities and the results are presented in Table 8.4-2. Radiation doses to the maximum exposed individual from accidental airborne radioactivity releases during decommissioning operations were calcu-lated to be quite low. Radiation dose to the maximum-exposed individual from accident radioactivity releases resulting from transportation accidents were calculated to be low for the most severe accident.

The biggest socioeconomic impact will have occurred before decommissioning started, following the acciden'. at the plant, namely the shutdown of the plant and the accident cleanup. The decommissioning staff will be approximately the same size as the accident cleanup staff. This GEIS does not consider the ad-visability or merit of whether a facility should be restarted or decommissioned following an accident.

8. 5 Comparison of Decommissioning Alternatives From examination of Tables 8.3-1 and 8.3-2, it appears that DECON or SAFSTOR are reasonable options for decommissioning a reactor following accident cleanup at a reactor that has experienced an accident. DECON costs less than SAFSTOR ,

and its larger occupational radiation dose is considered of marginal. significance to health and safety. Either of the two SAFSTOR options would be feasible since due to the long half-lives of the controlling radionuclides, there would be continued reduction in dose beyond the 30 year SAFSTOR. In addition, SAFSTOR may be a necessary alternative to account for the potential need to store -

accident generated wastes for an extended time period.

ENTOMB appears less desirable for the reasons discussed in Section 8.3.3.

Because of the large quantities of contamination and the long half-lives of the controlling nuclides it would be necessary to keep the reactor entombed for a period of time greater than 100 years in order for the facility radioactivity levels to decay to unrestricted use levels. This is not acceptable since it is not consistent with iecommended policy on reliance on institutional control for radioactivity confinement. Deferred decontamination of the entombed structure af ter 100 years would be difficult and result in the ENTOMB alter-native being more costly than DECON or SAFSTOR, generating more waste than -

DECON or SAFSTOR, and causing larger occupational exposures than SAFSTOR.

10/07/87 8-14 NUO586 CH 8

Table 8.4-2 Summary of radiation doses to the maximum-exposed individual from postulag '

releases due to industrial accidents during post-accident decommissioning Radiation dose to lung (rem) during:

Tctal DECON(d) Prg. fu sde storap release Incid:nt (pCi/hr) First year Fifty year First year Fifty year Explosion of LPG leaked from loader (b) 1.8 x 104 6.1 x 10 4 1.2 x 10 2 --(c) __

Explosion of oxyacetylene during vessel segmentation 3.6 x 102 6.1 x 10 8 1.2 x 10 2 -- --

Explcsion/ fire of ion exchange resin 1.9 x 102 6.5 x 10 8 1.3 x 10 5 -- --

Gr ss leak during decontamination - spray leak 1.1 x 102 3.8 x 10 8 7.5 x 10 8 3.8 x 10 8 7.5 x 10 s

- liquid leak 3.5 x 10 1 1.2 x 10 a 2.4 x 10 8 1.2 x 10 s 2.4 x 10 s S:gmenting undecontaminated RCS piping 1.1 x 101 7.3 x 10 7 7.9 x 10 7 -- --

Vrcuum bag rupture 5.0 x 100 -- --

1.7 x 10 7 3.4 x 10 7 Less of contamination control during vessel segmentation 2.3 x 100 3.9 x 10 10 4.4 x 10 8 -- --

i Accidental spraying of concentrated centamination with high pressure spray 6.0 x 10 1 -- --

2.0 x 10 s 4.1 x 10 s Filter loss during blasting of concrete bioshield 3.0 x 10 1 2.0 x 10 9 2.2 x 10 9 -- --

Less of portable filtered ventilation enclosure 1.5 x 10 1 5.1 x 10 8 1.0 x 10 8 __ __

Accidental break of contaminated piping 1.1 x 10 1 -- --

7.3 x 10 9 7.9 x 10 9 l Fire involving combustible radioactive wastes 3.0 x 10-2 1.0 x 10 9 2.0 x 10 9 1.0 x 10 9 2.0 x 10 8

(*)Raference 1, Table 14.3-3.

)All releases assumed to occur during a 1-hr period, for comparison purposes.

(c)A dash indicates the particular accident situation is not considered for the decommissioning alternative because either the accident situation does not apply to that alternative or a similar accident of

greater consequences is analyzed.

) (d) Corresponding doses for ENTOMB are assumed to be the same as those shown for DECON, with the deletion of these situations that arise from activities not undertaken during DECON (e.g., blasting, segmenting of 4

the vessel).

9

Table 8.4-3 Summaryofestimatedr:diaticndos:stothemaximum-exposedindgjdual from postulated transportation accidents during decommissioning Fifty year committed Total First year dose (rem) dose equivalent (rem)

Accident severity release (Ci/hr) (b) Total-body Bone Lung Total-body Bone Lung Minor 5 x 10 4 2.5 x 10 4 6.0 x 10 4 8.0 x 10 4 5.5 x 10 4 4.8 x 10 3 1.6 x 10-3 Severe 2 x 10 2 1.0 x 10 2 2.4 x 10 2 3.2 x 10-2 2.2 x 10 2 1.9 x 10 2 6.4 x 10 2

(*) Reference 1, Table 14.3-8.

(b) Releases assumed to occur in a 1-hr period for comparison purposes.

5

)

e t

I

l j

REFERENCES [

i i

1. E. S. Murphy, G. M. Holter, Technology, Safety and Costs of Decommissioning 1 at'a Reference Light Water Reactors Following Postulated Accidents, NUREG/

CR-2601, U.S. Nuclear Regulatory Commission, November 1982.

2. Final Programmatic Environmental Impact Statement related to decontamina-tion and disposal of radioactive wastes resulting from the March 28, 1979 accident at Three Mile Nuclear Station, Unit 2, NUREG-0683, U.S. Nuclear Regulatory Commission, March 1981.

3. Federal Register, 47 FR 13750, "Elimination of Review of Financial Qualifications of Electric Utilities in Licensing Hearings for Nuclear Power Plants," March -31,1982.

4. Federal Register, 46 FR 38081, "Licensing Requirements for Land Disposal of Radioactive Waste," July 24, 1981.

10/07/87 8-17 NUO586 CH 8

9 FUEL REPROCESSING PLANT A fuel reprocessing plant (FRP) is a facility for reclaiming plutonium and uranium from spent nuclear reactor fuel, so that the reclaimed plutonium and uranium can be later refabricated into new fuel elements. For the purpose of this section, it is assumed that the plant is to be operated 30 to 40 years. It is also assumed that any accidental releases of radioactive material are cleaned up immediately following the event. The generic site of a fuel reprocessing plant is described in Section 3.1.

This section is based primarily on a detailed study 1 of tne decommissioning of a fuel reprocessing plant conducted by Pacific Northwest Laboratory (PNL) for the NRC. In this study, PNL selected the Barnwell Nuclear Fuel Plant (BNFP),

located in Barnwell, South Carolina, as the reference FRP and assumed it to be located at the generic site. Although the Barnwell facility has never operated as an FRP, its design is considered to have characteristics typical of those present in any future FRPs. PNL then developed and reported information on the available technology, safety considerations, and probable costs for decom-missioning-the reference facility at the end of its operating life.

9.1 Description of Fuel Reprocessing Process and Facility 9.1.1 Process Description The reference plant uses the Purex process to recover plutonium and uranium from irradiated LWR fuels. A simplified block flow diag am of this process is shown in Figure 9.1-1.

The irradiated fuel is received in heavily shielded casks and is unloaded and stored underwater in the fuel receiving and storage station (FRSS). When ready for processing, each fuel assembly is transferred to the main process building where it is partly disassembled, chopped into pieces up to 10 cm long and 10/07/87 9-1 NUO586 CH 9

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i 10/07/87 9-2 NV0586 CH 9 L

dropped into a dissolver vessel where the fuel materials are dissolved with nitric acid. The undissolved fuel cladding hulls are packaged and taken to a bunker-type storage area onsite.

The nitric acid-fuel solution is then subjected to a solvent extraction process where the uranium, plutonium, and fission products are separated into individual streams, and the uranium and plutonium are purified and converted to uranium hexafluoride and plutonium oxide for offsite shipment. The fission products are stored in underground water-cooled tanks for about 5 years and then solidi-fled for disposal in a federal facility.

9.1.2 Flant Description The major facilities included in the reference reprocessing plant are: 1) the fuel receiving and storage station, 2) the main process building, 3) the high-and intermediate-level liquid waste storage area, 4) the waste solidification plant, and 5) the radioactive auxiliary service areas. Detailed descriptions of these facilities are presented in Reference 1.

The following is a listing of various operating parameters of the reference FRP:

Inputs to the FRP Spent Fuels from Light Water Reactors (Zircaloy or stainless steel cladding) with the following content:

UO2 (up to 3.5% enrichment when input to the reactor)

UO2 -Pu02 (Pu up to equivalent of 3.5% 23sU when input to the reactor)

Special fuels up to 5% initial enrichment under special operating conditions Spent Fuel Burnup(a);

From PWRs, average exposure of 31,800 MWD /MTHM (peak of 33,000 MWD /MTHM)

From BWRs average exposure of 25,300 MWD /MTHM (peak of 26,000 MWD /MTHM)

For total input, average total exposure of 29,300 MWD /MTHM Spent Fuel (a) Processing characteristics listed are different from those postulated for  !

near-term operation of BNFP. The information presented is currently l expected to be representative of long-term operating characteristics at a l plant such as BNFP. l 10/07/87 9-3 NUO586 CH 9

Out-of-Reactor Time prior to FRP input:

Minimum of 90 days prior to receipt at FRP

- Minimum of 1.5 years before reprocessing at FRP(a)

FRP Reprocessing Capacity (in MT of Spent Fuel)

- 1,500 MT/yr (30 yr lifetime)(a) average capacity 5 MT/ day peak capacity .

Products of Reprocessing Uranyl nitrate solution (converted to UFs for shipment from FRP to burial grounds)

Plutonium nitrate solution (converted to Pu0 for2 shipment from FRP to burial grounds)(a)

Wastes Resulting from Reprocessing High-level and intermediate-level wastes stored on an interim basis as liquids in underground tanks.

High and intermediate level liquid wastes converted within 5 years to a vitrified solid and shipped offsite to a Federal repository. -

- Fuel cladding hulls, failed equipment and other solid wastes stored onsite on an interim basis in concrete or stainless steel containers in engineered underground storage prior to shipment offsite for disposal.

(a) Processing characteristics listed are different from those postulated for near-term operation of BNFP. The information presented is currently expected to be representative of long-term operating characteristics at a plant such as BNFP.

10/07/87 9-4 NUO586 CH 9

Effluents from Reprocessing During Normal Operation

. Gases (only routine radioactive effluents are indicated):

asKr disharged up main stack (100 meters tall).

Majority of tritium and 14C discharged to main stack.

Excess water discharged up main stack as vapor.

Heat rejected to cooling tower via closed loop heat exchangers.

Process liquid wastes with low contamination diluted and discharged to river.

9.1.3 Estimates of Radioactivity Levels at FRP shutdown Estimates of radioactivity levels in the reference fuel reprocessing plant after reprocessing operations have been terminated (all spent fuel removed) and final operational cleanout flushings of the process areas have been completed are summarized in Reference 1.

9. 2 Fuel Reprocessing Plant Decommissioning Experience To date, there has been no experience in the decommissioning of a commercial FRP. Federal facilities at the Hanford, Savannah River, and Oak Ridge sites that have been involved with the reprocessing of irradiated fuels have been decontaminated and their equipment disassembled.2 A substantial amount of this information is directly relatable to decontamination of future fuel reprocessing'

, plants.

l The Nuclear Fuel Services (NFS) plant in West Valley, New York, is the only commercial reprocessing plant that has operated in the United States

(although it is not currently operating). The NFS situation is not directly translatable to the present or projected nuclear power industry because a national policy (10 CFR 50, Appendix F) requiring the solidification of high-level waste was not established until 1971, well after the plant began ,

operation. Therefore, since NFS has its reprocessing high-level wastes stored in large underground tanks in slurry form (similar to the practices followed 10/07/87 9-5 NUO586 CH 9 1

at the Hanford and Savannah River sites), the costs of decommissioning this plant are _ expected to be higher than that of newer FRPs.

9. 3 Decommissioning Alternatives Once a fuel reprocessing plant has reached the end of its useful operating life, it must be. decommissioned. As discussed in Section 2.3 this means safely removing the facility from service and disposing of all radioactive materials in excess of levels which would permit unrestricted use of the property. Alter-natives considered here as to-their potential for satisfying this general requirement for decommissioning include DECON and SAFSTOR (passive SAFSTOR and custodial SAFSTOR). ENTOMB is not considered a viable option because of long-lived transuranics present in the entombed structure resulting in radiation exposure which does not decrease with time. The, disposition of the nonradio-active buildings and facilities is left to the discretion of the facility owner and is not part of the decommissioning procedure. This section discusses the decommissioning alternatives evaluated for the FRP.

9.3.1 -DECON DECON is defined as the immediate removal and disposal of all radioactivity in excess of levels which would permit release of the facility for unrestricted use. Nonradioactive equipment and structures need not be torn down or removed as part of a DECON procedure. The end result is the release of the site and ,

any remaining structures for unrestricted use as early as the 5 years estimated for decommissioning after the end of facility operation.

DECON is advantageous because it allows for termination of the NRC license with tq

^

a relatively few years after cessation of facility operations and removes a radioactive site. DECON is advantageous if the site is required for other <

purposes, if the site is extremely valuable, or, if for some reason the site must be immediately released for unrestricted use. It is also advantageous in that the facility operating staff is available to assist with decommissioning and that continued surveillance is not required. An important disadvantage is i the higher occupational radiation dose which occurs during DECON compared to i

the SAFSTOR alternative.

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._ ]

Three important radiation exrosure pathw&ys need to be considered in the evalua-tion of the radiation safety of normal FRP decommissioning operations: Inhala-tion, ingestion, and external exposure to radicactive materials. For reasons similar to those discussed for PWRs in Section 4.3.1, during decommissioning the dominant exposure pathway to workers is external exposure while for the public the dominant exposure pathway is inhalation. During the transport of radioactive waste, the dominant exposure pathway is external exposure for both transportation workers and the public. A summary of the doses resulting from these pathways is presented in Table 9.3-2.

Occupational Radiation Oose The occupational radiation dose from external exposure to radioactive materials, not including transportation of radioactive waste, is estimated to be about 512 man rem over the 5 year period of DECON. Occupational radiation doses were calculated by PNL from estimated radiation levels in the various areas of the reference FRP and from man-hour estimates for performing the decontamination operations. Table 9.3-2 gives the estimated occupational external radiation exposure for DECON.

The reference FRP was designed to store high-level liquid waste (HLLW) for five years prior to solidification and then to store the solidified waste five years prior to shipment to a federal waste repository. It is expected that any future FRPs would be designed to solidify the HLLW continuously within the pro-cess building, and store only solidified waste. Therefore, future plants would ~

use a few smaller tanks instead of the large underground HLLW storage tanks and separate waste solidification plant. This would reduce the decommissioning occupational radiation exposure and costs by between 40 to 50 percent.

Public Radiation Dose The inhalation radiation dose to the public resulting from radionuclide releases during DECON, not including doses during transportation of radioactive waste, is estimated to be 10.2 man-rem (50 year population dose commitment to the whole-body). This radiation dose is very small compared to the background 10/07/87 9-7 NU0586 CH 9 t

I

1 radiation exposure normally receivcd by members of the public. Details of the methods used for calculation of doses are found in Reference 1.

Public Radiation Dose from Postulated Accidents During DECON DECON procedures were examined and potential accidents postulated that could lead to the release of radioactive materials. The largest radiation dose to the maximum-exposed individual from a postulated accident during DECON is the failure of the ventilation system HEPA filter during chemical decontamination of the high-level wasta tank. Approximately 60 mci of radioactivity are assumed to be released directly to the atmosphere. This release results in a maximum annual dose in the first year of 15 mrem to the lung and a 50 year dose commitment of 160 mrem to the bone of the maximum-exposed individual.

Transportation Safety During DECON Radioactive weste generated during the deconts.mination of an FRP must be packaged and shipped according to prescribed federal regulations to an offsite repository.

These wastes include transuranic (TRU) wastes that are shipped by rail to a Federal repository and non-TRU wastes that are shipped by truck to a commercial shallow-land burial ground. A summary of the wastes generated and shipped is given in Table 9.3-1.

Table 9.3-1 Packagingandshippig)informationforwastes generated from DECON -

Volume, Number of Number of Shipping Method m3 Weight, kg Container: Shipments Rail (TRU wastes) 4,600 3.7 x 108 3,200 180 Truck (non-TRU) wastes) 3,100 2.3 x 106 2,500 160 (a) Initial chemical decontamination wastes account for approximately 5%

of the total volume, 9% of the total shipments, and 99.9% of the total radioactivity 10/07/87 9-8 NUO586 CH 9

I The estimated radiation doses due to external exposure from rail and truck transport of radioactive waste are 20 man-rers to the transportation workers and 9 man-rem to the public.

The release of radioactive material from transportation accidents is estimated to be small. The more probable transportation accidents result in no release or one that is very small. For a severe truck accident, a hypothetical maximum-S exposed individual located 100 meter3 away is estimated to receive a 50 year dose commitment to the bone of 11 rem; however, this type of accidant has a very low probability of occurrence.

9.3.2 SAFSTOR SAFSTOR is defined as those activities required to place (preparation for safe storege) and maintain (safe storage) a FRP in such condition that the risk to safety is within acceptable bounds and that the facility can be safely stored and subsequently decontaminated to levels which permit release of the facility for unrestricted use (deferred decontamination).

Generally, the purpose of SAFSTOR is to permit residual radioactivity levels to decay to levels that will reduce occupational radiation exposure during decon-tamination. As indicated in Table 9.3-2 most of the occupational dose reduction due to decay occurs during the first 100 years aftet hutdown with less dose reduction thereafter. The public dose which is small to begin with, also experiences most of its reduction during the first 100 iears. Hence, in

~

contrast to DECON, to take advantage of this dose reduction, the safe storage period could be as long as 30 to 100. rears. The end result is the same as for DECON: release of the site and any remaining structures for unrestricted use.

SAFSTOR is advantageous in that it results in~ reduced occupational radiation exposure in situations where overriding land use considerations do not exist.

Disadvantages are that the licensee is required to maintain a possession only license and to meet its requirement: at all times during safe storage thes contributing to the number of sites dedicated to radioactive confinement ,or an extended time period. Other disadvantages are that surveillance is required, 10/07/87 9-9 NUO586 CH 9

i Table 9.3-2 Summarv cf radiation safety analysis for decommissioning the reference FRP (Man-rea)

DECON SAF5 TOR (Passive) 5AF5 TOR Custodial) 10 Years ,30 Yeaes 100 Years 200 Years 10 Years 30 Years 100 Years 260 Years Occupational Safety ,

Decontamination Operations 512 426I *) 2MI *) 124I *) s85I *) 423I *) 290I *) 113I *) s73I *I

• Transportation 20 17 12 5 +1 17 12 5 <2 Continuing Care 2 4 9 14 13 31 61 78 Total Occupa.ional Exposure 532 445 312 138 s100 453 333 179 s153 Pubile Safety Decontamination Operations le 8 5 2 <1 8 5 2 <1 Transportation -

9 7 5 2 <1 7 5 2 <1 neg.gg) neg.gc) neg.gg) gg) neg.gc) neg.gc) neo.gc) neg.gc)

Contiwing Care --

neo Total Pubile Exposure 19 15 10 4 <7 15 10 4 <2 he radiation exposures for the preparation for passive and tustodial safe storage are 81 and 69 man-rems, respectively and are included in the expsoures for Decontamination Operations.

) Radiation doses from postulated accidents are not included.

l IC) Meg. = negligible. Radiation doses to the public from normal continuing care activities are not analyzed in detaff, but are expected to be significantly smallei than those from decontamination operations.

t 4.fo

that dollar costs are higher than for DEC0;4, and that experienced operating staff may not be available at the end of the safe storage period to assist in the deferred decontamination.

The several subcategories of SAFSTOR are given in SectiJn 2.3.2. They are discussed in detail here as they pertain to FRP decommissioning.

Preparacion for Safe Storage Custodial SAFSTOR requires a minimum cleanup and decontamination effort during freparation for safe storage, followed by a period of continuing care with'the active protection systems (principally the ventilation system) kept in service throughout the storage period. Safe torage preparation procedures for passive (i.e. , hardened) safe storage are the same as those for custodial safe storage, with the exception of the following additional activities:

sealing all entrances to the radioactive portions of the facility, using welding techniques deactivating the ventilation systems deactivating all cranes and viewing windows Hardened safe storage requires slightly more extensive sealing of the structures than passive safe storage: however, the cost increase is estimated to be small. -

l Thus, passive and hardened SAFSTOR are considered the same for this assessment.

I l

The occupational radiation dost from passive and custodial safe storage prepa-ration, not including transportation, are estimated to be 81 and 69 man-rem, respectively, and are given in Table 9.3-2. The extra labor to prepare for passive storage results in the slightly higher dose.

The estimated inhalation radiation doses to the public from the release of rautonuclides dur?ng both passive and custodial safe storage preparation are estimated tt e 0.006 un-rem (bone dose) to the population. This dose is much below natural background radiation exposure.

NUO586 9-11 10/07/87

The maximum postulated accident for passive and custodial safe storage preparation is a fire in the ventilation system resulting in a maximum annual lung dose in the- first year of 0.006 mrem and a 50 year lung dose commitment of 0.008 mrem.

Estimated routine radiation doses from rail and truck transport of radioactive wastes from either passive or custodial storage preparations are 3 man-rem to transportation workers and 1.4 man-rem to the general public.

Safe Storage (Continuing Care)

Following completion of safe storage preparation, the facility is placed in a period of safe storage (continuing care). This safe storage consists of surveillance and mainte.1ance, detigned to ensure that the facility remains in a condition that poses minimum risks to the public. This phase includes routine inspections, preventive and corrective maintenance on operating equipment, and a regular program of radiation, effluent, and environmental monitocing. The status of all safety-related equipment is monitored throughout the continuing care period. Passive and custodial continuing care doses are listed in Table 9.3-2.

The release of radionuclides from accidents during the continuing care period is negligible. The combination of the low probability of the initiating evants and the immobility of the FRP radionuclide inventory minimizes the effect of poter.tial accidents during this period. -

Deferred Decontamination Deferred decontamination to residual levels permitting unrestricted use of the facility takes place after a number of years of safe storage. This decon-tamination is more thorough than the preliminary decontamination which was a part of the preparations for safe storage. The decontamination procedures are essentially the same following each of the different SAFSTOR modes; however, the steps necessary following passive safe storage are more extensivo. The additional activities include:

N00586 9-12 10/07/87 m .

removal of entrance barriers to contaminated areas reactivation of utilities, cranes, and ma ipulators insta11 scion of filters and reactivation of tae ventilation systems.

The principal advantage of deferred decontamination is that radioactive decay takes place duri.ig the continuing care period. Table 9.3-2 shows that decon-tamination at a deferred time reduces the occupational radiation exposure by a substantial amount. Deferred decontamination would also reduce the radiation dose commitment for public exposure as shown in '.able 9.3-2.

The radiation dose from transportation for deferred decontamination for both public and occupational exposures is expected to decrease because of radionuclide decay and also because of a reduction in materials needing transportation. A 100 year delay would result in a radiation dose reduction of about 75%. These doses are shown in Table 9.3-2.

9.3.3 Site Decommissioning The residual contamination of the FRP site e.,ulting from past operation and subsequent decommissioning is expected to be very low. This is as a result of continuous site surveys and the immediate removal of any contamination found during the life of the facility. Site cleanup is expected to be minimal, how-ever, this will be confirmed by the radiation survey.

9.3.4 Summary of Radiation Safety An advantage of DECON is that it results in the release of the site for unre-stricted use within about 5 years after shutdown of plant operations. However, DECON has higher estimated occupational radiation exposure (5" man-rems) than the other alternatives. Depending on the length of the continuing care period, both passive and custodial SAFSTOR can result in an occupational dose reduction the magnitude of which is considered to be of marginal significance in terms of health and safety (see Table 9.3.2).

l As shown in Table 9.3-2, radiation doses to the public from decommissioning operations and transportation of contaminated materials are all low, with a NUO586 9-13 10/07/87

maximum of 20 man-rem due to DECON. The maximum postulated accifent is esti-mated to give the maximum-exposed member of the public a 50 year dose commit-ment of 8.8 rem.

In summary, the radiation dose to the public is estimated to be quite low and to have little impact compared to natural background radiation. For decom-missioning workers, DECON results in larger radiological impact than the other alternatives. Reductions in this dose can be brought about by use of 30 year or 100 year SAFSTOR.

9.3.6 Decommissioning Costs An estimate of the costs of decommissioning the FRP by each of the principal alternatives is presented below. These costs are summarized and compared in Section 9.3.6.2.

9.3.6.1 Detailed Costs Reference 1 presents a discussion of decommissioning costs and their bases.

Costs are included for 1) direct labor ana subcontractor activities, 2) equip-ment and materials, 3) packaging, transportation, and disposal of contaminated waste, and 4) utilities, services, and other overheads. The details presented in Reference 1 inclede breakdowns for support staff labor, decommissioning worker labor, subcontractor activities, equipment and materials, shipping, waste disposal and utilities and taxes. -

The basic cost estimates presented assume relatively efficient performance of the decommissioning activities. A 25% contingency is added to the cost estimate totals as an allowance for unforeseen problems or scheduling delays that may arise during the decommissioning.

9.3.6.2 Sumury et Costs Table 9.3-4 summacizes the estimated costs in 1986 dollars for the decomais-sioning alternatives. As shown in the table, the costs for SAFSTOR are greater than the cost for DECON. All SAFSTOR modes increase in cost with increasing NUO586 9-14 10/07/87

years of continuing care. The continuing care cost following preparation for custodial and passive safe storage are estimated to be $1.05 million and

$262,200 per year, respectively. Costs for deferred decontamination after custodial and passive safe storage are estimated to be about $130 million.

Deferred decontamination is a comparatively large cost because it requires additional costs to refurbish auxiliary facilities, to reinstitute a trained decommissioning organization, and to provide a new safety analysis and an addi-tional license application. Other costs of deferred decontamination are lower than for DECON due to the decay of much of the radioactivity. As can be seen from Table 9.3-4, continuing care costs become more significant with time.

Waste management costs represent about two-thirds of the totel cost for decontamination of the reference FRP. Waste disposal costs for transuranic wastes, in turn, represent about 85% of the waste management costs. Since waste disposal costs are based on the volume of material placed in the deep geologic repository, reducing waste volumes has a significant effect in

. reducing decommissioning costs. Significant economic incentives exist to develop volume reduction techniques. For example, extensive use of electro-polishing, which has the potential to decontaminate metallic wastes to possibly releasaMe radioactive contamination levels or to levels that permit their dis;osal in shallow-land burial grounds, may offer cost reductions.

Decontamination of the liquid waste storage system represents about one-third of the total decontaminatior, costs. Alternative reprocessing plant designs might not employ large ligtid waste storage systems. These designs would have a significant decommissior,ing cost advantage (40 to 50%) over the design of the reference plant.

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l NUO586 9-15 10/07/87 l . _

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Table 9.3.4 Summary of estimated costs for decommissioning a fuel reprocessing plant (1986 $ millions)

DECOM SAFSTOR (passive) SAFSTOR (custodial)

Item 10 Years 30 Years 100 Years 200 Years 10 Years 30 Years 100 Years 200 Years Initial Decommissioning 168.6 46.6 40.3 46.6 46.6 '44.3 44.3 44.3 44.3 Continuing Care --

1.9 7.1 25.7 51.7 8.0 29.0 102.5 207.5 Deferred Decontamination -- 132.9 132.9 132.9 132.9 131.7 131.7 131.7 131.7 Total Costs (rounded) 169 181 187 205 231 184 205 278 384 1

t 9

It is assumed that radioactive contamination levels on the site from routine releases during facility operation do not require extensive site cleanup operations during decommissioning to meet the limits for release of the FRP for unrestricted use. A preliminary estimate of the costs to perforn, these activities, should they be required, is $10'),000. This would not appreciably change the decommissioning cost totals presented in Table 9.3-4.

9.4 Environmental Consequences The decommissioning of an FRP will have few negative environmental consequences.

By definition, the decommissioning of any nuclear _ facility is the removal of radioactive material to levels which are low enough to permit the facility to be released for unrestricted use. The decommissioning alternativt to be chosen depends to a large extent on the radiation dose and cost evaluations, on desired future use of the site, and on the time period involved.

The summaries of radiation safety and decommissioning cost analyses are given in Sections 9.3.5 and 9.3.6, respectively.

Demolition of remaining buildings (assuming prior decontamination to a level permitting unrestricted use of the FRP) is an optional owner and/or local government choice. Its major environmental impact on the surrounding population will be the resulting increase in noise level within the immediate vicinity of the plant (about 1 mile), primarily because of the use of explosives. However, "

n.st of this noise will be generated within the process building and will be ,

muffled by the building until the final removal of the building shell.

9.4.1 Wastes The management of wastes (i.e. , vitrified, chemical decontamination solutions, contaminated equipment and materials, and contaminated trash) resultinq from decommissioning is an important factor in the cost and environmental impact of decommissioning. The large volumes of waste generated during DECON, as shown in Table 9.4-1, require a large expenditure of money and energy. Complete decon-tamination of an FRP requires about 0.4 hectare (1 acre) of land for final storage of the contaminated materials removed from the site. The high-level 10/07/87 9-17 NUO586 CH 9

Table 9.4-1 Radioactive wastes resulting from decommissioning a reference FRP DECON Passive SAFSTOR(a) Custodial SAFSTOR(a)

Disposal Disposal Dispos 1 Cost, Cost, Cost, Disposition Volume, Millions Volume, Millions Volume, Millions of waste ma of $ m3 of $ m3 of $

TRU-Waste 4,600 86.8 210 20.9 210 20.9 non-TRU wastes 3,100 4.1 180 _0j 180 0.2 Totals 7,700 90.9 390 21.1 390 21.1 (a) Does not include deferred decontamination.

radioactive and TRU wastes will require about 4,600 m3 in an expensive deep geologic disposal facility. This is equivalent to 163,500 cubic feet mined from either salt or basalt. The low-level and non-TRU wastes will require about 0.16 hectares (0.4 acre) of shallow-land burial area. These are considered irre-trievable uses of land.

The volumes of waste for both passive and custodial safe storage represent the preparation state only. Deferred decontamination wastes increase each of these to values nearly that of DECON. However, although the overall waste volume may remain nearly constant, the amount sent to geologic storage will decrease with time, while shallow-land burial volumes will increase. Fcr example, if the continuing care period were extended for 100 years, there would be a reduction .

in. radioactivity and thus the total amount of waste to be disposed of to repositories would shift from deep geologic storage to shallow-land burial.

These changes could result in a substantial change of costs and repository use.

The deco min ioning of an FRP to levels which permit unrestricted use of the facilit; axes about 473 hectares (1,160 acres) of land available for reuse.

The value recovered from decommissioning depends on the value of the reclair2d land and the need the owner has for such property during the time period under consideration.

10/07/87 9-18 NUO586 CH 9

~

J If the plant site of about 20.4 hectares (50 acres) is restored to its original native condition, it will increase the natural habitat for flora and fauna by a relatively small amount. This is a favortble environmental impact, but one that is relatively insignificant.

An additional effect of decommissioning is that the decontamination of an FRP will require the use of expendable tools and materials that will be discarded as waste.

9.4.2 Nonradiological Safety Impacts The nonradiological hazards involved in the decommissioning of an FRP were reviewed on the basis of hazards to be found in both the chemical and construc-tion industries. These estimates are calculated to be conservative.

Potential chemical pollutants that could be released during the various decom-missioning alternatives were examined and found to be insignificant. The small quantities of hazardous chemicals used and the low likelihood of their dispersal into the environs indicate that potential chemical pollutants from decommis-sioning operations do not pose a significant public hazard.

The potei.tial lost-time injuries and fatalities are based on AEC/D0E operations data. T&ble 9.4-2, gives the lost-time injuries and the fatalities estimated for each decommissioning mode. The maximum potential for lost-time injuries and fatalities (1.9 and 0.01, respectively) is during the decontamination opera '

tions when the maximum amount of heavy equipment is being removed from its position, cut, boxed, and shipped to appropriate storage.

9.4.3 Socioeconomic Impacts The major societal impacts occur prior to decommissioning with the shutdown of the plant. The shutdown of the plant and DECON will reduce the work force from about 300 to 50 people over about a 2 year period and the 50 person decommis-sioning force will be reduced to near zero in 3 to 6 years. Thus, the total reduction in force will take place over a minimum period of 5 years and this should tend to mitigate the adverse impact of loss of jobs and income to the 10/07/87 9-19 HUO586 CH 9

4 Table 9.4.2 Summary of nonradiological safety impacts Source of SAFSTOR (Passive) SAFSTOR (Custodial)

Type of Safety Concern Safety Concern Units DECON 10 Years 30 Years 100 Years 200 Years 10 Years 30 Years 100 Years 200 Years ENT006 SeriousLogTime Decommissioning 1.7 1.9 1.9 1.9 1.9 1.75 1.75 1.75 1.75 0.85~

Injuries Operations no./ mode 0.17 0.10 Transportation no./ mode 0.17 0.17 0.17 0.17

• 0.17

- 0.17 0.17 0.17 Continuing Care no./ mode --

0.083 0.26 0.83 1.6 0.40 1.2 4.0 8.0 --

FrtalitiesI *)

Decommissioning 0.0096 9.005 Operations no./ mode 0.0091 0.010 0.010 0.010 0.010 0.00 % 0.00 % 0.0096 0.012 0.012 0.012 0.012 ~ 0.012 0.012 0.012 0.012 0.007 Transportation no./ mode 0.012 0.0008 0.0024 -0.0081 0.016 0.038 0.012 0.038 0.076 --

Continuing Care no./.aode --

(c) Estimates of lost-time accidents and fatalities for either passive or custodial safe storage preparation are 0.3 and 0.003, respectively. The transportation estimates of lost-time accidents and fatalities for either passive or custodial safe storage preparation are 0.03 and 0.002, respectively.

1 i

regional comunity. Since the planning stage preceding the shutdown will require about two years, the community will have an additional two years to plan for the reduction in jobs. Therefore, the impact from job loss (income loss of about $4 to $5 million annually) due to plant shutdown will be small because of-the period of time for the action to take place. Decomissioning tends to mitigate the impacts due to plant shutdown.

Tax revenues will also be lost to the local comunities and to the state, but here again, the impact is spread over a period of time and as employment reduces and_ people leave the area, public services will also reduce. Thus, decom-missioning tends to mitigate the impacts of plant shutdown.

9.5 Comparison of Decomissioning Alternatives Primary parameters that affect the selection of a decomissioning alternative are the radiation doses and the economic costs. These are sumarized in Tables 9.3-2 and 9.3-4.

Advantages of DECON are that the site and facility can be released for unre-stricted use 5 years after the shutdown of the plant and that the cost for DECON is less than for SAFSTOR, and therefore, DECON is considered to be a preferable alternative since occupational dose reduction by SAFSTOR is of an amount considered of marginal significance to health and safety. Both 30 year SAFSTOR and 100-year SAFSTOR may be reasonable options for reducing occupational exposure since additional radioactive decay occurs af ter 30 years. In 100 year ~

SAFSTOR, the occupational dose rates have decayed to about 30% of DECON and the costs, although increased by 20% over the 100 year period are still reasonable when evaluated against the reduced occupational dose.

9-21 NUO586 CH 9 10/07/87

2-a:

REFERENCES

1. K. J. Schneider and C. E. Jenkins, Technology, Safety, and Costs of Decommissioning a Reference Nuclear Fuel Reprocessing Plant, NUREG-0278, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, October 1977.

2. G J. Konzek arid C. R. Sample, De' commissioning of Nuclear Facilities--An Annotated Bibliography, NUREG/CR-0131, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, October 1978.

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t.

10 SMALL MIXED OXIDE FUEL FABRICATION PLANT A small mixed oxide (M0X) fuel fabrication plant is a manufacturing facility designed and constructed for the production of (U-Pu)0 2 pellets and incorporation of these pellets into clad fuel rods. The plant also has facilities for the recovery of plutonium from unirradiated scrap materials. This section considers the environmental consequences of decommissioning a small MOX plant.

This section is based primarily on a detailed study! of the decommissioning of a small mixed oxide fuel fabrication plant. In this study PNL selected the Cimarron Plutonium Facility located near Crescent, Oklahoma as the reference M0X plant and assumed it to be located at the generic site. The generic site is described in Section 3.1. Although not currently operating, Cimarron is considered to have characteristics related to many of the existing small M0X plants. Some operational features were added to this study to make it appli-cable to plants using other processes. PNL then developed and reported infor-nation on the available technology, safety considerations, and probable costs for decommissioning the reference facility at the end of its operating life.

10.1 Description of the Reference MOX Fuel Fabrication Plant The reference plant is assumed to have operated for 10 years at a production rate of 2 MT of heavy metals per year. The feed to the plant can be either the oxide powders or nitrate solutions of plutonium and uranium. The plant operation is l

l assumed to involve either mechanical blending of the oxide powders or coprecipi-l tation of the solutions, using ammonia. The plant consists of a single building l with a floor space of 2400 m2 that also contains offices, laboratories, and maintenance shops. Auxiliary facilities are a cooling tower, an electrical substation, ef fluent storage, and a gas supply. Processes include solvent i extraction, ion exchange, and oxalate precipitation for recovery of dirty scrap, and a two-stage liquid waste evaporation system followed by concreting of liquid wastes. The plant uses small, criticality safe vessels located in numerous glove 10/07/87 10-1 NUO586 CH 10

boxes distributed throughout nine rooms. Operation of most steps is on a batch basis.

The generic site (Section 3.1) for this plant is located in a rural area. The site occupies 470 hectares (1,160 acres) in a rectangular shape of 2 km (1.24 miles) by 2.35 km (1.46 miles). A moderate-size river runs through one corner of the site. The use of any part of this site for anything besides the MOX plant is prohibited. The plant is in a restricted area of about 1.2 hectares (3 acres) within the site.

As a part of the plant operations, it is assumed that a final inventory cleanout has been performed that included disposal of process materials, chemicals, trash, scrap, scrap solutions, and contaminated solutions. Empty product, scrap, and waste handling tanks have been flushed of remaining process solutions. The dominant remaining radionuclides that will contribute to organ doses are 23sPu, as9Pu, 240Pu, 242Pu, and 241Am. About 23 kg of plutonium are estimated to remain in the process building following the final inventory cleanout.

10.2 MOX Decommissioning Experience No direct experience exists in the decommissioning of licensed MOX fuel fabri-cation facilities because existing plants, which are not now operating, are being held in a standby or storage status. However, several government-owned plutonium fabrication facilities have been decontaminated. In all cases, the buildings still stand and contain radioactive contamination above unrestricted "

levels. Some are closed and sealed but others have been converted to new, related facilities involving the use of radioactive materials.

A list of these facilities, and a detailed discussion of decommissioning steps taken at two of them appear in Reference 1. This report also contains a discussion of lessons learned from decommissioning experiences.

i 10.3 Decommissioning Alternatives Once a M0X plant has reached the end of its useful operating life it must be decommissioned. As discussed in Section 2.3, this means safely removing the 10/07/87 10-2 NUO586 CH 10

facility from service and disposing all radioactive materials in excess of levels which would permit unrestricted use of the facility. Several alterna-tives are considered here as to their potential for satisfying this general requirement for decommissioning. The decommissioning alternatives considered and discussed here are DECON, SAFSTOR (custodial), and ENTOMB. Radiological effects and costs of each alternative are also discussed. After the radio-active inventory has been removed down to levels permitting unrestricted use of the facility and the contaminated equipment and structures decontaminated, demolition of the building would be left as an owner option.

The alternative used depends on such considerations as dose, cost, proposed use of the site, and desirability of terminating the license. A special consid-eration for decommissioning MOX plants is the half-lives of the radionuclides present in the facility. The radionuclides processed in a H0X plant are received from a reprocessing plant. Those radionuclides include plutonium and uranium and their de.uy products, but not fission products. There are several isotopes of these actinides, and the radioactivity of these isotopes is very high, par-ticularly that of the plutonium. These isotopes have such long half-lives that it is apparent that deferred decontamination for 10 or even 100 years would not result in reduced radiation doses to decommissioning personnel and, therefore, SAFSTOR would not appear to be a reasonable alternative without some other justification.

Safeguards will be required during each decommissioning alternative for protection of the public. Security is assumed to be similar to that needed during plant operation but on a smaller scale.

10.3.1 DECON DECON is defined as the immediate removal and disposal of all radioactivity in excess of levels which would permit release of the facility for unrestricted use. Nonradioactive equipment and structures need not be torn down or removed as part of a DECON procedure. The end result is the release of the site and any remaining structures for unrestricted use as early as the 5 years estimated for decommissioning af ter the end of f acility operation.

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l DECOW is advantageous because it allows for termination of the NRC license I with elatively few years after cessation of facility operations and removes a radioactive site. DECON is advantageous if the site is required for other purposes, if the site is extremely valuable, or, if for some reason the site must be immediately released for unrestricted use. It is also advantageous in that the facility operating staff is available to assist with decommissioning and that continued surveillance is not required.

The first step toward DECON is planning and preparation, which is initiated during the last 2 years of normal plant operation. During this time, detailed plans and procedures are prepared, a decommissioning staff is trained, safety and environmental impact reports are prepared if necessary, and effluent control systems modifications are started.

When the actual decommissioning work begins following shutdown, chemical decontamination of the wet process areas and physical cleanout of the dry process areas are started first. Physical decontamination of most plant areas proceeds next. Chemical decontamination involves flushing of internal surfaces of process piping and equipment, followed by spraying with chemical solutions the external surfaces of process equipment, piping, and internal surfaces of glove boxes.

Physical decontamination involves disassembly of equipment and enclosures and removal of the resulting materials. Physical decontamination also involves .

removal of contaminated parts of structural materials. These are packaged and transported offsite as waste, either as is or after chemical decontamination l to remove bulk quantities of radionuclides. For DECON, disassembly and removal l

l of equipment in some of the cleaner areas starts about 2 months after shutdown, l and proceeds in parallel with chemical decontamination of other areas. The facility and service systems are removed as the last steps. At this point, 1

the facility should be at or below acceptable levels of residual radioactivity and could be considered to be decommissioned. However, it may be desirable for nonradioactive reasons to remove the buildings, in which case the final phase would be demolition and restoration.

in /n i/n? in.a wnntnA rw in

I If demolition and restoration were used, all above grade portions of structures could be demolished using conventional methods such as explosives and impact balls. The site could then be graded and planted with vegetation to near pre-facility-conditions.

Analyses of radiation exposures and costs for DECON are presented in Section 10.3.4.

10.3.2 SAFSTOR SAFSTOR is defined as those activities required to place (preparation for safe storage) and maintain (safe storage) a MOX plant in such condition that the risk to safety is within acceptable bounds, and that the facility can be safely stored and subsequently decontaminated to levels which permit release of the facility for unrestricted use (deferred decontamination).

Generally, the primary purpose of SAFSTOR for most nuclear facilities is that it

. results in reduced occupational exposure compared to DECON. However, for the reasons given in Section 8.3 and as can be seen in Table 10.3-1 this is not necessarily the case for MOX plants. SAFSTOR could be advantageous in situations where there are overriding land use considerations. However, in addition to increased radiation exposure other disadvantages are that the licensee is required to maintain a material license and to meet its requirements at all times during safe storage thus contributing to the number of sites dedicated to .

radioactive confinement for an extended time period. Other disadvantages are that surveillance is required, the dollar costs are higher than for DECON, and the experienced operating staff may not be available at the end of the safe storage period to assist in the deferred decontamination.

Chemical and physical decontamination activities in preparation for custodial safe storage are similar to those performed for DECON, except that for custodial safe storage, initial decontamination is generally done to the point that loose radioactivity is removed.

Preparations for the continuing care period of custodial safe storage involve deactivation and isolation of contaminated areas, sealing of contamination by 10/07/87 10-5 NUO586 CH 10

adding dursble seals or covering with paint, refurbishing the plant ventilation system, and inst 611ing improved alarm and protection systems for fire, intrusion, or malfunctioning equipment.

Continuing care activities may include operation of the facility ventilation system, routine inspection, corrective and preventive maintenance of the ventilation and other safety systems, environmental surveillance. and prevention of unneeded intrusion by man.

For the M0X facility, custodial safe storage is terminated eventually by deferred decontamination to levels permitting unrestricted use of the facility. For this action, activities are generally similar to those for DECON, with allowances for the prior decontamination efforts and retraining of new decommissioning staff.

Analyses of radiation exposures and costs for SAFSTOR are provided in Section 10.3.4.

10.3.3 ENTOMB The ENTOMB alternative requires use of a structure to hold or confine the radio-activity until such time as it has decayed to levels which permit release of the facility for unrestricted use. ENTOMB would involve the encasement in concrete of heavily contaminated rooms within the reference MOX facility which would prevent the escape of radioactivity and prevent deliberate or inadvertent intru-f ~

l sion. The length of time the integrity of the entombing structure must be maintained depends on the inventory of radionuclides present.

The MOX plant will still contain the 23 kg of plutonium estimated to remain in the process building following final inventory cleanout at shutdown, (see Sec-tion 10.1) including 23SPu with a half-life of 24,390 years, and the entombed l

structure would in effect become a new surface high-level waste disposal site.

l This would be an undesirable situation in that it would be contributing to the l problems associated with increased numbers of waste disposal sites. Moreover, I the entombed structure would require surveillance in perpetuity which is well beyond the time that the required institutional control could be expected to be l

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effective (approximately 100 years is considered to be consistent with recom-sended EPA policy on reliance on institutional control of radioactivity confinement). Although the ENTOMB option does not appear viable for the reasons given, it will be discussed for comparative perspective with the other eptions.

10.3.4 Summary of Radiation Safety and Decommissioning Costs Each of the decommissioning alternatives has associated with it unavoidable radiation exposures, accident potential, and costs. As is seen from Table 10.3-1 none of these is appreciably reduced with time. This conclusion might change if technologies improve the reduction of accidental releases of radioactivity or the cost-efficiency of decontaminating the equipnent.

10.3.4.1 Radiation Safety Radiation safety for H0X plant decommissioning is discussed in detail in Reference 1. Oose calculations were based on maximum releases of radioactivity to maximize the consequences and thus present worst-case evaluations, Occupational radiation exposure of workers performing the decorissioning activities results from external exposure to surface contamination for reasons similar to that discussed for PWRs in Section 4.3.1. Dose calculations are based on the estimated radiation levels in various areas of the plant and the estimated labor requirements for decommissioning each of those areas. Many of the radionuclides remaining in a M0X plant af ter shutdown have long half-lives. '

l Generally, preparation for safe storage does not involve extensive decontami-nation of these radionuclides. Because the half-lives of these radionuclides are long compared to the time that the facility might be held in safe storage awaiting deferred decontamination, the occupational radiation exposures will not decrease as a result of using the SAFSTOR alternative. There will be a shift in nuclide content from 242Pu to 241Am while a plant is in continuing care, but this shift will be insignificant. In calculating the total doses received, there are additional exposures incurred under the custodial safe storage mode that must be considered. These are shown in Table 10.3-1, which is a summary of the radiation exposures that may result from each of the 10/07/87 10-7 NUO586 CH 10 l

Table 10.3-1 Summary of Radiation Safety Analyses for Routine Decommissioning of the Reference MOX Plant man-rem)(a)

SAFSTOR Occupational Exposure DECON 10 Years 30 Years ENTOMB Preparation NA 23 23 9.4 Continuing Care NA 64 206 neg De:ontamination(b) 70 70 70 NA Transportation 6.4 8 8 0.6 Totals 76- 165 307 10 Public Exposun (50 year dose commitment to critical organ)

Preparation NA 0.1 0.1 0.10 Continuing Care (b) NA 0.05 0.1 neg.

Decontamination 2.2 2. 2 2.2 NA Transportation 1.5 1.9 1.9 0.15 D DE

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Totals 4.3 43 (a) Adapted from Reference 1.

(b)For SAFSTOR, this is deferred decontamination.

3 decommissioning alternatives. It is to be noted again that the reference M0X plant for which the calculations were made is a small H0X plant.

The dominant radiation exposure pathway to members of the public during decom-missioning operations is inhalation of airborne radionuclides for reasons similar to those discussed for PWRs in Section 4.3.1. Emissions may result ,

from either routine decommissioning activities or from potential accidental releases. Total estimated public exposures during routine decommissioning activities are small, as shown in Table 10.3-1.

A wide range of possible accidents that would result in released radioactivity l is postulated. The largest releases are from failure of HEPA filters, cutting l of contaminated metal, and explosion and/or fire in the ion exchange resins.

l These would result in the same quantities of release and radiation doses and have the same probabilities of occurrence with either decontamination alternative.

A summary of the estimated doses to the public from accidents is shown in Table 10.3-2. The major postulated accident is the release of contaminated 10/07/87 10-8 HUO586 CH 10 l

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Table 10.3.2 Summary of radiation doses to the maximum-exposed individual from accidental airborne radionuclide releases during decossiissioning activities (a)

Fifty-Year Dose First-Year Dose, arem Commitment, erem Release to Expected Atmosphere frequency g Incident (pCi) Bone Lung Bone tung Occurrence Loss of Intermediate-Stage HEPA Filter After Exhaust Duct Decontamination 1.0 x 104 5.2 32 1.1 x 102 78 High Inadvertent Cutting of Undecontaminated Metal 1.6 x 102 8.5 x 10 3 5.0 1.8 1.3 High Explosion and/or Fire of Ion Exchange Resin 83 7.0 x 10 2 6.6 x 10 2 2.5 7.0 x 10 2 Medium l

l l Inadvertent Dumping of Contaminated Solid Wastes:

Abraded Firebrick 14 7.4 x 10 4 4.4 x 10 2 1.5 1.1 High Concrete Dust 1.4 7.4 x 10 5 4.4 x 10 3 1.5 x 10 2 1.1 x 10 2 High Condensed Metal Vapor 7.0 x 10 2 3.8 x 10 e 2.2 x 10 4 7.9 x 10 4 5.7 x 10 4 High Loss of Local Airborne Contaminaticn Control / Loss of Vacuum Filter 3.5 1.9 x 10 4 1.1 x 10 2 3.8 x 10 2 2.8 x 10 2 High Temporary Loss of Services:

Electricity (Normal and Emergency) 1.4 7.4 x 10 5 4.4 x 10 3 1.5 x 10 2 1.1 x 10 2 Medium Liquid Leak:

Chemical Decontamination 16 1.4 x 10 2 4.8 1.3 x 10 2 1.4 x 10 2 High Electropolishing 2.8 x 10 2 5.4 x 10 8 5.1 x 10 8 1.9 x 10 4 5.1 x 10 8 Medium Fire Involving Contaminated Clothing or Cumbustible Waste 0.11 9.6 x 10 5 9.2 x 10 5 3.4 x 10 3 9.2 x 10 5 Medium m . .

Table 10.3-2 (Continued)

Fifty-Year Dose First-Year Dose, mrem Commi % nt, aren Release to Expected Atmosphere Frequency g (pCi) Bone Lung Sone Lung Occurrence Incident Explosion of Hydrogen During Electropolishing 7.1 x 10 8 5.9 x 10 8 5.5 x 10 8 2.1 x 10 4 5.9 x 10 6 High Man Intrusion (c) 3.5 x 108 2.1 x 10 8 7.0 x 108 5.2 x 108 Low (8)This table is a summary of Table 11.2-3 in reference 1. It presents the highest dose from each of the decom-missioning alternatives.

(b) Frequency of Occurrences: High >1.0 x 10 2 to 1.0 5; Low <1.0 x 10 5 per year.

(C This accident is for the ENTOMB alternative only and is postulated to be a deliberate but ignorant intrusion by man into the facility after knowledge of the facility is lost after a period of several hundred years. The case postu-lated assumes a 40-hour exposure to an average air concentration of 290 pCi/m3 of mixed oxides containing plutonium.

e

I dust from an exhac,t duct by failure of a HEPA filter. Radiation doses to the public resulting from accidents are low enough to be insignificant. Even with the failure of a HEPA filter which, as stated above, would result in a major accidental release, the public would be partially protected by the other filters in the system.

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Radioactive waste materials are packaged and shipped offsite for burial during decommissioning of the reference MOX facility. These wastes include transuranic (TRU) contaminated wastes (a) that are shipped to a federal repository (deep geologic disposal) assumed to be located at 2,400 km (1,500 mi) from the plant site, and non-TRU wastes (b) that are shipped to a commercial shallow-land burial facility located about 800 km (500 mi) from the site. All wastes are assumed to be shipped by truck. To minimize the risk that radioactive shipments pose to the public and to transportation workers, federal and state regulations prescribe the containers, contents, packaging and handling, and burial requirements.

The cominant radiation exposure pathway to transport workers and the public during transportation of radioactive wastes is external exposure for reasons similar to those discussed for PWRs in Section 4.3 1. The external dose for routine transportation operations for all truck shipments, both high and low-level wastes, from DECON is conservatively estimated to be about 6.4 man-rem to transport workers and 1.5 man-rem to the general public. For SAFSTOR (custodial) the radiation dose is estimated to be 8 man-rem to handling and transportation workers and 1.9 man-rem to the public. These doses are based on regulations of the Department of Transportation governing radiation levels in shipments of radioactive materials and on estimates the distances of travel and .

lengths of time of exposure that workers and the public might expect These doses are summarized in Table 10.3-1.

The severity of accidents that may occur during transportation of radioactive waste dopends on a number of factors, such as speed, kind of accident, and accident locations. Regardless of the decommissioning alternative, the same total amount of radioactive material will be transported. Thus, the possible

(*)TRU wastes are assumed to be those contaminated with alpha radioactivity from transuranic materials at a level of 10 or more nCi/g of waste.

(b)Non-TRU wastes are assumed to have transuranic alpha radioactivity of less than 10 nCi/g of waste.

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release of radioactivity willibe dependent on frequency arl kind of accidents, as shown in Table 10.3-3.

10.3.4.2 Decommissioning Costs This discussion of the decommissioning costs is based on information in NUREG/CR-0129.2 Table 10.3-4 summarizes the estimated costs in 1986 dollars for the decommissioning alternatives analyzed in this report. All cost estimates include an added 25% for contingencies.

For DECON, the decommissioning costs are estimated to be $13.9 million. For custodial SAFSTOR the total decommissioning cost is estimated to be $27.6 million and $47.3 million for 10 year SAFSTOR and 30 year SAFSTOR, respectively. These SAFSTOR costs include $5.8 million for preparation for safe storage, $0.98 mil-lion per year for continuing care, and $13.0 million for costs of deferred ,

decontamination. A present value analysis of decommissionirig costs indicates a disincentive to defer decontamination for the reference case indicated, pri-marily because of the high cost of continuing care relative to DECON costs and the high cost of deferred decontamination due to the long half-lives of the radionuclides involved. For ENTOMB, the decommissioning costs are estimated to be $4.9 million.

Labor costs are about 60% of the total costs for the DECON and SAFSTOR alter-natives and about 50% for ENTOMB. Thus, there is considerable incentive to institute plans or techniques that could reduce labor, such as working around the clock for the total decommissioning activities to reduce support labor and license and miscellaneous costs. The deferral of decontamination requires additional costs to refurbish auxiliary facilities, to reinstitute a trained decommissioning organization, and to provide a new safety analysis and appli-cation for amended license.

Costs of management of the wastes from decontamination range from about 7% to about 20% of the total costs of decommissioning, depending on the decommission-ing alternative. Thus, there is a modest economic incentive to reduce these costs. A potentially major economic factor favoring DECON is the value of the land or facility when released for productive uses. A facility in safe storage 10/07/87 10-12 NUO586 CH 10

Table 10.3.3 Estimated frequencies, radioactivity releases and doses for selected truck transport accidents Frequency of Accident per Facility Radiation Dose for Maximum Exposed Individual (rem) 50 Year Oose Severity of Accident Release of 1st Year Commitment (in Closed Van) GECON SAFSTOR Radioactivity (b}Ci Bone Lung Bone Lung Minor 7.4 x 10 2 9.9 x 10 2 No Release - - - -

Moderate 1.8 x 10 2 2.4 x 10 2 1 x 10

• 6.8 x 10 3 2.6 x 10 2 2.4 x 10 1 6.5 x 10 2 Severe 4.7 x 10 4 6.3 x 10 4 1 x 10 2 6.8 x 10 1 2.6 2.4 6.5

(*) Table adapted from NUREG/CR-0129, Table 11.4.3.

( ) Assumes a shipping inventory of 100 Ci of dispersable radioactive material.

Table 10.3-4 Summary of estimated costs for decommissioning the reference small MOX fuel fabrication plant Estimated Costs in Million of 1986 Dollars SAFSTOR (Custodial)

Item DECON 10 years 30 years ENTOMB Initial Decommissioning (*) 13.9 5.8 5.8 4.3 Continuing Care NA 8.8 28.5 NA Deferred Decontamination (a) NA 13.0 13.0 NA Onsite Burial NA NA NA 0.6 Total Costs (Rounded) 13.9 27.6 a7.3 4.9 (a)Costsarebasedontenshifts/weekformostofthedecommissioning.

Decommissioning on a 24-hour / day basis would reduce costs and time requirements.

will provide economic return only as a tax write-off during the years before deferred decontamination, while a facility and land that have unrestricted use can be put to productive uses.

With the exceptions of the possible use of the process building and economic considerations, there is little or no advantage to either decommissioning alternative over the other regarding short-tern and long-term uses. Once the facility has been prepared for custodial safe storage, tho only area of concern for exposure to radionuclides is inside the exclusion area and, depending on the perceived potential accident risks, the rest of the property may be released for unrestricted use. In the reference facility and site, the building is sited in an exclusion area of 1.2 hectares (3 acres). This exclusion area represents about 0.25% of the total site area of 470 hectares (1160 acres).

However, in view of the fact that SAFSTOR of fers no advantages from reduced radioactivity (in fact, a small increase in potential hazard from a buildup of 2"Am), it appears that DECON would b6 the more acceptable of these two decomissioning alternatives f ar MOX plants.

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10.4 Environmental Censequency The decommissioning of a M0X plant has few r + '4ve environmental consequences.

As was defined in Section 2.3, the decommissir,- , c.' ary nuclear facility insolves the removal of radioactive material to It <1 (hich permit release of the facility for unrestricted use. The decornmiss' g alternative to be chosen depends to a large extent en the radiation dose, cost evaluations, desired future use of the site, desirability of terminating the license and the time period.

The summaries of radiation safety and decommissioning costs are given in Section 10.3.4.

Demolition of remaining buildings (assuming prior decontamination to a level permitting unrestricted use of the F'JX plant) is an optional owner and/or local government choice. Its major environmenta ' impact on the surrounding population

-will be the resulting increase in noise level within the immediate vicinity of the plant (about 1 mile), primarily because of the use of explosives. Ilowever, most of the noise will be generated within the process building and will be muffled by the building until the final removal of the building shell.

10.4.1 Waste A major environmental consequence of decommissioning is the commitment of land area to the disposal of radioactive waste. PNL made t.,e estimates shown in Table 10.4-1 of the waste disposal volume required to accommodate radioactive '

waste and rubble removed from the facility and transported to a licensed site for disposal. The volume for ENT0MB does not include tte volume of the entomb-l ing structure or the wastes entombed within it. The entombing structure is effectively a new shallow high level radioactive waste burial ground, separate j and distinct from the ones in which the wastes in Table 10.4-1 are buried.

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Table 10.4-1 Burial volume of radioactive waste and <

rubble resulting from decommissioning l a reference M0X plant (m3)

SAFSTOR (Custodial) ENTOM8 Disposition of Waste DECON 10 Years 30 Years Deep Geologic Disposal 164 205(a) 205(a) 21 Shallow Land Burial 267 267 267 5 Total 431 472 472 26(D)

(a) Includes 52 m3 of waste from preparation for safe storage.

(b)Does not include vo Nme of entombing structure or entombed waste.

If shallow land burial of radioactive waste in standard trenches is assumed, then a burial volume of 267 m3 of radioactive waste can be accommodated in less than 0.02 acres. An additional 164 m3 would be required in a high-level waste repository for DECON. An additional 52 m3 of high-level waste disposal space would be required for SAFSTOR.

These land use req'Jirements for waste disposal are not large in comparison with the approximately 1160 acres used as the site of the reference M0X plant which could now be returned to unrestricted use ,

An additional effect of decommissioning is that the decontamination of a M0X plant will require the use of expendt.ble tools and material that will be dis- ,

carded as low-level waste.

10.4.2 Nonradiological Safety Two potential nonradiological safety considerations are recognized. These are releases of chemicals used to decontaminate the plant and accidents in trans-porting meterials to and from the plant.

Chemicals used in decontamination are detergents, oxidizing agents (acids),

reducing agents, chelating agents, acids, caustics, and electropolishing solu-tions. Fumes from these chemicals will not be a safety hazard to workers pro-vided there are adequate precautions and ventilation. Possibly the greatest 10/07/87 10-16 NUO586 CH 10

i I

pctential for gaseous emissions is frorg the electropolishing process. Hydrogen an:: oxygen will be evolved in amounts that are proportional to the applied cur-rest and the surface area. For example, if a current of 10,000 A is applied te an area of 6 m2 at an electropolishing station, hydrogen gas will be evolved at the rate of 4.5 m3 per minute and oxygen at half that rate, for a total of 6.8 m2 per minute. At this rate of release, these gases will entrain 10 mg of liquid electrolyte per m3 of gas. The air filtering system operating for the removal of radionuclides will also remove this entrained liquid. Adequate ven-tilation will keep a fire or explosion from developing by preventing the hydrogen concentration in the air from building up to exceed the lower flammability level of 4.1%. This consideration will be very important when electropolishing a closed container such as a tank.1 Shipment of materials in and out of the plant will inherently have the same risk of accidents as any other shipping activities. Since transport is assumed to be by truck, the probability of accidents can be estimated from highway travel statistics. Assuming 630 round trips of 1600 km (1000 miles) to a shallow land burial site and 32 round trips of 4800 km (3000 miles) to a deep geologic burial site, there may be expected about 0.61 injuries and 0.036 fatalities per facility.1 10.4.3 Socioeconomic Effects An immediately felt >.on-decommissioning effect of closing a M0X plant will be .

the loss of employnent. A plant that has act been operating (as is tne case with some of the existing plants) will rec. 1 that a number nf people be hired and trained, thus providing short-term employment (1 to 5 years). If decom-sissioning follows immediately af ter shutdown, some of the operating personnel will be used in the decommissioning work, thus providing a reduced level of employment for a short time. In the case of DECON, the staff size will remain at about a constant level until the decontamination activities near completion l

j nearly 3 years after shutdown. On the case of custodial SAFSTOR, the staff

! will decrease as soon as initial chemical decontamination is completed. TVough-out the period of continuing care, only maintenance, monitoring, and security personnel will be required. At the end of the continuing care period, the staff size will again increase to accomplish the final decontamination. Unless l

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decontamination is performed by a contractor with a trained staff, a decontami-nation crew will have to be recruited and trained before this work begins.

Changes in employment levels will not occur suddenly aut will happen over the decommissioning a:.riod regardless of the decommissioning alternative. The custodial SAFSTOR alternative will require a small staff throughout the con-tinuing care period, but this will be a small part of any local economy.

One possible benefit to the community will result from the removal of restric-tions on the use of the land, which may happen if the facility is not used for other nuclear activities.

10.4.4 Noise and Aesthetics One environmentas effect will result from noise. Noise levels during decontami-nation will increase over operation levels because of the physical removal of concrete surfaces. Because these activities will be inside the buildings and because the buildings are some distance from the site boundary, these noises will not likely be heard offsite.

Aesthetic effects will not likely be a result of the deconaissioning process per se, but will rather depend on the final disposition of the building and site. Removal of the M0X building will allow the site to be returned to its preconstruction state or be used for any other purpose. A building that is being held in continuing care may not require limitations on the use of the remainder of the site. The ENTOMB alternative will result in a large mound of '

earth whose blending into its surroundings will depend largely on the local terrain. This mound could be nuite conspicuous in a flat area. In addition, the earthen fill must be taken from some borrow area and careful planning will be required to prevent this from creating another set of aesthetic problems.

Thus, the aesthetic impact of ENTOMB is potentially greater than that for one of the other decontamination alternatives.

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10.5 Comparison of Decommissioning Alternatives The decommissioning alternatives as discussed here apply to a small M0X plant.

Economics and radiation exposures may change somewhat for a facility with dif ferent characteristics.

The alternatives considered viable are DECON and custodial SAFSTOR. The dif ferences between these alternatives are very small in matters of environment, ecology, and aesthetics. The major differences occur in occupational radiation exposure and decommissioning costs. Due to the long-lived nature of the radio-nuclides present in the M0X plant, doses and cos'.s are not reduced even when decontamination is deferred for 30 years, as can be seen from Tables 10.3-1 and 10.3-4. Since the cost and doses of continuing care are major items and con-tinue to-increase with increasing safe storage time, the doses and costs asso-ciated with the complete SAFSTOR process exceed those for DECON. Thus, DECON would seem to be the most advantageous alternative.

Over the short-term, ENTOMB appears to offer some economic advantage in that initial costs are lower than for other alternatives. This advantage disappears, however, over the long-term because of the naed to maintain surveillance of the site in perpetuity. Major societal concerns of this alternative include the problems associated with increased numbers of nuclear waste sites, holding long-lived hazardous materials near man's environment, and maintaining financial responsibility. All of these concerns combine to make ENT0MB an unacceptable .

alternative.

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REFERENCE 1 .C. E. Jenkins, E. S. Murphy and K. J. Schneider, Technology, Safety, and Costs of Decsmissioning a Reference Small Mixed Oxide Fuel Fabrication Plant, NUREG/CR-0129, Vol. 1 and 2. Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Comission. February 1979.

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11 URANIUM HEXAFLUORIDE CONVERSION PLANT l The function of a uranium hexafluoride (UFe) conversion plant is to convert uranium concentrates, received from various uranium mills, to the purifiea ura-nium hexafluoride that is used as the feed material for the gaseous diffusion enrichment of 23sU. Currently there cre five conversion plants in operation in the United States. Their names and locations are:

Allied Chemical Metropolis, Illinois Kerr-McGee Sequoyah County, Ohlahome Fernald DOE Cincinnati, Ohio Paducah, 00E Paducah, Kentucky Portsmouth,00E(a) Portsmouth, Ohio Three other plants have been shut down: the Mallinckrodt Chemical Company Plant at Welden Springs, Missouri, the NUMEC Plant at Apollo, Pennsylvania, and the Oak Ridge Enrichment Plant.

The plant described here is a reference plant that is assumed to have processed 10,000 metric tons (MT) per year of natural uranium and to have been in opera-tion for about 30 years. A detailed report on the decommissioning of a UFs plant, similar to those prepared for other facilities discussed in this EIS -

was issued in October 1981 (Ref. 1). The reference plant discussed here is based on the latest technology. For t:1e plants listed above, currently oper-ating plant processes vary from the reference plant in the type of equipment that is being u;ed to perform the same process steps. However, from a decom-missioning standpoint, the differences in the amount and size of equipment for various plant processes and the reference plant are small. Therefore, this l decommissioning description is considered representative.

(a)The large hexafluoride conversion plant was put into safe storage in the 1961-62 period. It has since been converted to another use. There is l currently a small hexafluoride plant for converting returned and reclaimed uranium compounds to feed for the cascade enrichment plant.

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11.1 Uranium Hexafluoride Conversion Plant Description 11.1.1 Plant and Process Description The reference UFs plant is assumeci to occupy about 30.4 hectares (75 acres) within the generic site described in Section 3. The plant consists of three buildings containing approximately 120,000 ft 2 of ficor area. The buildings are of normal industrial construction, with heavy concrete floors to support equipment. In addition, there are a series of retention ponds for sanitary waste and process raffinates. The plant is designed to receive U30s or yellowcake in 208-liter (55 gallon) drums from various uranium mills located in the western United States and to convert the feed stock to uranium hexafluoride (UFs). Two processes, which differ only in the method of purification, are in use today.

The major steps in either process are:

1. pre process handling, weighing, sampling, and storage

2. conversion of the U3 0s or yellowcake tc, uranium trioxide (UO3 ) by roasting

3. reduction of the U03to UO2 with hydrogen

4. hydrofluorination of the UO2 to UF4 with hydrogen fluoride

5. fluorination of the UF 4 to UFs with elemental fluorine

6. storage of the purified UFs in shipping cylinders The purification step is added either at the beginning using a solvent extrac-tion process or at the end by fractional distillation of the UFs. The use of the solveat extraction purification step (the wet process) results in the radio-active uranium daughters (2soTh and 22sRa) and impurities being left in the solvent extraction raffinate. The acidic raffinate is neutralized and the slurry is retained in lagoons. The dried slurry would be disposed of in a shallow-land burial ground or returned to a mill for uranium recovery and disposal with the tailings (see Figure 11.1-1). The dry process, on the other hand, removes the impurities from the UFs product stream by fractional distillation and incorporates O

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them with other waste products for dispo331 as solid waste in a shallow-land '

burial ground (see Figure 11.1-2). All gaseous effluent streams are filtered, and those containing fluorine compounds are scrubbed with potassium or calcium hydroxide solution.

The plant equipment, fabricated mostly of monel, is mainly a series of fluidized bed chemical reactors with intermediate vessels, such as storage bins, air clasrifiers, product filters, cold traps. and air effluent purification systems.

The plant facility has lagoon areas for neutralized liquid effluents and a burial arca for disposal of defunct equipment.

The purified UFs is placed in cylinders for storage and future shipment to one ,

of the Department of Energy's enrichment plants.

11.1.2 Estimates of Radioactivity Levels at UFs Plant Shutdown The reference UFs plant processes 10,000 metric tons of natural uranium per year in the form of ore concentrate (yellowcake) produced by domestic uranium mills. The feed to the reference UFs plant is assumed to be a composite product of uranium, produced 85% from acid leach and 15% from alkaline leach, which has aged at least six months in sealed drums after milling. The radionuclides of primary concern are teatural uraniur, 22sRa, 23oTh, ?3*Th, 234mPa, and 222Rn.

The daughter products of radon are not listed as radionuclides of primary con-cern either because they have half-lives of less than 2 hoirs and do not accumu-late in the bioenvironment (21spo, 214Pb, 21481, and 2 oPo) or because they ~

individually contribute less than 0.02% of the total relative hazard (2 oPb, 2:o8i, and 2 oPo). Analysis of the plant feed at the Allied Chemical Plant at Metropolis, Illinois 2, indicates that there are 2,800 picocuries of 23oTh and 200 picocuries of 22sRa per gram of natural uranium. This amounts to 28 curies of 2 aoth and 2 curies of 22sRa entering the plant each year, the majority of which is recycled at the mills by wet processing or is sent to low-level waste burial as solid waste from dry processing. Natural uranium is the most abundant radionuclide present. The predominant health and safety consideration is not radiological, but rather the effect that heavy metal (uranium) chemical toxicity has on the human kidney.

10/08/87 11-3 NUO586 CH 11

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11.2 Uranium i;e<afluoride Conversion Plant Decommissioning Experience OCE has terminated UFs conversion at the Oak Ridge, Tennessee and Portsmouth, Ohio Enrichment Plants and at the Mallinckrodt Chemical Company Plant at Welden Springs, Missouri. The Welden Springs Plant is currently undergoing decommis-sioning, and the knowledge gained from this experience will be useful in the planning and decommissioning of similar plants The status of decommissioning of the Oak Ridge Plant is not known at this time 11.3 Decommissioning Alternatives Once a UFs plant has reached the end of its useful operating life, it must be decommissioned. As discussed in Section 2.3, this means safely removing the facility from service and disposing of all radioactive materials in excess of levels which would permit unrestricted use of the facility. Several alterna-tives are considered here as to their potential for satisfying this general requirement for decommissioning. The decommissioning alternatives primarily considered and discussed here are DECON and SAFSTOR. ENTOMB is not considered a realistic alternative, and is included only for completeness.

The alternative used depends on such considerations as cost, dose, and the pro-posed use of the site. Special considerations involved in decommissioning the reference UFs plants include the following general assumptions:

~

1. natural uranium and its radioactive daughters are the only radioactive materials handled at the plant,

2. uranium spills that occur during the life of the plant, both inside and outside, are cleaned up immediately, and

3. safety reasons dictate that the maximum amount of uranium be removed from the plant prior to decommissioning.

Other considerations include the f act that decontamination of equipment is com-paratively easy since most uranium found at the UFs conversion plant is quite 11-5 NUO586 CH 11 10/08/87

soluble in nitric acid (HNO 3) and aluminum nitrate. The cleanout of the plant following shutdown removes essentially all of the uranium. Decommissioning following this cleanout should be equivalent to the cleanup of any chemical processing plant. An extensive radiation survey of the buildings and equipment would pinpoint any contaminated areas and thus allow an estimate to be made of the time and money needed for decommissioning. This radiation survey may show that all of the buildings and equipment can be released for unrestricted use, although it is more probable that some are releasable and some need further decontaminaticn. Because of the low specific activity of uranium, radiation exposures of the public are negligible and therefore are of little concern, the owner could choose the most economical alternative for decommissioning with NRC concurrenca. The most practical choice of decommissioning alternatives based on economics, appears to be basically only one: DECON. However, the other options listed above are briefly discussed here.

11.3.1 DECON DECON is defined as the immediate removal and disposal of all radioactivity in excess of levels which would permit release of the facility for unrestricted use. Nonradioactive equipment and structures need not be torn down or removed

as part of a DECON procedure. The end result is the release of the site and any remaining structures for unrestricted use as early as 8 months after the end of facility operation.

j OECON is advantageous because it allows for termination of the NRC license l shortly af ter cessation of f acility operations and removes a radioactive site.

DECON is advantageous if the site is requi ed for other purposes or if the site is extremely valuable. It is also advantageous in that the facility operating staff is available to assist with decommissioning and that continued surveillance is not required.

1 l

Because of the low radiation exposures from natural uranium, DECON could start i

l at once fs11owing the final operational equipment cleanout and radiation survey, Salvageable equipment would be decontaminated as necessary by water or nitric acid flushing, hand scrubbing, or by vibratory or electropolishing techniques.

I Nonsalvageable or hard-to-decontaminate contaminated equipment would be shipped i

11-6 HUO586 CH 11 10/08/87 l

to a low-level waste burial ground for fisposal. The structures used to house the UFs process would be decontaminated as necessary and then demolished or used for another purpose at the discretion of the owner. The site would be surveyed and any contamination would be removed. Most contaminated materials would be disposed of in a low-level waste burial ground.

The disassembly of the equipment would result in valves and piping being boxed for disposal. The larger vessels will be cut into pieces for disposal. The vessels could act as their cwn containers and have all openings bolted or welded closed. Trash would be stuffed into these vessels for disposal.

Ten percent of the concrete floor is assu;ned to be contaminated and 10 cm (4 in.)

of the top of this surface is chipped away and disposed of as rabble. This estimate accounts for building materials that might need to be disposed of in a shallow-land burial site.

The removal of the uranium from the process equipment removes any significant radiation exposure to either the public or to the decommissioning worker. The radiation dose for the dismantling crew is expected to be less than for the initial cleaning. Average radiation dose rates in the plant during the initial cleaning are expected to be much less than 2 mrem /hr, which is the radiation dose rate from bulk quantities of uranium. Thus, the decontamination of the plant, packaging of contaminated wastet, and transporting of this material to a low-level waste burial ground is estimated to result in negligible radiation l

exposure to the public (see Tab'e 11.5-1). An additional 17 man-rem is estimated for transportation of contaminated waste, including disposal of lagoon waste.

I i

Table 11.5-1 summarizes the estimated costs in 1986 dollars for the decommission-ing alternatives analyzed in this report. The OECON costs are estimated to be

$12.1 million. These costs include costs for labor, equipment and materials, l waste disposal and other expenses. Lagoon waste is assumed be disposed of at a uranium mill. If lagoon waste must be disposed of at a waste burial ground the cost is estimated to be $53 million. All cost estimates include an added 25%

for contingencies. A time period of about 1 year is estimated for DECON.

l l

l 11-7 NUO586 CH 11 10/08/87 I _ ,_ _. _

Once DECON is complete, i.e., once the facility is decontaminated to levels permitting release of the fr.ility for unrestricted use. the radioactive mate-rials license would be terminated and the owner would be free to dispose of the site as he wished.

11 3.2 SAFSTOR SAFSTOR is defined as those activities required to place (preparation for safe storage) and maintain (safe t age) a UFs plant in such condition that the risk to safety is within acceptable bounds, and that tM facility can be safely stored and subsequently decontaminated to levels which permit release of the facility for unrestricted use (deferred decontamination).

Generally, the primary purpose of SAFSTOR for most nuclear facilities is that it results in reduced occupational exposure compared to DECON However for the reasons given in Section 11.3 and as can be seen in Table 11.5-1 this is not the case for UFs plants. A disadvantage of SAFSTOR is that the licenste is required to maintain a material license and to meet its requirements at all times during safe storage. Other disadvantages are that SAFSTOR contributes to the number of sites dedicated to radioactive confinement, surveillance is re-quired, the dollar costs may be higher than for DECON, and the experienced operating staff may not be available at the end of the safe storage period to assist in the deferred decontamination.

Safe storage preparation is the same as the initial decontamination. The build-ings and areas would be secured, but because of the small amount of radiation (les: than 1 mrem /hr) and minimal danger to an intruder, only periodic surveil-lance would be necessary (twice per week). The length of the continuing care period would then be at the option of the owner. Continuing care would cost approximately $125,000 per year. A safe storage period of 10 years would result l

in total SAFSTOR costs of $15.1 million, which is larger than for DECON. This l

would take place with no inc%ase or decrease in total radiation dose to the l

public or workers. Deferred decontamination could take place at any time, l

I would require the same steps as DECON and would result in similar costs and doses as for DECON.

10/08/87 11-8 NUO586 CH 11

For the reasons discussed in Section 11.3 radiation dose to the public would be negligible (see Table 11.5-1).

11.3.3 ENTOMB ENTOMB of a UFs plant untti its radioactivity has reached levels permitting release of the facility for unrestricted use requires its encasement in con-crete to protect the public from radiation exposure. Because the radiation levels from the trace amount of natural uranium in the equipment and buildings are nearly zero and because the process buildings are not suitable for ENTOMB, this is a very expensive and unnecessary decommissioning alternative and is not considered a viable option.

11.3.4 Site Decommissioning No site decommissioning other than a radiation survey is expected to be necessary since it is assumed that each spill will be cleaned up immediately. If failed equipment or other contaminated solids have been buried onsite, they will have to be removed to a low-level burial ground. However, the removal of onsite buried materials is expected to be a minor effort compared to the rest of the decommissioning 11.4 Environmental Consequences The environmental consequences of decommissioning a UFs conversion plant are -

small. The largest environmental impact is postulated to be tne use of about 0.2 hectare (0.5 acre) of irretrievable land for shallow-land burial and the consumption of materials (gasoline, wood, metal tools, etc.) during the decom-missioning activities. Decommissioning would make the 30.4 hectares (75 acres) of plant-site land available for unrestricted use. Reactivation of the site as another industrial endeavor would be advantegeous to the local residents, about 100 of whom worked at the plant. The occupational and public radiation doses which are negligible, are discussed in Section 11.3. Discussion of costs are also included in Section 11.3.

10/08/87 11-9 NUO586 CH 11

11.4.1 Waste Disposal The volume cf low-leval waste to be disposed of is estimated on the basis that all process equipment is discarded. The volume estimated, 1,259 m 3 , is con-sidered to be a maximum that requires about 0.4 hectare (1 acre) of a shallow-land burial site. Any equipment that can be reused or released for salvage will reduce the volume sent to burial. The land used for burial is considered irretrievable. These land use requirements for waste disposal are not large in comparison with the approximately 1160 acres used as the site of the reference UFs plant which could now be returned to unrestricted use.

11.4.2 Additional Effects of Decommissioning The socioeconomic impacts are mainly from the shutdown (not decommissioning) of the facility and associated loss of about 100 jobs. Since the main attributes of an industrial site are still available, it would be in the best interests of the local communities to establish a new industry that would supply jobs and money through taxes. On the basis of economics, this use of the site would probably be preferred to returning it to its original condition.

11.5 Comparison of Decommissioning Alternatives Table 11.5-1 presents a summary of the decommissioning alternatives discussed in this section. The choice of an alternative generally depends on such consid-erations as dose, cost, and proposed use of the site. As discussed in Sec-tion 11.3 3, ENTOMB is not considered a viable option ' is not listed in Table 11.5-1. Of the two remaining alternatives, DECON and SAFSTOR, DECON appears to be the more advantageous option. This is because the radiation doses are small for either alternative, while DECON has lower costs and results in release of the facility for unrestricted use in a fairly short time period.

10/08/87 11-10 NUO586 CH 11 l

Table 11.5-1 Summary of Decommissirning Alternatives SAFSTOR DECON 10 Years 30 Years 100 Years Total Cost (millions ofconstgg}1986 dollars) 12.1 15.1 17.6 26.4 Occupational Radiation Dose (man-rem) 62 63 65 67 Transportation Radiation Dose (man-rem) 17 17 17 17 Public Radiation Dose (man-rem) 5.7 5.7 5.7 5.7 Potential Industrial Accidents - Injuries 1.8 1.9 2. 0 2.6 Fatalities 0.094 0.095 0.096 0.10 Manpower Expenditures (cumulative man years) 43.4 60.3 80 9 150 Land Area Committed (acres) 0 75(b) 75(b) 75(b)

(a) Lagoon waste assumed to be shipped to a uranium mill. If disposal of lagoon waste at a commercial waste burial ground is necessary, add

$53 million.

(b)Part of the site might be decontaminated, surveyed, and released for unrestricted use while the facility is put in safe storage, if desired, l .

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l 10/08/87 11-11 NUO586 CH 11

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tEFERENCE

. 1.' H. K. Elder, Technoloqy,: Safety and Costs of Decommissioning a Reference Uranium HexafTuoride Conversion Plant, NUREG/CR-1757, Prepared by Pacific Laboratory for U.S. Nuclear Regulatory Commission, October 1981.

4 l

.s 10/08/87 11-12 NUO586 CH 11

12 URANIUM FUEL FABRICATION PLANT A uranium fuel fabrication plant (U-fab plant) is a facility in which enriched uranium, received as uranium hexafluoride (UFs), is converted to UO2 and formed into fuel pellets that are inserted into fuel rods. These fuel rods are, in turn, assembled into fuel bundles. There are two kinds of U-fab plants: high-level enriched U-fab plants which produce fuel for reactors that power naval 3 vessels and for reactors that serve other special purposes, and low-level enriched U-fab plants which produce fuel for commercial nuclear power reactors that generate electricity. Plants that fabricate fuel for the U.S Navy are outside the scope of this EIS, but their decommissioning impact would be similar to the decommissioning of low-level enriched U-fab plants.

Some low-level enriched U-fab plants perform the whole operation, i.e. , they receive UFs and produce fuel bundles. Other facilities operate in two stages, i.e., one plant receives UFs and produces 002 powder or pellets, and a second plant assembles the fuel rods and bundles. The reference plant for this study performs the whole operation.

This section presents an assessment of the environmental effects that may be expected from the decommissioning of such a facility. This s3ction is based primarily on information from a study! of the decommissioning of a uranium fuel fabrication plant. In this study PNL selected the General Electric Plant located at Wilmington, North Carolina as the reference U-fab plant and assumed it to be located at the generic site. The generic site is described in Section 3.1. As part of this study, PNL developed information on the available technology, safety considerations, and probable costs for i decommissioning the reference facility at the end of its operating life.

i 12.1 U-Fab Plant Description l

The reference U-fab plant is assumed to have operated for 40 years, processing l an average of 1000 MT of uranium per year. Production consists of three general kinds of activities: conversion of slightly enriched UFs to U0 ;2 mecSanical 12-1 NUO586 CH 12 10/08/87 l

production of fuel pellets and assembly of fuel rods and bundles; and recovery of uranium from scrap, wastes, and off-standard pellets.

Conversion of UFs, as received from an enrichment facility, to U02 is accom-plished by either a chemical or a direct process. In the chemical process, the UFs is first hydrolyzed to UO 2F2 and amonium hydroxide is added to precipitate the uranium as ammonium diuranate (ADU). Then the ADU is reduced and calcined to produce UO2 powder. In the direct process, conversion reactors convert UFs directly to U 0s, 3 which is then reduced to U0 2-In the production of pellets, the U02 is pulverized and compacted to granules of a desired density. The granules are pressed into pellets which are sintered at high temperature in a reducing atmosphere. The pellets are then ground to proper size and loaded into zircaloy or stainless steel tubes which are dried, evacuated, filled with helium, and welded closed. The tubes (now called fuel rods) are tested for leaks, assembled into fuel bundles, inspected and stored for shipment.

The building is a two-story, windowless structure of concrete and steel.

Interior walls, typically of concrete block, divide the building into discrete operations areas that house .ach of the production steps. When the plant is shut down and the final inventory cleanout has been perfomed, it is antici-pated that there will be ' total of about 270 kg of unrecovered uranium remain-ing in the plant. ",f this amount, approximately 150 kg is in the equipment and

~

120 kg is in tha ventiiation system. This uranium has enrichments that range from 2% to less than 5% 2ssU. CaF 2 is a waste product that is produced by treating the fluoride wastes with Ca(OH)2 The CaF2 is stored in waste ponds.

Those CaF2 waste ponds will contain some enrichad uranium and will therefore require some decommissioning activity. Although CaF2 has low solubility, the toxicity of inorganic fluorides in general suggests that these wastes may be a s

biological hazard.2 12.2 U-Fab Plant Decommissioning Experience

-Several U-fab plants have ceased operation and are in various stages of decommis-sioning. At some facilities a high-level enriched U-fab operation has been 10/08/87 12-2 NUO586 CH 12

shut down, leaving a low-level enriched U-fab operation still in production.

Examples are a Babcock and Wilcox Plant at Apollo, Pennsylvania, and a Combustion Engineering Plant at Hematite, Missouri. At the Combustion Engineering Plant, there has been a partial cleanup, but at neither plant has the facility been completely decommissioned. Babcock and Wilcox also has a high-level enriched t plant at Leechburg, Pennsylvania, that has been shut down and partially decommis-l sioned. Some equipment has been removed but the ventilation system is still intact. United Nuclear closed a high-level enriched U-fab plant at New Haven, Connecticut, several years ago and U.S. Nuclear Corporation decommissioned a high-level enriched U-fab test and research facility at Oak Ridge, Tennessee.

Among the low-level enriched U-fab plants, two facilities which have been shut down are examples of decommissioning experience. A Kerr-McGee Plant at Crescent, Oklahoma, has been partly decommissioned. The plant is still intact, but the

) waste ponds have been cleaned up. This waste was loaded into drums and shipped to a burial ground.

Perhaps the best experience in decommissioning a low-level enriched U-fab pia nt was with a General Electric U-fab Plant in San Jose, California. At shutdown, the area was cleaned to administrative control levels not exceeding 1000 dpm/

100 cm2 . Decommissioning was accomplished by dismantling and removing all of the process equipment and ventilation system and cleaning the building. Pipes, lighting fixtures, etc., were cleaned; fluorescent tubes were replaced; ceilings, walls, pipes, and lighting fixtures were damp-wiped, baseboard moldings and tile floors were removed, and concrete floors were vacuumed and mopped. Pump

~

basins that had been formed by constructing concrete berms were cleaned up by removing the berms and wet grinding hot spots. The decommissioning effort was more extensive than should have normally been necessary, because on one occasion an accident occurred that released a large amount of UFs inside the plant. This accident contaminated not only all the building and fixture surfaces in the production areas but also the otherwise clean areas, such as offices.

I 10/08/87 12-3 NUO586 CH 12

i l

l 12.3 Decommissioning Alternatives Once a U-fab plant has reached the end of its useful operating life it must be decommissioned. As discussed in Sectior 2.3, this means safely removing the facility from service and disposing of all radioactive materials in excess of levels which would permit unrestricted use of the facility. Several alternatives are considered here as to their potential for satisfying this general requirement for decommissioning. The decommissioning alternatives considered and discussed here are DECON, SAFSTOR, and ENTOMB. The alternative used depends on such considerations as cost, dose, proposed use of the site and desirability of terminating the license.

Most of the residual radioactivity in a U-fab plant following shutdown is surface contamir.ation,3 although concrete in some areas of the plant may be contaminated to a shallow depth. It is assumed that a complete radiological survey of the plant and its equipment will be made as a normal operational procedure at the time of shutdown and that nitrote wastes have been removed and reprocessed as a part of norw,al operations. Thus, preparing the facility for unrestricted use will involve removal of the equipment, decontamination of the building, removal of some concrete surfaces as indicated by the survey, disposal of chemical wastes, and disposal of the CaF2 wastes in the lagoons Discussions of the decommissioning alternatives follow:

12.3.1 DECON l DECON is defined as the immediate removal and disposal of all radioactivity in i excess of levels which would permit release of the facility for unrestricted l

use. Nonradioactive equipment and structures need not be torn down or removed as part of a OTCON procedure. The end result is the release of the site and

any remaining structures for unrestricted use as early as the 9 months estimated for decommissioning after the end of facility operation.

DECON is advantageous because it allows for termination of the NRC license shortly after cessation of facility operations and removes a radioactive site.

DECON is advantageous if the site is required for other purposes or if the site 10/08/87 12-4 NUO586 CH 12 l

is extremely valuable. It is also advantageous in that the facility operating staf f is available to assist with decommissioning and that continued surveillance is not required.

DECON of a U-fab plant presents few problems. The equipment and ventilation systems are removed and the building surfaces are damp-wiped. The equipment and vents most highly contaminated will be in the calciner, press, hammer mill, blender, and grinder areas. Some of this equipment and the furnaces can be reclaimed by replacing the parts that were exposed to the uranium. While the same may apply to the vent systems, it is likely that much of this material will be~ discarded. The replaced and discarded material will be shipped to a low-level waste burial ground. In some parts of the building, particularly the chemical processing areas, there will be places, such as pump basins, where it will be necessary to remove concrete floor surfaces. This will be accomplished by grinding, chipping or spalling. with the removed concrete being sent to a low-level waste burial ground.

The major problem in decommissioning a U-fab plant may be with the waste ponds and other areas where the soil is contaminated. Wastes in the nitrate ponds will have been removed, shipped to another plant, and reprocessed; but the calcium fluoride waste may have to be removed and shipped to a ?ow-level waste burial ground. It is also possible that the CaF2 waste may be removed and reprocessed at another plant to recover uranium. The CaF2 would then be dis-posed of by th2 new owner. The nonradioactive chemical wastes will be sent to a chemical waste burial ground. -

Analyses of radiation exposure and costs for DECON are presented in Section 12.3.4.

12.3.2 SAFSTOR (Custodial)

SAFSTOR is defined as those activities required to place (preparation for safe storage) and maintain (safe storage) a U-fab plant in such condition that the risk to safety is within acceptable bounds, and that the facility can be safely stored and subsequently decontaminated to levels which permit release of the facility for unrestricted use (deferred decontamination).

10/08/87 12-5 NUO586 CH 12

Ge'erally, the primary purpose of SAFSTOR for most nuclear facilities is that

.it results in reduced occupational exposure c.ompared to DECON. However for the reasons given in Section 12.3.4.1 and as can be seen in Table 12.3-1 this is not necessarily the case for U-fab plants. SAFSTOR could be advantageous in the event that there is a shortage of immediate waste burial acccmmodation. If this is the case it may be desirable to place the facility in custodial safe storage prior to deferred decontamination leading to release of the facility for unrestricted use. Custodial SAFSTOR for a U-fab plant would require only minimal cleanup with continuing maintenance and security. The CaF2 wastes may have to be sold and removed for reprocessing or removed to a permanent waste burial ground. The chemical wastes will be removed to a chemical waste disposal area.

Table 12.3-1 Summary of radiation safety analyses for routine decommiss{ggingofthereferenceU-fabplant (man-rem)

SAFSTOR Occupational Exposure DECON 10 years 30 years Preparation NA 0.4 0.4 NA 6 18 Continuing Care (b)

Decontamination 16 16 16 Transportation 2.6 2.6 2.6 Totals 18.6 25 37 (50 year dose commitment to the critical crgan)

Public Exposure Preparation NA 0.06 0.06 Continuing Care (b) NA 0.05 0.15 Decontamination 0.06 0.06 0.06 Transportation 0.53 0.53 0.53 Totals 0.6 0.7 0.8

)

( ) Adapted from Reference 1 For SAFSTOR, this is deferred decontamination 10/09/87 12-6 NUO586 CH 12

Another disadvantage of SAFSTOR, in addition to increased radiation ex;,;sure, is that the licensee is required to maintain a material license and to meet its requirements at all times during safe storage thus contributing to the number of sites dedicated to radioactive confinement for an extended time period.

Other disadvantages are that surveillance is required, the dollar costs are higher than for DECON, and the experienced operating staff may not be available at the end of the safe storage period to assist in the deferred decontamination.

Over the short-term, custodial SAFSTOR might be temporarily expedient, but neither the cost of eventual decontamination nor the occupational radiation dose would be decreased by delaying decontamination due to the long half-lives of the radionuclides involved. It appears that the viability of this alternative will be determined on a case-by-case basis and will be dependent on the needs and resources of the UFs plant owner and the requirements of HRC.

Analyses of radiation exposures and costs for SAFSTOR are presented in Section 12. 3.4.

12.3.3 ENTOMB ENTOMB of a U-fab plant requires its encasement in concrete to protect the public from radiation exposure until its radioactivity has reached levels permitting release of the facility for unrestricted use. It is a possible but not very reasonable alternative. The building is not structurally suited to entombment, therefore, the initial entombing process would be costly. Because the radio-I nuclides present in the UFs plant have very long half-lives, the structure would have to be monitored and maintained in perpetuity, which is well beyond the l time that required institutional control, could be expected to be effective l (approximately 100 years is considered to be consistent with recommended EPA policy on institutional control for radioactivity confinement). Also. there would be no cost or safety advantage to ENTOMB, because DECON is simple, safe, and relatively inexpensive. In any event, the waste ponds would have to be l removed and could not be entombed. ENTOMB is not a viable decommissioning alternative.

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12.3.4 Summary of Radiation Safety and Decommissioning Coits 12.3.4.1 Radiation Safety Residual radioactivity following inventury removal at a U-fab plant will be confined mainly to the interior parts of equipment and the ventilation system.

The CaF2 waste, containing some uranium, may have to be reprocessed or sent to a low-level waste burial ground.

The radioactivity in a U-fab plant is mostly due to 2ssU and 234U. External dose to decommissioning workers will be at plant background, which is about 1 arem/hr. Because of the long half-life of 23sU, (approximately 7 x 10s years) this background will not be decreased appreciably by placing the plant in custodial safe storage for a time before deferred decontamination.

The approximately 270 kg of uranium that are still in the plant at shutdown contain about 8 kg of 23sU, which will be thinly dispersed over very large surface areas of the equipment and ventilation system. The possibility is remote that a worker at any particular location would contact a large concentration of 2ssU. Nevertheless, some pieces of equipment will be more highly contaminated than others and the possibility exists that dust can be dislodged and suspended in the air where it will be inhaled. For this reason, appropriate protective clothing and face masks will likely be needed for decommissioning selected parts of the plant.

Occupational radiation exposure of werkers performing the decommissioning activities results from. external exposure for reasons similar to that discussed for PWRs in Section 4.3.1. Table 12.3-1 presents a summary of the radiation exposures t" t may result from each of the d uommissioning alternatives. As can be seen from the table, the occupational expn ures de not decrease as a result of using the SAFSTOR alternative. This is because of the long half-lives l of the radionuclides present in the facility compared to the time the facility might be held in safe storage awaiting deferreo decontamination. As can also

( be seen from Table 12.3-1, total estimated public exposures from decommissioning activities are very small. If the CaF2 waste has not been removed and shipped to another plant for reprocessing, it may have to be packaged and shipped to a 10/08/87 12-8 NUO586 CH 12

_ =

low-level waste burial ground for disposal. This would result in additional occupational and public radiation doses of 20 and 0.4 man rem, respectively.

A range of possible accidents that would result in released radioactivity is postulateo. The largest releases are from loss of HEPA filters. This would result in the same quantities of release and radiation doses and have the same probabilities of occurrence with either decontamination alternative.

A summary of the estimated doses to the public from accidents is shown in Table 12.3-2. Radiation doses to the public resulting from accidents are low enough to be considered insignificant.

Radicactive waste materials are packaged and shipped offsite for burial during decommissioning of the reference U-fab plant. The dominant radiation exposure pathway to transport workers and the public during transportation of radioactive wastes is external exposure for reasons similar to those discussed for PWRs in Section 4.3.1. The external dose for transportation is conservatively estimated to be 2.6 man-rem to transportation workers and 0.53 man-rem to the public for either DECON or SAFSTOR. These doses are based on regulations of the Department of Transportation governing radiation levels in shipments and on estimates of the distances of travel and lengths of time of exposure that workers and the put,lic might expect. These doses are sumarized in Table 12.3-1.

The severity of accidents that may cccur during transportation of radioactive waste depends on a number of factors, such as speed, kind of accident, and accident locations. Regardless of the decomissioning alternative, the same total amount of radioactive material will be transported. Thus, the possible release of radioactivity will be cependent on frequency and kind of accidents, as shown in Table 12.3-3.

12.3.4.2 Decomissioning Costs Table 12.3-4 sumarizes the estimated costs in 1986 dollars for the decomis-sioning alternatives analyzed in this report. All cost estimates include an added 25% for contingencies. For DECON, the decomissioning costs are estDated i

to be $8.8 million. For custodial SAFSTOR, the total decommissioning cost u l

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Table 12.3-2 Summary of radiation doses to the maximum-exposed individual from accidentalairborneradionuclidereleasesduringpgjommissioning activities for either decommissioning alternative Release to Fifty year committed Expected Atmosphere First-Year dose, mres dose equivalent, mrem Frequency of Incident (pCi) Bone Lung Bone Lung Occurrence Loss or It.ter-mediate HEPA Filter After Duct Decon-2.7 2.3 x 10 3 7.6 x 10 2 4.5 x 10 3 1.9 x 10 1 High tamination Loss of Local Airborne Con-tamination Control, Loss of Vacuum Filter 0.70 6.0 x 10 4 2.0 x 10 2 1.1 x 10 3 4.9 x 10 2 High Liquid Leak During Chemi-cal Decon-tamination 4.5 x 10 3 3.7 x 10 8 1.3 x 10 4 7.3 x 10 8 3.1 x 10 4 High (a) Adapted from Reference 1

- -- -_-_a

d

.c Table 12.3-3 Estimated f.equencies and radioactivity releases for selected transpertation accidents (d)

RadiationDoseformaxigge-Exposed Individual (rem)

Accidents per Releate First-rair Dose Committed Dose Equivalent Accident Description Dismantlement (Ci)(b,c) Bone Lungs Bone Lungs Minor Accident  ? w 10 2 No Release -- -- -- --

Moderate Accident 5.4 x 70 3 1 x 10 7 3.9 x 10 8 1.3 x 10 4 7.7 x 10 8 3.2 x 10 4 Severe Accident 1.4 x 10

• 1 x 10 5 3.9 x 10 4 1.3 x 10 2 7.6 x 10 4 3.2 x 10 2 U) Maximum-exposed individual is assumed at 100 m from the site of the accident.

(b) Based on an inventory of 100 mci, the expected maximum per truck shipment.

(c) Release fraction for respirable material for moderate and severe accidents are assumed to be 10 8

. and 10 4, respectively.

( Adapted from Reference 1.

l 4

1 l .

Table 12.344 Summary of estimated costgor decomissioning the reference U-fab plant Estimated Costs in Millions of 1986 Dollars SAFSTOR (Custodial)

Item DECON 10 year 30 year Preparation NA 1.4 1.4 Continuing Care (b) NA 4.6 14.0 Decontamingon 8.8 9.3 9.3

, Total O 1D 24T7 (a) Adapted from Reference 1.

(b)For SAFSTOR, this is deferred decontamination.

(C) Total does not include additional potential cost of contaminated CaF2 disposal. This would add approximately 36.8 million to the total.

estimated to be $15.3 million and $24.7 million for 10 year and 30 year SAFSTOR, respectively. These SAFSTOR costs include $1.4 million for preparation for safe storage, $0.47 millicn per year for continuing care, and $9.3 million for deferred decontamination. A present value analysis of decommissioning costs indicates a disincentive to defer decontaminatien for the reference case indicated, primarily because of the high cost of continuing care relative to DECON costs and the high coct of deferred decontamination due to the long half-lives of the radionuclides involved. Therefore, from a cost standpoint, it is probably to an operator's advantage to choose the DECON alternative and convert the building to other uses.

Most of the cost of decommissioning a U-fab plant will oe for labor. A large portion of the labor costs will be fo- handwashing the ceiling, wall, and floor surfaces of the building. Equipment that is still serviceable will also be damp-wiped or flushed with detergent iolutions or weak acid where hand wiping is not possible. Some spalling of concrete floors may be required in areas such as pump basins which have had contact with uranium soluticns. Deferring ,

decontamination adds to the total cost because of the cost of labor for ron-tinuing s.tre, of reactivating full utility service and of holding licenses.

It does not decrease the cost of eventual decontamination.

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Of the total costs listed in Table 12.3-4, the cost of waste management is 53.5 million. This includes $1.7 million for low-level waste burial of con-tarinated equipment, building components, and concrete, and $1.8 million for d'sposal of the chemical waste sludge (nonradioactive) in a chemical waste burial ground. The CaF2 waste will potentially be disposed of ir, a low-level waste burial ground, and removal, packaging, shipment, and burial would cost an additional $36.8 million.

12.4 Environmental Consequences Because radiological effects are quite small, the potential nonradiological effects will have the greater impat.t on the environment.

12.4.1 Nonradiological Safety The area of greatest concern for the welfare of decommissioning workers is the calcium fluoride lagoons and storage pits. The very caustic nature of CaF2 makes it necessary to protect the workers from contacting it on their skin and breathing the dust. The workers will therefore require protective clothing and respirators. The trucks used for tran port to the burial ground will have the same risk of traffic accidents as with any other trucking operation, and the probability of accidents can be estimated freein highway safety statistics, This probability is estimated to be 1.5 x 10 6 accidents per kilometer of travel.4

~

12.4.2 Commitment of Resources The largest commitment of resources will be for space in chemical and low-level waste burial grounds. The burial volume of contaminated equipment, building companents, and concrete is 1100 m8 , the burial volume of CaF2 waste would be 29,600 t8 (accounts for almost 3 acres of burial ground), and the buria'l volume of other chemical waste is 5300 m8. Materials used up in decontaminating a U-fab plant will include cleaning supplies, such as detergent 3, clothes, sops, and brushes.

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a

.i 12.4.3 Socioeconomic Effects ir decommissioning a U-fab plant, many of the same people that operated the plant can da the cleaning, but the dismantling and moving of equipment will be done by electricians, plumbers, mechanics, and equipment operators, most of wnom will be hired or contracted. The socioeconomic effects of decommissioning, then, will come from the employment of these craf tsmen. The total decontami-nation crew may be larger than the operatii.g crew, and so for the period of decontamination, the economic input to the community will increase. In the case of custodial safe storage, the work force may decrease to a security and maintenance crew for the period of continuing care.

Because of the planning time needed to precede the decommissioning, changes in the number of employees will not be sudden or without warning, and people w;11 have time to find other employment. .

12.5 Comparison of Decommissioning Alternatives The options of OECON and SAFSTOR (custodial) both eventually end with the same results: a decontaminated facility that can be released for unres cicted use.

The choice of an alternative generally depends on such considerations as dose, cost, proposed use of the site, and desirability of terminating the license.

For a U-fab plant, due to che long lived nature of the radionuclides present, doses and costs are not redaced esen when decontamination is deferred for 30 '

years, as can be seen from lables 12.3-1 and 12.3-4. In addition since the cost and doses of contineirg ca.e is a major item and continues to increase with increasing safe storege time, the doses 6nd cost associated with the cemplete SAFSTOR prtcess exceed those for CECON Therefore, DECON appears to be a more advantageous option. For the reasons given in Section 12.3.3, EhTOFD is not considered a viable alternative.

12-14 NUO586 CH 12 10/08/87

REFERENCES

1. Technology, Safety and Costs of Decommissioning a Reference Uranium Fuel Fabrication Plant, NUREG/CR-1266 Prepared by Pacific Northwest Laboratory

- for U S. Nuclear Regulatory Commission, October 1980, 2.

Sax, N. b >ing, Dangerous Properties ~of Industrial Materials, Fifth Ed.,

Van Nostrand Reinhold Co., 1979.

3 Cleaning up the Remains of Nuclear Facilities - A Multibillion Dollar Problem, deport to the Congress by the Comptroller General of the United States, June 16, 1977.

4. Reactor Safety Studies - An Assessment of the Accident Risks in U.S.

Commercial Nuclear Power Plants, WASH-1400 (NUREG-75/014), U.S. Nuclear Egulatory Commission, Washington, DC 1975; 1,$ #  !

e 4

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13 INDEPENDENT SPENT-FUEL STORAGE II;STALLATION An independent spent fuel storage installation (ISFSI) is a facility for handling and storing irradiated spent fuel assemblies fron nuclear power reactors until they can be permanently disposed of as high-level waste. The two basic design categories of I'WS!s t,re wet storage of the fuel and dry storage of tae fuel.

The design of the wet storage ISFSI is similar to that of reactor spent fuel storage pools except that the storage capacity is significantly greater. There are different designs for dry storage ISFSIs, however the four basic types thu are considered here are drywell storage, silo storage, vault storage, and cask storage. For cooling the fuel, these dry, storage designs depend on such means as air currents, heat dissipation in the soil, and metal heat transfer fins.

This section presents an assessment of the environmental, fincncial, and socio-economic effects that may be expected from the decommissioning of an ISFSI.

This section is based primarily on information from a studyt of the decommission-ing of five different reference ISFSIs corresponding to the five different designs noted above. In the study, each reference ISFSI design was assumed to be located at the generic site. The generic site is described in Section 3.1.

As part of this study, PNL developed infwmation on the available technology, safety considerations, and probable costs for decommissioning the five reference ISFSIs at the end of their useful operational lives. -

13.1 Description of an Independent Spent Fuel Storage Installation (ISFSI) 13.1.1 Wet Storage ISFSI In this design, spent nuclear fuel is stored under water in a large pool. The reference wet ISFSI is tia installation at the General Electric Company's Morris, Illinoic plant. The reference wet ISFSI has a capacity of 2000 metric tons of fuel. The facility is a below ground, open pool, four-section water basir for receiving and storing spent nuclear fuel.

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The major structures of the reference wet ISFSI includes: (1) the rain building which houses the cask receiving and decontamination areas, fuel unloading basin, control room, fuel storage facilities (i.e., the storage basins), basin support systems,'and low-activity waste facility; (2) the ventilation filter building; (3) the plant stack, and (4) other minor support structures.

13.1.2 Dry Storage ISFSI 13.1.2.1 Reference Drywell ISFSI In this design, spent nuclear fuel is stored dry in individual wells or caissons.

The top of the well, located near ground elevation, is covered with a shielding plug. The drywell ISFSI uses the physical properties if soil to provide both a thermal sink and radiation shielding while the spent fuel is stored in under-ground drywells. The spent fuel is sealed in canisters that are placed in the drywells.

The reference drywell consists of three major components: (1) a hot cell facil-ity for receiving, inspecting, and packaging spent fuel; (2) an onsite trans-porter; and (3) the fuel storage area which holds an appropriate number of under-ground drywells. The reference drywell ISFSI has a fuel storage area which is 16 hectares in area that will accommodate 4705 drywells holding 2000 MT or more of spent fuel as well as the other necessary components. The drywells within the fuel storage area are laid out on a grid that allows the necessary heat dissipation. The hot cell facility includes an area to receive pent fuel,'

an area to receive containers to be filled with spent fuel, an operating floor for packsging the fuel, a room for equipment maintenance, underground transfer tunnels, a control rcum, and a waste handling facility.

13.1.2.2. Reference Silo ISFSI In this design, packaged spent fuel assemblies are stored in cylindrical, above-ground concrete silos. The silo ISFSI uses the massive thickness of the concrete silo for shielding and uses convective air currents for heat dissipation.

Individually packaged spent fuel assemblies are placed in baskets in the silos.

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The reference silo ISFSI consists of three major comporants: (1) a hot cell facility for receiving, inspecting, and packaging sent fuel; (2) an onsite trans-po-ter; and (3) a fuel storage area that holds an appropriate number of silos.

The reference silo ISFSI has a fuel storage area which is 16 hectares in area that will accommodate 1120 silos holding 2000 MT of spant fuel as well as the other necessary components. The silos within the fuel storage area are laid oct on a grid that allows the necessary heat dissipation. The hot cell is as described above in 13.1.2.1.

13.1.2.3 Reference Vault ISFSI In this design, cpent nuclear fuel is stored dry in arrays of tubes housed in massive concrete buildings of enclosed canyon structures. Convective or forced air flow rovides r cooling for the stored fuel.

The reference vault ISFSI consists of two major components: (1) a fuel storage area and (2) a hot cell. The fuel storage area is below ground canyon-type construction, w!th encapsulated fuel and ratural convective air circulation.

It has a capacity of 2000 MT of spert fuel. The Mt cell is as described above in 13.1.2.2.

13.1.2.4 Reference Cask ISFSI l In this design, spent nuclear fuel is stored in iron casks whose designs are .

similar to those commonly used for rail transportation of spent fuel in the U. S. The casks use thick tron-wall construction for shielding and use iron l heat transfer fins for heat dissipation.

The reference cask ISFSI basically consists of a fuel storage area that contains a collection of specially shielded and sealed metal containers (casks) that l are located in a security area on the generic site. The casks themselves serve l as the storage medium. The reference facility has a capacit,y of approximately 2000 MT.

(

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l L 13.2 ISFSI Decommissioning Experience l

At present, no ISFSIs have undergone decommissioning in the U.S. However, de:ommissioning information is available from facilities that have similar structural or design use characteristics, such as reactors and nuclear feel handling facilities since portions of their decommissionings are similar in nature to that anticipated for ISFSIs.

The PNL reports on decommissioning PWRs 2 and BWRs3 contain information on the technology, safety and costs of the decommissinning of, among other things, the spent fuel pool and ancillary equipment located at the reactor. Many sim-ilarities exist between this facility and the reference wet ISFSI.

Data taken from the GE Morris Plant have indicated that Co-60 radionuclide levels in the storage pool have varied between 1 x 10 8 and 2 x 10 4 pci/ml and that radiation surveys of fuel baskets indicated doses of 50 mrem /hr or less.

There has been de:ommissioning experience at several not cells which, as dis-cussed above in 13.1, form part of certain of the dry reference ISFSIs. This experience includes hot cells and transfer tunnels at the Santa Susana Field Laboratory, the Hot Fuel Examination Facility hot cell at Idaho National Engi-l neering Laboratory, and the hot cell facility at the Canyon Building at Oak Ridge National Laboratory. Activities at these facilities included decontamination and dismantling of the hot cell facilities.

13.3 Decommissionina Alternatives Once an ISFSI has reached the end of its useful operating life it must be decom-missioned. As discussed in Section 2.3, this means safely removing the facility from service and disposing of all radioactive materials in excess of levels which r uld permit unrestricted use of the facility. Several alternatives are coniidered here as to their potential for satisfying this general requirement for decommissioning. The decommissioning alternatives considered and discussed tiere are DECON, SAFSTOR, and ENTOMB. The alternative used depends on such con-siderations as cost, dose, physical design of the facility, types of residual 10/08/87 13-4 NUO586 CH 13

~

radioactivity present, propose d use of the site, and desirability of terminating the license.

Discussion of the decommissioning alternatives follows:

13.3.1 DECON DECON is defined as the immediate removal and disposal of all radioactivity in excess of levels which would permit release of the facility for unrestricted use. Nonradioactive equipment and structures need not be torn down or removed as part of a DECON procedure. The end result is the release of the site and any remaining structures for unrestricted use shortly after the end of facility operation. DECON is estimated to be completed in 13, 24, 10, 10, and 6.5 months for the reference wet, drywell, silo, vault, and cask ISrSIs respectively.

DECON is advantageous because it allows for termination of the NRC license shortly after ccssation of facility operations and removes a radioactive site.

DECON is advantageous if the site is required for other purposes or if the site is extremely valuable. It is also advantageous in that the facility operating staff is available to assist with decommissioning and that continued surveil-lance is not required.

To accomplish DECON, all potentially contaminated systems must be disassembled and removed and all contaminated material must te removed from the facility and .

be transported to a regulated disposal site. The simplicity of ISFSI design and the low levels of surf ace contamination anth.ipated to remain after opera-tions are terminated make the DECON alternative advantageous. It appears that l the decommissioning of an ISFSI can be done with a relatively small commitment

( of resources, thereby encouraging the selection of DECON as e decommissioning l

alternative.

Analyses of radiation exposure and costs for DECON are presented in i

Section 13.3.4.

l l

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13.3.2 SAFSTOR SAFSTOR is defined as those activities required to place (preparation for safe storage) and maintain (safe storage) an ISFSI in such condition that the risk to safety is within acceptable bounds, and that the facility can be safely stored and subsequently decontaminated to levels which permit release of the facility for unrestricted use (deferred decontamination).

SAFSTOR is normally considered to be an acceptable decommissioning alternativt, for facilities that contain short-lived radionuclides so that the residual radio-activity decays to acceptable levels for unrestricted release within a period of a few years. Even if unrestricted levels are not reached by decay alone, SAFSTOR might be acceptable for ISFSIs if the decay of snort-lived radioactivity is followed by decontamination to remove the remaining long-lived radionuclides.

A disadvantage of SAFSTOR is that the licensee is required to maintain a material license and to meet its requirements at all times during safe storage thus con-tributing to the number of sites dedicated to radioactive confinement for an extended time period. Other disadvantages are that surveillance is required, the dollar costs are higher than for DECON, and the experienced operating staff may not be available at the end of the safe storage period to assist in the deferred decontamination.

l

! Analyses of radiation exposures and costs for SAFSTOR are presented in

' ~

Section 13.3.4.

13.3.3 ENTOMB i

ENTOMB requires encasem, *. of the ISFSI in concrete to protect the public from radiation exposure until the contained radioactivity has decayed to levels per-mitting release of the facility for unrestricted use. The relatively low levels

of radioactive contamination anticipated to be present in retired ISFSIs, cc,upled with the physical designs of the facilities, makes ENTOMB an unlikely choice for decommissioning for most of the reference ISFSIs. The use of ENTOMB for a drywell or silo facility appears untenable. The cur.struction of an i

above ground entombment structure vould not give the required assurance that i

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radianuclide leakage would not occur. 'NTOMB is generally consider 6' a method for consolidating radioactive materials within a single structure that can be set aside until radioactive decay has reduced radionuclide levels to those acceptable for unrestricted use. The wide dispersior, of individual drywells or silos around an ISFSI site makes such a decommissioning alternative not viable.

Similarly, use of ENTOMB for a cask ISFSI does not appear viable.

From the standpoint of physical design, ENTOMB is a potential alternative when a concrete monolith is already utilized as part of the operational features of a fa:ility. Entombment would be accomplished by sealing the entrances to the exis'ing facility. Howeve- ENTOMB at a wet ISFSI would require either the expct.se of filling the pool completely with concrete or canstructing a structur-ally sound thick concrete cap across the pool and hence would not appear to be a viable alternative.

Analysis of radiation exposurcs and cost for the case for which ENTOMB was examined in Reference 1 are presented in Section 13.3.4.

13.3.4 Sumary of Radiation Safety and Decommissioning Costs 13.3.4.1 Radiation Safety Estimates are n;ade of the external occupational radiation doses that are accumu-l lated by workers conducting decommissioning tasks. The dominant radioactive ,

l species contributing to occupational exposure during DECON is Co-60 and the dominant species af ter 10 years of SAFSTOR will be Cs-137. Occupational radia- '

tion exposure of workers performing the decommissioning activit'es results from external exposure for reasons similar to that discussed fe,r PWRs in Section 4.3.1.

Table 13.3-1 presents a summary of the radiation exposures that may result for the decommissioning alternatives cr.asidered in Ref1rence 1. The dose resulting from ENTOMB at a vault ISFSI is estimated to be 45.5 man-rem, not including additional doses accumulated from swveillance and maintenar.:e or potential I delayed decontamination of the facility. The dose to the public from routine effluents de-ing decommissioning activities for any of the reference ISFSIs is less than 1 x 10 5 man-rem for any of the decommissioning alternatives considered.

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Table 13.3-1 Summary of occupational radiation safety analyses for routine decommissioning of the reference ISFSIs (man-rem)(a)

SAFSTOR Type DECON 10 Yr 30 Yr 50 Yr 100 Yr Wet 53.4 35.1 28.0 27.6 27.6 Drywell 120.0 62.3 33.3 24.7 16.6 Silo 116.0 60.4 32.5 24.3 16.4 Vault 155.0 86.5 50.5 41.9 34.2 Cask 12.5 -- -- -- --

(a) Adapted from Reference 1.

Thus the estimated public exposures from decommissioning activities are very small.

Radioactive waste materials are packaged and shipped offsite for burial during decommissioning of the reference ISFSIs. The dominant radiation exposure path-way to transport workers and the public during transportation of radioactive wastes is external exposure for reasons similar to those discussed for PWRs in Section 4. 3.1. The external dose for transportation is conservatively estimattJ to be less than 0.28, 0.073, an 0.14 man-rem to transport workers for DECON, SAFSTOR, and ENTOMB, respectively, for any of the reference ISFSIs, and less than 2.7 x 10 2, 7,1 x 10 3, and 1.4 x 10 2 man-rem to the public for DECON, SAFSTOR, and ENTOMB, respectively, for any of the reference ISFSIs. These doses, are based on regulations of the Department of Transportation governing radiation levels in shipments and on estimates of distances of travel and lengths of time of exposures that workers and the public might expect.

13.3.4.2 Decommissioning Costs Table 13.3-2 summarizes the estimated costs in 1986 dollars for the decommission-ing alternatives analyzed. All cost estimates include 25% for contingencies.

The cost of SAFSTOR for the wet ISFSI include $2.79 million for preparations,

$93,700 per year for continuing care during safe storage, and the remainder of the cost during deferred decontamination. The cost of SAFSTOR for the drywell 10/68/87 13 8 NU0586 CH 13

Table 13.3-2 Summary of estimated costs for decommissioning the reference ISFSIs(a) (3gg))4,g3)

SAFSTOR Type DECON 10 Yr 30 Yr 50 Yr 100 Yr Wet 7.18 9.63 9.69 10.34 14.62 Drywell 16.65 18.28 20.57 22.86 28.60 Silo 4.44 6.04 8.34 10.65 16.40 Vault 3.90 6.59 9.56 12.56 20.34 Cask 2.25 -- -- -- --

(a) Adapted from Reference 1. Values include a 25% contingency a,1d are in constant 1986 dollars.

ISFSI includes $1.92 million for preparations an. $114,650 per year for con-tinuing care; for the silo ISFSI, the costs include $1.91 million for pre-parations and $115,100 per year for continuing care; and for the vault ISFSI the costs include $2.43 million for preparations and $156,125 per year for continuing care. All the costs include the costs of decommissioning all com-ponents associated with the ISFSI as described above in Section 13.1. For axample, the costs for a wet ISFSI include costs for decommissioning the fuel storage area r ad associated equipment and strectures, while costs for a drywell ISF51 include costs for decommissioning the fuel storage area, the hot cell, and the transporter. The cost for ENTOMB at the vault ISFSI is $2.8 million, plus a cost of $31,740 per year for annual surveillance and maintenance.

l .

13.4 Environmental Consequences j

13.4.1 Waste Disposal l

l The volume of low-level radioactive waste to be dispased of for DECON is estimated to be 2720, 6700, 920, 500, and 42 a8 for the wet, drywell, silo, vault, and l cask ISFSIs, respectively. The volume of waste for SAFSTOR at the dr>vell and l silo ISFSIs is not expected to decrease below that for DECON because the dose

( rates for the contaminated drywells and silos do not decay to low enough values

to permit release of these materials to unrestricted use and hence they must be

! disposed of in much the same manner as for DECON. Waste volumes at the wet and 10/08/87 13-9 NUO586 CH 13 i-

vault ISFSIs will decrease due to reduced quantities of radionuclides and cor-responding waste quantities and for the wet ISFSI are estimated to be 1460, 620, and 350 m3 for 30, 50 and 100 years respectively, and for the vault ISFSI are estimated to be 440, 400, and 390 m3 for 30, 50, and 100 years, respectively.

The waste volumes requiring burial would represent a use of less than one acre of' land for the disposal. This amount is not large in comparison with the size of the JSFSI site (which is approximately 100 acres) which could now be returne '

to unrestricted use.

13.4.2 Socioeconomic Effects The socioeconomic impacts are mainly from the shutdown (not decommissioning) of the storage facility, which would reduce the income of the community and region because of the loss of about 33 to 40 jobs.

In decommissioning a' ISFSI, many of the same people that operated the plant can do the cleaning, but the dismantling and moving of equipment will be done by electricians, plumbers, mechanics, and equipment operators most of whom will be hired or contracted. The socioeconomic effect of decommissioning then, will come from the employment of these craftsmen. The total decontamination crew may be larger than the operating crew, and, if so, for the period of decon-tamination, the economic input to the community will increase. In the case of safe storage, the work force may decrease to a security and maintenance cr?w for the period of continuing care. Because of the planning time needed to precede the deccmmissioning, changes in the number of employees will not be sudden or without warning, and people will have time to find other employment.

l 13.5 Comparisor of Decommit,sioning Alternatives 3

! The decommissioning alternatives eventually end with the same results: a decon-taminated facility that can be released for unrestricted use. The' choice of an alternative generally depends on such considerations as dose, cost, the physical i design of the facility, the desirability of terminating the license, and availa-bility of waste disposal capacity. Based on the relatively simple design of the ISFSI, the low levels of surface contaminatior, anticipated to remain after 10/08/07 13-10 NUO586 CH 33

l operations are terminated, and the fact that occupational doses at the refererce ISFSIs are much less significant and much easier to manage than for power reae-l tors, DECON appears to be a more advantageous option. DECON also costs less than the SAFSTOR options. SAFSTOR may be justifiable in some cases where there l is a problem with off-site waste disposal since there is some reduction in occu-pational exposure for ISFSIs and reduction in waste disposal volumes for certain

( types of ISFSIs. ENTOMB is not expected to be viable for ISFSIs both because l of the physical design of the ISFSIs and because of the long-lived radionuclides at the ISFSIs which would mean that there would have to be maintenance and sur-

. veillance at the facility well beyond the time that required institutional con-trol could be expected to be effective (approx!utely 100 years is considered to be consistent with recommended EPA policy on institutional control for radio-

'tivity confinement). Also, there does not appear to be a cost or safety am,antage to ENTOMB, because when the costs of maintenance and surveillance are included the total cost of ENTOMB soon becomes larger than DECON and the occupa-tional exposure is approximately the same as 30 or 50 year SAFSTOR. Hence ENTOMB is not expected to be viable for ISFSIs.

(

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REFERENCES I

1. J. D. Ludwick and E. B. Moore, Technology, Safety and Costs of Decommis-sioning Reference Independent Spent Fuel Storage Installations, NUREG/CR-2210, Prtpared by Pacific NorthwestAaboratory for the U.S.

,, Nuclear Regulatory Commission, January 1984.V

2. R. I. Smith, G. J. Konrek, and W. E. Kennedy, Jr. , Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station,-NUREG/CR-0130, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, June 1978 P*

3. H. D. Oak, et al., Technology, Safe h and Costs of Decommissioring a Reference Boiling Water Reactor Power Station, NUREG/CR-0672, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, June 1980.*

p H. ,

fCopiesofallreferenceddocumentsmaybepurchasedthroughtheU.S.

Government Printing Office by calling (202) 275-2060 or by writing to the i U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082.

Copies may also be purchcsed from the National Technical Information Service, U.S. Department of Commerce, 5285 Port Royal Road, Springfield, VA 22161. A copy is avail 6ble for inspection or copying for a fee in the

, NRC Public Document Room, 1717 H Street NW., Washington, OC 20555.

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14 NON-FUEL-CYCLE NUCLEAR FACILITIES Non- fuel-cycle facilities are those facilitios which handle byproduct, source and/or special nuclear materials but which are not involved in the production of power as outlinet, in Figure 2.1-1 of Section 2 of this EIS. These non-fuel-cycle facilities must be licensed by the NRC or the Agreement States.

There are thousands of non-fuc~i-cycle facilities in the United States at which byproduct, source, and special nuclear materials are handled under specific licenses of the NRC and the agreement states. These facilities house operations that vary from the occasional use of a few short-lived radionuclides by a doctor to the larg6 scale processing of radioactive materials (gaseous, liquid, and particulate forms). The operations include a wide range of applications in industry, medicine, and research such as manufacture of smoke detectors, radiation therapy equipment, and manufacturing quality control instruments.

Tables 14.0-1 and 14.0-2 give the number of NRC specific material licenses and of agreement state licenses, respectively as of June 1978. Approximate numbers of those which are not connected with the fuel cycle are given in parentheses in Table 14.0-1. These numbers do not exactly represent the number of existing facilities since some of the commercial establishments are licensed under more than one part of the regulations and thus have more than one license. -

A large majority of the non-fuel-cycle material licensees have facilities which do not require a major decommissioning effort. However, a few of the non-fuel-cycle facilities will require significant decommissioning procedures which may present some unique problems and which may have rather large decommissioning l costs and significant environmental impacts. A detailed technical report on the decommissioning of non-fuel-cycle nuclear facilities 1 has been prepared and published in February 1981 by Battelle Pacific Northwest Laboratories and is the basis for the information in this section. The emphasis in that report, and in this EIS, is on some selet.ted facilities which are considered to involve significant decommissioning activities. Examples of these facilities are:

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v Table 14.0-1 NRC Materia' Licenses as of June 1978 Byproduct Medical 2,239 Academic 384 Industrial 4,205 Livil Defense 104 Other 27 Total Byproduct 6,959 (6,924)(a) 5.urce 400 ( 332)

Special Nuclear 720 ( 583)

Total 8,079 (7,839)

C') Licenses not connected with the nuclear fuel cycle are in parenthesis. These numbers were obtained by subtracting fuel cycle facilities and also export /

import licenses which are, in effect, paper trans-actions and do not represent separate facilities.

Table 14.0-2 Agreement State Licenses (June 1978)

Medical 4,749 Academic 867 Industrial 5,030 Civil Defense 185 ,

Other 900 Total 11,731 10/08/87 14-2 NUO586 CH 14

manufacturers of sealed sot rces, manufacturers of radiochemicals, research and development institutions, and processors of ores in which the tailings Contain licensable quantities of radionuclides. Costs and radiological impacts of decom-missioning have been estimated for individual reference facilities such as labor-atories for the manufacture of labeled compounds. One licensee's facilities may include a number of such individual facilities. Decommissioning of reference site components has also been studied.

14.1 Facilitier Descri9tions 14.1.1 Selected Types of Materials Facilities Brief descriptions of selected types of non-fuel-cycle nuclear facilities are given in the following subsections. Reference individual facilities and sites have been selected which are representative of facilities for these types of operations in order to facilitate estimates of costs and radiction doses due to decommissioning. Descriptions of these individual reference facilities and sites are given in 14.1.2.

14.1.1.1 Sealed Source Manufacturer Sealed sources are manufactured for such uses as reference standards, moisture probes, quality control instruments, therapy units, and smoke detectors. In general, these uses require long-lived isotopes, but fairly weak sources, except for 80Co therapy units in which high-energy, high-intensity gamma ray einission is the most important consideration. The manufacturing process is a hand opera-tion that does not lend itself to mass production. Alpha and beta emitters are plated on platinum, stainless steel, or aluminized mylar film and mounted in aluminum rings to form standard disc sources. Liquid gamma sources are sealed in plastic or glass vials, and solid gamma sources are mounted in rods or plastic discs.2 The materials are handled in hoods, glove boxes, or hot cells, depending on the kind and energy of emissions (exposure potential of the isotope).

Contaminated glassware and equipmer.t that cannot be economically reclaimed are discarded into drums for shipment to a waste burial ground. Spills are cleaned up when they occur, and the area and equipment are monitored regularly.

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Ventilation syftems utilize absolute filters, and contamination is thus generally confined to the interiors of the hoods, glove boxes or cells, and the ducts and filters.

The reference individual facilities in sealed source production are based on l

laboratories at New England Nuclear Corporation (NEN) of Boston, Massachusetts.

! NEN has manufacturing facilities at both Boston and Billerica, Massachusetts.

I These buildings contain a number of small laboratories, each of which is devoted to a specific process ar.d/or isotope. Each laboratory contains one or more hoods, glove boxes, and/or hot cells. People entering the laboratcry areas change shoes or put covers over their shoes; when exiting, they change again and monitor their har.ds and shoes for radioactivity. Radioactive wastes are placed in drums and stored in separate building: until shipped to a waste burial ground or, in the case of short-lived isotop like 32P, the drums are held on the premises until the isotope has decayed to a suitable level of activity.

14.1.1.2 Radiochemical and Radiopharmaceutical Manufacturers Manufacturing facilities for radioactively labeled chemicals and pharmaceuticals are auch the same as those for the manufacture of sealed sources in that opera-tions are carried out in ventilated enclosures. Chemical manufacturing, however, requires more extensive and complicated laboratory equipment to perform the inorganic reactions and organic syntheses. The isotopes are either shipped in from an outside supplier or are produced in onsite cyclotrons.

The basis for reference Individual facilities for the manufacturing of labeled chemicals is also New England Nuclear Corp. Chemical syntheses are carried out at both their Boston and Billerica plants. The physical facilities for these operations are similar to those for sealed source manufacturing.

Syntheses are performed in small batches in hoods, glove boxes, or hot cells equipped with absolute filters. Each chemical is produced in a separate labora-tory, which is a restricted area. As compounds progress through their synthesis, they are moved from hood to hood through connecting doors and are packaged in lead shipping containers before being removed from the hood. Radioactive solid waste, including glassware, is placed in plastic-lined drums for disposal.

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E fore being removed from the restricted area, liquid wastes are put in leak-proof, shatterproof containers filled with absorbent materials and are labeled as to quantity, type of activity, date, and sur' ace dose rate.

All wastes are placed in drums and moved to a separate building where the short-lived isotopes, such as 32P, are allowed to decay to negligible levels. Wastes with long-lived isotopes are shipped to waste burial grounds.

14.1.1.3 Broad Research and Development (R&D) Program Facility R&D facilities using nuclear materials cover an e; tremely broad range of activi-ties. A large university is representative of many of these R&D activities.

An example is the University of Washington in Seattle, Washington. There are about 400 laboratories or health treatment areas on the university campus that have used or are using radioisotopes. Radioisotopes are used in chemisuy and physics laboratories to conduct basic experiments and in biological laboratories to investigate absorption and metabolic phenomena. These laboratorics, in gen-eral, present no decommissioning problems because the isotopes used are short-lived and are of low activity. The university also uses radioisotopes for vari-ous medical purposes in a university hospital and a health services complex.

These uses include both radiation exposure from sealed sources and injections of short-lised isotopes. Most of these isotopes are produced elsewhere, but l SSTc is produced from SSMo in a technetium generator.

Probably the highest intensity source used is the sealed 80Co source used in biological irradiation studies of fish. This source is on the order of 40,000 Ci, so shielding requirements are extensive, and these shielding requirement:; must be considered in decommissioning activities.

The longest lived isotopes normally used are 8H and H C, both of which are low-erwrgy beta-emitting isotopes. Other isotopes that are commonly used as tracers include itsy, ssp,, asC1, zeAj, ssCr, and ass. Radioactive wastes are packaged for shipment to a waste burial ground.

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A reference institutional laboratory has been studied. It was not taken directly from the University of Washington but is a small complex of rooms designed to represent the types of f acilities typical of an R&D facility.

14.1.1.4 Ore Processors Non-fuel-cycle processing facilities that deal with ores containing appreciable concentrations of radionuclides are licensed to store their mill tailings.

There are relatively few such facilities in the U.S. , but the volumes of tailings they generate are sufficient to require a significant decommissioning effort, The reference rare-metals refinery is a plant that refines raw material for the recovery of the tantalum and niobium. The raw material is the slag produced by tin smelters located on the Malay Peninsula. This slag consists of glassy flakes or pellets that contain 0.1 to 0.5 wt % U 0s3 and Th02 . In one building the slag is ground. roasted, and digested with hydrofluoric acid.

The hydrofluoric acid is filtered off and passed to a facility for the chemical extraction of niobium and tantalum. The sludge, which contains essentially all of the thorium and uranium, is pumped to a settling pond located about 100 m from the refinery. In the settling pond, the water is allowed to evaporate, converting the sludge to a glassy solid.

At some facilities the settling pond is unlined. At newer facilities it is lined with a fluorocarbon-type material, and at one facility the tailings are dried and stored in above ground concrete buildings.

In such a facility, the radioactivity is primarily in the tailings, nowhere else in the operation is there significant radioactive contamination. Costs for decommissioning the remainder of the facility and site would be primarily that of the termination survey. The operational problem is that there is cur-rently no satisfactory place to ship the tailings for disposal. Storage in specially made aboveground structures becomes expensive and cumbersome, and in addition, the operating license may limit the amount of tailings that can be stored onsite.

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5ince the main decommissioning task involves the disposition of the tailings pile or pond, a reference ore tailings pile has been studied.

14.1.2 Reference Facilities and Sites 14.1.2.1 Radioactive Material Processor Laboratories Five example laboratories for the manufacture of sealed sources or radiochemicals were included in the PNL study,1 each limited to the processing of one radio-active nuclide:

1. 3H Laboratory - The reference laboratory for the manufacture of 3H - labeled compounds is 120 m2 in area and contains five fume hoods and six glove boxes, each separately vented through roughing and HEPA filters, 20 linear meters of laboratory workbenches, refrigerators, a freezer, and a storeroom.

2. 14C Laboratory - The reference laboratory for the manufacture of 14C -

labeled compounds is 80 m2 in area and has four fume hoods, four glove boxes, 15 linear meters of workbenches limited to nonradioactive opera-tions, refrigerators, freezers, a storage room, a sink through which low levels of radioactivity are discharged to the sanitary sewer, and vacuum manifolds and distillation equipment typical of an organic chemistry laboratory, i

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3. 12sl Laboratory - The reference laboratory for the manufacture of 12s] .

1abeled cumpounds is 48 m2 in area and has four fume hoods with a specially designed glove box located inside each hood. Each hood and glove box is equipped with an activated charcoal filter. In addition, there are 8 linear meters of workbenches, a refrigerator, storage cabinet, and a sink. Liquid l effluent from this sink is discharged to a tank where it is held for decay, monitored, and diluted before discharge to the saattary sewer; wastes from l processing operations are not discharged to the sink, but are packaged and shipped to a commercial shallow land burial site for disposal.

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4. 137Cs Laboratory - The reference laboratory for the manufacture of 137Cs sealed sources is 48 m2 in area and has two small hot cells, two fume hoods, 4 linear meters of workbenches, locked storage casks, and a sink with holdup tank which is not used for discharge of wastes from processing.

5. 241Am laboralory - The reference laboratory for the manufacture of 241Am sealed sources is 60 m2 in area and contains two fume hoods, seven glove boxes, 6 of these being connected in series by transfer tunnels, a storage cabinet for nonradioactive supplies, a small workbench, and a change area.

14.1.2.2 Institutional User Laboratory The reference institutional user laboratory is representative of the type of facility a broad research and development licensee might have. It contains a room for synthesizing labeled compounds and for preparing radioactive samples, a small-animal laboratory, a counting room, office space, and an equipment and storage room. The radioisotope room is approximately 49 m2 in area and contains a glove box, three fume hoods, two sinks, a lead storage unit, a refrigerator, and workbenches. The animal laboratory contains two fume hoods, a sink, animal cages, and workbenches.

14.1.2.3 Reference Sites Three examples of contamination onsite were studied:

1. An underground drain line and holdup tank - 20 m2 of o,1.m - diameter cast iron pipe and 1.5-m diameter by 2 m2 high cylindrical steel tank buried 1.5 m below the ground surface.

2. Contaminated ground surface - a 40,000 m2 site with 1000 m3 of soil contaminated with residue from uranium processing operations.

3. Rare metals refinery tailings pile - an unlined settling pond 100 m long by 50 m wide by 5 m deep with a 2\ to 1 slopi. on each side dug into a clayey silt on a 20,000 m2 site.

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t 14.2 Non-Fuel-Cycle Materials Facilities Decommissioning Experience Decommissionings of non-fuel-cycle facilities have been many and varied, and a 1

large number of these operations have had little cost or environmental inpact.

Because of their unique sizes, locations, and conditions, no two facilities had identical decommissioning problems or conditions. Documentation on these decom-missionings is fragmentary. However, a number of things, as discussed below, are apparent from the documentation that is available on the decommissioning of these facilities.

First, a large variety of facilities, both commercial and others, have been successfully decommissioned without unreasonable occupational exposures or significant public exposures. The decommissioning approach has generally been to decontaminate the facility to radioactivity levels low enough to permit release of the facility for unrestricted use.

Each facility can present problems that are unique to its decommissioning. In some cases, these problems can lead to uncertainties in estimating costs for decommissioning, even at the time of shutdown. This is particularly true for a facility where a number of operations involving processing of a variety of nuclides have been carried out and an adequate history of operations and events has not been documented. However, what is also apparent is that the same basic approach to decommissioning applies to all facilities and that knowledge obtained from experience in decommissioning, in general, including some methods of

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facilitation can be applied as appropriate to any facility.

There has also been some decommissioning experience specifically relevant to the types of facilities chosen as references. Manufacturers of sealed sources and labeled chemicals carry out their operations in small batches in glove boxes, hoods, or remote operation cells, and contamination outside these structures is limited almost entirely to the ventilation ducts and filters. The isotopes creating the worst problems in these facilities are "C, which requires tedious inspection and cleanup, 8H, which is easily dispersed and requires many washes to remove; and gases of 12sy, tall, and asKr. Equipment for handling cesium and strontia becomes so thoroughly contaminated that it is normally sent to waste burial without any attempt to clean it up.

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New England Nuclear Corp. has had a great deal of experience w th these kinds of structures and has decommissioned an entire five-story building plus basement which had contained biochemical and organic chemical laboratories for the manu-f acture of compounds labelled with 3H,14C, and 32P, and is now being put to other, non-nuclear uses. Decommissioning criteria used by NEN are given in Ref. 3. This decommissioning consisted of removing all the isotope-handling equipment and ventilation ducts, decontaminating them when possible, and if not economically recoverable, disposing of them to low-level waste burial grounds.

In practically all cases, it was not considered economically feasible to decon-taminate ductwork. The entire facility was surveyed for radioactivity and any areas with contamination levels of 900 or more dpm per 100 cm 2 were cleaned to reduce contamination by at least a factor of 2. The walls and ceilings were steam cleaned. The floors consisted of vinyl tile laid over plywood on top of the original floor. Where contamination occurred, the floor tiles were replaced and, if necessary, sections of the plywood were cut out and replaced. Some of the worst areas of contamination were under the laboratory benches, which were not accessible for routine cleaning. Glove boxes that were not to be reclaimed were spray painted, loaded with contaminated equipment, filled with a quick-setthg foam material, and shipped to a low-level waste burial ground. Lead bricks were etched with hcl, and areas contaminated with liC were washed with NaOH and NH40H. These same procedures are followed on a continuing basis as NEN rearranges and remodels other laboratories.

Experience with decommissioning of commercial non-fuel-cycle ore processing facilities is limited, primarily because there are few such facilities in the U.S. The ores handled in these facilities have such low levels of radioactivity that the machinery can be readily decontaminated and surveyed to confirm that radioactivity levels are low enough to allow unrestricted use. Therefore, the main problems with decommissioning are disposal of the slag or tailings and cleaning up of spills. Kawecki Berylco Indur,tries, Inc. has one such site in which the contaminated surface soil was scraped into a single pile and stabi-lized with vegetation. The matter of final disposition of the sludge from current operations containing the unextracted uranium and thorium has not been re solved.

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Also relevant to the decommissioning of this typt of facility is the ongoing work to decontaminate some sites which had been used some time ago fu similar processes and subsequently abandoned. Two of these are: Reed Keppler Park in West Chicago, where thorium-containing wastes from a rare earth processing plant had been deposited in the 1940s, and a plant in Parkersburg, West Virginia, where ore had been processed for the recovery of zirconium and hafnium.

Experience in dealing with uranium mill tailings piles is also relevant to decommissioning this type of operation since they present similar problems.

14.3 Decommissioning Alternatives Decommissioning alternatives likely to be used for non-fuel-cycle materials facilities are discussed in the following subsections, first as they apply in general and then as applied specifically to the reference facilities. The general section describes each of the alternatives presented in Section 2.4 as they apply to non-fuel-cycle facilities. The specific section for each reference facility discusses only those alternatives considered viable for that facility.

14.3.1 Decommissioning Alternatives for Non-Fuel-Cycle Facilities Once a non-fuel-cycle facility has reached the end of its useful operating life it must be decommissioned. As discussed in Section 2.3, this means safely removing the facility from service and disposing of all radioactive materials in ,

excess of levels which would permit unrestricted use of the facility. Several alternatives are considered here as to their potential for satisfying this general requirement for decommissioning. The decommissioning alternatives considered and discussed here are DECON, SAFSTOR, and ENTOMB.

Since there is such a large range in the type and size of facilities and opera-tions licensed to handle radioactive materials, the level of effort'. required to decommission these facilities varies greatly. The necessary actions can vary from essentially administrative procedures for small facilities (in addition to a final certification survey which could be similar to operational surveys) to a multi-million dollar effort for the more significantly contaminated facilities.

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For many materials handling facilitis s it may be quite straightforward to deter-mine what actions are necessary; for some, however, detailed consideration of more than one viable alternative may be required. Any of the decommissioning alternatives listed above may be viable for some of the non-fuel-cycle facilities.

For a large number of non-fu::1-cycle facilities some variation or combination of these alternatives will be the best choice. Discussion of the decommissioning alternatives follow.

14.3.1.1 DECON DECON is defined as the immediate removal and disposal of all radioactivity in excess of levels which would permit release of the facility for unrestricted use. Nonradioactive equipment and structures need not be torn down or removed as part of DECON procedures. The end result is the release of the site and any remaining structures for unrestricted use. A large number of non-fuel-cycle facilities will require some positive action in order to reduce radioactivity to levels considered acceptable for releasing the facility for unrestricted use. The procedures necessary for DE',0N vary greatly with the type of f acility and its operation. Any procedure, whether involving only removal of sealed sources, decontamination, or dismantling, will follow the general concepts defined for DECON in Section 2.4.2. DECON can include dismantling, removing, and disposing of any contaminated equipment, as well as decontaminating or removing any contaminated parts of the building.

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For many non-fuel-cycle facilities, the most appropriate decommissioning alternative will be DECON. This will involve decontamination of the f acility; most licensees will not : Teed to dismantle the facility.

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In the case of an ore processing facility, removal of sludge also follows the general concept of DECON. An extension of this option is chemical extraction of the radionuclides, in which case the depleted sludge can be disposed of in l a landfill and the radionuclides taken to a waste burial site or sold.

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14.3.1.2 SAFSTOR SAFSTOR is defined as those activities required to place (preparation for safe storage) and maintain (safe storage) a non-fuel-cycle facility in such condition that the risk to safety is within acceptable bounds, and that the facility can be safely stored and subsequently decontaminated to levels which permit release of the facility for unrestricted use (deferred decontamination).

For some of the materials facilities, SAFSTOR may be an acceptable and desirable decommissioning alternative. The simplest case illustrating the advantage of SAFSTOR would most likely be if most or all of the radioactivity in a specific facility is from relatively short-lived nuclides that will decay to levels permitting unrestricted use of the facility in a short time. In this case, little action, in some cases just a radiation survey, is expected to be required at the time of deferred decontamination. During the safe storage period, the facility would have to be made secure against intrusion. Limited surveillance and monitoring would also be required.

Stabilization may be a decommissioning alternative considered for the tailings pile remaining at ore processing facilities. At this time, the NRC has not determined whether this will be acceptable; but currently its acceptability would be considered on a case-by-case basis. Stabilization of tailings piles would be considered as preparation for safe storage and would require monitoring until final disposition.

14 3.1.3 ENTOMB ENTOMB requires the encasement of a facility in concrete to protect the public from radiation exposure until its radioactivity has decayed to levels permitting unrestricted use of the facility. For a non-fuel-cycle facility, ENTOMB would require the construction of a heavily reinforced concrete building in advance l

of licensing in which the facility operations would be cor. ducted. Given the expense of construction and the low radioactivity level of most of the isotopes l to be handled, ENTOMB does not appear to be a viable alternative, l

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14.3.2 Decommissioning Alternatives for Sealed Source and Radiochemical Manufacturers The same kinds of facilities are used in the manufacture of sealed sources and radio-labeled chemicals. Since the methods for decommissioning these facilities are the same, they are combined in this discussion. The alternatives considered for decommissioning these facilities are DECON and SAFSTOR. These are discussed below.

14.3.2.1 DECON DECON is a logical alternative for facilities such as those of New England Nuclear Corp. which have been established for the manufacture of sealed sources and radio-labeled chemicals. It is relatively uncomplicated, will eliminate a need for continued monitoring, and will release the facility for other uses.

Decontamination activities will include the removal of hoods, glove boxes, hot cells, laboratory benches, and ventilation systems. Room surfaces will be washed and floor coverings removed as needed to eliminate hot spots that may have resulted from spills.

In planning a decommissioning action, it is important to know the history of the operation, how diligent the operators were in keeping the rules regarding con-tasnination and releases, and how good a record of accidents and spills was kept.

Meth'ods of disposal of equipment will depend on what isotopes are involved and on future use of the equipment. Hoods that have been used for strontium and cesium may be so badly contaminated that they cannot be reasonably and econom-ically cleaned for further use. These will be shipped to low-level waste burial.

Other hoods may be decontaminated to a suitable radioactivity level for reuse in a nuclear facility by removing the baffle and washing the hood surfaces, or, if they are easily decontaminated or have been used with short-lived isotopes, they may be cleaned and possibly made suitable for unrestricted use. It may be economically attractive to decontaminate stainless steel equipment by electropolishing.

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Hoods that are to be discarded as low-level waste will be painted to seal in the radioactivity, filled with other contaminated equipment, such as ductwork and filter boxes, and packaged in plywood boxes for shipping to a burial grouhd.

Glove boxes will be filled with a quicksetting foam material, packaged and shipped to a burial ground. Hot cells and manipulators will be disassembled, and compressed into steel drums. The actual handling and disposal methods will depend on the quantity of activity and the radiation characteristics. These methods will also determine the number of barrels needed for packaging, which in turn will greatly influence the disposal cost. An estimate of costs and manpower requirements for decommissioning (by DECON) various individual labora-tories described in Section 14.1.2.1 is shown in Table 14.3-1. Decisions on the extent of dismantling and on discarding specific items will depend on the dollar value of the item and the cost and degree of difficulty of decontaminating it.

These will be case-by-case decisions.

Actual packaging and shipping costs depend on the isotope involved. Iodine hoods, for example, may be decontaminated by wiping, but all the wastes have to be placed in packages that are surrounded by activated charcoal in a steel drum.

Decommissioning costs for manufacturing licensees with a large complex of faci-lities could be in excess of one million dollars.

Exposures to decommissioning workers will depend on the isotopes processed in a particular laboratory and on whether respirators and protective clothing are .

worn. At New England Nuclear, waste barrels are packed to measure no more than 250 mR/tr on the surface, or, if the waste has a very high radioactivity level, the barrel is kept to no more than 5 R/hr and it is kept shielded during hand 1-ing and loading. Exposure of decommissioning workers is generally kept within operational exposure levels, and in no case is a worker allowed to receive more than 300 arem/wceki.

The critical exposure time in decommissioning a laboratory is during the removal of the hoods, ventilation system, and hot cell. During this time, external exposure can be as high as 100 mrem / week. The remainder of the decommissioning 14-15 NUO586 CH 14 10/08/87 v.-

--r . -- , , , . - - - - - - - - - ,-

Table 14.3-1 Summary of estimated requirements and costs for DECON of six reference laboratories that process or use radioisotopes Requirement or Cost for Reference Laboratory (*)

3H 84C 22sy is7c3 24 sam Institutional Laboratory Laboratory Laboratory Laboratory Laboratory Laboratory Parameter 62 61 60 81 70 Time (days) 71 279 235 230 226 336 270 Manpower (man-days) ,

Occupational Dose (b) 0.1 0.001 0.1 6 40 0.1 (man-rem)

Cost ($ thousands)(C}

65.4 55.3 53.6 53.3 78.6 63.3 Staff Labor 4.4 3.9 3.3 6.9 4.0 4.4 Equipment 8.1 10.0 9.2 9.1 11.2 8.9 Supplies 66.4 52.1 39.9 32.2 49.6 52.2 Waste Management 121 106 102 143 129 Totals 144

(*)The listed value represents the requirement of cost for both planning and preparation and the actual decommissioning of the laboratory.

(b) Estimated on the assumption that workers do not use protective respiratory equipment. Doses could be reduced by 1 or 2 orders of magnitude through the use of this equipment. This is a likely alternative for the 24 tam laboratory.

(C) Costs are in 1986 dollars and include a 25% contingency.

time is spent in scrubbing hot spots. During this time, d>se levels are at or below those encountered in operation of the laboratory (about 3 mrem / day).

Occupational dose estimates for the reference individual facilities are also given in Table 14.3-1.

Examples of contamination which might exist at manufacturing facility sites were also considered. The manpower, cost, and dose estimates for decommissioning the reference contaminated sites described in Section 14.1.2.3 are given in Table 14.3-2.

Table 14.3-2 Summary of estimated manpower requirements, costs, and radiation doses for decommissioning three reference sites Requirement or Cost Occupational Time Manpower Cost (*) Radiation Site (days) (man-days) ($ thousands) Oose (man-rem)

Underground Drain Line & 17 72 67 0.04 Hold-up Tank Contaminated Ground 42 203 1889 0.14 Surface Tailings Pile Stabilization Option 32 174 251 0.08 Removal Option 139 1660 32,690 1.0 (a) Costs are in 1986 dollars and include a 25% contingency.

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14.3.2.I SAFSTOR SAFSTOR is a reasonable alternative for decommissioning if the isotopes involved

, at a particular facility are short-lived and the facility has no other immediate planned usage. Use of a safe storage period of a few days to a few months may allow the radioactisity to decay to low enough levels that no further decon-tamination is required and that little action, perhaps only a radiation survey and some administrative action is necessary for releasing the facility for unrestricted use.

14.3.3 Decommissioning Alternatives for Broad Research and Development Program Facilities Decommissioning a large R&D facility is a piecemeal operation because of the many separate working areas involved, although each area is relatively uncom-plicated. The major activity in preparation for decommissioning will be the elimination of inventory. An accurate accountability system is difficult when such a large variety of laboratories and uses may be involved. Some labora-tories may have small amounts of 14C compounds, for example, lef t over from oxperiments conducted several years previously. Preparation for decommissioning must include an exhaustive inventory to discover these. The elimination of any inventory is the next step of decommissioning, which is carried out before the rest of the facility is decommissioned. The decommissioning alternatives considered are: DECON and SAFSTOR. These are discussed below. .

14.3.3.1 DECON A viable alternative for decommissioning an R&D laboratory is DECON. For many of the laboratories, this will not require discarding equipment. Most hoods, glove boxes, and ventilation systems can be decontaminated by washing. For laboratories where long-lived isotopes (8H and 14C) have been used over a period of several years, it may be sufficient to wash and paint the exposed surfaces or it may be desirable to discard some of the equipment as lo -ievel waste. If they are to be discarded, the hoods and glove boxes will be painted to stabilize the surface contamination before aismantling. Ducts and other ventilation 10/08/87 14 18 NUO586 CH 14

equipment parts will be placed inside the *oods and packaged for disposal at a low-level burial site.

Manpower, cost, and exposure estimates for the reference laboratory are included in Table 14.3-1.

A large university such as the University of Washington may have as many as 400 rooms where radioactive material is used. These include preparation rooms, experimental rooms, counting rooms, teaching laboratories, offices and storage rooms.

For many of these where only short-lived nuclides or sealed sources are used the major decommissioning action is a certification survey which would involve a couple ir.an-days of ef fort.

Although it is unlikely that the entire complex would be decommissioned at one

, time, the total impacts for such a decommissioning would be on the order of ten times those of the reference facility with costs in the range of $250,000 -

$1,000,000, and occupational dose of only about 1 man-rem.

14.3.3.2 SAFSTOR For most of the laboratories at an R&D facility, this is the decommissioning alternative most likely to be employed. Except for 3H and 150, the isotopes used at such a facility have short half-lives and a wait of a few days to a few .

month: will allow the radioactivity to decay so that no further cleaning or dismantling is necessary. SAFSTOR assumes either that a laboratory can be left unoccupied for a time or that a survey indicates that the kinds and/or levels of radiation will permit people to work safely in the laboratory. The total cost of decc >nissioning will be that for extensive surveys to monitor decay of the radioactivity. This option will not apply to laboratories with long-lived isotope contamination. For a laboratory that has handled only 3H or 14C, DECON is probably the more viable alternative since these isotopes will not decay for many years. If several isotopes have been used in this same facility, it may be desirable to let the short-lived ones decay before decontaminating.

Personnel exposure under this option will be negligible.

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14.3.4 ' Decommissioning Alternatives for Processors of Radioactive Ore

,The milling of nonradioactive metals by Kawecki Berylco Industries from ores containing uranium and thorium may contaminate the handling or milling equip-ment where the materials are retained by machinery. A simple cleanup and a survey are the only decommissioning actions required. As the materials are processed, most of the uranium and thorium remain with the sludge from the initial extraction, and the following decommissioning alternatives are con-sidered for the sludges: removal (DECON), and neutralization and stabilization for long-term care.

14.3.4.1 Removal (DECON)

A potential decommissioning alternative is removal of the sludge from the milling site and disposal of it at a low-level waste burial ground. The effectiveness of this action could be enhanced by mixing lime into the sludge to neutralize any acid in it before depositing it where it might be contacted by wate . Draw-backs to this option are the great amount of material that must be handled for the sake of a relatively small amount of radioactivity and the long distances that the material must be transported. Costs to transport and dispose of the sludge at a low-level waste burial ground 500 miles away, assuming that there are 90 million pounds of sludge, will be approximately 33 million in 1986 dollars (Table 14.3-2). The costs for transporting and burial are the major costs of disposal. .

Radiation exposure to workers handling this sludge will be very similar to that of people working with uranium mill tailings piles. Radiation levels are 0.5 to 1.0 mrem per hour. Wearing respirators will reduce any problems from inhala-tion of particulates and leave only 222Rn as a concern. Radon levels at the sludge site are also similar to levels at a tailings pile. Exposures and dose estimates to the workers and public are shown in Table 14.3-2.

This sludge could be disposed of in a local landfill if it did not exceed an acceptable residual radioactivity dose limit, which has yet to be determined.

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[

Decontamination ef the sludge by chemical removal of the uranium and thorium seems an attractive alternative, especially if the extraction costs are low enough that sale of the recovered uranium would return a profit or at least reduce the net cost of disposal. Previous milling practices may have affected the chemical nature of the uranium and tharium so that conventional milling methods will be ineffective. Any extraction process would have to remove thorium as well as uranium to make the sludge acceptable at a landfill.

14.3.4.2 Neutralization and Stabilization This alternative is similar to preparation for safe storage and is followed by long-term care. The steps to accomplish this are to remove the roof, cover the pile with lime to neutralize residual acid, cover the entire structure with backfill, add a clay cap, c ver with topsoil, and plant vegetation. The requirements for the kind and depth of cover will be similar to that for uranium tailings piles. However, while uranium mills and their tailings piles are generally located in the semi-arid western part of the U.S., the ore processing plants are likely to be found in areas where humidity and rainfall are much higher and the water table shallower. This will likely increase the need for protection against erosion, but vegetation to stabilize the surface will also grow better in this moister climate. This alternative may not be viable over a long term and would have to be considered on a case-by-case basis. Cost and radiation dose estimates for this alternative are shown in Table 14.3-2.

14.4 Environmental Consequences There are other possible environmental consequences from decommissioning these kinds of facilities that cannot be reasonably discussed on a generic basis but have to be assessed for indiiidual facilities. These include the effects on a local work force and on a local economy. The greatest 9ppacts of this type will have occurred when the operations ceased and the effects of decommissioning will be minor by comparison.

The greatest terrestrial disturbance will come from decommissioning an ore processing facility, because of the large quantity of material involved. The alternative of stabilizing the tailings will require a large amount of earthen 10/08/87 14-21 NUO586 CH 14

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fill, the obtaining of which will necessitate digging up another area. Both the ,tabilized site and the borrow area will likely require reclamation and monitoring to prevent problems with erosion and surface water sedimentation.

Of great concern with these facilities will be potential chemical toxicity from the processing chemicals and mobilized heavy metals in the tailings.

Both occupational and public exposure to radioactivity will be small for decom-missioning a single facility. Although there are a large number of facilities, the potential dose .' rom decommissioning all of the facilities is still expected to be relatively sma31.

14.5 Comparison of Decommissioning Alternatives A comparison of decomr.issioning alternatives is highly specific for each kind of non-fuel tycle f6cility. For most of the facilities that come under this designation, a removal of inventory will eliminate nearly all of the possibility of radiation exposure. The facilities discussed here are those that are perceived to have the greatest need for decommissioning action.

The most likely alternative for decommissioning most non-fuel-cycle f acilities is DECON. In these facilities, radioactive contamination is low. Therefore, cleanup is not difficult. In some facilities, or parts of facilities where only short-lived isotopes have been used, delaying decontamination for a few weeks or months (SAFSTOR) may, allow all the radioactivity to decay and eliminate the need for actual decontamination operations leaving only a final survey to -

be done. Facilities where chemicals and pharmaceuticals have been formulated will require extensive cleaning of the inside building surfaces after the equip-ment has been removed. ENTOMB is not a practical decommissioning alternative for any of the kinds of facilities discussed here.

Stabilization with long-term care may be a viable alternative for disposal of radioactive tailings from an ore processing facility. These tailings are similar to uranium mill tailings and should be subject to the same requirements for stabilizing in place in comparable settings. The disposition of radioactive ore tailings (other than stabilization) has limited possibilities. Removal of the tailings to a low-level waste burial ground will be expensive but is feasible.

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i Reprocessing to remove the radioactive elements from the sludge lacks practicality, mainly because the volumes and rates of production are not attractive to commercial processors.

Although there are thousands of non-fuel-cycle nuclear facilities and the ref-erence facilities discussed here have significant costs and impacts, the overall impact of decommissioning non-fuel-cycle facilities is small. The reference facilities represent only the very few existing facilities which have significant impact while the large majority of the remaining facilities have impacts which are small or nonexistent. For example, approximately half of all the licensees are users of sealed sources and the environmeatal impacts of decommissioning these facilities are negligible. Also, most medical licensees (about 35% of all licensees) are for use of short-lived isotopes (and sealed sources), and the environmental impacts of these decommissionings would in most cases be very small. Hence, because most facilities have small environmental impacts due to

' decommissioning, the cumulative impact of decommissioning all of them is not significant.

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I i-1 REFERENCES >('

1. Technology, Safety, and Costs of Decommissioning ReTerence Non-Fuel-Cycle Nuclear Facilities, Prepared by Pacific Northwest Laboratory for the U.S.

Nuclear Regulatory . Commission, NUREG/CR-1754, February 1981.

2. New Englar.J Nuclear Radioisotope Catalog, New England Nuclear Corporation, Boston, MA.

3. Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material U.S. Nuclear Regulatory Commission, November 1976.

4. Handbook of Radiation Protection, Required Rules and Procedures, New England Nuclear Corporation, 1976.

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a 15 NRC POLICY CONSIDERATIONS At the end of the useful life of a licensed nuclear facility, the facility must be decommissioned. For such a facility, removal of the radioactivity to levels which permit unrestricted use of the facility (including the site) through decommissioning is necessary for full license termination. Present policy and regulatory guidance which addresses nuclear facility decommissioning is not specific enough to adequately assure that this desired objective is accomplished in a manner consistent with protection of the public health and safety. The NRC has been reevaluating its decommissioning policy 3 and considering amending its regulations to provide more specific requirements relating to the decommis-sioning of nuclear facilities. On February 11, 1985, the Commission published a Notice of Proposed Rulemaking on Decommissioning Criteria for Nuclear Facilities.2 Addressed in this notice and the proposed rulemaking are reactors and associated fuel cycle facilities, and non-fuel-cycle facilities. Excluded from specific consideration in this plan and rulemaking are: (1) low-level waster burial facilities, which are separately addressed in regulations in 10 CFR Part 61; (2) Uranium mill and mill tailings, for which a Final EIS8 is currently available and amended regalations have been promulgated; (3) High-level waste repositories, which will be covered in separate rulemaking; (4) Uranium mines and currently existing government owned enrichment plants, which are not under NRC jurisdiction. -

As part of the deconaissioning policy reevaluation and development of a series of NUREG reports (4-26), reports by Battelle Northwest Laboratory, Oak Ridge National Laboratory, other contractors, and by NRC staff have been developed.

These reports are intended to serve as an infomation base for the development of decommissioning regulatory activities and contain information on technology, safety, and costs of decommissioning, on radiation termination surveys, and on financial assurance for decommissioning. In relation to such regulatory activ-ities, an attempt has been made to maintain a dialogue with the public during development of rulemaking. This included public meetings, issuance of a draft environmental impact statement for public coment, and issuance of proposed rules 10/08/87 15-1 NUO586 CH 15

I for public comment. Based on the above information base and on consideration of the regulatory role NRC must provide in protecting public health and safety, the following conclusions appear evident:

(1) The technology for decommissioning nuclear facilities is well in hand and, while technical improvements in decommissioning techniques are to be ex-pected, decommissioning at the present time can be performed safely and at reasonable cost. Radiation dose to the public due to decommissioning activities should be very small and be primarily due tu transportation of decommissioning waste to waste burial facilities. Radiation dose to decom-missioning workers should be a small fraction of their exposure experienced over the operating lifetime of the facility and usually be well within the occupational exposure limits imposed by regulatory requirements. Decommis-sioning costs are reasonable and are, at least, for the larger facilities such as reactors, a small fraction of the present worth commissioning costs (i.e., less than 10%).

- (2) Decommissioning of nuclear facilities is not an imminent health and safety problem. However, planning for decommissioning as an integral activity prior to commissiening as well as during facility life, is a critical item that can have an impact on health and safety as well as cost. Essential to such planning activity is reasonable assurance that funds will be available for performing required decommissioning activities at cessation of facility operation and of the Dcilitation of decommissioning, the ~

decommissioning alternative to be usea, as well as consideration of accept-able residual radioactivity levels for unru tricted use of the facility.

(3) Decommissioning of a nuclear facility generally has a positive environ-mental impact. At the end of f acility life, termination of a nuclear

license is a required objective. Such termination requires decontamina-tion of the facility such that the level of residual radioactivity remaining in the facility or on the site is low enough to allow unrestricted use of the facility and site. Commitment of resources, compared to operational aspects, is generally small.

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'The major environmental impact of decommissioning is the commitme.t of small amounts of land for waste burial in exchange for reuse of the facility and site for other neclear or nonnuclear purposes. Since in many instances, such as at a reactor fa;;ility, the land has valuable resource capability, return of this land to the commercial or public sector is highly desirable. In decommission-ing of nuclear facilities, the objective of NRC regulatory policy is to ensure that. for the comercial sector, proper and explicit procedures are followed in major key areas to mitigate any potential for adverse impact on public health and safety or on the environment.

In the following sections, major recomended regulatory positions are described with resoect to decommissioning alternatives, planning, financial assurance, and residual radioactivity. In the final section, the manner in which such recomendations are to be explicitly incorporated into the regulatory process is discussed. A sumary of the estimated radiation doses from decomissioning and costs of decomissioning for the facilities covered in this EIS is found in Tables 15.0-1 and 15.0-2.

15.1 Major Regulatory Particulars 15.1.1 Decommissioning Alternatives Decommissioning means to remove a facility safely from service and to reduce residual radioactivity to a level that permits release of the property for '

unrestricted use and termination of the license. This can be accomplished by decontamination and dismantling the facility for unrestricted use soon after cessation of operations. Alternatively, in certain situations for certain facilities, where the potential exists for occupational exposure and waste volume reduction (resulting from radioactive decay), or where there is an inability to dispose of waste because of lack of capacity, or for other site-specific factors which may affect health and safety, safe storage or entombment may be feasible.

Categorization of decommissioning alternatives is broken into three major clas-sifications which are referred to in this EIS by the pseudoacronyms DECON, SAFSTOR, and ENTOMB. These terms have been used to discuss potential decommissioning 10/08/87 15-3 NU0586 CH 15

Table 15.0.1 Summary of Estimated Radiation Dose. from Decommissioning Huclear Fuel Cycle Facilities (in man-rem) o SAFSTOR DECO' ENTOMB 10 Years 30 Years 100 Years Occupational Exposure / Facility (a)

Pressurized Water Reactor (PWR) 1,183(D) 652 329 304 920(c 025(d Boiling Water Reactor 1,955 931 442 320 1,624 ,1,753 Fuel Reprocessing Plant 532 453 333 179 175 Small Mixed 0xide Plant 76 165 307 (e) 10 UFe, Conversion Plant 1 1 1 1 (e)

Uranium Fuel Fabrication Plant 18.6 30 62 (e) (e)

Non-Fuel-Cycle Facility 72 77 87 122 Independent Spent Fuel Storage 1,091 621 318 295 900(c)15,1,000(d)

Installation (ISFSI)

Multiple Reactor 72 77 87 122 15 Public Exposure / Facility (a)

)

Pressurized Water Reactor (PWR)

Boiling Water Reactor 21(D) 10 5 7

2 3

2 2

5 h),7, Fuel Reprocessing Plant 19 15 10 4 3 Small Mixed Oxide Plant 4 4 4 1 UFg Conversion Plant 4(I) 0 0 0 0 (e)

Uranius Fuel Fabrication Plant 1 1 1 (e) 0(g)(e)(d)

Non-Fuel-Cycle Facility 0 0 0 0 ,O Independent Spent Fuel Storag? 0 0 0 0 0 Installation (ISFSI)

Multiple Reactor 0 0 0 0 0 I") Data in this table calculated for the reference facilities as defined in the specific '

(b)EIS section for that facility.

Includes does due to transportation of wastes.

(c)With reactor internals included.

(d)With reactor internals removed.

Not calculated.

Means neligible dose.

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Table 15.0.2 Summary of Estimated Costs for Cecomissioning Nuclear Fuel Cycle Facilities (in Millions - based on 1986 Dollars)(a) g SAFSTOR ENTOMB DEC05 10 Years 30 Years 100 Years Facility (b)

Pressurized Water Reactor (PWR) 103.5 97.7 100.5 80.3 70.5(d) 60.2(C)

. Boiling Water Reactor (BWR) 131.8 128.3 131.4 106.1 97.0(d),84.7(c)

Fuel Processing Plant 169.0 181.0 187.0 205.0 (e) (e)

Small Mixed Oxide Plant 13.9 27.6 47.3 (e) 4.9 UFs Conversion Plant IP.1 15.1 17.6 26.4 (e)

Uranium Fuel Fabrication Plant 8.8 15.3 24.7 (e) (c)

(a) Costs for specific facilities are based on References 1 through 8. Table includes costs for equipment, supplies, power, materials, waste, labor and services plus a 25% contingency factor. Costs do not include cost for demolition of nonradioactive structures.

(b) Data in this table calculated for the reference facilities as defined in the specific EIS section for that facility.

(g)With reactor internals included.

(d)With reactor internals removed.

I')Not calculated.

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alternatives in the nuclear facility studies presented in this report. Briefly,  !

l they have the following meanings:  ;

DECON is the alternative in which the equipment, structures and portions of a fa:ility and site containing radioactive contaminants are removed or decontamin-ated to a level that permits the property to be released for unrestricted use shortly af ter cessation of facility operations.

SMSTOR is the alternative in which the nuclear facility is placed (preparation for safe storage) and maintained (safe storage) in such condition that the nuL: lear facility can be safely stored and subsequently decontaminated to levels which permit release of the facility for unrestricted use (deferred decontamina-tion). Depending on the radioactivity level at the end of the safe storage period, decontamination at the final stage may consist of only a radiation survey to verify that the radioactive constituents have decayed to an appropriate unrestricted access level.

ENTOMB is the alternative where at the end of facility life the equipment con-taining radicactive contaminants is encased in a structurally long-lived mate-rial, such as concrete; the entombed structure is appropriately maintained and continued surveillance is carried out until the entombed radioactive contamina-tion decays to a level permitting release of the facility for unrestricted use.

Based on an analysis of the technical data bass, (4-26) decommissioning can ,

be accomplished safely and at reasonable cost sho-tly af ter cessation of facil-ity operation. DECON has certain benefits in that it would prepare the property for unrestricted use in a much shorter time period than SAFSTOR or ENTOMB with acceptable effects on occupational and public health and safety. Completing decommissioning and releasing the property for unrestricted use eliminated the potential problems that may result from an increasing number of sites con-taminates with radioactive material, as well as eliminating potential health, safety, regulatory, and economic problems associated with maintaining the nuclear facility. The use of DECON assumes the availability of capacity to handle waste requiring disposs1. The Federal and State governments have activ-ities undervey to assure that there will be this capacity.

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,1 Delay in the camoletion of decommissioning, as in the case of SAFSTOR or ENTOMB, would be acceptable primarily for reasons for occupational health and safety, since it is recognized that with delay there will be reduction in occupational dose and radioactive waste volume for some nuclear facilities due to radioactive decay. In addition, SAFSTOR may have some advantage where there are other operational nuclear facilities at the same site, and may also become necessary in other cases if there is a shortage of radioactive waste disposal space off-site. The appropriate delay will depend on the type of facility and the con-taminant isotopes involved. One of the difficulties with ENTOMB for any complex structure such as a reactor is that the radioactive materials remaining in the entombed structure would need to be characterized well enough to be sure that they will have decayed to acceptable levels at the end of the surveillance period. If this cannot be done adequately, deferred decontamination would become necessary, which could make ENTOMB more difficult and costly than DECON and SAFSTOR.

The issue of timing concerns what amount of time would be appropriate to allow for completion of decommissioning including the entire period between final shutdown and license termination. The primary consideration is the decay of radioactivity which may result in reductions in occupational exposure and waste needing disposal. Facilities differ regarding the particular radionuclides most critical to decommissioning. For light water power reactors Co-60, with a half-life of 5.3 years, is the nuclide that has the most effect on decontamina-tion efforts and is referred to as the critical / abundant nuclide. Other isotopes that can affect decommissioning efforts are Cs-137 (30 year half-life) ~

and the long-lived isotopes Nb-94 and Ni-59.

As discussed above, a review of the technical data shows that, for DECON, occu-pational exposure can be kept reasonable. For example, studies indicate that occupational doses from decommissioning light water power reactors would be about 300 man-rem per year (1200-1900 man-rem over 6 years for large reactors).

This is generally less than current annual doses at operating reactors. SAFSTOR will result in reduced occupational dose and amount of radioactively contamin-ated waste. Based on the half-life of the critical / abundant nuclide: the reduction of occupational doses beyond about 30 years would be marginally signif-icant although a significant volume reduction in contaminated waste would result 10/08/87 15-7 NUO586 CH 15

~

• om 60 years in safe storage. It appears that DECON or SAFSTOR up to 60 years are reasonable options for decommissioning light water power reactors. Generally for reactors, the overall impact of either of these alternatives is similar, with the lower occupational dose and wastes with SAFSTOR compensating for the costs and uncertainties of controlling the site for a long period. The choice of alt.rnative in individual cases will depend on a number of factors specific to the particular reactor, site, and time of decommissioning, for example, a longer SAFSTOR period may be acceptable if the safety of an adjacent reactor might be affected by dismantlement procedures or if there is an inability to dispose of waste due to lack of disposal capacity.

With regard to the ENTOMB alternative, long-lived activation products contained in reactor interrals, such as Nb-94 (20,000 years half-life) and Ni-59 (g000 years half-life), would probably preclude the use of ENTOMB for power reactors unless reactor internal; were removed. If reactor internals are removed, some method would have to be provided to demonstrate that the entombed radioactivity will decay to levels permitting release of the property for unrestricted use within about 100 years, which, as noted above, would be difficult.

For research and test reectors and ISFSIs, occupational doses would be much less significant and much easier to manage than for power reactors. Thus, DECON is considered the most reasonable option. SAFSTOR could be justified in some cases. ENTOMB is not expected to be viable for ISFSIs and it also unlikely to be a reasonable option for non power reactors as the cost would not '

be justified.

For materials facilities associated with licenses under Parts 30, 40, and 70, occupational doses are also quite low in most cases, and DECON the most likely option. SAFSTOR is possible for short-lived saterials, but any extended delay would rarely be justifiable. For these reasons the amendments to Parts 30, 40, and 70 do not mentien alternatives or have special requirements for extended

! delays. If after disposing of inventory and some preliminary decontamination, i

contamination from relatively short-lived materials is reported, the Commission

( will detemine whether allowing a period for decay is an appropriate means of completing decommissioning. It is expected however, that for most licenses i

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under these parts it will be practical to complete decontamination to levels suitable for unrestricted release prior to reporting levels of residual radio-activity to the Commission. A survey must be carried out and reported on promptly af ter the end of operations and prior to the expiration of the license.

15.1.2 Planning 15.1.2.1. Preliminary Planning Planning for decommissioning is a critical item for ensuring that the decommis-sioning activities can be accomplished in a safe and timely manner. Develop-ment of detailed plans at the application stage is not possible becauie many factors (e.g. , technology, regulatory requirements, economics) will change before the license period ends. Thus, most of the planning for the actual decommissioning will occur near final shutdown. However, a certain amount of preliminary planning should be done at the application stage.

The availability of adequate funds is important in assuring that decommissioning will be carried out in a safe and timely manner. There are also aspects of design and operations that could affect decommissioning in terms of improved health and safety and reduced radioactive waste such as ready access to major contaminated equipment.

Information on decommissioning funding provisions, described in section 15.1.3 must be submitted with an application for an operating license for a production '

or utilization facility. An application for an independent spent fuel storage installation will also include funding provisions. In the case of existing Part 50 licensees, information on funding provisions would need to be submitted within a reasonable time period following the ef fective date of this rule.

This information should include the method of assuring funds for decommission-ing and ar. indication of the amount being set aside. Provision should be made to adjust cost levels and associated funding levels over the life of the faci-lity. In particular, Part 50 licensees must submit 5 years prior to the pro-jected end of the operation an up-to-date cost estimate on which to base financial assurance, in this manner, it is expected that the amounts being assured by the funding method will reach a level at the end of life which is 10/08/87 15-9 NUO586 CH 15

approximately equal to the actual costs of decommissioning. In particular, the ecst estimate submitted at 5 years prior to end of operation would be based on a current assessment of major factors that could affect decommissioning costs.

The requirement is intended to assure that Part 50 licensees shall consider relevant, up-to-date information which could be important to adequate planning and funding for decommissioning well before decommissioning actually begins.

For most facilities associated with licenses under Parts 30, 40, and 70, decommissioning is much less involved, and has much less impact than the decommissioning of a reactor, for example. However, for larger facilitien decommissioning funding provisions similar to those for reactors are necs

• rv although for most materials facilities with small decommissioning costs2, submittal of information is not necessary.

The studies performed as part of the policy reevaluation have shown that faci-litation of decommissioning in the design of a facility or during its operation can be beneficial in reducing operational exposures and waste volumes requiring disposal at the time of decommissioning. In addition, facilitation can improve financial assurance by keeping actual costs of decommissioning in line with the estimated costs on which the levels of financial assurance are based. The effects of operational procedures on decommissioning should be considered by licensees as part of their program to maintain radiation exposures and effluents "as low as is reasonably achievable" in existing 10 CFR Part 20. The facilita-tion of decommis,ioning in the design of facilities can be considered under the general standard for issuance of license that equipment and facilities be ade- '

ouate to protect the health and safety of the public contained in $$ 30.33(a)(2),

40.32(c), 50.40(a), 70.23(a)(3), and 72.76. Suggestions for facilitation are presented in the PNL studies and in a preliminary study on facilitation of reactor decommissioning.

In particular, experience has shown that an important aspect of operation is the maintenance of adequate information on the design and current condition of the facility and site, so that decommissioning can be carefully planned and carried out. Records of relevant operational information helpful in facili-tating decommissioning must be kept by all reactor and materials licensees.

t 30/08/87 15-10 NUO586 CH 15

x Plans should be developed to collect, maintain, and recall records and archive files which include as-built and as-revised drawings and specifications and operational occurrences which could significantly affect decommissioning so that important information is kept until termination of license and that it be readily accessible when needed.

15.1.2.2 Final Planning Final decommissioning planning will involve greater technical detail than pre-liminary planning. Decommissioning plans should be submitted in a timely way for review and approval prior to the initiation of any major decommissioning activity to avoid delay of decomissioning af ter facility shutdown. For a power reactor, review and approval could take up to a year. Decommissioning plans must address the following:

(1) Decommissioning alternative - A descriptien of the alternative to be used

, for decomiLioning must be presented. Plans for processing and disposing of radioactive waste must also be described. Plans must assess the avail-ability of waste disposal facilities. If wasta disposal space is unavail-able, then plans must address use of available temporary above ground waste storage or other method. Depending on a variety of circumstances, tempor-ary above ground wasta storage may be accomplished offsite or onsite and may require NRC review and approval.

(2) Technical and environmental plans - Controls and limits on procedures and '

equipment to ensure occupational and public safety and to protect the

.nvironment during decomissioning must be proposed by the licensee.

(3) A plan for a final radiation survey must also be presented to ensure that remaining residual radioactivity is within levels permitted for releasing the property for unrestricted use. Although the SAFSTUR or ERTOMB alter-natives may have been selected, which would require a complete termination survey at some future time, unrestricted access to portions of the property may be desirable prior to full decommissioning. A separate termination survey would be necessary for these areas.

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(4) An updated cost estimate must be included along with a plan to ensure t. hat adequate decor.dssioning funds are available to carry out decommissioning operations.

(5) Quality assurance and safeguards - As appropriate for a particular facility, quality assurance and safeguarda provisions during ciecommissioning must be addressed.

The NRC's evaluation of the information submitted in the decommissioning plan and the licensee's subsequent conduct of decommissioning activities can be based on existing regulatior.s applicable to reactors and other facilities undergoing decommissioning. These regulations include 10 CFR Parts 20, 50, 61, 70, 71 and

73. For example,10 CFR Part 20 contains standards for protec*.bn against radiation and is applicable to all licensees during operation a well as decommissioning.

15.1.3 Financial Assurance The primary objective of the NRC with respect to decommissioning is to protect the health and safety of the public. An important aspect of this objective is that there is reasonable assurance that, at the time of termination of facility operations, adequate funds are available to decommission the facility in a safe and timely manner resulting in its release for unrestricted use and that lack of funds does not result in delays in decommissioning that may cause potential ,

health and safety problems for the public. The need to provide this assurance arises from the fact that there are uncertainties concerning the availability of funds at the time of decommissioning. The nuclear facility licensee has the responsibility for completing decommissioning in a manner which protects public health and safety. Satisfaction of this objective requires that the licensee provide reasonable assurance that adeqaate f Jnds for performing decomissioning will be available at cessation of facility ope:ation.

In providing rer' )le assurance that funds will be available for decommission-ing, there are < , possible financing mw.hanisms which are available to applicants ano ,c- sees. The w'de diversity in dif ferent types of nuclear 10/08/67 15-12 NUO586 CH 15

l fa-ilities necessitates that the NRC allow latitude in the implementation of these financing mechanisms. In analyzing funding methods, the NRC has developed the following major classification of funding alternatives.

(1) Prepayment - The deposit prior to the start of operation into an account segregated from licensee assets and outside the licensee's administrative control of cash and liquid assets such that the amount of funds would be sufficient to oay decommissioning costs. Prepayment could be in the form of a trust, escrow account, government fund, certificate of deposit, or deposit of government securities.

(2) Surety bonds, letters of credit, lines of credit, insurance, or other guarantee methods - These mechanisms guarantee that the decommissioning costs will be paid should the licensee default. The licensee still must provide funding for decommissioning through some other method. It appears l questionable that surety methods of the size necessary and for the time involved with power reactors will be available. However, they appear to be available for facilities that involve smaller costs and periods. The contractual arrangement guaranteeing the surety methods, insurance, or guarantee must , include provisions for insuring that these methods will in fact result in funds being available for decommissioning. It should be kept in mind that sureties would only be called if at the time of cessation of facility operation or impending surety loss, licensee decommissioning funds were inadequate or unavailable.

l l

l (3) External Sinking Funds - A fund established and maintained by setting funds l

aside periodically in an account segregated from licensee assets and outside l

the licensee's administrative control in which the total amount of funds would be sufficient to pay decommissioning costs at Q e time termination of operation is expected. An external sinking fund could be in the form of a trust, escrow account, government fund, certificate of deposit, or deposit of government securities. The weakness of the sinkirg fund approach is that in the event of premature closure of a facility the decom-missioning fund would be insufficient. Therefore, the sinking fund would have to be supplemented by insurance or surety bonds, or letters or lines of credit mechanisms of item (2).

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(4) Internal Reserve or Unsegregated Sinking Fun - A fund established and maintained by the periodic deposit or crediting of a .orescribed amount into an account or reserve which is not segregated from licensee assets and is within the licensee's administrative control in which the total amount of the periodic deposits or funds reserved plus accumulated earn-ings would be sufficient to pay for decommissioning at the time termina-tion of operation is expected.

In this mechanism, the funds are not segregated from the utility's assets, rather they may be invested in utility assets and at the end of the faci-lity life, internal funds are used to pay for decommissioning by, for example, issuance of bonds against licensee assets and the funds raised are used to pay for decommissioning. An internal reserve may also be in the form of an internal sinking fund which is similar to an external sink-i g fund except that the fund is held and invested by the licensee. Such a mechanism is generally considered to be less expensive in terms of net present value than the options listed above, although, as discussed in Section 2.6, whichever funding mechanism is used should not have a signi-ficant impact on the revenue requirements. The problem with the internal or unsegregated funding method is the lesser level of assurance that funds will be available to pay for decommissioning than the other mechanisms.

Because this method depends on financing internal to the licensee, and therefore is vulnerable to events that undermine the financial solvency of a utility.

• 1 l

The NRC has considered the use of all of these methods (16-18, 20), and in particular internal reserve in several documents and has reviewed public comments on the proposed rule 2 and the draft GEIS. Based on these docu-l l

ments and on the discussion presented in more detail in Section 2.6.2 of i this EIS, using a standard of providing reasonable assurance that suffi-cient funds are available for decommissioning, licensees may use the methods listed as #(1) to #(3) above singly or in combination. In addi-tion, electric utility licensees owning more than one generating facility may use the method listed as #(4).

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As discussed in 15.1.2.1, inforrn ition on funding assurance provisions must be su'omitted by an applicant prior to licensing the facility. This information must include the method of assuring funds for decommissioning a and an indication of the amount being set aside. To minimize administrative effort while still maintaining reasonable assurance of funds for certain' facilities, the financial provisions may be based on an amount which is at least equal to amounts prescribed in the amended NRC regulations. These amounts vary for the ditterent facilities covered by the regulations. Provisions should also include means for adjusting cost levels and associated funding levels over the life of the facility.

15.1.4 Residual Radioactivity Levels for Unrestricted Use of a Facility Decommissioning requires reduction of the radioactivity remaining in the facility to residual levels that permit release of the facility for unrestricted use and NRC license termination.

The Commission is participating in an EPA organized interagency working group which is developing Federal guidance on acceptable residual radioactivity for unrestricted use. Proposed Federal guidance is anticipated to be published by EPA. NRC is planning to implement this guidance through rulemaking as soon as possible. The selection of an acceptable level is outside the scope of rule-making supported by this EIS. Currently, criteria for residual contamination levels do exist and research and test reactors are being decommissioned using present guidance contained in Regulatory Guide 1.86 for surface contamination plus case-by-case considerations for direct radiation. As an example, NRC pro-vided such criteria in letters to Stanford University, dated 3/17/81 and 4/21/82 providing "Radiation criteria for release of the dismantled Stanford Research Reactor to unrestricted access." The cost estimate for decommissioning can be based on current criteria and guidance regarding residual radioactivity levels for unrestricted use. As discussed in Section 2.5 of this EIS, the information in the studies by Battelle Northwest Laboratory and Oak Ridge National Laboratory on decommissioning have indicated that in any reasonable range of residual radioactivity limits, the cost of decommissioning is relatively insen-sitive to the radioactivity level and use of cost data based on current criteria should provide a reasonable estimate.

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Even in situations where the residual radioactivity level might have an effect on decommissio. 'ng cost by use of update provision in the rulemaking, it is expected that the decommissioning fund available at the end of facility life will approximate closely the actual cost of decommissioning.

15.1.5 Environmental Impact Statement Generally, the major environmental impact from decommissioning, especially for power reactors, occurs at commissioning, where the decision to operate the reactor is made. Provided the provisions of this rule are in place and based on the conclusion of Chapters 4 and 5 regarding impacts from reactor decommis-sioning alternatives, it is not expected that any significant environmental impacts will result from the choice of alternatives. Therefore current 10 CFR Part 51 needs to be amended to delete the mandatory EIS requirement for decom-missioning of power reactors. An EIS may still be required but this should be based on site specific factors. Therefore a licensee should submit a supple-mental environmental report and safety analysis and based on these submittals, the NRC should consider issuance of a negative declaration of impact, wnich is expected to be reasonable for most situations.

It is imperative that these decommissioning rule amendments in 10 CFR Parts 30, 40, 50, 51, 70 and 72 be issued at this time because it is important to establish financial assurance provisions, as well as other decommissioning planning pro-vision, as soon as possible so that funds will be available to carry out decom ,

missioning in a manner which protects public health and safety. Based on this need for the decommissioning rule and provisions currently existing and those contained in the rule amendments, the Commission believes that the rule can and should be iswed now, i

15.2 Regulations As discussed in Section 15.1, consideration must be given to decomissioning of X a facility riuring the design construction j

and operating stages of a nuclear facility lifetime. Regulations which have relevance for decommissioning planning and accomplishment are contained in Title 10 of the Code of Federal Regulations (10 CFR), Parts:

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Part % Title 30 Rules of General Applicability to Domestic Licensing of Byproduct Material 40 Domestic Licensing of Source Material 50 Domestic Licensing of Production and Utilization Facilities 51 Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions 70 Domestic Licensing of Special Nuclear Material 72 Licensing Requirements for the Storage of Spent Fuel in an Independent Spent Fuel Storage Installation (ISFSI)

Many of the regulatory requirements contained in the aforementioned regulations do not contain the explicit consideration of necessary decommissioning require-ments discussed in this section (although many of the explicit decommissioning requirements have been required as a condition of NRC licensing in case-by-case instances). Development of a separate regulation which specifically addresses decommissioning was considered. However, such a separate regulation would be

cumbersome because it would need to contain many of the requirements already

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l presented in 10 CFR Parts 30, 40, 50, 51, 70, and 72. Since decommissioning requirements are an integral consideration in nuclear facility licensing and operation, it is appropriate in terms of simplicity, efficiency and reduction of regulatory burden, to amend the pertinent parts of the existing regulations to explicitly include appropriate decommissioning requirements.

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0 REFERENCES

1. Plan for Reevaluation of NRC Policy 'on Decommissioning of Nuclear Facilities, NUREG-0436, Revision 1, U..S. Muclear Regulatory Comission,.

December 1978, and Supplement 1, July 1980, and Supplement 2, March 1981.

1

2. 50'FR 5600, February 11, 1985..

3. .FinalGenericEnvironmentalI'mpactState.dentonUraniumDrilling, NUREG-0706, U.S. Nuclear Regulatory Commission, September 1980.

4. R. I. Smith, G. J. Konzek, and W. E. Kennedy, Jr. , Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station NUREG/CR-0130, Prepared by Pacific Northwest Laboratory for th? U.S. Nuclear Regulatory. Commission, June 1978, Addendum 1, August 1979, Addendum 2, .luly 1983, and' Addendum 3, September 1984.

5. H. D. Oak, et al. , Technology, ' Safety, and Costs of Decommissioning a Reference Boiling Water Reactor Power Station, NUREG/CR-0672, Prepared by-Pacific Northwest Laboratory for the.U.S. Nuclear Reg'ulatory Commission; June 1980, Addendum.1, July 1983, .and Addendum 2, September 1984.

6. G. J. Konzek, Technology, Safety, and Costs of Decommis'sioning Reference Nuclear Research and Test Reactors, NUREG/CR-1756, prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, February 1982, and Addendum, July 1983.

7. Norm G. Wittenbrock, et al. , Technology, Safety, and Costs of Decommis-sioning Light Water Reactors at a Multiple Reactor Station, NUREG/CR-1755, pr2 pared by prepared by Pacific Northwest Laboratory for the U.S. Nuclear l Regulatory Commission, January 1982.

8. Eramett B. Moore, Jr. , Facilitation of Decommissioning of Light Water Reactors, NUREG/CR-0569, prepared by Pacific Northwest Laboratory for the l U.S. Nuclear Regulatory Commission, December 1979.

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9. E. S. Murphy, Technology, Safety, and Costs of Decommissioning Reference Light Water Reactors following Accidents, NUREG/CR-2601, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, November 1982.

1

10. K. J. Schneider and C. E. Jenkins, Technology, Safety, and Costs of Decommissioning a Reference Nuclear Fuel Processing Plant, NUREG-0278, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, October 1977.

11. H. R. Elder and D. E. Blahnik, Technology, Safety, and Costs of Decommissioning A Reference Uranium Fuel fabrication Plant, NUREG/CR-1266, Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, October 1980.

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1

)

i

12. H. R. Elder, Technology, Safety, and Costs of Decommissioning a Reference )

, Uranium Hexafluoride Conversion Plant, NUREG/CR-1757, Prepared by Pacific l Northwest Laboratory for the U.S. Nuclear Regulatory Commission, October 1981.

13. C. E. J' enkins, E. S. Murphy, and K. J. Schneider, Technology, Safety, and Costs of Decommissioning a Reference Small Mixed Oxide Fuel Fabrication j Plant, NUREG/CR-0129, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, February 1979.

14. E. S. Murphy, Technology, Safety, and Costs of Decommissioning Reference Non-Fuel-Cycle Nuclear Facilities, NUREG/CR-1754, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, February 1981.

15. J. D. Ludwick and E. B. Moore, Technology, Safety, and Costs of Decomissioning Reference Independent Spent Fuel Storage ' Installations, NUREG/CR-2210, Prepared by Pacific Northwest Laboratory for the U.S.

Nuclear Regulatory Commission, January 1984,

16. Robert S. Wood, Assuring the Availability of Funds for Decomissioning Nuclear Facilities Draf t Report, NUREG-0584, Revision 3, U.S. Nuclear Regulatory Commission, March 1983.

17. Financing Strategies For Nuclear Power Plant Decommissioning, NUREG/CR-1481, Prepared by Temple, Barker, and Sloan, Inc., for the New England Conference of Public Utilities Commissioners, Inc. , for U.S. Nuclear Regulatory Commission, July 1980.

18. P. L. Chernick, et al. , Design, Costs and Acceptability of an Electric Utility Pool for Assuring the Adequacy of Funds for Nuclear Power Plant Decomissioning Expense; NUREG/CR 2370, Prepared by Analysis and Inference, Inc. , for U.S. Nuclear Regulatory Comission, December 1981.

19. C. F. Holow'y a and J. Witherspoon, Monitoring for Compliance with x Decomissioning Terminatien Survey Criteria, NUREG/CR-2082, Prepared by

. Oak Ridge National Laboratory for the U.S. Nuclear Regulatory Comission, .

~

June 1981.

20. J. J. Siegel, Utility Financial Stability and the Availability of Funds for Decommissioning, NUREG/CR-3899, Prepared by Engineering and Economics Research, Inc. , for the U.S. Nuclear Regulatory Commission, September 1984, and Supplement 1 (to be published).

21. J. P. Witherspoon, Technology and Cost of Termination Surveys Associated with Decommissioning of Nuclear Facilities, NUREG/CR-2241, prepared by Oak Ridge National Laboratory for U.S. Nuclear Regulatorf Commission, January 1982. .

22. H. K. Elder, Technology, Safety, and Costs of Decommissioning Reference Nuclear Fuel Cycle and Non-Fuel-Cycle Facilities Following Postulated Accidents, NUREG/CR-3293, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, May 1985.

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23. H. K. Elder, Technology, Safety, and Costs of Decommissioning Reference Nuclear Fuel Facilities, NUREG/CR-4519, Prepared by Pacific Northwest Laboratory for U.S. Nuclear Regulatory Commission, May 1986.

24 J. C. Evans, et al. , Long-Lived Activation Products in Reactor Materials, NUREG/CR-3474, Prepared by Pacific Northwest Laboratory for the U.S.

Nuclear Regulatory Cocimission, August 1984.

25. K. H. Abel, et al. , Residual Radionuclide Contamination Within and Around Commercial Nuclear Power Plants, NUREG/CR-4289, Prepared by Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, February 1986.

26. T. S. LaGuardia and J. F. Risley, Identification and Evaluation of Facilitation Techniques for Decommissioning Light Water Power Reactors, NUREG/CR-?S87, Prepared by TLG Engineering, Inc. for the U.S. Nuclear Regulatory Commission, June 1986.

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l 4

GLOSSARY Abbreviations, acronyms, terms, and definitions used in this study and directly relateo to decommissioning work and related technology are defined and explained in this section. The section is divided into two parts, with the first part containing abbreviations and acronyms, and the second part containing terms and definitions (including those used in a special sense for this study). Common terms covered adequately in standard dictionaries are not included.

ABBREVIATIONS AND ACRONYMS AEC Atomic Energy Commission ALAP As Low As Practicable ALARA As Low As is Reasonably Achievabic(a)

BEIR Biological Effects of Ionizing Radiation CFR Code of Federal Regulations (a; Ci Curie (a)

DF Decontamination Fac, ga )

DOE Departm nt ur Energy DOT Department of Transportation DPM Disintegrations per Minute (a) a See the following section Glossary Definitions, for additional information or explanation.

10/09/87 G-1 NUO586 GLOSSARY

i EPA Environmental Protection Agency HE A High Efficiency Particulate Air (Filters)(a) 1 HLW High Level Waste (a)

HVAC Heating, Ventilation, and Air Conditioning ICRP International Commission on Radiological Protection LLW Low Level Waste (a) m3 Cubic Meters mR Milliroentgen (a) arad Millirad (a) arem Millirem, also see rem MT Metric Ton (a)

MT11M Metric Ton of Heavy Metal W d/MTU Thermal Megawatt-day per Metric Ton of Uranium, the Burnup(a)

MWe Megawatts electric MWt Megawatts thermal NEPA National Environmental Policy Act NRC Nuclear Regulatory Commission a

See the following section Glossary Definitions, for additional information or explanation.

10/09/87 G-2 NUO586 GLOSSARY

ORNL Oak Ridge National Laboratory OSF Overall Scaling Factor PNL Pacific Northwest Laboratory R Roentgen (a) rad Radiation Absorbed Dose (a) rem Roentgen Equivalent Man (a)

SW Special Nuclear Material (a)

T Half Life, Radiological (a)

V2 TRU Transuranic UF 6

Uranium hexafluoride U0 2

Uranium dioxide a

See the following section, Glossary Definitions, for additional information ,

or explanation.

GLOSSARY DEFINITIONS Actinides--A series of heavy radioactive metallic elements of increasing atomic number (Z) beginning with antinium (89) or thorium (90) through element hahnium of atomic number 105.

Activation--The process by which a material is cade radioactive by its exposure to neutrons or protons. Material in the primary coolant of a reactor may become activated in its passage through the reactor core. Also, the 10/09/87 G-3 NV0586 GLOSSARY

internals of a reactor may become radioac'ive due to their exposure to neutrons.

Activity--See Radioactivity.

Agreement State--A state with which the NRC has entered into an agreement, under provisions of the Atomic Energy Act of 1954 and its amendments, in which States assume regulatory respo;isibility over bygroduct, source material, and small quantities of special nuclear material.

Airborne Radioactive Material--Radioactive particulates, mists. fumes, and/or gases in air.

ALARA--A regulatory design philosophy to maintain radiation exposure A,s Low As is Reasonchly Achievable.

Atomic Number (2)--The number of protons in the nucleus of an atom; also its positive charge. Each chemical element has its characteristic atomic num-ber, and the atomic numbers of the known elements form a complete series from 1 (hydrogen) through 105 (hahnium).

Background--The level of radioactivity from sources other than the one directly under consideration, in this case those existing without the presence of the nuclear facility. .

Beta Decay--Radioactive decay in which a beta particle is emitted or in which an orbital electron capture occurs.

Bio-availability--The degree to which radionuclides are available for transmittal through the food chain to the exposed individual.

Burial Grounds--Areas designated for storage of packaged radioactive wastes in j soils below the surface.

Burnup, Specific--The total energy released per unit mass of a nuclear fuel.

l It is commonly expressed in megawatt-days per metric ton of fuel material.

l 10/09/87 G-4 NU0586 GLOSSARY

Byproduct Material--Any radioactiie material (except special nuclear material) yielded in or made radioactive by exposure to the radiation incident to the process of producing or utilizing special nuclear material.

Cask--A neavily shielded shipping container for radioactive materials. Some casks weigh as much as 100 metric tons.

Certification survey--See terminal radiation survey.

Chemical decontamination--Decontamination accomplished by the use of chemical solutions to remove surface films containing radioactive materials.

Code of Federal Regulations (CFR)--The Code of Federal Regulations is a documen-tation of the general rules by the Executive departments and agencies of the Federal Government. The Code is divided into 50 titles that represent broad areas subject to Federal regulation. Each title is divided into Chapters that usually bear the name of the issuing agency. Each Chapter is further subdivided into Parts covering specific regulatory areas.

Commissioning--The licensing and startup of a nuclear facility.

Container--See cask.

Contamination--Undesired radioactive materials that have been deposited on the surfaces, or are internally ingrained into structures or equipment, or that havt been mixed with other materials.

Continuing care--See safe storage.

Critical Facility--A non-reactor facility that handles, tests or processes fissile material.

10/09/87 G-5 NUO586 GLOSSARY

Curie--A special unit of radioactivity. One curie equals 3.7 x 1020 nuclear transformations per second. (Abbreviated Ci.) Several fractions of the curies are in common usage:

Millicurie. One-thousandth of a curie. Abbreviated mci (3.7 x 107 d/s).

Microcurie: One-millionth of a curie. Abbreviated pCi (3.7 x 104 d/s).

Nanocurie: One-billionth of a curie. Abbreviated nCi (37 d/s).

Picocurie. One-millionth of a microcurie. Abbreviated pCi (0.037 d/s).

Custodial SAFSTOR--A minimum cleanup and decontamination followed by a period of safe storage with active protection systems in service and completed by deferred decontamination. The active protection systems (i.e., principally ventilation) are kept in service, the site is secured against intrusion by physical barriers and by guards, and use of the facility and site is limited to nuclear activities.

Decay, Radioactive--A spontaneous nuclear transformation in which a particle, gamma radiation, or x-ray radiation is emitted.

Decommissioning--To remove a facility safely from service and reduce residual radioactivity remaining to a level that permits release of the property fer unrestricted use and termination of license.

Decommissioning insurance--A mechanism for assuring the funding of decommission-ing which could provide funds for all decommissioning expenses, including those for premature closure of the facility, or alternatively, funds to cover costs of premature decommissioning in the event that other mechanisms provided by the insureds were insufficient.

10/09/87 G-6 NUO586 GLOSSARY

DECON--Tae alternative in which the equipment, structures, and portions of a facility and site containing radioactive contaminants are removed or de-contaminated to a level that permits the property to be released for unrestricted use shortly af ter cessation of operations.

Decontamination--Thoce activities employed to reduce the levels of contamination in or on structures, equipment and materials.

Decontamination Factor (OF)--The ratio of the initial concentration of an undesired material to the final concentration resulting from a treatment process. The term may also be used as a ratio of quantities.

Deferred Decontamination--Those actions required af ter the safe storage period of SAFSTOR to disassemble and remove sufficient radioactive or contaminated materials from the facility and site to permit release of the property for unrestricted use.

Design Basis Accident--A postulated accident believed to have the most severe expected impacts on a facility. It is used as the basis for safety and structural design.

Disintegration, Nuclear--The transformation of the nucleus of an atom from one element to another, characterized by a definite half-life and the emission of particles or electromagnetic radiation.

Disintegration Rate--The rate at which disintegrations occur, characterized in units of inverse time, i.e., disintegrations per minute (dpm), etc.

Dismantlement--Those actions required to disassemble and/or remove radioactive or contaminated materials from the facility and site.

l Dispersion--A process of mixing one material within a larger quantity of another.

For example, the mixing of material released to the atmosphere with air l

causes a reduction in concentration with distance from the source.

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10/09/87 G-7 NU0586 GLOSSARY ,

1 Disposal--The disposition of materials with the intent that the materials will )

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not enter man's environment in sufficient amounts to cause a health hazard.

{

Dese, Absorbed--The mean energy imparted to matter by ionizing radiation per unit mass of irradiated material at the place of interest. The unit of absorbed dose is the rad. One rad equals 0.01 joules / kilogram in any medium (100 ergs per gram).

Dose commmitment--The integrated dose that results unavoidably from an intake of radioactive material starting at the time of intake and continuing to a later time (usually specified to be 50 years from intake).

Oose, Equivalent--Expresses the amount of radiation that is effective in the human body, expressed in rems. Modifying factors associated with human tissue and body are considered. Equivalent dose is the product of absorbed dose multiplied by a quality factor multiplied by a distribution factor.

Referred to as Dose in this report.

Oose, Occupational--The exposure of an individual to radiation as a result of his employment, expressed in rems.

Dose Rate--The radiation dose delivered per unit time and measured, for instance, in rem per hour.

Enrichment--The ratio (usually expressed as a percentage) of fissile isotope to the total amount of the element (e.g. , the % of 235U in uranium).

ENTOMB--The alternative in which radioactive contaminants are encased in a struc-turally long-lived material, such as concrete; the entombed structure is appropriately maintained and continued surveillance is carried out until the radioactivity decays to a level permitting unrestricted release of the property. '

10/09/87 G-8 NUO586 GLOSSARY

Exposure--The condition of being made subject to the action of radiation; also frequently the quantity of radiation received. The special unit of exposure is the roentgen (see Roentgen).

Exposure Pathway--The mechanisms by which radioactive material passes from the source of the material through the environment to an exposed individual.

External exposure--As used in this EIS, an exposure pathway in which an individ-ual is externally exposed directly to radioactive materials dispersed in the air (immersion) or is exposed directly to surfaces containing radioactive materials.

Facilitation--As used in the context of decommissioning, consideration to be given to facility design and normal operational procedures, as well as decommissioning procedures, with the primary purpose of reducing occupa-tional and public radiation dose and waste volumes during the decommission-ing process.

Facility--The physical complex of buildings and equipment within a site.

Final Inventory Cleanout--An extensive inventory cleanout and special nuclear material audit conducted upon termination of normal facility operations.

Since these cleanout operations are also conducted periodically during nor-mal operation for audit and contamination control purposes, this procedure is not considered part of decommissioning and its cost is not included as -

a decommissioning cost.

Fission--The splitting of a heavy atomic nucleus into two lighter parts (atomic nuclides of lighter elements), accompanied by the release of a relatively large amount of energy and generally one or more neutrons. Fission can occur spontaneously but usually it is caused by nuclear absorption of neutrons or other particles.

Fissile Materials--Materials that are capable of fission.

10/09/87 G-9 NUO586 GLOSSARY

Fission Products--The lighter atomic nuclides (fission fragmentd formed by the fission of heavy atoms. It also includes the nuclides formed by the fission fragments' radioactive decay.

Food Chain--The pathways by which any material passes through man's environment through edible plants and/or animals to man.

Fuzi Assembly--A grouping of fuel elements (hollow rods filled with nuclear fuel for LWRs) that supply the nuclear heat in a nuclear reactor. A fuel element or rod is the smallest structurally discrete part of a reactor or fuel assembly that has nuclear fuel as its principal constituent.

Fuel Cycle--The series of steps involved in supplying fuel for nuclear power reactors and handling spent fuel and radioactive waste, including transportation. These steps are usually divided up as the head end and back end as follows:

Head end: Mining, milling, conversion, enrichment, and fabrication of fuel.

Back end: Includes reactors, spent fuel storage, spent fuel reprocessing, mixed-oxide fuel fabrication, and waste management.

Fuel Element--A rod, tube, or other form into which nuclear fuel is fabricated to use in a reactor.

Gamma Rays--Short-wavelength electromagnetic radiation. Gamma radiation fre-l quently accompanies alpha and beta emissions and always accompanies fission.

Gamma rays are best stopped or shielded against by dense materials such as l lead or uranium. These rays originate from within the nucleus of the atom.

1 f

l l Gaseous--Material in the vapor or gaseous state, but can include entrained liquids and solids. A gas will completely fill its container regardless of container shape or size.

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l 10/09/87 G-10 NU0586 GLOSSARY

r Half-Life, Radioactive--The time in which half the atoms of a particular sub-stance disintegrates to another nuclear form. Each radionuclide has a unique half-life. Measured half-lives vary from millionths of a second to billions of years.

Heavy Metal--Terminology used in reference to metals with atomic numbers 90 and greater. It usually refers to nuclear fissile or fertile fuels such as thorium, uranium, and plutonium.

HEPA filter--A filter used in facility ventilation systems whose purpose is to remove particulate material from the ventilation air stream.

High-Level Wastes--Intact fuel assemblies that are being discarded after having completed their useful lives in a nuclear reactor (spent fuel) or the por-tion of the wastes generated in the reprocessing of spent fuel that contain virtually all of the fission products and most of the actinides not separated out during reprocessing.

Hot Spots--Areas of radioactive contamination higher than average.

Ingestion--As used in this EIS, an exposure pathway in which radioactive mate-rials reach the exposed individual through the ingestion of food and water.

Inhalation--As used in the EIS, an exposure pathway in which radioactive materials reach the exposed individual through the breathing process. -

Institutional Control Reliance--The degree to which reliance can be placed on the ability of man-made institutions to both safely confine the radio-activity in and prevent the intrusion into a nuclear facility while it is in safe storage or while it is entombed.

Insurance for decommissioning--See decommissioning insuranca.

10/09/87 G-11 NU0586 GLOSSARY

ji Internal reserve--A mechanism for the funding of decommissioning in w iich a fund is established and maintained by the periodic deposit or crediting of a prescribed amount into an account or reserve which is not segregated from licensee assets and is within the licensee's administrative control in which the total amount of the periodic deposits or funds reserved plus accumulated earnings would be sufficient to pay for decommissioning at the time termination of operation is expected.

Ion Exchange--A chemical process involving the selective adsorption or desorption of various chemical ions in a solution onto a solid material, usually a plastic or resin. The process is used to separate and purify chemicals, such as fission products from plutonium or "hardness" from water (i.e.,

water snftening).

Licensed Material--Nuclear source material, special nuclear material, or nuclear by product material received, possessed, used, or transferred under a license issued by the Nuclear Regulatory Commission.

Long-Lived Nuclides--For this study, radioactive isotopes with long half-lives typically taken to be greater than about ten years. Most nuclides of interest to waste management have half-lives on the order of one year to millions of years.

Low-level Wastes--Wastes contaminated with radioactive materials emitting pri-

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marily beta or gamma radiation, nct high-level waste (see high-level wastes) and which art. not transuranic wastes, i.e. , they contain less than 10 nano-curies per gram of transuranic elements (see transuranic waste).

Management (Waste)--The planning, execution, and surveillance of essential func-tions related to radioactive waste, including treatment, solidification, packaging, interim or long-term storage, transportation, and disposal.

l l Man-rem--A measure of radiation dose distributed to a population. To calculate l radiation dose to the population, the dose equivalent in rem received by j each persen in the population is summed.

I I 10/09/87 G-12 NUO586 GLOSSARY l .. - .. .. .

Mass Number--The number of nucleons (protons and neutrons) in the nucleus of an atom. (Symbol: a).

Maximum Exposed Individual--The hypothetical member of the public who receives the maximum radiation dose. For the common case where exposures from air-borne radionuclides result in the highest radiation exposure, this individ-ual resides at the location of the highest airborne radionuclide concentration and eats food grown at that location.

Megawatt-day--A unit for expressing the energy generated in a reacter; specifi-cally, the number of millions of watt-days of heat output per metric ton of fuel in the reactor. Also, the net electrical output in millions of watts of e~lectrical energy averaged over one day.

Megawatt Days per Metric Ton of Uranium--Amount of thermal megawatt-days produced per metric ton of uranium; also called burnup. (See also specific power.)

Metr,ic, Ton--1000 kilograms.

Mixed Oxide--A mixture of uranium dioxide and plutonium dioxide.

Monitoring--Taking measurements or observations for recognizing the status, or significant changes in conditions or performance, of a facility or area.

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Negative Net Salvage Value Depreciation--An accounting procedure which allows depreciation to be collected in a manner that considers that the salvage value of a nuclear facility is actually negative, i.e., the price of any salvageable equipment is outweighed by the cost of decommissioning. Thus the net depreciation value of a nuclear facility is its original capital cost plus its decommissioning cost.

Net present worth--As used in this EIS, the cost of decommissioning in terms of 1986 dollars.

10/09/87 G-13 NUO586 GLOSSARY

Normal Operating Conditions--Operation (inc uding startup, shutdown, and main-tenance) of systems within the normal range of applicable parameters of an operating facility.

Nuclear Reaction--A reaction involving a change in an atomic nucleus, such as fission, fusion, particle capture, or radioactive decay.

Of fsite--Beyond the boundary line marking the limits of plant property.

Onsite--Within the boundary line marking the limits of plant property.

Operable--Capable of performing the required function.

Package--The packaging plus the contents of radioactive materials.

Packaging--The assembly of radioactive material in one or more containers or other components necessary to assure compliance with prescribed regulations.

Passive SAFSTOR--A partial cleanup and decontamination effort initially, followed by a period of safe storage and completed by deferred decontamination.

Daring the period of safe storage, all systems are deactivated, the structures are secured by strong physical barriers and continuous remote monitoring, and the plant is limited to nuclear use only, while the site may have non-nuclear uses.

Physical decontamination--Decontamination accomplished by the use of mechanical cleaning means or by the removal of the surface itself.

Plant--The physical complex of buildings and equipment, including the site.

Preparation for Safe Storage--Those cleanup and decontamination activities required during the initial stages of SAFSTOR in order to prepare the facility for the safe storage period.

10/09/87 G-14 NUO586 GLOSSARY

Prepayment--A mechanism for tN funding of decommissioning in which there is a deposit prior to the start of operation into an account segregated from licensee assets and outside the licensee's administrative control of cash or liquid assets such that the amount of funds would be sufficient to pay decommissioning costs.

Protective Clothing--Special clothing worn by a person in a radioactively con-taminated area to minimize the potential for contamination of his body or perstnal clothing and to control the spread of contamination.

Quality Assurance--The systematic actions necessary to provide adequate con-fidence that a material, component, system, process, or facility performs satisfactorily, or as p'anned, in service.

Quality Control--The quality assurance actions that control the attributes of the material, process, component, system or facility in accordance with predetermined quality requirements.

Pg--A unit of 5 sorbed dose. The energy imparted to matter by ionizing radia-tion per t,..it mass of irradiated material at the place of interest. One rad equals 0.01 joule / kilogram of absorbing material.

Radiation--(1) The emission and propagation of radiant energy through space or through a material medium in the form of waves; for instance, that of electromagnetic waves or of sound and elastic waves. (2) The energy of such waves; and (3) corpuscular emissions, such as alpha and beta radiation, or rays of mixed or unknown types.

Radiation Background--See Background.

Radioactive Material--Any material or combination of materials which spontane-ously emit ionizing radiation, generally alpha or beta particles, often accompanied by gamma rays.

Radioactivity--The number of nuclear transformations occurring in a given quan-tity of material per unit of time with the emission of particles, gamma radiation, or x-ray radiation. Often shortened to "activity."

10/09/87 G-15 NUO586 GLOSSARY

1 l

l Ra:ioactivity, Nat Jral--The property of radioactivity exhibited by ;nare than fifty naturally occurring radionuclides.

Ra iological Protection -Protection against the effects of internal and external exposure to radiation and to radioactive r sterials.

Rate of return--As used in this EIS, the rate that investment by decommise,ioning funding mechanisms will increase in value.

Regulatory Guides--Regulatory Guides are issued by the NRC, to describe and make available to the public, methods acceptable to the NRC staff, for implement-ing specific parts of the NRC's regulations, to delineate techniques used by the staff in evaluating specific problems or postulated accidents, or to provide other guidance to applicants for nuclear operations. Guides are not substitutes for regulations and compliance with them is not explic-itly required. Methods and solutions different from those set out in the guides may be acceptable if they provide a basis for the finding requisite to the issuance or continuance of a permit or license by the NRC.

Ree--A unit of radiation dose equivalence. The radiation dose equivalence in rem is numerically equal to the absorbed dose in rads multiplied by the quality factor, the distribution factor, and any other necessary modifying factors.

Respository (Federal)--A site owned and operated by the Federal Government for ~

long-term storage or disposal of radioactive materials.

Residual Radioactivity Levels--As used in this EIS, the amount of radioactively contaminated material remaining in a nuclear facility after decommissioning has been completed and the facility license terminated. To be acceptable, this level must be low enough to permit the facility to be released for unrestricted use.

Restricted Area--Any area to which access is controlled for protection of l

individuals from exposure to radiation and radioactive materials.

I 10/09/87 G-16 NUO586 GLOSSARY L

Risk- As used in this EIS, quantitative risk estimation of potential health effects.

Roentgen--A unit of exposure to ionizing radiation. It is that amount of gamma or x-rays required to produce ions carrying one electrostatic unit of elec-trical charge (either positive or negative) in one cubic centimeter of dry cir under standard conditions. One roentgen equals 2.58 x 10 4 coulombs per kilogram of air. (See also Exposure.)

SAFSTOR--The alternative in which the nuclear facility is placed (preparation for safe storage) and maintained in a condition that allows the nuclear facility to be safely stored (safe storage) and subsequently decontaminated to levels that permit release for unrestricted use (deferred decontamination).

Safe Storage--A period of time starting after the initial decommissioning activ-ities of preparation for safe storage cease and in which surveillance and maintenance of the facility takes place. The duration of time can vary from a few years to on the order of 100 years.

Sealed source--Radioactive material that is encased in a capsule designed to prevent leakage or escape of the radioactive material.

Segregated funding mechanism--As used in this EIS, a term to indicate that the funding mechanism being employed deposits funds in accounts separate from ~

company assets and under control of a party other than the license?.

Shield--A body of material used to reduce the passage of particles or electro-magnetic radiation. A shield may be designated according to what it is intended to absorb (as a gama ray shield or neutron shield), or according to the kind of protection it is intended to give (as a background, or ther-mal shield). It may be required for the safety of personnel or to reduce radiation enough to allow use of counting instruments for research or for locating contamination or airborne radioactivity.

10/09/87 G-17 NUO586 GLOSSARY

Sheat-Lived Radionuclides--For this study, those radioactive isotopes with half-lives less than about 10 years.

Shotdown--The time during which a facility is not in productive operation.

Sinking Fund--A mechanism for the funding of deconsissioning in which a pre-scribed amvJnt of funds, subject to periodic revision, is set aside at regular intervals such that the fund plus accumulated interest would be sufficient to pay for decommissioning costs at the end of facility operation.

Site--The geographic area upon which the facility is located that is subject to controlled public access by the facility licensee (includes the restricted area designated in the NRC license).

Solid Radioactive Waste--Material that is essentially solid and dry but may con-tain sorbed radioactive fluids in sufficiently small amounts as to be immobile.

Solidification--Conversion of radioactive wastes (gases or liquids) to dry, stable solids.

Special Nuclear Material--Plutonium, uranium enriched in the isotopes 233 or 235, and any other material as defined in 10 CFR 70 by the NRC.

Surety bond--A mechanism for the f+Jnding of decommissioning which guarantees that decommissioning costs will be paid should the bond purchaser default.

Surface Contamination--Contamination that is the result of the deposition and attachment of foreign materials to a surface.

Surveillance--Those activities necessary to assure that the site remains in a safe condition (including inspection and monitoring of the site, maintenance of barriers to access to radioactive materials left on the site, and prevention of activities on the site that might impair these barriers).

10/09/87 G-18 NU0586 GLOSSARY

Survey--An evaluation of the radiation hazards incident to the production, use, release, disposal or presence of radioactive materials or other sources of radiation under a specific set of conditions.

Technical Specifications--Requirements and limits that encompass nuclear safety but are simplified to facilitate use by plant operation and maintenance personnel. They are prepared in accordance with the requirement of 10 CFR 50.36, and are incorporated by reference into the amended license issued by the NRC.

Terminal Radiation Survey--The radiation survey conducted near the end of the decommissioning period the ,wrpose of which is to certify that decommission-ing of the facility has resulted in residual radioactivity levels acceptable for releasing the facility for unrestricted use.

Transuranic Elements--Elements with atomic number (Z number) greater than 92.

Transuranic Waste--Any waste material measured or assumed to contain more than a specified concentration (i.e. , proposed as 10 nanocuries of alpha emitters per gram of waste, or more presently proposed as 100 nanocuries/cm 3 of waste 239U) of transuranic elements.

Unfunded reserve--See internal reserve.

Unrestricted access--The condition of a nuclear facility after decommissioning ~

is complete and the facility license is terminated. At this time the general public would be allowed use of the facility without radiation protection controls.

Unsegregated sinking funds--See internal reserve.

Volumetric contamination--Contamination that is contained within the volume of the contaminated material, such as activation products.

t i

10/09/87 G-19 NUO586 GLOSSARY

t Wastes, Radioactive--Equip.- 3. wt. ri:1s (from nuclear operations) that are radioactive and for w?";h the-' o- fu,ther known use.

WholeBodyDoseEquiv@ 4- csed i> - 1. report for the discussion of residual radioactivity lever ,c- ^ '

.e equivalent number that is a summation of dose equivalent fi a m8 a.. segans multiplied by respective weighting factors related to cancer oducing risk.

a I

i 4

10/09/87 G-20 NUO586 GLOSSARY

APPENDIX A. DISCUSSION OF COMMENTS ON THE DnAFT GENERIC ENVIRONMENTAL IMPACT STATEMENT In a Federal Register notice, 46 FR 27, dated February 10, 1981, the Commission announced the availability of a Draft Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, and invited public comment on the state-ment. Coments received on the Draft GEIS are reproduced in Appendix B of this Final GEIS.

The staff's consideration of the coments received and its disposition of the issues involved are reflected in part by revisions in the pertinent sections of this Final GEIS and in part by the following discussions. This section is organized according to major identified questions or subject areas. These areas are those indicated in Chapters 1 and 2 of the Draft GEIS. These subject areas and the sections of Appendix A in which they are covered are as follows:

Subject Area Section i

l General Questions about Decommissioning A-1 Regulations and the GEIS l

~

Planning for Decomissioning A-2 Decommissioning Alternatives and Other A-3 l Design Issues l

l Residual Radioactivity A-4 l

l Financial Assurance A-5 Waste Disposal A-6 l

Other general questions A-7 I

10/09/87 A-1 NUO586 APPENDIX A

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These general subject areas are broken down in more detail in each of the sec- )

tions. Discussions on the coments on similar topics are grouped together. The comment letters to which the discussions apply are referenced by the number following the title of each response; these numbers are keyed to the letters in Appendix Bj Table B-1.

On February 11, 1985, the NRC published a Notice of Proposed Rulemaking on Decommissioning Criteria for Nuclear Facilities (50 FR 5600). The proposed amendments covered a number of topics related to decommissioning that would be applicabie to 10 CFR Parts 30, 40, 50, 70, and 72 applicants and licensees.

These topics included decommissioning alternatives, planning, assurance of funds for decomissioning, environmental review requirements, and residual radioactivity. A total of 143 different organizations and persons submitted comments on the proposed rule. Detailed responses to those individual coments dre documented in NUREG-1221 entitled "Summary, Analysis and Response to Public Coments on Proposed Rule Amendments on Decomissioning Criteria for Nuclear Facilities." Many of the coments made on the proposed rule are similar to those made on the Draft GEIS and hence the responses are the same. To minimize repetitiveness, in responding to the Draft GEIS comments in this Appendix A, reference is made to NUREG-1221, as appropriate, for a more complete discussion.

A.1 General Questions About Decommissioning Regulations and the GEIS A.1.1 Need for Regulations Coment No.1 - Questions why the GEIS does not present the needed cecomis-sioning regulations. (1, 3, 32, 36, 40)

Discussion The gel $ itself does not present the decomissioning regulations. However, as indicated in Section 1.1 of the Draft GEIS, the purpose of the GEIS is to assist the NRC in developing new policies and in promulgating regulations with respect to decomn.issioning of licensed nuclear facilities. In Chapter 15, the GEIS contains recomended policy items that should be included in decommissioning 10/09/87 A-2 NU0586 APPENDIX A

i regulations. On February 11, 1985, the NRC published a Notice of Proposed Rulemaking on Decomissioning Criteria for Nuclear Facilities (50 FR 5600).

The proposed amendments would be applicable to 10 CFR Parts 30, 40, 50, 70, and 72 applicants and licensees and covered decommissioning alternatives, planning, assurance of funds for decommissioning, and environmental review requirements.

Final regulations based on the proposed rule and public comment on that rule and incorporating conclusions of the Final GEIS will be issued as effective at the time that the Final GEIS is published.

Coment No. 2 - Raises the question that current regulations on decomissioning are adequate, that the NRC has not indicated why new or amended regulations are needed, and that decommissioning criteria should be applied on a case-by ase basis. (16,23,25,34,35)

Discussion Currently, regulations and guidance pertaining to decommissioning of the facil-ities covered by this EIS are contained only within 10 CFR Parts 50 and 72, and in Regulatory Guide 1.86 and in similar NRC staff guidelines. However,15 discussed in the Draft GEIS Section 15 many of the existing regulatoly require-ments do not contain sufficiently specific consideration of necessary decommis-sioning requirements to assure that decommissioning is accomplished in a manner l which protects the health and safety of the public (although many of the requirements have been required as a condition of NRC licensing in case-by-case instances). There is need for more specific guidance especially 'n such areas '

as. assurance of funding, decommissioning alternatives, planning for decommis-sioning and environmental review requirements.

In the area of funding, the Commission has recently deleted requirements (see 46 FR 40) for financial qualification for electric utilities from 10 CFR 50.33(f) i and 10 CFR Part 50 Appendix C, with the proviso that there be rulemaking on specific requirements for funding of decommissioning in the near future. In addition, there is a need for funding requirements for materials facilities because of problems arisiiig from licensees' lack of funds for decommissioning and abandoning contaminated facilities, e c m :t In the area of planning for decomnissioning, there is a need for recordkeeping requirements so that 10/09/87 A-3 NUO586 APPENDIX A

1 decommissioning can be carried out in a manner which keeps occupational and

~

public radiation exposures as low as reasonably achievable. In the area of decommissioning alternatives, there is a need for criteria as to what alter-natives for completing decommissioning are considered acceptable.

l l It is the intention of the amended regulations to provide for specific guidance and consistent licensing effort for all facilities licensed by NRC. More detail on these areas are contained in NUREG-1221, Sections B.3.1, C 7.1, 0.8.1, E.1, and G.I.

Comment No. 3 - Indicates that there should be flexibility in the proposed decommissioning rules and that rules for reactors may not be applicable to materials facilities. (8,23,31)

Discussion It is NRC's intention that the rule amendments on decommissioning contain suf-ficient flexibility to take into account individual situations while still maintaining consistency in the overall licensing criteria. That this is the intention should be evident in such Draft GEIS sections such as 15.1.1 (which indicates the bases upon which different decommissioning alternatives could be used), and in Section 15.1.3 (which indicates that NRC will allow latitude in the implementation of financing mechanisms due to the wide diversity in differ-ent types of nuclear facilities). More detail on these areas are contained in NUREG-1221, Sections B.4.2, 0.3.1, and G.3. -

Comment No. 4 - Questions whether regulations are needed at this time since there is not a large number of facilities now nearing the end of their useful lives. (23)

Discussion Regulations are needed at this time to ensure that certain activities are initiated that are needed at this time to prepare for decommissioning.

Specifically, this includes such activities as providing assurance for the 10/09/87 A-4 NU0586 APPENDIX A

i funding of decommissioning, (for all types of facilities including reactors, fuel cycle facilities, and materials f acilities) and planning for the facili-tation of decommissioning, specifically recordkeeping. In addition, there is a sufficient number of different types of facilities that are now, or in the near future, undergoing decommissioning .nd hence consistent criteria for accom> aing their decommissioning it. needed.

Comment No. 5 - Raises the question that there should be separate rulemaking for premature decommissioning including that resulting from accidents. (23)

Discussion The proposed amendments apply to nuclear facilities that operate through their normal lifetime, as well as to those that may be shut down prematurely. This is consistent with the definition of decommissioning as presented in EIS Sec-tion 2.3. However, the activities following premature shutdown of a facility as a result of an accident are somewhat different than those of a routine decommissioning. There are three stages involved: a stabilization period, during which accident conditions are brought under control if necessary; an accident cleanup period: and a decommissioning period. During the accident cleanup, the major portion of contamination resulting from the accident is cleaned up and the associated wastes are processed. Following accident cleanup, the facility may either be recovered for reuse or be decommissioned. A detailed study of reactor decommissioning following accident cleanup (NUREG/CR-2601-Reference 7) indicated that there may be differences in some of the specific ~

aspects of decommissioning such as the spread of contamination, waste volumes, exposures, and costs. However, the report also indicates that the technology exists to accomplish the decommissioning and that the safe +.y and costs of decom-missioning tallowing the accident cleanup do not vary significantly from that following normal operations.

Comment No. 6 - Questions whether a separate decommissioning regulation should be prepared rather than incorporating into existing parts.

(34) 10/09/87 A-5 NUO586 APPENDIX A

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Discussion Se: tion 15.2 of the Draft GEIS and reprinted in this Final GEIS indicates the reasons for incorporating the regulations into existing parts.

A.I.2 Applicability of Regulations to Existing Facilities Comment No.1 - Questions whether proposed regulations should be applied to existing facilities and indicates that less stringent criteria should be applied. (16)

Discussion The general criteria of the regulations will be applicable to all facilities.

Thus the general provisions of funding, alternatives, and planning are appli-cable so that there is consistency in criteria. Specific requirements in these areas will allow for a reasonable period of time before funding assurance provisions must be instituted at existing facilities while recordkeeping pro-visions should be instituted following the rule becoming effective. Specific problems related to situations at existing facilities will be considered, most likely in regulatory guidance. More detail with regard to facilities already l shut down is contained in NUREG-1221, Sections C.9 and D.4.6.3.

A.1.3 Need for Cost-Benefit Analysis in Regulations Comment No. 1 - Raises the question that the regulations being considered have not been supported by an adequate cost-benefit analysis or value/ impact l ar.alys i s. (15,16,24,34) l Discussion A separate regulatory analysis has been submitted with the proposed decommis-sioning rulemaking (issued February 11, 1985) and a modified regulatory analy-sis supporting final rule requirements will be issued dealing with appropriate cost benefit analysis resulting from implenentation of the rule.

l 10/09/87 A-6 NUO586 APPENDIX A l

Comment No. 2 - Raises the question that the proposed regulations will have an adverse impact on the nuclear industry. (16)

Discussion Based on the conclusions of the DGEIS and FGEIS (Chapter 15) and the regulatory analysis in support of the rulemaking referred to in response to Comment No. 1 immediately above, it is concluded that rulemaking can be optimally implemented to assure health and safety requirements with a minimum of impact on the nuclear industry and generally will have a beneficial impact.

A.1.4 General Comments on GEIS Document Comment No. 1 - Questions whether the GEIS document should treat so many dif-ferent facilities. (16)

Discussion As discussed in the response to Comments No. 1, 2 and 3 in Section A.1.1, the purpose of the GEIS is to assist the NRC in developing new policies with re-spect to decommissioning all licensed nuclear facilities, and specifically in such a manner that these policies be implemented so that there is consistency in overall licensing and regulatory criteria while still maintaining sufficient flexibility to take into account the diversity in types of facilities.

Comment No. 2 - Questions whether the GEIS should establish standards properly within the province of EPA. (16)

Discussion As discussed in Section 2.5 of the FGEIS, selection of an acceptable residual radioactivity level is outside the scope of the rulemaking supported by this EIS. Proposed Federal Guidance is anticipated to be puolished by EPA and the NRC is planning to implement this guidance through rulemaking as soon as pos-sible af ter publication by EPA.

l l

10/09/87 A-7 NUO586 APPENDIX A

Coenent No. 3 - Questions whether the issue of assurance af the funding of decommissioning should be treated in the GEIS. (24, 30)

Discussion As indicated in Section 1.1 of the Draft GEIS, the purpose of the GEIS is to assist the NRC in developing new policies and in promulgating regulations with respect to decommissioning of licensed nuclear facilities. In Section 15 of the GEIS, policy matters recommended for inclusion in proposed regulations are indicated, one of which is assurance of funding for decommissioning. As is stated in Section 15.1.3, providing reasonable assurance of availability of funds ensures that decommissioning can be accomplished in a safe and timely manner and that lack of funds does not result in delays in decommissioning that may cause potential public health and safety problems. Hence, the issue of financial assurance is, in this instance, appropriate to treat in this GEIS.

Comment No. 4 - Raises the question that the GEIS should include discussion of rulemaking on issues related to decommissioning, as well as more detailed dis-cussion on the need for decommissioning regulations and their scheduled preparation. (23,38)

Discussion The GEIS discusses principally those issues related to decommissioning that are the subject of the regulations being amended. These include decommis-siening alt (rnatives, planning, financial assurance, and environmental review requirements. In addition, the GEIS discusses related areas, including waste management, safeguards and socioeconomic effects. The need for amended regula-tions is discussed in detail in Section 15 of the FGEIS. Pertinent regulations related to decommissioning are discussed in Sections 2 and 15 of the FGEIS.

10/09/87 A-8 NUO586 APPENDIX A

A. 2 Questions Related to Planning for Decormissioning A.2.1 Initial Plans Comment No. 1 - Some commenters question the usefulness of initial plans, specifically whether they have any use for facilities which would not be 1 decommissioned for several years, and indicating that therefore they should not be too rigid and should allow for change (2, 7,11,14, 23, 28, 31, 34, 35),

while other commenters raise the question that initial plans should be detailed, especially in the area of cost estimates. (32, 40)

Discussio_n The terminology of a specific requirement for submission of "initial plans" has been dropped from use. Those provisions necessary to be addressed in planning for decommissioning early in facility life have been retained. These are financial assurance and facilitation. In the area of financial assurance, applicants and licensees need to indicate the provisions for providing reason-able assurance of the availability of funds for decommissioning. These pro-visions include the method of funding and the amount of funds to be set aside, ac well as provisions for updating periodically over facility life. Specific criteria for the various types of facilities are different and are contained in the amended regulations. In the case of facilitation, the aspect of Tacilita-tion covered in the rule is recordkeeping. Licensees are to retain records important to decommissioning. However, submittal of information is not neces- '

sa ry. Other aspects of facilitation are net contained in the rule but are expected to be addressed in accordance with existing regulations and with regulatory guidance related to facilitation being considered.

With regard to the commenters requesting detailed initial plans, the require-ments in the final regulations are very specific regarding funding methods, funding amounts, and recordkeeping requirements. In addition to the specific requirements early in facility life, the rule contains update provisions.

Specifically, reactor licensees must submit updated detailed provisions for decommissioning five years prior to expected end of operations to take into account then current conditions related to decommissioning, as for example, 10/09/87 A-9 NUO586 APPENDIX A

waste disposal conditions. Witt, the specific requirements for preliminary planning early in facility life indicated above and update provisions, it is expected that decommissioning can.be carried out in a manner which protects public health and safety.

Comment No. 3 - Raises the question that the initial plan should not be required because it could delay licensing cases. (2. 7, 10)

Discussion As discussed above, initial decommissioning planning consists of financial assurance provisions and facilitation requirements, specifically recordkeeping for decommissioning. With regard to recordkeeping, licensees would be required to maintain but not submit records important to decommissioning. With regard to financial assurance, applicants and licensees would be required to submit provisions for funding as a reporting requirement in accordance with specific provisions contained in the rule. Further details in effect on pending licenses is contained in NUREG-1221, Section 0.4.1.2.

Comment No. 4 - Raises the question how the matter of initial plans should be applied to existing plants. (3, 40)

Discussion The primary purpose of financial assurance and recordkeeping requirements which -

make up the preliminary (or initial) planning part of the amended regulations is to provide information to establish adequate financial assurance provisions and to include consideration of facilitating decommissioning. As such, the need for these requirements are as necessary for operating plants as for new plants. As discussed in Section A.1.2 implementation procedures which are reasonable for decommissioning planning are contained in the amended regulations.

, 10/09/87 A-10 NU0586 APPENDIX A l

l

A.2.2 Updating of )lans Comment No.1 - Raises the question that the periodic updating of the initial plans should not be more frequent than once per five years and should not be the occasion for public hearings. (7)

Discussion As discussed in the replies to the previous commenter, the initial requirements are reporting ones and do not require explicit periodic update by the licensee.

Since they are entirely prescriptive, they do not offer occasion for public hearings.

A.2.3 Final Plans Comment No. 1 - Raises the question of the contents of the final plan. (7)

Discussion Final decommissioning plans would be submitted at the time of written notifi-cation that the licensee desires to terminate the license and would contain sufficient detail to permit an NRC determination that decommissioning can be accomplished safely. The content of the decot.11ssioning plan is discussed in Section 15.1.2.2 of this GEIS.

A. 3 Questions Related to Decommissioning Alternatives and to the Definition of Decommissioning A.3.1 Conversion of Facilities to Other Uses Comment No.1 - Raises the question that the GEIS should consider conversinn of l facilities to nuclear or non-nuclear uses, or the reuse or refurbishment of the existing facility. (2, 5, 7, 23, 34, 37) l l

10/09/87 A-11 NUO586 APPENDIX A L

s Discussion As indicated in Section 2.4 of the GEIS, conversion to a new or modified use, or refurbishment and reuse of a facility, is not ,onsidered in detail in the GEIS. This is because conversion, itself, is not considered to be a decommis-cioning alternative, whether the new use involves radioactivity or not, accord-ing to the definition of decommissioning as presented in GEIS Section 2.3. If the intended new use involved radioactive material and thus was under NRC licensing authority, an application for license renewal or amendment or for a new license would be submitted and reviewed according to appropriate existing regulations. If the intended new use does not involve radioactive materials, i.e., unrestricted public use, and does not come under NRC licensing authority, then such application for a new use would be reviewed as a request for decom-missioning and termination of license. In this case, the new use is not im-portant except as it affects the decommissioning alternative chosen. For these reasons, conversion to a new or modified facility is not considered further in this GEIS.

Comment No. 2 - Questions whether the conversion of a facility to a low-level disposal site should be considered. (35)

Discussion l

In general, the GEIS does not treat this issue for the same reasons as discussed in the response to Comment No. I above. With regard to the specific question -

of whether a nuclear reactor site could be converted to a low-level waste dis-posal site, this would involve licensing questions outside the scope of this GEIS. These questions would include the problem of evaluating whether the i reactor site was environmentally suitable as a low-level disposal site.

l i A.3.2 Use of a "No Action" Alternative l

l Comment No.1 - Questions whether there should be more detail on the "No Action" Alternative in the GEIS. (23, 30) l 10/09/87 A-12 NUO586 APPENDIX A

Y Discussion As discussed in Section 2.4.1 of the GEIS, "No Action" is not considered viable for any facility discussed in this GEIS, and hence it is not considered in any detail. The reasoning for this is discussed in Section 2.4.1.

A.3.3 Initiation of Decommissioning Comment No.1 - Questions whether the GEIS should discuss NRC authority to re-quire the initiation of decommissioning and identify NRC criteria under which decommissioning will be required. (37)

Discussion The question of NRC's authority to require the initiation of decommissioning is outside the scope of this GEIS. The purpose of the GEIS is to assist the NRC in developing regulations which will ensure that decommissioning is properly planned for and that, once begun, that decommissioning is carried out in such manner as to protect the health and safety of the public.

The rule amendments would require decommissioning plans for production and utilization facilities and ISFSIs to be submitted within two years following permanent cessation of operation or one year prior to operating license expira-

, tion. The decision as to whether a shutdown will be permanent is, of course, the licensee's. This provision does not limit how long a licensee may have a -

facility shut down under his operating license, but means only that when a facility is permanently removed from operational status, plans need to be made as to how the ultimate termination of license will be attained. Upon approval of the plans, the license will be modified to reflect the approved decommis-sioning alternative authorizing continued possession until the approved alter-native has beec carried out.

10/09/87 A-13 NUO586 APPENDIX A

A.3.4 Decommissioning Alternatives Comment No.1 - Raises the question that, in general, any regulations on alter-natives would have to be flexible, taking into account site-specific concerns; and in fact, alternatives should not be covered by a rule. (:.1, 23, 35)

Discussion As discussed in the Overview section of the GEIS, it is the responsibility of the NRC, in protecting the public health and safety, to ensure that after a nuclear facility permanently ceases operation the facility is decommissioned in a timely manner consistent with *.he particular nature of a specific facility.

Hence, general requirements regarding decommissioning alternatives must be included in decommissioning regulations. It is NRC's intention that proposed decommissioning rules provide sufficient flexibility to take into account individual situations while still maintaining consistency in the overall cri-teria and protecting public health and safety. Specifically, this approach can be seen in Section 15.1.1 of the GEIS which discusses the bases upon which different decommissioning alternatives could be used.

Comment No. 2 - Some comenters indicated that DECON (immediate dismantlement) should be the preferred alternative, and that if SAFSTOR is used, in no case should it be longer than 30 years especially for fuel reprocessing plants. (4, 31, 32, 36, 37) Other commenters indicated that the GEIS preference for DECON needs to be better justified; and that specifically there are health and safety '

implications for DECON, and that during DECON there should be delay time allowed for decay. (8,23,34) Other commenters indicated that there is insufficient justification of the problems indicated in the GEIS with delaying decommissioning. (2, 11, 23, 35) One commenter questions whether SAFSTOR shouldn't be allowed, at least in the case in which an owner maintains control of the site (8), and one questions why SAFSTOR is not allowed for greater than 30 years, especially since there could be technological improvements in the future which could further reduce the dose beyond 30 years. (11) 10/09/87 A-14 NU0586 APPENDIX A

Discussion ,

The advantages and disadvantages of DECON and SAFSTOR for the various types of facilities discussed in this GEIS are discussed in detai! in Sections 2.4 and 15.1.1 of the GEIS as well as in the specif1c sections for each facility (Sections 4 through 14 of the GEIS). Based on the analysis in those sections, Section 15.1.1 concludes that DECON or 30 to 50 years SAFSTOR are reasonable options for decommissioning a light water power reactor. Delay beyond that time would have to be justified based on unavailability of waste disposal capacity or site specific f actors affecting safety such as presence of other licensed facilities on the site. Section 15.1.1 also concludes that for research and test reactors and independent spent fuel storage facilites, DECON ,

is the most reasonable option although SAFSTOR could be justified in some cases. For fuel cycle and non-fuel-cycle facilities associated with licenses under 10 CFR Parts 30, 40, and 70, Section 15.1.1 indicates that DECON is the most reasonable option and although SAFSTOR is possible for short-lived saterials, any extended delay would rarely be justifiable. More detail on areas of DECON and SAFSTOR is contained in NUREG-1221, Section B.4.

Comment No. 7 - Questions whether there should be special considerations for allowing SAFSTOR for ore processing facilities. (15)

Discussion In the case of tailings piles, SAFSTOR may be justifiable until provision for

~

l removal of tailings, if necessary, can be accomplished. At the present time, tailings disposal would be on a specific case basis and could possibly be accommodated at phosphate or mill tailings piles that would ultimately require stabilization.

Coannent No. 8 - Expresses the opinion that use of ENTOM8 at power reactors should be acceptable, especially in light of cost concerns and the ability to store wastes in the entombed structure. (11, 23, 25, 30, 35) 10/09/87 A-15 NV0586 APPENDIX A

Discussion Sections 4.3.3 and 5.3.3 of the GEIS discuss the advantages and disadvantages of the ENTOMB alternative, and Sections 4.5 and 5.5 compare the ENTOMB alter-native with the other decommissioning alternatives. These discussions are based to a '.arge extent on information and data developed on ENTOMB by Battelle PNL for the NRC. In aodition, Section 2.4 and Section 15.1.1 analyze the ENTOMB alternative. The GEIS sections indicate that ENTOMB, with the internals entombed, does not appear to be a viable alternative due to the presence of the long-lived nuclides NiS9 and Nb95 which would be present for thousands of years.

If a facility were entombed with the internals removed, it may be possible to release the site for unrestricted use at some time within the order of a hundred years. However, one of the dif ficulties with ENTOMB for any complex structure such as a reactor is that the radioactive materials remaining in the entombed structure would need to be characterized well enough to be sure that they will have decayed to acceptable levels at the end of the surveillance period. Some method would have to be provided to demonstrate that the entcabed radioactivity will decay to levels permitting unrestricted use which would be difficult. The ENTOMB alternative appears to be less desirable than either DECON or SAFSTOR based on consideration of the fact that ENTOMB results in higher radiation exposure and higher initial costs than 30 year SAFSTOR, that the overall cost of ENTOMB over the entombment period is approximately the same as DECON, and the fact that regulatory uncertainty after the long entombment period might result in additional costly decommissioning activity in order to release the More detail in this area is contained in facility for unrestricted use.

NUREG-1221, Section B.S.

Comment No. 9 - Raises the question that health and safety differences between alternatives are not great and that costs and alternative uses of the facility should be considered, especially those uses which do not require full decommis-sioning (as DOE has done with some of its facilities). (34)

Discussion See discussion of answer to item A.3.1 "Conversion of facilities to Other Uses", comment 1.

10/09/87 A-16 NU0586 APPENDIX A

~ - ___ _ _ _. -

l Coment No.10 - Points out that during SAFSTOR or ENTOMB only a v iry small j

. portion of the land area originally covered by plant buildings would need to l

be restricted. (34)

Discussion Provided that NRC licensing conditions were suitably modified to redefine the radioactive constituents of the facility requiring restricted use categoriza-tion, only the small portion of land originally covered by the plant buildings could be controlled and the rest be classified as unrestricted.

Coment No.11 - Questions why the NRC has indicated a 100 year period on institutional controls for radioactivity confinement. (16)

. Discussion Although the DGEIS indicated a 100 year period for institutional controls of radioactive confinement, based on an old EPA draft policy, the FEIS has removed this comment and replaced it with a more general recommendation that institutional control reliance could be reasonable for the order of 100 years.

This is also consistent with the section on institutional controls in 10 CFR 61 concerning low-level waste burial grounds.

Comment No.12 - Some comenters question the definition of decomissioning which requires that the facility be returned to unrestricted use (11,12,16, 23, 30, 34, 35). One comenter agrees with the requirements that the facility be released for unrestricted use, but raises the question that more detail be given as to what facilities be released. (36)

Discussion The definition of decommissioning as expressed in the GEIS prevides a descrip-tion of the process in a regulatory framework. Specifically, it is the process of removing a facility safely from service and reducing residual radioactivity 10/09/87 A-17 NUO586 APPENDIX A

to a level which permits release of the facility for ;nrestricted use and termination of the license. This definit'on expresses the complete process of decomm :sicaing and puts it into the context of reaching a safe point.

It is tb Comission's belief that there is nothing in +he definition which would inhibit future use of the site once t"e license is terminated. Unre-stricted use refers to the fact that from a radiological standpoint, no hazard exists at the site, the license can be terminated, and the site can be considered an unrestricted area. This definition is consistsnt with the definition of an unrestricted area as it exists in 10 CFR 20.3 as being "any area access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials and any area used for resid2ntial quarters." The specific futura use of the site after the license is terminated is outside the scope of the GEIS. With regard to reuse of the site for nuclear purposes, there is nothing in this GEIS preventing such reuse. As ir.dicated above, reuse of the nuclear facility for other nuclear purposes is not considered decommissioning. Therefore, a licensee would not be required to submit a decom-missioning plan or apply for termination of license.

The rule also does not limit the use of alternative decomissioning methods which delay the completion of decomissioning thereby not releasing the site for unrestricted use during a period of radiological decay as long as the methods provide reasonable assurance of protection of public health and safety and there is a benefit in the use of the delay. The definition of decomis-sioning as well as the definitions of the alternatives contained in Sections 2.4 and 15.1.1 of this GEIS indicate that, if permanent cessation of nuclear activity occurs at the facility, the licensee is to propose to NRC the method that it intends to use in decomissioning the facility in a manner ulti-mately leading to the return of the site to an "unrestricted area" according to the definition of 10 CFR 20.3 and the termination of the facility license.

10/09/87 A-18 NU0586 APPENDIX A

A.4 Questions Related to Acceptable Resitual Radioactivity levels at Decommissioned Facilities A.4.1 General Requirements for Setting Residual Radioactivity Levels Comment - Several commenters raise questions regarding setting of residual radioactivity levels. Some (1, 16, 31) said EPA has authority to set such criteria and NRC should, therefore, not precede EPA in setting such criteria, while some (23, 34, 40) said that regulations covering residual radioactivity are not needed now, especially in light of lack of high-level waste disposal criteria, and one (10) said residual limits shculd be set by the Radiation Protection Council. Several commenters made specific comments on the numerical

.value of the residual limit, how it should be chosen, and the dose pathway modeling which should be used one commenter indicated that residual limits for ore processing facilities should be set on a case-by-case basis. (2, 4, 7, 8, 9, 11, 15, 16, 23, 30, 32, 33, 34, 35, 37, 40).

Discussion Comment Analysis and Response The selection of an acceptable level is outside the scope of the rulemaking supported by this GEIS. Proposed Federal guidance is anticipated to be published by EPA. NRC is planning to implement the EPA guidance through rulemaking as soon as possible after it is issued. The Commission is participating in an EPA -

organized interagency working group which is developing Federal guidance on acceptable resic'ual radioactivity for unrestricted use. Currently, criteria for residual contamination levels do exist and research and test reactors are being decommissioned using present guidance contained in Regulatory Guide 1.86 for surface contamination plus 5 pr/hr above background measured at I reter for direct radiation. As an example, NRC provided such criteria in letters to Stanford University, dated 3/17/81 and 4/21/82 providing "Radiation criteria for release of the dismantled Standard Research Reactor to unrestricted access."

10/09/87 A-19 NUO586 APPENDIX A

l l

A. 4. 2 Termination Survey Conment No.1 - Raises the question as to what nuclides will be considered in determining what are the principal nuclides for surveying, with concern that certain nuclides, which have longer half lives but may be initially insignificant, would be ignored. (37)

Discussion The principal nuclides for surveying should be those that offer the best signature for detection (e.g. , strong gamma emitters such as 80Co for reactors).

Generally those nuclides will also be the greatest dose contributors. Based on some reasonable nuclide spectrum analysis, it should be possible to demonstrate that removal of these signature nuclides to some acceptable level will result in adequate removal of non-signature nuclides with longer half-lives so that the dose contribution from those that remain will be acceptable. Of course careful spectrum analysis of a few representative cleaning areas should be performed to provide additional assurance that radioactive contamination has been properly performed.

Comment No. 2 - Questions whether there is measurement detection capability which is cost effective to measure concentrations corresponding to the acceptable residual radioactivity levels. (10,16,26)

Comment No. 3 - Questions whether the cost to decontaminate facilities to residual radioactivity levels corresponding to 10 mrem / year has been adequately addressed in the GEIS. (30, 34)

Discussion See discussion in answer to Comment A.4.1.

10/09/87 A-20 NUO586 APPENDIX A

A. 5 Questions r elated _to Financial Assurance A.S.1 Costs of Decomissioning Comment 1 - Raises the question that the GEIS should indicate more clearly the uncertain nature of the cost estimates made. (23,30) Raises the question that the NRC should require that licensees obtain detailed cost estimates specific to their facilities and location rather than have them rely on Battelle PNL reports and their subsequent sensitivity analyses. (57)

Discussion Sections 2.6 2 and 15.1.3 of the GEIS include discussions which recognize the uncertainties in the cost of decommissioning various nuclear facilities.

Table 15.0-2, which is a summary of estimated costs for decommissioning nuclear facilities, indicates that the cost figures include a 25% contingency factor which can account for unforeseen events that might impede the conduct of the decommissioning work. In addition, the GEIS sections on LWRs (Sections 4 and

5) include sensitivity analyses which assess the variability in costs of decom-missioning depending on various factors such as reactor size, plant design, contamination levels and waste disposal considerations. Also, it is indicated in Sections 2.6.2 and 15.1.3 of the GEIS that the funding levels will require updating over the life of a facility to assure that adequate funds are available for decommissioning.

It is not the intention of the GEIS to indicate that when cost estimates are submitted, NRC will accept cost estimates based solely on the Battelle PNL reports. However, due to limited experience in decommissioning, the Battelle PNL reports are useful for preliminary cost estimating. In using these reports to make cost estimates, a licensee must make suitable adjustments to account for facility differences and to make periodic revisions to his cost estimates.

More detail in this area is found in NUREG-1221, Sections D.1.1 and 0.2.1.

Comment 2 - Raises the question that since costs are given in 1978 dollars, how would escalation affect costs. (3) l 10/09/87 A-21 NUO586 APPENDIX A

7 Dis ussion The costs in the final GEIS are given in 1986 dollars and account for the effects of escalation since 1978. Costs are given in present value dollars with the intention that decommissioning funds will be set aside in such manner that the principal plus accumulated interest, plus adjustments as necessary, will cover the effects of inflation on decommissioning costs.

Comment 3 - Raises the question that the GEIS has not clearly presented the type of cost being listed in tables, namely whether they are discounted or 4

undiscounted, so as to be able to properly compare costs of alternative plans which would take place over different time frames. (9)

Discussion Comment 4 - Questions where in references 1 and 3, cited on page 0-7 and listed on page 0-46 of the draft GEIS, that there is a discussion of the sensitivity of the cost of decomissioning to the dose level from residual radioactivity.

(38)

Discussion In Section 9.1.1.2 of NUREG/CR-0130 there is a discussion which indicates that the cost of decontamination of surfaces as estimated in that report is '

essentially independent of the level to which it must be decontaminated as long as that level is in the range of 10-25 mrem /yr to an exposed individual.

Section 6.4 of NUREG/CR-0278 indicates that the report considers decommis-sioning activities necessary to release the facility for unrestricted use for both 10 and 25 mrem / year values. In the cost analysis of decommissioning in NUREG/CR-0278, only one set of cost estimates is presented since the report assumes that the values are essentially the same whether the acceptable residual level is 10 or 25 mrem / year.

In addition to those discussions, reference 1 of this section presents addi-tional discussion of the basis for the statement that a difference in the 10/09/87 A-22 NUO586 APPENDIX A

acceptable residual radioactivity level between 10 and 25 mram/ year would have relatively small impact on the total decommissioning cost. (For additional discussion, see response to comment 3 in Section A.7.2.)

Comment 5 - Raises the question that the cost of decommissioning should include the cost of having the decommissioning effort performed by a contractor. (33)

Discussion Sections 4.3.4 and 5.3.4 of the GEIS have been revised to include the impact on decommissioning costs (included in References 2 and 3) of having contractors perform the bulk of the decommission effort at reactors while the licensee retains certain overview and control functions. These references indicate that use of such contractors is likely for these large facilities.

For material facilities, the cost estimates do not specifically include the assumed use of contractor costs because: amounts listed are considered reasonable in providing adequate funds so that a facility does not become a concern to public health and safety. The additional expense associated with requiring all material licensees to set aside in their funding method the added costs of assuming use of a contractor is not justified compared to the small number of licensees expected to have to use contractors. The increased cost of use of a contractor is not expected to be as large as suggested by the commenter. -

l Comment 6 - Raises the question that the GEIS should provide better detail of costs for certain material facilities where the survey costs may be significant.

(38) i Discussion -

Survey costs for five typical material facilities are presented in NUREG/CR-2241, which presents estimated survey costs for various types of nuclear facilities.

The costs of the termination surveys for the material facilities considered can be compared to the overall costs of decommissioning these facilities which are 10/09/87 A-23 NU0586 APPENDIX A

o ,

presented in NUREG/CR-1754 In general, for these material facilities, the cost of the terminal survey is estimated to be approximately 5% of the overall cost of decommissioning assuming an acceptable residual radioactivity level in the range of 10 to 25 mrem / year to an individual.

For material facilities which require little or no decontamination effort, either because the source of radioactivity is sealed, or short-lived, or there has been no spread of contamination, it is intended that the survey effort will be minimal and of low cost.

Comment 7 - Questioned the accuracy of the cost estimates in the GEIS stating that they are too low, especially in light of the high cost of the operational decontamination at Dresden 1. (14)

Discussion The cost information contained in the GEIS is a summary of costs developed in a series of reports prepared by Battelle-PNL on the technology, safety, and costs of decommissioning nuclear facilities. The purpose of these reports has been to develop a data base on decommissioning nuclear facilities to support an NRC reevaluation of its decommissioning policy. The PNL reports are detailed engineering evaluations of the activities involved in decommissioning nuclear facilities. The reports consider: (1) the detailed design and layout of the reference facility; (2) estimated conditions in the facility at the time of shutdown (just prior to decommissioning) including estimates of radionuclide inventory and radiation dose rates; (3) techniques for decontamination and dismantling which are current and proven; and (4) radiation protection require-ment for workers and the public. Based on these considerations, the PNL reports develop detailed work plans and time schedules to accomplish decommissioning, including those for planning and preparation, decontamination, and component disassembly and transport. In ' king costs estimates of decommissioning, the PNL reports include such matters as work scheduling estimates, staffing requirements, specialty contractors, essentici systems, radioactive disposal, and supplies.

10/09/87 A-24 NUO586 APPENDIX A

Although it may be dif ficult to make comparisons between different cost estimates for different facilities because of site-specific considerations, it can be said that the PNL estimates represent reasonable approximation of the range of decommissioning costs, in particular because they use eigineering assumptions and are based on decommissioning experiences. Other estimates, made independent from PNL and made using engineering assumptions, are in the same general cost range as PNL. Other estimates may be higher but careful review of the assumptions used should be made such as whether they use engineer-ing assumptions or only extrapolations, whether they are in current milars or future year dollars, i.nd whether they include they cost of demolition and site restoration in the cost estimate. The PNL costs presented in this GEIS are in 1986 dollars and do not include the costs of demolition of nonradioactive struc-tures and site restoration after termination of the NRC license.

More detail on the basis of the PNL studies and comparison with other estimates is contair.ed in NUREG-1221, Section 0.1.1.

Specifically, with regard to cost estimates made for the operational decon-tamination of Dresden 1, it is incorrect to compare the cost of decomissioning a plant to the cost of decontaminating an operating plant with the intention of returning it to service. Specifically, in the case of the operating plant it is necessary to do extensive testing and analysis to check material compati-bility with decontamination solutions for eventual restart of the reactor. It is also necessary to run the decontamination process under very controlled conditions so as not to damage pressure boundary material. In addition, there -

will be additional system flushings necessary to ensure that the system is free of decontamination solutions before it is restarted. These additional system flushings can generate large volumes of additional radioactive waste which must be processed, packaged, and disposed of. These additional activities, which can be costly, are not necessary for a decommissioning in which the intent is to dismantle the plant and material compatibility is not as large a concern.

In addition, the Dresden 1 facility is an atypical situation. The Dresden 1 project was a research and development study for the purpose of demonstrating the feasibility of decontaminating plant systems to reduce occupational exposure prior to a plant resuming further operations. When returning a plant to field 10/09/87 A-25 NUO586 APPENDIX A

service, great care hac to be taken to ensure that the decontamination solut ons and procedures used do not adversely affect the plant's systems. Therefore the precedure used at Dresden was relatively costly since it was high'y controlled as a research project. Conversely, decontamination solely for the purpose of reducing the worker dose prior to the initiation of decommissioning would not require the same level of system protection since the systems would never be intended for further use. The system at Dresden consisted of a much larger, more complex set of systems than the portable systems employed today for primary system decontaminations by several nuclear service companies. The costs of a single primary system decontamination is estimated by the service companies to be in the range of $1 to $3 million, depending upon site-specific circumstances.

The system decontamination described in NUREG/CR-0130, including waste treatment but excluding waste disposal, was estimated to cost about $484,000 in 1978 dollars. When escalated to 1984 dollars, that cost becomes $1.07 million, in reasonable agreement with the prices currently quoted by nuclear service companies.

Comment 8 - Questioned the higher cost of decommissioning BWRs vs. PWRs as given in the GEIS, siating that the higher BWR costs are based on more restrictive assumptions regarding allowable occupational dose thus resulting in higher costs, and that higher costs for special equipment for BWRs are estimated. (6)

Discussion The PNL studies for PWRs and BWRs have been updated and a summary of the results' is contained in Sections 4 and 5 of this GEIS. In the updating, the assumptions regarding allowabla occupational dose have been put on a common basis thus allowing better comparison of results.

Comnent 9 - Raises the question that the GEIS should contain more detail on the matter of unforeseen expenses should there be cost overruns at low-level waste burial sites due to engineering and/or management control problems. (36) 10/09/87 A-26 NUO586 APPENDIX A

Discussion Se:tions 4 and 5 of the GEIS have been revised to include an evaluation of the technology, safety, and costs involved if it is ne:essary to store wastes onsite past the expiration of a facility operating license due to pro 51 ems at disposal sites.

Costs of waste disposal are included as part of the decommissioning costs in Ta:1e 15.0-2. These costs are based on data developed in the Battelle PNL reports. Tt.e reports develop waste disposal costs by determining the volume of waste which must be buried, the turie content of the waste, and the costs of burial. The costs include a 25% contingency factor to account for unforeseen difficulties in carrying out the activities.

Specific details on problems at low level burial sites which may cause burial costs to increase in the future were beyond the scope of the original Battelle PNL reports. It is intended that future revisions of these reports will consider updated considerations of burial site costs.

As discussed in Section 15.1.2.1, licensees will be required to submit 5 years prior to the projected end of operation up-to-date cost estimate on which to base financial assurance. In particular, this estimate would be based on a current estimate of major factors that could affect decommissioning costs, as for example, then-current problems at low-level waste burial facilities. This requirement is intended to ensure that the licensees consider relevant up-to- -

date information which could be important to adequate planning and funding for decommissioning well before decommissioning actually begins.

A.5.2 NRC Authority in the Area of Financial Assurance Comment 1 - Quastioned the authority of the NRC to write regulations in the financial assurance area, and specifically to allow certain funding methods, while precluding others. (2,16,24,34,35) r 10/09/87 A-27 h00586 APPENDIX A

{L

{/

Discussion TN/ Commission's statutory mandate to protect the radiological health and safety of the public and promot.' 'he common defense and security stems principally from the Atomic Energy Act and Energy Recrganization Act. In i carrying out its licensing and related regulatory responsibilities under these acts, the NRC has determined that there is a significant radiation hazard associated with nondecommissioned nuclear facilities and that the public health and safety can best be protected by promulgating a rule requiring reasonable assurance that at the time of termination of operations adequate funds are evailable so that dec'mmissioning can be carried out in a safe and timely manner and that lack of funds does not result in delays that may cause potential health and safety problems, Although these acts do rfst permit the NRC to regulate rates or to interfere with the decisions of State or Federal agencies respecting the economics of nuclear power, they do authorize the NRC to take whatever t-egulatory actions may be necessary to protect the public health and safety, including the promulgation of rules prescribing allowable funding methods for meeting decommiscioning costs.

More detail on this area is contained in NUREG-12n, Section 0.8.1 and 0.8.3.

Becaus.) of the diversity of NRC licensees and facility types, as discussed in SectiNs 2.6.2 and 15.1.3 of the CEIS, the NRC will allow latitude in implementation of funding methods to provide reasonable assurance of funding.

Comment 2 - Questioned the authority of NRC to require sureties, stating that Congress granted that authority only for the regulation of uranium mills in Section 203 of the Uranium Mill Tailings Radiation Control Act of 1978. (16)

Discussion As discusred in Comment 1 of Section A.S.2, NRC has authority to require reasonab'.e assurance of the availability of funds to decommission a facility based on its responsibility as stated in the Atomic Energy Act to protect the health and safety of the public. NRC has used its authority not only to reqeire sureties for the decommissioning of uranium mills, but has used its i

10/09/87 A-28 NUO586 APPENDIX A

authority under the Atomic Energy Act .o require sureties to provide assurance of funds for the closure and stabilization of low-level waste burial grounds in Section 61.62 of 10 CFR Part 61, "Licensing Requirements for Land Disposal of Radioactive Waste."

A.S.3 The Level of Assurance Required Comment 1 - Disagreed with the GEIS statement that a "high" level of fir.ancial assurance was necessary for decommissioning, and indicated that the NRC should require "reasonable" levels of financial assurance. (7, 23, 24, 3J)

Discussion The GEIS has been revised to indicate that the NRC will require that there be reasonable assurance that funds for decommissioning will be available when necessary.

A.S.4 Acceptable Funding Methods

, A.S.4.1 Need for Flexibility in Funding Methods Comment 1 - Raises the question that because of the different types of reactor licensees, that NRC requirements must be flexible, and that it would be better to have case-by-case evaluations based on the specific licensee situations and general guidelines. (12, 24, 27, 29) -

Discussion The staff agrees with this comment. As discussed in Sections 2.6 and 15.1.3 of this final GEIS, the NRC is allowing latitude in the use of funding methods, based on two criteria. The first and most important criterion from the Com-mission's standpoint is reasonable assurance that funds will be available in a timely manner for safe decommissioning. Based on this criterion, certcin funding methods are deemed acceptable in the proposed rule for providing reasonable assurance of funds. Latitude for choosing among these methods is permitted by the amendments to take into account other issues which are normally 10/09/87 A-29 NUO586 APPENDIX A

4 l

l l

l outside NRC's jurisdiction including rate collection, ratepayer cost, taxation ef fects, whether a method is equitable to ratepayers, and other local concerns.

l A.S.4.2 Commenter Opinions Regarding Funding Methods Comment - Some commenters indicated support for use of prepayment and external funds as the only allowable funding methods (1, 5, 16, 27, 32, 36, 37, 38, 40).

Other commenters indicate that there should be more flexibility in NRC rules and that internal reserves should also be allowed since there is a significant cost advantage to the internal reserve and that the internal reserve provides reasonable assurance of funds for decommissioning.

Discussion A revised discussion of acceptability of funding methods in terms of providing reasonable assurance of funds for decommissioning is contained in Sections 2.6 and 15.1.3 of the Final GEIS. The NRC has considered the use of various fund-ing methods, and in particular internal reserve, in several documents and has reviewed public comments on the proposed rule and the draft GEIS. Based on these documents and on the discussions presented in Section 2.6.2 and 15.1.3 of this GEIS, and presented in more detail in NUREG-1221, Sections 0.3.2.1 and 0.3.2.2, using a standard of providing reasonable assurance th .t suf ficient funds are available for decommissioning, electric utility lictusees may use prepayment or external reserve, and electric utility licensees owning more than one generating facility may also use internal reserves. As noted above, more

~

l.

detail in this area is contained in NUREG-1221, Section D.3.2.1 and D.3.2.2 A.S.4.3 Procedural Questions on Funding Comment 1 - Indicated that, because current financial provisions for decommissioning are inadaquate, that collectioe of funds should begin promptly and that there should be more detail on requirements for existing plants in the GEIS. (9, 27) l l

I

(

10/09/87 A-30 NV0586 APPENDIX A

f i

Discussion Revised Sections 2.6 and 15.1.3 contains additional discussion concerning financial assurance requirements for operating plants. Upon issuance of an effective rule on decommissioning, current licensees will indicate to the NRC their provisions for providing funds for decommissioning within two years after the issuance of the final rule. Additional discussion on how existing licensees should carry out these activities is contained in Section A.1.2 of this Appendix.

Comment 2 - Raised the question that the regulatory approach of the NRC has not been able to deal with sufficiant specificity on financial matters. (1)

Discussion Sections 2.6 and 15.1.3 indicates funding methods considered acceptable to the NHC in assuring availability of funds for decommissioning. Section 15.2 indicates the intent of the NRC to publish decommissioning regulations covering the issues presented in Section 15.1. These regulations will contain specific requirements on allowable funding methods and on setting funding levels.

Comment 3 - Raises the question of how NRC will work with the state PUC's in assuring availability of funds. (40)

Discussion NRC has included in its amended regulations funding provisions considered acceptable in protecting public health and safety. This is similar to other health and safety matters contained in the Commission's regulations. State

! PUC's are responsible for setting a utility's rates so that all reasonable costs of serving the public are satisfied, including costs of adhering to NRC regulations concerning decommissioning. Provisions contained in the amended regulations are very specific and NRC does establish specific requirements for indicating to NRC how reasonable assurance will be provided that funds will be available for decommissioning. Specific financial and local issues, such as rate of fund collection, procedures for fund collection, cost to ratepayers, taxation effects, equitableness, accounting procedures, ratepayer versus 10/09/87 A-31 NUO586 APPENDIX A

itockholder considerations, and responsiveness to change, will not be ad:ressed by. NRC but will be lef t to state PUCs to determine. The final rule recognizes that funding for decommissioning of electric utilities is also scject to the regulation of State and Federal agencies (e.g. , FERC and state PUCs) having jurisdiction over rates, and that the NRC requirements are in addition to, and not substitution for, other requirements, and are not intended to be used, by themselves, by other agencies to establish rates.

Hence, NRC does not intend to become involved as part of the decommissioning rate regulation process. More detail in this area is contained in NUREG-1221, Section D.8.3.

A.S.4.4 Opposed to GEIS Funding Recommendations for Fuel Cycle and Non-Fuel-Cycle Facilities.

Comment 1 - Two commenters raised the question that financial requirements will impose a financial burden on non-fuel-cycle facilities engaged in radiopharmaceutical medical research and development and clinical laboratory facilities, and also on tantalum manufacturers placing them at a disadvantage to foreign competitors. (15,31) Another commenter (16) raised the question l

that self-insurance should be allowed since there is no evidence that it is not suitable, and that certain licensees are at least as financially sound as bonding or insurance companies or banks. This comenter also raised the question that sureties should not be required because there is no evidence in the GEIS that any licensee has ever defaulted in carrying out pertinent decom-missioning requirements, and because they may not be available to licensees, ~

! and that they are not necessary since the NRC would not issue or renew a license if a licensee were not prepared to carry out decomaissioning.

l l

Discussion The types of funding methods discussed in this GEIS, and allowed for materials I

licensees in the amended regulations, are consistent with those contained in earlier NRC promulgated rules in 10 CFR Part 40, Aopendix A, regarding require-ments for funding the decommissioning of uranium mills and mill tailings, and in 10 CFR Part 61 regarding funding for closure of low-level-waste burial facilities. The Commission found in developing those requiremen's that self 10/09/87 A-32 NUO586 APPENDIX A

insurance for a private sector applicant or licensee would not be an acceptable fors of surety. Even if a private sector applicant or licensee is currently adequately capitalized, a lack of funds at the time of decommissioning, which may not occur for several years in the future, can cause problems with complete decommissioning. Problems such as bankruptcy have arisen in recent years with NRC licensees and Agreement State materials licensees not having sufficient funds for decommissioning.

As part of the effort involved in preparation of the proposed rules, NRC pre-pared a Regulatory Analysis, which evaluated the benefits and costs associated with the requirements contained in the proposed rules. The Regulatory Analysis indicates that the large majority of NRC licensees are exempted from the specific requirements on demonstrating financial assurances, although they are neverthe-less financially responsible for paying for decommissioning as well as carrying out decommissioning. Those exempted include those possessing smaller quantities of radioactive material than prescribed in the regulations, those using sealed saterials and those using material with half life less than 120 days. In addi-tion, for many of those remaining licensees who must demonstrate funding assur-ance, a certification of an amount and funding method as prescribed in the rule would be sufficient. For those remaining licensees who must submit a funding plan, the plan would only be required at the time of license renewal at which time it is much more efficient for the licensee and staff to implement as part of the overall renewal effort. The regulatory analysis evaluated the costs associated with submittal of these funding plans. Based on these costs and on the number of exempted licensees, the regulatory analysis concluded that the moderate increase in overall costs to the NRC and the industry is balanced by l the important increase in the effectiveness of decommissioning activities that will assure that impacts on health, safety, and the environment are minor.

As an additional effort to minimize impacts while maintaining reasonable assurance that funds are available for decommissioning, the NRC has decided to modify the proposed rule to permit the use of parent company guarantee when accompanied by financial tests for licensees. This is consistent with NRC's Policy Guidance Regarding Parent Company and Licensee Guarantees for Uranium Recovery Licenses issued in December 1985.

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10/09/87 A-33 NUO586 APPENDIX A l

Th's area is discussed in more detail in NUREG-1221, Section D.6.

Comment 5 - Raised the question that financial assurance provisions should not be extended to facilities currently undergoing decommissioning. (16)

Discussion Pe response in Section A.1.2 of the Appendix.

A.5.5 Funding for Premature Decommissioning, for Reactors Including Post-Accident Decommissioning Comment 1 - Disagrees with the GEIS discussion or finds insufficient detail on funding for premature decommissioning, in particular post-accident cleanup and decommissioning. (7, 12, 23, 24, 28, 29, 32, 35, 37, 38) Also raised the question of how funding will be available for non-accident premature decommissioning. (5, 7, 12, 23, 24, 28, 29, 32, 35, 37, 38)

Discussion Revised Section 8 of the GEIS entitled "Decommissioning of Reactors Which Have Been Involved in Accidents," based on a Battelle-PNL report on post-accident cleanup and decommissioning, contains information on the technology, safety, and costs of prematurely decommissioning a reactor which has been involved in

, an accident. -

The availability of funds for post-accident cleanup is related to financial assurance for decommissionin0 The costs of post-accident cleanup can be sub-stantially larger than the costs of decommissioning. Assurance of funds for post-accident cleanup activities is more properly covered by use of insurance.

Post-accident cleanup activities are broader in scope than decommissioning, that is, they can lead ultimately to either reuse or decommissioning. Accord-ingly, the funding requirements for accident cleanup are not included in the GEIS or in these amended rules but are contained in 10 CFR 50.54(w) which requires that utility licensees for production and utilization facilities 10/09/87 A-34 NUO586 APPENDIX A

obtain insurance to cover decontamination and cleanup costs associated with onsite property damage resulting from an accident.

With regard to the funding of decommissioning activities which would occur prematurely either following an accident or if an. accident did not occur, NRC has had several studies done to address this issue. These include NUREG-0584, NUREG/CR-1481, NUREG/CR-3899, NUREG/CR-3899 Supplement 1, and NUREG/CR-2370.

These documents address the question of assurance provided by the various funding methods, including prepayment, external reserve, internal reserve, and insurance. In particular, as discussed in Section 2.6 of the EIS and in more detail in NUREG-1221, Section 0.3.2.1.1, NUREG/CR-3899 notes that the market value of utilities, even those involved in the most extreme financial crises, is still far in excess of decommissioning costs and that the value of the assets of a utility both tangible and intangible are more than adequate to cover future projected decommissioning costs. These considerations must also be viewed within the context of the Commission requirements for onsite property damage insurance in 10 CFR 50.54(W), discussed above, the proceeds from which a utility could use to decontaminate its reactor af ter an accident. Although these insurance proceeds would not be used directly for decommissioning, they would go a long way toward reducing the risk of a utility being subject to a tremendous demand for funds af ter an accident. Because most utilities are now carrying insurance in excess of $1 billion and the Commission has implemented its proposed requirement in 10 CFR 50.54(w) for insurance at this level, a sajor threat to long term utility solvency will have been substantially reduced.

In' addition to the factors discuused in Section 2.6 of the EIS and in more detail in NUREG-1221, Section D.3.2.1.1, the considerations in NUREG/CR-3899

! and the presence of the accident insurance provided by 10 CFR 50.54(w) one needs to balance the benefit of the reasonable assurance criteria against the l cost or practicality of assurance. Methods that could be usec to handle

( premature decommissioning include prepayment of funds, external reserve, insurance, and sureties. However, prepayment of funds has been recognized by several studies as being significantly more costly than the other methods.

Furthermore, in view of the unlikely nature of the events and the potential problems being considered, prepayment has a cost too high for the benefit 10/09/87 A-35 NUO586 APPENDIX A i l

4 ,that would te realized. External funding would not by itself provide aeditional assurance for premature shutdown. Earlier studies in NUREG-0584 found that surety bonds were not generally available in the amounts necessary fcr decommissioning power reactors. Use of insurance for nonaccident related decommissioning was found in an earlier study performed for the NRC, NUREG/CR-2370, to have potentially serious problems of insurability and moral hazard and .is not cur ently available. (Moral hazard is a term used in the insurance industry to indicate a situation of lack of loss prevention or loss control because those insured have access to risk prevention.)

In light of the factors considered, including the assurance provided by the various methods, the unlikely nature of the various events and the cost and practicality of providing more absolute assurance by certain methods, it is concluded that the funding methods provided in the proposed rules are adequate.

More detail in this area is found in NUREG-1221, Section 0.3.2.1 and 0.3.2.2.

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A.6 Questions Related to the Effect on Decommissioning of tie Unavailability of Waste Disposal Capacity 1

Comment 1 - Some commenters raised the question that the NRC's decommissioning regulations must consider the effect that the unavailability of high-level waste and low-level waste disposal capacity will have on the capability to decom-sission a facility. Two commenters raised the question that there needs to be low-level waste and high-level waste disposal regJlations before questions about I

decommissioning can be resolved. One commenter (15) questioned the ability to assess realistically the impact of deco,missioning criteria that call for disposal of high volume, low-level radioactive sludges because there are no I

sites available now or in the foreseeable future to accept this waste. (15)

(3, 11, 16, 23, 30, 34, 36)

Discussion Disposal of decommissioning wastes is covered by existing regulations and is beyond the scope of the rulemaking action supported by this GEIS. Disposal of spent fuel will be via geologic repository pursuant to requirements set forth in NRC's regulation 10 CFR Part 60. In addition, storage of spent fuel in independent spent fuel storage installations is covered by 10 CFR Part 72.

Disposal of low-level wastes is covered under NRC's regulation 10 CFR Part 61.

These regulations are all in effect. Because low-level wastes cover a wide range in radionuclide types and activities,10 CFR Part 61 includes a waste classification system that established three classes of waste generally suitable for near-surface disposal: Class A, Class B, and Class C. This classification system provides for successively stricter disposal requirements so that the potential risks from disposal of each class of waste are essentially equivalent to one another. In particular, the classification system limits to safe levels the concentrations of both short- and long-lived radionuclides of concern to low-level waste disposal. The radionuclides con-sidered in the waste classification system of 10 CFR Part 61 include long-lived activation products, such as Ni-59 or Nb-94, as well as "intense emitters" such as Co-60.

i 10/09/87 A-37 NU0586 APPENDIX A

Wastes exceeding Class C limits are considered .o be not generally suitable for near-surface disposal, and those small quantities currently being generated are g being safely stored pending development of disposal capacity. The recently enacted Low-Level Radioactive Waste Policy Amendments Act of 1985 (Pub. L.99-240, approved January 15,1986, 99 Stat.1842) provides that dis-posal of wastes exceeding Class C concentrations is the responsibility of the Federal Government. The Act also requires a report by DOE to Congress with recommendations for safe disposal of these wastes.

As far as decommissioning wastes are concerned, technical itudies coupled with practical experience from decommissioning of small reactor units indicate that wastes from future decommissionings of large power reactors will have very similar physical and radiological characteristics to those currently being generated from reactor operations. Two of the studies performed by NRC include NUREG/CR-0130, Addendum 3, and NUREG/CR-0672, Addendum 2, which specifically address classification of wastes from decommissioning large pressurized water reactor (PWR) and large boiling water reactor (BWR) nuclear power stations.

These studies indicate that the classification of low-level decommissioning wastes from power reactors will be roughly as follows:

Waste Class PWR (Vol. %) BWR (Vol. %)

A 98.0 97.5 -

0 1. 2 2.0 C 0.1 0.3 Above C 0.7 0.2 l

As shown, the great majority of the waste volume from decommissioning will be classified as Class A waste. Only a small fraction of the wastes will exceed Class C limits.

Disposal capacity for Class A, Class B, and Class C wastes currently exists.

Development of new disposal capacity under the State compacting process is covered under the low-level Radioactive Waste Policy Amendments Act referred 10/09/87 A-38 NUO586 APPENDIX A

l to at,ove. This Act provides for i centives i for development of such capacity, I as well as penalties for failure to develop such capacity. For wastes exceeding Class C concentrations, 00E has of fered to accept such waste for storage pending development of disposal criteria and capacity. For spent fuel, a detailed schedule for development of monitored retrievable storage and geologic disposal capacity is provided in the Nuclear Waste Policy Act of 1982.

Li:ensees will have to assess the situation with regard to waste disposal as part of the decommissioning plan which they submit according to the requirements of the amended regulations. In addition, the rule amendments require that at or about five years prior to the projected end of operation, each reactor licensee submit a cost estimate for decommissioning based on an up-to-date assessment of the actions necessary for decommissioning. This requirement is intended to assure that consideration be given to relevant, up-to-date informa-tion which could be important to adequate planning and funding for decommission-ing well before decommissioning actually begins. These considerations would likely include an assessment of the then current waste disposal conditions. If for any reason disposal capacity for decommissioning wastes were unavailable, there are provisions in Section 50.82 of the amended regulations to allow delay in compl' sn of decommissioning which would permit temporary safe storage of decommisuoning waste. In addition, Section 50.82 contains requirements to ensure that adequate funding is available for completion of delayed decommissioning.

Although the DECON decommissioning alternative assumes availability of capacity -

to dispose of waste, alternative methods of decommissioning are available including delay in completion of decommissioning (such as SAFSTOR) during which time there can be storage of wastes. Delay in decommissioning can result in a reduction of occupational dose and waste volume due to radioactive decay.

Comment 2 - Raises the question that the NRC should consider the decommissioning of low level waste storage facilities erected at reactor sites. (8)

Discussion Battelle PNL, as part of development of the c'ata base for this GEIS, prepared '

an evaluation of the technology, safety and cost of decommissioning a nuclear 10/09/87 A-39 NUO586 APPENDIX A

facility for the case in which waste must be stored at the site after expira-tion of the operating license. That evaluation also includes an evaluation of the decommissioning of the temporary low-level waste storage facilities. This evaluation showed the additional impact was not significant.

One commenter (15) questions the ability to assess realistically the impact of decommissioning criteria that call for disposal of high volume, low-level radioactive sludges because there are no sites available now or in the foreseeable future to accept this waste. (15)

Comment 3 - Raises the question that on-site, low-level waste disposal is the most likely and most reasonably available method for decommissioning. (16)

Discussion l -As indicated in the response to Comment 1, decommissioning regulations will contain provisions for use of delayed decommissioning alternatives, such as SAFSTOR, for facilities which must store low level waste at the site past the expiration date of the facility operating license. However, it is assumed that this storage at the site will be temporary. Permanent conversion of sites to a low level waste burial facility is not considered a decommissioning alternative because, as is stated in Section 2.3 of the GEIS, decommissioning of a facility leads to unrestricted use. Conversion and use of a facility for a LLW disposal site after its operating life is over is outside the scope of the rulemaking supported by this GEIS and would have to be reviewed on a case by case basis by-NRC.

Comment 4 - Disagreed with the statement made in Section 2.7 of the draft GEIS that the quantity of waste from operating reactors will considerably exceed that generated by facilities being decommissioned, although one commenter indi-cated agreement with the statement. (1, 35) 10/09/87 A-40 NUO586 APPENDIX A

Discussion The basis for the statement in Section 2.7 of the draft GEIS is that it hos been estimated that an operating 1000 MWe rea: tor will generate approximately 1300 m 3 /yr of low level waste.

Thus for 100 reactors, the total waste volume generated would be approximately 130,000 m3 /yr. DECON of a reactor is estimated to generate less than 5000 m /yr over a 4 year period.

3 It is recognized in Section 2.7 in thc GEIS that, for any one reactor, decom-missioning will generate an appreciable fraction of the low level waste gene-rated by a reactor over its lifetime. It is also recognized in the GEIS that there is a need for burial capacity of this low level radioactive waste.

Comment 5 - Questioned the validity of the comparison made in Section 0.4.4 of the GEIS of 17900 m3 of waste volume generated to 1160 acres that the plant originally occupie1.

Discussion The comparison was between the 2 acres which would be used at a low level waste burial ground for the 17900 m3 and the 1160 acres originally used as the site of the reference PWR. The comparison is valid because the operating nuclear facility with restricted use covering 1160 acres has been converted to 2 acres of waste disposal space following termination of license.

A.7 General Technical Questions About Deconmissioning A.7.1 Questions on the Information Base Developed for the GEIS Comment 1 - Questions the adequacy of the information base developed for the GEIS, in particular the lack of completed reports on research and test reactors, multiple reactors, non-fuel-cycle facilities, UFs conversion plants, and post-accident decommissioning. (1, 5, 7, 16, 23, 30, 36, 38) 10/09/87 A-41 NUO586 APPENDIX A

l l

Discussion The technical data base upon which the GEIS is based represents an extensive study of the decommissioning of nuclear fuel cycle and non-fuel-cycle facilities.

Since the draft GEIS was issued, reports on the technology, safety and costs of

. decommissioning the following facilities have been completed: multiple reactors, non-fuel-cycle facilities, UFs conversion plants, research and test reactors, independent spent fuel storage installation and decommissioning of reactors and fuel cycle facilities involved in accidents. Reports that had already been completed at the time the draft GEIS was published include the technology, safety and costs of decommissioning the following facilities: pressurized water reactors, boiling water reactors, reprocessing plants, fuel fabrication plants, and mixed oxide fuel fabrication plants. Also reports on the technology and costs of termination surveys to verify that residual radioactivity levels meet acceptable levels have been completed.

Based on the above, it is the NRC's judgment that the development of the data base is sufficiently complete to develop the GEIS and subsequent rules. More details on this area are in NUREG-1221, Section D.1.1.

Comment 2 - Raises the question that the GEIS does not provide enough technical detail or historical detail to provide information on decontamination and decow.;;issioning performance. (30)

~

! Discussion i

As stated in the GEIS, Section 1.1, the purpose of the GEIS is to assist NRC in promulgating revisions to regulations on decommissioning. As such, the GEIS presents a summary of the technical data base. The full technical data base including detailed information on decontamination and decommissioning techniques and experience is contained in over 20 volumes of reports prepared by Battelle-Pacific Northwest Laboratories. These reports are referenced through out the GEIS and should be used if detailed technical information is necessary to a user. These include large amounts of technical and historical detail on decontamination and decommissioning performance.

10/09/87 A-42 NUO596 APPENDIX A 1

Comnent 3 - Raises the question tnat waste volumes, and content, and risks ha* e been underestimated in the GEIS. (1)

Discussion The quantities and radioactivity of the wastes arising from decommissioning are

! developed in the Battelle-PNL reports through an analysis of the radionuclide inventory in the plant at the time of plant ' shutdown, the types and quantities j of wastes that must be disposed of and the decor.tamination procedures that l generate waste volumes. The pressurized water reactor and boiling water reactor reports also provide details concerning the sensitivity analysis of impact of differing plant conditions, including different amounts of contamination than those initially estimated and different reactor sizes.

)

This is based on the data base currently existing on decontamination and decom-missioning and on estimated plant condition at the time of shutdown. Based on the detailed technical analysis completed, the waste volumes and risk associated with decommissioning are not considered to be underestimated. More <ietail on this area is found in NUREG-1221, Sections D.1.1 and H.1. I Comment 4 - Raises the question of why decommissioning of HTGRs is not considered in the GEIS. (35)

Discussion The purpose of developing the technical data base is to provide support for development of a generic rule on decommissioning which can provide consistent licensing basis and remove the need for case-by-case licensing decisions.

Since there is only one HTGR currently in commercial operation and none are currently planned to be built, there is no currently sufficient r.eed to study, in a generic manner, the decommissioning of HTGRs. Review of the decommission-ing for that facility can be undertaken on a case specific basis. Of course, i the existing HTGR will be required to conform to general proposed rule require-l ments, namely financial assurance, planning, and decommissioning alternatives, although specific details will have to be considered for the plant.

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1 Cocment 5 - Questions the adequacy of the NRC analysis on the decommissioning of low level waste burial grounds and questions whether the NRC analysis of fuel reprocessing plants is based on realistic models. (32)

Discussion l Detailed evaluation the decommissioning of low level waste burial grounds is I outside the scope of this GEIS. This evaluation is contained in NUREG-0782, Draft Environmental Impact Statement on 10 CFR Part 61 "Licensing Requirements for Land Disposal of Radioactive Waste," September 1981, and in the proposed rule on "Licensing Requirements for Land Disposal of Radioactive Waste,"

46 FR 38081, July 24,1981.

The Battelle-PNL study on the technology, safety and costs of decommissioning a fuel reprocessing plant is based on the Barnwell Nuclear Fuel Plant, located in Barnwell, South Carolina. Although the Barnwell Plant has never operated as a fuel reprocessing plant (FRP), its design is considered to have characteristics typical of those present in any future FRP. In addition, because the existing portions of the Barnwell plant do not include f acilities for high-level liquid waste solidification (which any future FRPs would contain) the Battelle-PNL study included a conceptual facility of this type added on to the Barnwell plant and analyzed its decommissioning.

Although the Nuclear Fuel Services (NFS) plant in West Valley, New York, is the only commercial reprocessing plant that has operated in the United States (although it is not currently operating) it is not used as the reference plant.

The NFS situation is not directly translatable to the present or projected nuclear power industry because a national policy (10 CFR Part 50, Appendix F) requiring the sol;dification of high-level waste was not established until 1971, well after the plant began operation. Therefore, since NFS has its reprocessing high-level wastes stored in large underground tanks in slurry form', the costs of decommissioning this plant would be expected to be higher than that of any new FRPs if they were to be constructed and hence West Valley was not usett as the reference plant.

10/09/87 A-44 NV0586 APPENDIX A

At the present time no commercial spent fuel is being considered for reprocessing.

Coment 6 - Raises the question that the draf t GEIS has not adequately handled the problems of decommissioning following an accident, specifically the costs, financial considerations and procedures. (7, 35, 37, 39)

Discussion See revised GEIS Section 8 for a discussion of decommissioning of a reactor which has been involved in an accident.

A.7.2 Technical Details Comment 1 - Raises the question that the NRC assumption that "good housekeeping" practices have been employed is not valid. (1)

Discussion The full sentence quoted above from page 2-13 of the draft GEIS is "Most rooms should not be mildly contaminated with radioactivity in excess of levels which are acceptable for unrestricted facility use since it is assumed that good housekeeping and ALARA practices will be used during facility operations to control the spread of contamination." The context of this sentence is that rast rooms will either be highly contaminated, thus requiring extensive decon- ~

tamination efforts, or will have very low contamination levels because of the need to control occupational exposures during operations. These exposures during operations must be kept "as low as reasonably achievable" in accordance with 10 CFR 20.1 and with Regulatory Guide 8.8.

l The draft GEIS, Section 2.5.3, goes on to state that, if necessary, decontam-l ination of these mildly contaminated rooms during decomissioning can be accom-plished at a low cost and with low expenditures of manpower.

10/09/87 A-45 NU0586 APPENDIX A i

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l I

Coment 2 - Questions how decommissioning will vary as a function of the year of design of the LWR and the effect of different levels and durations of plant operation. (3)

Discussion The development of the technical data base on decommissioning included an analysis of the sensitivity of the technology, safety and cost of decommission-ing to several different parameters considered to be potentially significant in their effect. These parameters included: (1) plant size (thus considering the level of plant operations, as well as the year of design of the reactor since older reactors are generally a small power level while newer reactors are larger); (2) the degree of radioactive contamination (thus considering the duration of plant operatien since with longer lifetimes there could be greater contamination); (3) waste disposal charges; (4) contractual arrangements; and (5) for BWRs, different containment designs. The results of these sensitivity analyses are contained in Sections 4.3.4 and 5.3.4 of the GEIS. These sections point out that, while there were some differences in results, the conclusion of the sensitivity analyses is that the differing parameters do not substan-tially affect the original cost and dose conclusions.

Comment 3 - Disagrees with GEIS statement (made in the draf t GEIS, Sec-tion 2.5.3) that decontamination costs for a facility are essentially indepen-dent of the level to which it must be decontaminated as long as it is within the range of 1-25 mrem /yr. (23, 30, 34, 38) ~

Discussion The context of Section 2.5.3 of the draf t GEIS is that cott-benefit considera-tions are involved in the evaluation of the extent of facility decontamination necessary to decommission the facility, i.e., to release it for unrestricted use. In estimating the costs of decommissioning, it is assumed that all neutron-activated material and all potentially contaminated piping and equip-ment is removed and disposed of as radioactive waste. The question of unre-stricted release levels becomes important when the final cleanup of the struc-tures is begun. In the PNL analyses costs of decommissioning were computed 10/09/87 A-46 NU0586 APPENDIX A

on the basis that in all areas anticipated to have contaminated concrete sur' aces, two inches of concrete were removed and disposed of as radioactive material.

These surfaces included such areas as the walls and floor behind stainless-steel-lined pools and the walls and floors of process areas. Where an eight-inch concrete block wall was involved, the analyses postulated removing the entire wall, not just the two-inch surface layers.

There will, in practice, be situations where contamination has penetrated more deeply than two inches. At the same time it should be recognized that most of the concrete surfaces will be contaminated to a depth of about one-half inch or less. Thus, the approach of evaluating the cost of removing and disposing of a two-inch layer is generally conservative and should adequately cover the instances where additional material must be removed locally to obtain a clean surface.

Even if additional concrete must be removed it will not have significant impact on the overall costs of decommissioning. The incremental cost of removing twice as much concrete in releasing the facility for unrestricted use has been esti-mated as adding approximately 2% to the cost of decommissioning. This is within the 25% contingency factor which is included in the cost estimates in Tables 4.3-1 and 5.3-1.

Based on the preceding discussion the cost of decommissioning the facility, i.e., reducing the contamination to unrestricted use levels, is essentially independent of the unrestricted use level, as long as that level is in the -

range of 1-25 mrem / year to an exposed individual.

Comment 4 - Comments that existing operational ALARA considerations are adequate, and comments that any NRC proposed facilitation requirements should be justified on a rigorous cost-benefit basis. (10,34)

Discussion The studies performed as part of the policy reevaluation have shown that facili-tation of decommissioning in the design of a facility or during its operation can be beneficial in reducing operational exposures and waste volumes requiring 10/09/87 A-47 NUO586 APPENDIX A

disposal at the time of decommissioning. In addition, facilitation can improve financial assurance by keeping actual costs of decommissioning in line with the estimated costs on which the levels of financial assurance are based. A specific recuirement on facilitation was contained in the proposed rule (recordkeeping),

the effects of operational procedures on decommissioning should be considered by licensees as part of their program to maintain radiation exposures and effluents "as low as reasonably achievable." The facilitation of decommissioning in the design of facilities can be considered under the general standard for issuance of license that equipment and facilities be adequate to protect the health and safety of the public contained in SS 30.33(a)(2), 40.32(c), 50.40(a), 70.23(a)(3),

and 72.31(a)(10). Suggestions for facilitation are presented in the PNL studies and in a preliminary study on facilitation of reactor decommissioning prepared for NRC.

In particular, experience has shown that an important aspect of facilitation during operations is the maintenance of adequate information on the design and current condition of the facility and site, so that decommissioning can be care-fully planned and carried out. The amended rule does specifically require that records of relevant operational information helpful in facilitating decommis-sioning be kept by all reactor and materials licensees. Plans should be developed to collect, maintain, and recall records and archive files which include as-built and as-revised drawings and specifications and operational occurrences which could significantly affect decommissioning. The amended rule specifically allows the use of references to relevant information and locations in order to avoid unnecessary duplication of record.c kept for other purposes and also specifies that referencing of drawings need not include indexing of each individual relevant document. The intent of this requirement is to assure that all important information is kept until termination of license and that it be readily accessible when needed.

Ccament 5 - Disagrees with the GEIS statement that the technology for decommissioning is well in hand, because technology has not been developed to remotely dismantle a reactor af ter 30 years, and because reusable decommission-ing equipment and equipment which could further lower costs and occupational exposures has not been developed. (32) 10/09/87 A-48 HUO586 APPENDIX A

Discussion The context of the statement in GEIS Section 15.0, referred to in Comment 5 above, is that the technology for decommissioning nuclear f acilities is well in hand and, while technical improvements in decommissioning techniques are to be expected, decommissioning at the present time can be performed safely and' at reasonable cost.

Radiation dose to the public due to decommissioning activities should be very small and be primarily due to transportation of decommissioning waste to waste burial grounds. Radiation dose to decommissioning workers should be a small fraction of their expos'Jre experienced over the operating lifetime of the facility and usually be well within the occupational exposure limits imposed by regulatory requirement s. Decommissioning costs are reasonable and are, at least for the larger faci'ities such as reactors, a small fraction of the present worth commissioning costs (i.e., less than 10%). This statement is not meant to imoly that there won't be technical improvements in the future and as decommissioning experience is obtained these improvements will be made, however as is stated decommissioning can be performed safety. Based on the statements in Section 15.0, regulations can be written containing requirements for decommissioning.

t Comment 6 - Raises the question that the GEIS should contain more detail of the impact of wastes from decommissioning activities on waste disposal sites. (30)

Discussion See Section A.6 of this Appendix. In addition the environmental impact asso-ciated with waste disposal sites is contained in the Final Environmental Impact Statement on 10 CFR Part 61 "Licensing Requirements for Land Disposal of Radioactive Waste."

Comment 7 - Raises the question that the GEIS should consider the impact that variations in residual radioactivity criteria would have on projected waste volumes. (30) 10/09/87 A-49 NUO586 APPENDIX A

Discussion As discussed in Section 2.5 of the Final GEIS the impact of differences in residual radioactivity limits, within a reasonable range, is not significant in times of cost. Hence the impact on waste disposal is not expected to be significant.

A.7.3 Socioeconomic,and Human Factors Comment 1 - Questions whether the GEIS should contain more detail concerning the socioeconomic impacts of shutting the plant down and decommissioning it.

(8,40)

Discussion As discussed in GEIS Sections 4.4, 5.4, and 7.4.3, the major socioeconomic impact occurs prior to decommissioning, namely at the time of thc owner's decision to shutdown the nuclear facility, thus removing a source of employ-ment and tax income for the community. Treatment of thase effects is outside the scope of this GEIS. Decommissioning activities tend to mitigate the impact of job and tax income reduction for a period of time af ter shutdown, and hence those effects are not treated in detail in the GEIS.

Comment 2 - Questions why the GEIS does not consider "human error" in its analysis. (40) ~

Discussion Because the reactor is not operating during decommissioning, the analysis contained the GEIS and the information base prepared by Battelle-PNL does not include the significant impacts which can result from human error at operating facilities.

l Nevertheless, the GEIS and the information base reports do contain in their analysis several considerations of human error. For example, Tables 4.3-1 and

! 5.3-1 indicate that costs for decommissioning include a 25% contingency factor

! 10/09/87 A-50 NUO586 APPENDIX A

which can account for unfareseen events that might impede the conduct of the decommissioning work. The costs listed in the table also include scheduling and cost allowances for inefficiencies associated with working in radiation envi ronments.

In addition, the GEIS includes (see for example, Tables 4.4-2 and 4.4-3) an analysis of the radiation dose impact to the public from accidents resulting from various causes, including human error. The GEIS found that even for the most severe accident that the doses were moderate (see for example, Section 4.4).

The information base developed by Battelle-PNL also includes an analysis of injuries to workers resulting in lost time from the job, and worker fatalities.

This analysis was based on industrial type accidents during the decommissioning.

It was found, for example, that for boiling water reactors that less than 10 lost-time injuries to workers would occur, and that essentially no fatalities due to industrial accidents would occur during the decommissioning or the transportation of decommissioning wastes.

In order to minimize human error, Section 15.1.2.2 recommends that quality assurance provisions during conduct of decommissioning be described in the decommissioning plan. This would involve describing the equipment and proce-dures requiring QA procedures during decommissioning. As another means of minimizing human error, Section 15.1.2.1 recommends that records of information important to a decommissioning be kept over the lifetime of the facility. These records would include records of spills and unusual occurrences involving spread of contamination in the facility and would also include as-built drawings and -

modifications of structures and eqt.ipment in high radiation areas. Maintenance and availability of such records at the time of decommissioning will assist plant staff in conducting work in radiation areas and minimize radiation exposure and human error.

Comment 3 - Questions why the GEIS does not mention the impact that the disposal of decommissioning waste will have on communities surrounding the waste burial grounds. (40) 10/09/87 A-51 NUO586 APPENDIX A

Discussion The GEIS includes an analysis of population exposure from truck transport of decommissioning waste to burial grounds. (See, for example, Tables 4.3-2 and 5.3-2). The evaluation of the impact of waste at the burial grounds is outside the scope of this GEIS, but analysis of the environmental impact of waste is included in the Final Environmental Impact Statement on 10 CFR Part 61 "Licensing Requirements for Land Disposal or Radioactive Waste," November 1981.

A.7.4 Occupational Exposures Coenent 1 - One commenter (38) questions the basis of the GEIS statement that the occupational radiation dose resulting fron DECON of boiling water reactors and fuel reprocessing plants is of marginal significance to health and safety while-another agrees with GEIS statement that decommissioning of a reactor can be accomplished with reasonable occupational radiation exposure and virtually no public radiation exposure.

A.7.5 Non-Fuel-Cycle Facilities Comment 1 - Raises the question that decommissioning considerations for non-fuel-cycle facilities should be different from those for fuel cycle facilities, because of the different nature of the facilities; and also that there should be separate consideration for difficult types of non-fuel-cycle facilities and that for some processing facilities, decommissioning considerations should be on' a case-by-case basis. (15)

Discussion The GEIS recognizes the unique nature of the different types of facilities by treating them in separate analyses and by analyzing the costs, waste disposal concerns, and decommissioning alternatives for the different types of facilities.

Specifically revised Section 14 of the GEIS discusses the alternatives, cost, dose impacts, and waste disposal of the different major type of non-fuel-cycle facilities requiring significant decommissioning action, including sealed source 10/09/87 A-52 NUO586 APPENDIX A

nanufacturers, radiochemical and radiopharmaceutical manufacturers, ore processors, ana broad research and development facilities. ,

Despite the different and unique nature of the non-fuel-cycle facilities, the general NRC policy consideration outlined in Section 15 of the GEIS can apply in general to all facilities. These policy considerations include planning, financial assurance, and decommissioning alternatives. All facilities con-sidered in this GEIS, and in subsequent rulemaking, which have a significant decommissioning effort, need to plan for decommissioning; need to establish a fund; need to consider which of the decommissioning alternatives is most appro-priate and what the timing of that alternative should be; and need to have criteria for acceptable levels of residual radioactivity.

The GEIS recognizes the unique nature of the different facilities in several GEIS sections. Section 15.1.1 recognizes that dif ferent decommissioning alter-natives may be more logical for certain facilities than others. Section 15.1.3 recognizes that because of the diversity of facility types, different funding methods may be acceptable.

Certainly in any decommissioning, including that for an ore processing facility, there will be case-by-case considerations but it is expected that these will fit into the general guidelines of these amended regulations. Regulatory guides under consideration will treat such considerations.

Comment 2 - Raises the question that the cost of decommissioning ore processing -

facilities is significantly underestimated and that the GEIS does not adequately treat the cost of transporting waste from ore procassors to low level burial facilities.

Comment 3 - Raises the question that the GEIS has underestimated the complexity of decommissioning an ore processing facility and not provided sufficient basis for the statement that decommissioning of an ore processing plant has only minor adverse impact. (15) 10/09/87 A-53 NUO586 APPENDIX A

Discussion As stated in the GEIS, Chapter 14, the major problem with the ore processing facility decommissioning is the tailings pile disposal problem. The GEIS recognizes many options for handling tailings, such as possible disposal in a local landfill, depending on an acceptable residual level (p. 14-20), in place stabilization (p.14-22), through to a removal option for which the major costs of transportation and burial for the example case is 33 million in 1986 dollars (p. 14-20). Thus the GEIS recognizes that decommissioning of an ore processing facility can, depending on circumstances at the time, be reasonably simple or very complex in terms of cost.

10/09/87 A-54 NUO586 APPENDIX A

c*

i APPENDIX B. COMMENTS RECEIVED'ON.THE DRAFT GENERIC ENVIRONMENTAL IMPACT STATEMENT Table B1 lists the source of comment letterc on the Draft GEIS. l i

f i

I 10/09/87 B-1 NUO586 APP B

g, =2& 2 l

Table B-1 Comment Letters on the Draft GEIS Docket No. Commenter

, 1 Marvin Lewis 2 San Diego Gas and Electric 3 Wisconsin Public Service Commission 4 Jay Gertz 5 Consolidated Edison Company 6 General Electric Company 7 Commonwealth Edim Company 8 Tennessee Valley Authority 9 baited States Environmental Protection Agency 10 Cembustion Engineering 11 Detroit Edison Co.

12 Deloitte, Haskins and Sells 13 Klevorn, Oreyer and Dubois 14 Houston Light and Pcwer Company 15 Baker and Hostetler 16 Kerr-McGee Corp.

17 Hesslin 18 Mallinckrodt, Inc.

19 Nichigan Public Service Commission -

23* Atomic Industrial Forum 24 Debavoise and Liberman 25 Duke Power Company 26 General Electric Co. - Nuclear Fuels and Set , ces Division 27 Consumers Power Company -

28 Public Service Electric and Company 29 Arizona Public Service Company 30 U.S. Department of Energy 31 Health Industry Manufacturers Assn.

32 Sierra Club Radioactive Waste Campaign 10/09/87 B-2 NUO586 APP B

l 33 Texas Dept. of Health 34 Public Service of Indiana 35 Arkansas Power and Light Co.

36 Texas Dept. of Water Resources 37 California Energy Commission 38 .New York State Dept. of Environmental Conservation 1 39 Ohio Citizens for Responsible Energy .

40 J. A. Savage l

l l

l

• Comments from General Electric, Consolidated Edison, and Commonwealth Edison l were inadvertently docketed twice as nurabers 6, 20 and 5, and also as numbers f.'., 7 and 22, respectively. Therefore, docket numbers 20, 21, and 22 are not listed here.

10/09/87 B-3 NUO586 APP B L_