ML20148B692
| ML20148B692 | |
| Person / Time | |
|---|---|
| Issue date: | 08/31/1978 |
| From: | Briggs R, Buck J, Glaser M Atomic Safety and Licensing Board Panel |
| To: | Chilk S NRC OFFICE OF THE SECRETARY (SECY) |
| References | |
| RULE-RM-50-3 NUDOCS 7810310345 | |
| Download: ML20148B692 (141) | |
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' UNITED STATES i
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MEMORANDUM FOR: {SamuelJ a Chilk, Secretary FROM: Fuel' Cycle Rulemaking Hearing / Bompd Michael L. Glaser r (/ P, j' .h Dr. John H. Buck R. Beecher Briggs j
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SUBJECT:
.(1) REPORT BY THE NEARING BOARD TO THE I NUCLEAR REGULATORY COMMISSION REGARDING J ; i THE ENVIRONMENTAL ' EFFECTS OF THE r
URANIUM FUEL CYCLE / 7DOCRET NO. RM 50-3 (2) VIEWS OF THE[ HEARING BQARD ON WHETHER THE RECORD IM RM 50-3 pUPPORTS AN EXTENSION OF THE_EEESENT INTERIM RULE (TABLE S-3) FOR A REASONABLE PERIOD, - e PENDING ACTION ON A FINAL RULE $P g y .- m w, m By Memorandum dated July 6', 1978, the Hearing Board a'dirised ' q l the Commission that a Summary of the Record and Qttbline of O the Significant Issues raised in the reopened Uralii,um' Fuel "I Cycle'Rulemaking-Proceedings, Docket No. RM 50-3; Vould be 3 completed and submitted to the Commission by August 3T,1978 d We also informed the Commission that our recommendatibns J regarding the reprocessing and waste management as'plec{ti of the fuel cycle rule would be completed by October 15, 1978. We indicated that we would advise the Commission of the status of our recommendations at the time we submitted our report on the summary of the record and outline of the sig-1: nificant issues. Finally, we stated that we would express our views on whether the rulemaking record supports an extension of the Interim Fuel Cycle Rule for a reasonable period, pending preparation of our recommendations and Com-mission consideration prior to action on a final Rule. f The Hearing Board has completed its Summary of the Record and Outline'of the Significant Issues raised in the reopened rulemaking proceedings. Accordingly, we are submitting our Report to the' Commission containing the Summary and Outline as an attachment.to this memorandum. We are also submitting'our views on whether the rulemaking record supports an extension of the Interim Rule, pending completion of our recommendations regarding the reprocessing 1 y " c Q W 9 L78/ M lb 3 W Q3
. 9
, Samuel J. Chilk August 31, 1978 l
! s N j and waste management aspects of the fuel cycle rule and I- Commission consideration of this proceeding prior to action I on a final rule. l Finally, the Hearing Board advises the Commission that we l may take until October 31, 1978, rather than October 15, ! 1978, to complete the preparation of our recommendations. l Additional time may be needed because of the extensive .l record developed in the reopened rulemaking proceedings. We do anticipate recommending some additions in and modi-
, fications to the fuel cycle rule. .These additions and modifications, however, will not, in our judgment, change the effects of the present Interim Rule.
l l l . j' FUEL CYCLE RULEMAKING HEARING BOARD o I - l lo[G2d[ Michael L. GYaser
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, <' 3G f*
John H. Buck
$( LL w bW l R. Beecher Brigg Attachments - 2 1 cc: Chairman Hendrie Commissioner Gilinsky Commissioner Kennedy 'j Commissioner Bradford Commissioner Ahearne I
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UNITED STATES OF AMERICA. [~ x ' NUCLEAR REGULATORY COMMISSION' ; i
.]
I
. HEARING' BOARD .
Michael L. Glaser, Chairman Dr. John H. Buck 4 R. Beecher Briggs
?
) I In the Matter of ) l
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AMENDMENT.0FL10'CFR PART'51 ) }
- LICENSING OF PRODUCTION.AND ) Docket No. RM 50-3 ;
UTILIZATION FACILITIES ) ;
) !
(Environmental Effects of the ) t Uranium-Fuel Cycle) )
)
REPORT OF THE HEARING BOAPO TO THE NUCLEAR REGULATORY COMMISSION
'REGARDING ENVIRONMENTAL EFFECTS OF URANIUM FUEL CYCLE,.
DOCKET-NO. RM 50-3 August 31, 1978
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TABLE OF CONTENTS 4 I Page^ FART 1 Preliminar'y Statement in the. Reopened Rulemaking Proceedings Involving Impacts.from Spent Fuel Reprocessing and Radioactive' Wastes in the Uranium. Fuel Cycle, Docket No. RM-50-3 3 I
.PART 2 .
Summary of the Record I. Puel Reprocessing , 17 High'LeveI Waste Disposal II. 68
- III. Low Level Waste' Disposal 90 ,
IV. Interim Storage 97 V. Transportation 105 VI. -Accidents , 107
<VII. ' Sabotage 110 f
.VIII. Decoramissioning ' 113 IX. -Socioeconomic Effects- 117 I
.X. Economic Feasibility 124
. PART 3 Outline of Significant Issues' 135 L
O li _ L L_._L.-- _-___ _ . -._ . . _ - . ,n
e a PART 1 PRELIMINARY STATEMENT IN THE REOPENED RULEMAKING PROCEEDINGS INVOLVING IMPACTS FROM SPENT FUEL REPROCESSING AND RADIOACTIVE WASTES IN THE URANIUM FUEL CYCLE, DOCKET NO. RM-50-3 This reopened rulemaking proceeding arises from the decisions of the United States Court of Appeals in the - District of Columbia Circuit in Natural Resourqps Defense Council v. NRC, 547 F.2d 633 (D.C. Cir. 1976) and Aeschliman
- v. NRC, 547 F.2d 623 (D.C. Cir. 1976) wherein the Court reviewed the NRC's fuel cycle rule and the NRC's policy of excluding fuel cycle issues from individual power reactor licensing proceedings before it. The following discussions summarize the procedural and evidentiary record in the reopened rulemaking proceedings before this Hearing Board.
A. The Purpose of this Reopened Rulemaking Proceeding This reopened rulemaking proceeding is limited to gener-ically quantifying the reprocessing and waste management portions of the uranium fuel cycle relevant to individual light water power reactors. Therefore, this reopened rule-making proceeding does not consider environmental impacts from other portions of the fuel cycle. The purpose of this reopened rulemaking proceeding is to select an appro-I l priate model for assessing the environmental impacts of reprocessing spent fuel and waste management, not to select reprocessing and waste management systems for licensing. I L . _ .- .
.. e 4 ,
( For this limited purpose the staff examined various alternative technologies in reprocessing and waste manage-ment, attempted to select a credible model, and analyzed that model to determine the environmental costs of licens-ing a nuclear reactor attributable to reprocessing and waste management. 1! The staff's model was selected on the basis of a number of factors including availability, likelihood of use, amount of information available, and the degree of impact. This reopened rulemaking proceeding also does not consider health effects. The consideration of health effects in reprocessing and waste management was clarified by the NRC in an amendment to Table S-3 on April 14, 1978. 1! The amendment stated that Table S-3 did not include health effects from the effluents described in S-3 and it removed all dose estimates attributable to the gaseous effluents in the Table. Radiological dose values were not previously included in Table S-3 because the radiological impacts of the uranium fuel cycle were considered in the cost-benefit analyses in.the Environmental Impact Statements for indi-
- vidual reactors in connection with comparing uranium to l
other fuel cycles. i e ( _1/ NUREG-Oll6, pp. 1-2. 2/ 43 F.R. 15613. l
p .. . L L j L (,,) N B. Procedural Background
.l. Original Rulemaking Proceedings I
l Under the National Environmental Policy Act of
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1969 (NEPA), the NRC is required to prepare an environmenta1 impact statement (EIS) in connection with the proposed issuance of a construction permit or operating license for each light water nuclear power reactor. The EIS includes detailed evaluations of the environmental impacts of construc-tion and operation of reactors, and assesses the costs and [ benefits of the proposal. In November 1972, the Atomic } Energy Commission (AEC), predecessor of the NRC, published ( an " Environmental Survey of the Nuclear Fuel Cycle" S!' L (Environmental Survey) to establish a basis for ccnsidering the environmental effects of the uranium fuel cycle in Environmental Impact Statements for individual light water power reactors. The nuclear fuel cycle was treated generically in order to provide an overview for the entire nuclear power industry. The Environmental Survey evaluated model fuel l cycle facilities rather than specific site and design details. Estimates of effluent concentrstions, radiation dose rates, 1 and human population densities based on model fuel cycle i facilities were made. l l 1 i i f\
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3/~An updated report was published as Environmental Survey
- of the' Uranium Fuel Cycle, WASH-1248, April 1974. I m.e m
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The AEC first sought public comment on the Environ-mental Survey. Thereafter, a legislative-type rulemaking hearing was held in February 1973 to consider amending the regulations to specify, in a rule, the costs and benefits of the nuclear fuel cycle to be included in environmental impact statements for individual light water reactors. After considering the record and written comments in the legistive-type rulemaking hearing, in April 1974 the AEC promulgated a fuel cycle rule which was embodied in its regulations as a table entitled Table S-3. d! In accord with National Environmental Policy Act requirements, the AEC intended Table S-3 to represent a complete assement of the environmental costs of the fuel cycle.
- 2. Cudicial Review of Original Rulemaking Proceedings In Natural Resources Defense Council v. NRC, supra, and Aeschliman v. NRC, supra, the Court approved the AEC's i
overall methodology of promulgating the fuel cycle rule. It found that the AEC's underlying Environmental Survey adequately described most of the processes involved in the fuel cycle. The Court noted that the Environmental survey _4/.39 F.R. 14188. O
O assembled data on consumption of resources, discussed the risks of accident and other hazards in detail, and provided numerous references to scholarly literature and technical reports in support of its conclusions respecting environ-mental impacts. However, the Court found that the record was inadequate on radioactive waste management and the impacts of reprocessing spent fuel. The Court remanded the rulemaking proceeding.
- 3. Remand Proceedings In response to the Court's decisions, the NRC issued a general statement of policy announcing reopened rulemaking proceedings to supplement the existing recora on reprocessing and radioactive waste management, and to determine whether the rules should be amended, and if so, in what respects.
The NRC directed its staff to prepare, on an expedited basis, a well-documented Supplement to the Environmental Survey to establish a basis for identifying the environmental impacts of fuel reprocessing and waste management associated with licensing a model light water reactor. The staff published its Supplement to the Environmental Survey, NRC publication NUREG-Oll6, in October 1976. 5! 5YE nvironmental Survey of the Reprocessing and Waste Manage-O- ment Portions of the LWR Fuel Cycle, NUREG-Oll6, October 1976. 6 L-
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Based on the staff's Supplement, the NRC announced a proposed interim rule on October 13, 1976, covering the environmental impacts of fuel reprocessing and waste manage-ment. The proposed interim rule was described in the form of a table assessing the environmental impacts in these two areas, and was published in the Federal Register on ! October 18, 1976. 5/ On the same date, the NRC also pub-lished in the Federal Register a notice of proposed revi-sions to Table S-3. The NRC also announced the public availability of the staff's Supplement, and solicited comments on the proposed rule changes. The NRC distributed approximately 2,400 copies of the Supplement, and more than 50 interested parties submitted their comments on it. After further review, the NRC published the staff's response to
.the comments, as well as additional information on environ-mental impacts of reprocessing and waste management which either had become available to the staff after its supple-ment was issued, or which specifically responded to requests for clarification of material in the Supplement. 1/ The NRC published an interim rule based on the staff's Supple-ment and responses to comments on March 14, 1977.-S/
45849, _6/ 41 F.R.
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7/ This record publication was denoted NUREG-0216, "Public Comments and Task Force Responses Regarding the Environ- [~)
\-- mental Survey' of the Reprocessing and Waste Management l Portions of the LWR Fuel Cycle." 4 8/ 42 F.R. 13803. ;
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.4. ' Reopened Rulemaking Proceedings The NRC issued its notice of reopened rulemaking proceedings on May 20, 1977 (42 F.R. 26987). It ordered the legislative-type rulemaking- proceedings which gave .
rise to Table S-3' reopened to determine (1) the environmental effects of reprocessing spent fuel and radioactive waste l management attributable to light water reactors, and (2) ! whether the interim rule should be made permanent, or if it l l should be' altered, in what respects. The NRC adopted the ! legislative-type procedures used in the original rulemaking hearings of February 1973. This Hearing Board was appointed to conduct the reopened rulemaking proceedings. The NRC specifically authorized.this Hearing Board to interrogate witnesses w'ho appeared in the reopened proceeding. The NRC also gave the participants in the reopened proceedings an
' opportunity to suggest questions to the Hearing Board, which the Board in turn might wish to ask the witnesses. On January 26, 1978,- the NRC modified the procedures governing this reopened proceeding by allowing the participants to request cross-examination of witnesses on specific factual issues,where it was shown with particularity that cross-examination was necessary to develop an adequate record. This Hearing-Board was given sole discretion to determine whether q
( -- cross-examination should be permitted in any particular case. l e l b _ _ _ _ _ _ _ _ _ _ . _ _ _ __ __ll___1
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This Hearing Board held a prehearing conference with all interested participants on July 28, 1977, to discuss further proceedings. A prehearing conference order was issued on August 12, 1977 scheduling procedural dates for the reopened proceedings. In the order, this Hearing Board provided for the submission of written direct testimony by the participants, written questions and answers based on such testimony, written follow-up questions, and dates for commencement of hearings. We directed the participants to provide us with a list of proposed witnesses in order to facilitate the~ proceedings. The procedural schedule provided for the following dates: Date Procedural Action September 30, 1977 Submission of written statements by par ties October 31, 1977 Submission of first round of written questions based upon written state-ments- 2-December 2, 1977 Responses to written questions December 16, 1977 Submission of follow-up written questions January 16, 1978 Commencement of hearings The following parties participated in this reopened proceeding: the staff of NRC; the Environmental Protection () Agedcy; the Department of Interior; the U.S. Geological Survey;
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rm. the States of California (California Energy Resources h Conservation and Development Commission), Delaware, Maryland, Ohio, Wisconsin and New York; Baltimore Gas and Electric Co. , et al (a group of 16 utilities) ; Commonwealth Edison Co., et al (a group of 8 utilities); the Tennessee Valley Authority; the Allied-General Nuclear Services, Co.; Exxon Nuclear j' Company; Westinghouse Electric Corporation; the Atomic l Industrial Forum; the Natural Resources Defense Council; the l l Pacific Legal Foundation; Environmentalists, Inc.; the Sierra Club; the Union of Concerned Scientists; Mr. Marvin Lewis; and Dr. Chauncey Kepford. Actual oral hearings began on January 16, 1978. After the hearings commenced the Hearing Board made several procedural rulings which expanded the opportunity to participate in the proceedings. By Memorandum and Order of May 4, 1978, we permitted the participants to file written rebuttal testimony prior to the submission of final written statements. During the course of the proceedings, we also ruled that a party's failure to attend actual hearing sessions would not result in a forfeiture of its right to participate further, including cross-examination upon the proper showing (Tr. 102-104). In addition, we permitted a proposed witness to respond to our questions in writing rather than appear for live interrogation (Tr. 19-20, 35-36, 104-106). Further, we attempted to accommodate: the schedules of various expert witnesses who wished to testify (q) (Tr. 40-41,.61-62, 97-102, 1074-1075). . Finally, upon the i
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petition of the State of New York, S! we expanded the scope of this proceeding to consider the economic feasibility of the model facilities on which the staff based the Table S-3 values.
- 5. The Evidentiary Record The evidentiary record developed in the reopened rulemaking proceedings is extensive. The record includes:
the staff's NUREG-Oll6 and NUREG-0216,bS! the staff's testimony on the economic feasibility of its model facilities, 1 the direct testimony of participants which exceeds 1,100 pages, two rounds of written questions propounded by participants j which resulted in several hundred pages of responses, more than 1,200'pages of transcript of oral hearings which consumed 10 days, written rebuttal testimony of the parties, and final concluding statements of the parties. The Hearing Board conducted all of the questioning during the oral hearings. We also heard oral arguments on various requests for cross-examination. Although the Hearing Board denied these requests S! Response of State of New York to the NRC comments on Scope of proceedings received December 2, 1977. (Memoranda and Orders, December 23, 1977; February 9, 1978) 10/ NUREG-Oll6 briefly describes model fuel reprocessing and waste management facilities and their environmental impacts. The details of the processes and calculations and information 1 about alternatives are contained in many of the reports i referenced in NUREG-Oll6. For the purposes of this proceedingi we consider these principal references, and the principal
-[ )
references' relied on by the other participants to constitute a supporting record. l [ _
for cross-examination-on grounds that the requesting parties had not rade the requisite showing, we did permit the parties to submit written rebuttal testimony prior to concluding statements. C. Issues Raised by the Participants At the outset of the oral hearings, the participants raised a number of issues regarding the staff's proposed Table S-3. The Natural Resources Defense Council (NRDC) identified nine deficiencies in Table S-3.
- 1. NRDC stated that a table of numbers provides no meaningful disclosure of the impacts of a radioactive waste disposal system.
- 2. The staff's Supplements and responses to questions failed to indicate the uncertainties and assumptions underlying its figures and calculations.
- 3. Radiological impacts expressed as numbers of curies released are not meaningful health effects.
- 4. Predicting the cumulative impacts of radiological wastes is important.
- 5. The staff failed to analyze and present a single identifiable waste management system.
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'6. ' By includir.g reprocessing, the staff.used an unrealistic waste management system. ENRDC suggested that consideration of reprocessing should be deleted from Table S-3.
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- 7. NRDC also contended that an assessment of the econom-ics of implementing; hypothetical waste disposal' systems is necessary to predict whether these systems will be implemented if they become available.
- 8. NRDC stated that the use of mod,el facilities that j .would be developed in the future is too optimistic.
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- 9. And finally, that Table S-3 was deficient because it did not provide guidelines for using its information in the future.
The California Energy Resources, Conservation and Development Commission (CERCDC) questioned (1) whether facilities.will or will not be available and (2) the maturity of-the waste management industry. It suggested that the inexpe'rience of the industry could have an environmental impact. ; i The State of New York requested the Hearing Board to focus on the economic feasibility of reprocessing and waste management.11/' 11/.As a' result of New York's request, the board re-examined the matter of economic impacts of reprocessing and waste management portions of fuel cycle, and. reaffirmed that only the economic feasibility =of the model facilities proposed by the staff were relevant.to this reopened-rulemaking' proceeding. See Memorandum and Order, dated February 9,;1978. -See also, supra.
r The states of Ohio and Wisconsin stated that Table S-3
\- failed to adequately consider international implications of nuclear waste management, socioeconomic factors such as physiological stress and perpetual care of nuclear wastes, and the total dollar costs of nuclear waste disposal.
Baltimore Gas and Electric Company, et al, contended that this reopened rulemaking proceeding was limited to determin-l ing the environmental effects of reprocessing'and waste manage-ment in a generic sense, and that this proceeding should not be extended beyond this limited purpose. The Pacific Legal Foundation also maintained that this reopened proceeding was properly limited to the environmental j impacts of reprocessing spent fuel and waste management, but questioned whether the model systems and technologies used by the Staff could reasonably be expected to comply with e'xisting regulations. The Environmental Protection Agency believed three principal issues had not been properly resolved. First, EPA believed that its standard for nuclear operation in 40 C.F.R. 190 should be reflected in Table S-3. Second, EPA stated that calculations of the effects of radioactive releases to the environmental were seriously deficient because (1) the l l environmental impacts of contaminants persist for periods i (~) R> e E
much longer than 50 years, (2) the assessment had been arbitrarily limited to impacts on U.S. population, and (3) the assessment of radiation source terms was not complete. Finally, epa believed that Table S-3, in its present form, represented neither an adequate nor conservative assessment of the environmental impact of the reprocessing and waste management portions of the fuel cycle. The Department of the Interior through the U.S. Geological Survey (USGS) questioned whether the wide range of approximate data and results could lead to. reasonable risk evaluation. Mr. Marvin Lewis raised issues related to health effects and safeguards for reprocessing and waste management. The staf f identified three principal issues. These con-sisted of (1) the reasonableness of the model' systems, including the availability of technology, the likelihood of its imple-mentation, the conservatism of the staff's analysis, and the disclosure of other issues related thereto, (2) uncertainties in the staff's analysis including the long term risk of the failure of a waste depository, the risks from sabotage in both reprocessing and waste management, the risks from unforeseen accidents, and uncertainties in disposing of spent fuel or plutonium, and (.3 ) evidence which might tend to change.the values.in Table S-3. O 9
i A ly PART 2 -
SUMMARY
OF THE RECORD i [ t I. Fuel Reprocessing A. The'Model Facility' The staff's values for the environmental impacts of fuel reprocessing that are incorporated in Table S-3 are based on analysis of the performance of a model reprocessing facility described in the GESMO proceeding.12/ - The facility uses the Purex solvent extraction process and is similar in concept to the Allied Gulf Nuclear Services plant (AGNS or Barnwell) at Barnwell, South Carolina. The model facility was specified to have a capacity of 2000 MTEM/yr13/ when operating with a capacity factor of 0.8.14/ The facility is intended to reprocess the spent fuel from l 57 reference reactors 15/ each of which was specified to operate at'1000 MWe.with a capacity factor of 0.8 and to discharge 35 MTEM/yr of spent fuel with a burnup of 33,000 MWD /MTEM.b5! { 12f NUREG-Oll6, p. 2-3. 13/ MTEM = metric tons of heavy metal (uranium and plutonium) MT = metric tons, MTU = metric tons of uranium and MTHM have been used interchangeably and with the same meaning in this proceeding. 14/ NUREG-0116, p. 3-9; Tr. 179. 15/ A reference reactor is an " average" reactor based on (') an industry consisting of 1/3 BNR's.and 2/3 PWR's ('j A L reference Reactor Year PIRY) indicates operation and resulting impacts from 1 years operation of the reference , reactor. ! 16/ NUREG-0116, p. 3-14.
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18 - No significant issues have been raised regarding the staff's choice of processes fcr the reprocessing facility. However, issues have been raised concerning whether the staff had made a conservative analysis of the performance of the facility. Specifically, participants in the pro-ceeding questioned whether the capacity and capacity factor specified by the staff are warranted, and, whether the - estimates of radioactive releases and occupational expo-sures are conservative. The economic feasibility of the model reprocessing facility was also questioned. In its analysis, the staff assumed that after removal from the. core of a reference reactor, the spent fuel assemblies are stored at the reactor site for at least 150 days and then are shipped to a model fuel reprocessing facility. At the reprocessing facility the spent fuel is unloaded into a storage basin from which it is removed within a few days for reprocessing.17/ - The first step in reprocessing is to shear the spent fuel assembly, which is a bundle of long metal tubes con-taining ceramic pellets of fuel, into small pieces. The fuel pellets are then dissolved in hot nitric acid. The acid solution of uranium, plutonium, other transuranium () ~~17/ NUREG-Oll6, p. 4-5; Baltimore Gas and Electric Co., Commonwealth Edison Co., TVA, et al., Initial Written ' Statement (BG &E/IWS) , p. II-4, II-5.
_ 19.. l (TRU) nuclides, and fission products is withdrawn from the dissolver for treatment by the Purex solvent extraction process leaving behind the sheared pieces of cladding ) (hulls) and assembly hardware.18/ - In the Purex process the acidic aqueous solution is ? contacted with an immiscible organic solvent, a hydrocarbon containing tributyl phosphate, in several stages. In the first stage, the uranium and plutonium dissolve in the organic phase. Most of the fission products remain in the aqueous phase. The uranium and plutonium are separated further from fission products and from each other in addi- ~ tional stages of solvent extraction. Equipment is provided in the plant for recovering and recirculating acid and solvent.19/- The primary products of the Purex process are an aqueous solution of uranyl nitrate and an aqueous solution of plutonium nitrate. These products contain more than 99 percent of the uranium and plutonium. The uranyl nitrate is converted into uranium hexafluoride and shipped to an enrichment plant for re-enriching and recycling in 18/ NUREG-Oll6, p. 4-5 thru 4-7; BG&E/IWS p. II-5. 19/ NUREG-0116, p. 4-5 thru 4-7; BG&E/IWS p. II-6 thru II-8; Tr. 210.. O .
t L' reactor fuel. The plutonium nitrate is converted into solid plutonium oxide, packaged, and stored at the repro-
. cessing facility for as long as five years before being 20 shipped off-site for disposal as a waste.- /
4 The major byproduct of reprocessing is an acid' solu-tion of fission products. This high-level liquid waste (HLLW) is stored in underground tanks for as'long as five years and then converted into a solid. The liquid is evaporated in a spray calciner to produce particle:s of , solid-that are melted into a massive glass in cans. The l cans of glass are sealed and stored at the reprocessing l 21 f acility -for an additional period of five yearr 3r more- / and are shipped off-site to a Federal repository for permanent disposal.22/ Other' radioactive wastes are also produced. These include the cladding hulls and hardware, degraded solvent from the Purex process, liquids and solids from treating off-gas from the dissolver and other process vessels and air from the plant ventilation system, failed equipment, .
--20/ NUREG-Oll6, p. 4-6, 4-7, 4-102; BG&E/IWS, p. II-9; Staff Responses to Questions on Economic Data p. 1-103.
--21/ The total storage time for HLW, as liquid and solid, at the reprocessing facility is 10 years.
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) 22/ Id., p. 4-8.
< /' . 1 l (q l and miscellaneous liquid and solid wastes arising from i day-to-day plant operation and maintenance. As almost all these wastes are contaminated by more than 10 nanocuries/ gram of TRU nuclides, they are classified as TRU wastes 23 and must be disposed of in a Federal repository.- / Only solid, non-combustible TRU wastes are shipped from the reprocessing facility. Combustible liquids and solids are incinerated. Aqueous solutions are concentrated. Incinerator ash and other dispersible solids and aqueous concentrates are mixed with cement or other solidification agents and packaged. Cladding hulls and hardware and other non-combustible solid waste are packaged.24/ - The ; staff assumes that these TRU wastes are then stored for less than 20 years at the reprocessing plant before being 2
' shipped to the Federal repository.-5/
The radioactive gases generated during the reprocess-ing are treated in an off-gas system to remove iodine and particulates and are discharged to the atmosphere. Con-densate from the various liquid waste concentrators and 23 / Id. , p. 4-8. 24/ Id., pp. 4-39'thru 4.58, 25 / Id. , p. 4-6 ,
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- p. 1-52. '
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/~N U' commercial plant has been built and operated in the U.S.
to reprocess spent oxide fuel from commercial power reactors. That plant was the Nuclear Fuel Services (NFS) plant _at West Valley, New York. Designed to reprocess 1 MT/ day (300 MT/yr) , it operated from 1966 through 1971 and. reprocessed 630 MT of which 250 MT was commercial fuel i 1 I ranging in burnup from 12000 to 32000 MWD /MT. 28/ - The NFS i plant was shut down for modifications to overcome a variety of operational problems and to increase the capacity. After some study, the owner decided that modifying the small existing plant to meet the present regulatory require-l ments would be uneconomic and terminated the operation.29/ -
.In 1964-1974, the General Electric Co. built and tested a 300 MT/yr facility, the Midwest Fuel Recovery Plant (MFRP) at Morris, Illinois. A process much different from Purex was used and the tests showed that extensive modifications should be made before attempting to reprocess radioactive fuels. General Electric decided to terminate the project rather than make the modifications.30/ -
28/ BG&E/IWS, pp. II-13 thru II-17. 29/ Id., pp. II-20 thru II-22. ,
~~30/.Id., II-22 thru II-25; Joint Responses of TVA, Common-wealth Edison, et al., and Baltimore Gas and Electric,
-g(') et al. to First Round Questions (BG&E/FRA) p. 13.
fs . i Since 1971,.the Allied Chemical Company and associates have built the Barnwell plant. This facility uses the
. Purex process, has a design capacity of 5 MT/ day (1500
. 1 MT/yr), and. carries the reprocessing through the produc- i l
tion of purified plutonium nitrate and uranium hexafluoride. It does not have facilities for converting the-plutonium l I to the oxide or for solidifying the RLLW. Construction has been completed and the plant is in the final phases l of cold (non-radioactive) testing. 3J/ In 1976, Exxon Nuclear Co. submitted to the NRC a preliminary safety analysis report and an environmental report for a Nuclear Fuel Recovery and Recycling Center 1 (NFRRC or Exxon) to be constructed at Oak Ridge, TN. , I The NFRRC is designed for an initial capacity of 1500 MT/yr and an ultimate capacity of 2100 MT/yr. The facility uses the Purex process and includes conversion of plutonium to the oxide, uranium to the hexafluoride and ELLW to a , glass.}2/ The current position of the Federal government i with regard to reprocessing and plutonium recycle has interrupted plans for the facility. 3}/ BG&E.IWS, II-25 thru II-27. 32/ BG&E/IWS, II-27 thru II-30.
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The Purex process is in use for reprocessing oxide fuels, or its use is planned, in other reprocessing plants around the world. The Windscale plant in England has a capacity to reprocess 300 to 400 MT/yr of oxide fuel as part of its 1500 to 2500 MT/yr total capacity. A 1000 MT/yr commercial plant is planned for 1984.33/-- The La Hague plant in France, total capacity 800 MT/yr, reprocesses cxide fuels at rates that are planned to increase from 150 MT/yr in 1976 to 800 MT/yr in 1980. A 1000 MT/yr commercial plant is planned for 1985.34/ Pilot plants have operated in Germany, Japan, Belgium, Italy and India. Germany is planning a 1500 MT/yr commercial oxide reprocessing plant for 1985.35/ C. Capacity of Model Facility California Energy Resources Conservation and Development Commission (CERCDC) witness, Dr. Anderson, l 1 contended that the staff had chosen too large a capacity for the model facility. The values in Table S-3 are 1 intended to apply to nuclear power plants that are being l l I 33 / NRC/FRA, p. 2-12. l 34/ Id., Sierra Club exhibit "L'usine de la Hague", p. 15, 3 5 / NRC/FRA, pp. 2-13, 2-14
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l
p) 6
'v licensed at the present time. Dr. Anderson concluded that, at'least until the 1990's, the fuel from these reactors would be reprocessed in the first commercial reprocessing plants, which would have a capacity of about 1500 MT/yr.
Some impacts of these plants, on a per reactor year basis, l would be larger than for a 2000 MT/yr plant and could ) l exceed the staff's estimates for Table S-3.36/ - l l According to the staff, 62 reactors had operating l licenses and 143 plants had construction permits in j October 1976. The NRC projected that within the next five years 117 reactors would be operating and 151 would be , under construction.37/ Barnwell and NFRRC would fall far I short of providing the capacity needed to reprocess the fuels from this number of reactors. Noting the trend
'toward increased capacity in commercial reprocessing facil-ities, the staff selected 2000 MT/yr as a likely capacity 38 for additional plants.- / Experience with the U. S. gov-ernment plants was cited as evidence that the technology is presently available to build and operate a Purex 36/ CERCDC/IWS, p. 23.
37 / NUREG-Oll6, p. 2-20. Also see CERCDC/IWS, p. 21, for spent fuel discharge and backlog. 38
/ NRC/FRA, p. 1-77; Tr. 171, 172. (An example is the - NFRRC which has 2100 MT/yr proposed capacity.)
U
reprocessing facility with a capacity well in excess of 2000 MT/'yr. 31/ Some of the processes required for a com-mercial facility, such as the shearing of the fuel elements and solidification of the HLLW, have not yet been demon-strated on so large a scale. The Barnwell plant, the NFRRC design, and results of development programs (pp. 37-39, infra) were cited as evidence that equipment will be avail-able to carry out these operations on a 2000 MT/ year scale. Dr. Anderson also maintained that the environmental impacts of reprocessing fuel with burnups over a range, say-from'20,000 to 40,000 MWD /MT would have to be analyzed in order to ensure that the impacts included in Table S-3 l l are conservative.40/- A reprocessing facility of given j capacity would reprocess the fuel from fewer reactors if l the burnup were lower than assumed by the staff, and some impacts would increase on a per reactor basis. The utilities contend that, by using 35 MT/RRY as the reprocessing rate and 33,000 MWD /MT as the burnup for spent fuel, the staff has made its analysis so conservative that 39/ Although the government plants reprocess metal fuels,
~~
Dr. Anderson acknowledged that the differences between metal and oxide. fuels do not necessarily make the latter moref difficult to reprocess. CERCDC/FRA, pp. 20, 21. 40/-CERCDC/FRA, p. 5. [
/N V the study of impacts for a range of burnup is unnecessary.
.A typical light water nuclear power plant generates elec-trl?ity.with a thermal efficiency of 31 percent or more.
A reference reactor would actually irradiate the fuel to only 27,000 MWD /MT if 35 MT/RRY were discharged. The actual discharge rate would be 29 MT/RRY for a burnup of 33,000 MWD /MT.A1! CERCDC witness, Dr. Anderson, believed that the staff had chosen a burnup considerably above current experience with nuclear plants. However, his analysis was based on inform-ation from the early operation of the reactors.12/ The design burnup for the ':quilibrium fuel cycle is 27,500 MWD /MT for BWRs and 33,000 MWD /MT for PWRs. Current actual rates are 25,000 MWD /MT for BWRs and 33,000 MWD /MT for PWRs.b$! .The realistic burnup for a reference reactor on the equilibrium fuel cycle would then be about 31,000 MWD /MT and the corresponding reprocessing rate would be about 30 MT/RRY. The fission product and TRU concentrations in the fuel increase with increasing burnup.bb/ According to 41/ BG&E/IWS , pp. II-34, II-35; CERCDC/IWS, p. 9. 42 / CERCDC/IWS, pp. 7-8; CERCDC/FRA, p. 10. 4 3 / NRC/FRA , . p . 6-5.
,- 4 4 / Id. , pp. 1-86, 1-87.
I ['T \J l l the' staff and the utilities combining the higher-than- ) actual burnup and discharge rate provides a conservative basis for assessing impacts related to radioactivity and to the amount of fuel that must be reprocessed. D. Predicted Performance of the Model Facility
- 1. Capacity Factor The conservatism of the staff's evaluation of the environmental impacts of reprocessing also depends on whether the reprocessing facility actually operates at a capacity factor high enough to' serve 57 reference reactors.
Testimony of several partiesAE/ questioned the 0.8 capacity factor assumed by the staff. If the facility cannot be operated at a high enough
' capacity factor, more plants will be required for a given number of reactors; therefore, the impacts related to plant size would increase per RRY. Also, low capacity factor 1 implies difficulties with plant and equipment and the need for extra maintenance. Extra maintenance can be expected I'
to result in greater occupational exposure and larger vol-umes of maintenance-generated radioactive waste per RRY.A5! i i
'45/ Tr. 1211-1214; Pohl and Resnikoff, App. A, pp. 89-169. l
/j~ 46/.Tr. 151, 171, 187.
-- f
\_/
A
v) A major issue regarding the staff's assessment of impacts is the extent which past experience should be used in projecting the capacity factor of future plants. As noted earlier, the staff's model facility was primarily based on Barnwell. Participants who contend that the staff has underestimated the impacts of reprocessing point out that Barnwell has not operated. They say that more weight.should be given to the actual experience at NSF than to the improvements projected for Barnwell and the model facility.A1! The staff realizes that Barnwell might not operate with the 0.8 capacity factor expected by its designers. But, the staff believes that 0.8 is a reasonable capacity factor for future plants, such as the model facility, because the designers will have knowledge
' gained from resolving any difficulties experienced at Barnwell.AS! Although the overall capacity factor at NFS was only 0.33,AE! the fractional down time for the U. S.
government plants is reported to be 0.2 to 0.255S! which indicates that a capacity factor of 0.75 to 0.8 can be 47/ Pohl and Resnikoff 5; Tr. 941-950, 1211-1214. 48/ Tr. 189-190. 49/.Tr.-180. 50/ NRC/FRA, p. 1-52. i u . 1
(~' (_ achieved.- On a realistic basis, a reference reactor would discharge fuel at a rate of 30 MT/yr (supra, p. 28 ) . At this rate a model facility with a capacity factor of 0.7 could service 57 reactors.
- 2. Releases of Radioactivity Several parties believe that the staff has underesti-mated the amount of radioactivity that will be released in effluents from the model facility because too little weight was given to past experience.51! The Sierra Club contended that the staff had underestimated some of the releases by several orders of magnitude.
Testimony by Sierra Club witnesst. Dr. Resnikoff, described problems encountered at NSF and his evaluation ; of the improvements made at Barnwell.5 ! As evidence of the fallibility of predictions that are not founded on historical experience, Dr. Resnikoff showed that the 51/ Natural Resources Defense Council Intial Written State-
~~
ments (NRC/IWS) (Terry R. Lash), pp. 21-22; Testimony of Ruth Thomas on Behalf of Environmentalists, Inc., i
- p. 1; Statement of' Union of Concerned Scientists, p. 3; Pohl and Resnikoff, p. 1.
52/ Pohl and Resnikoff, Appendix A. This testimony by Dr. Resnikoff was prepared originally for the GESMO _ proceeding, Docket No. RM 50-5, and included in the testimony for this proceeding. ? - l
/~S ( 4 .V radioactivity in the liquid effluents from IFS had reached concentratior.s more than 1000 times greater than had been predicted.52! He contested the staff's use of decontamina-tion factors for the model facility that are 1000 times greater than were experienced at NFS.5A!
Dr. Resnikoff agreed that the releases of radioactivity at NFS had not exceeded the limits imposed by 10 CFR 20. He agreed that Barnwell probably could also meet those limits.b5! However, he pointed out that 10 CFR 20 concerns the concentration of radioactive nuclides in the effluents but not the quantity. He argued that industry could comply with the regulations, while discharging many more curies 1 than has been predicted by the staff, by pumping more air l or water for dilution.bb/ According to testimony by the staff and the utilities, there will be substantial differences between the model l
~
53/ Pohl and Resnikoff, Appendix A, pp. 19-20. 54/ Pohl and Resnikoff, p. 6, Appendix A. p. 20. The
~~
decontamination factor, as used here, is the ratio of the amount of radioactivity in the fuel entering the reprocessing facility to the amount discharged in the, effluents. 55/ Tr. - 956-957.
~~
The staff estimates that the releases from Barnwell will be less than five percent of the 10 CFR 20 limits. Tr. 216-226. 56/ Tr. 958. V i
y~ ( l reprocessing facility and the NSF or U. S. government plants which were built many years before the NRC regula-
- tions required an ALARAE! finding. Those plants would not satisfy current regulations. S! Improved effluent l control systems were incorporated in Barnwell and were assumed for the model facility in order to satisfy the ALARA requirement. In calculating the radioactive releases from the model facility, the staff used the lower range of of the expected performance of each component in the ef fluent treatment systems. - ! The staff believes that the values obtained are conservative and appropriate for l use in Table S-3, even though they might not be found to be ALARA when licensing an actual plant.- !
The EPA objected to some of the release rates predicted by the staff because they exceed the rates that will be allowed by the regulations in 40 CFR 190.6_1__/ Testimony presented by the utilities compared the releases predicted by the staff for the model facility with those predicted E/ As low as reasonably achievable. , 58/ NUREG-0216, p. 3-9; Tr. 222. 59/ NUREG-0216, p. 3-14. - 60/ Id., Tr. 154, 155. j 61/ Tr. 71, 1029-1032. See infra pp. 51, 55. l n) L
l 1 for NFRRC by its designers. The utilities concluded that the staff had overestimated the release rates by factors that varied from 1.1 to more than 1000 depending on the 1 l nuclide. 6 2/ The testimony of the utilities cites two basic 1 l assumptions that tend to make the staff's calculations conservative. First, the amount of radioactivity processed I and'the-amount released per RRY are proportional to the product of the reprocessing rate per RRY times the burnup.
'The staff's assumption of a reprocessing rate of 35 MT/RRY verst- a realistic value of 29 MT/RRY for a burnup of 1
35,000 MWD /MT causes the staff's estimates to be 1.2 times ; realistic estimates. Second, the assumption that the fuel will be reprocessed after 160 days cooling, as opposed to 1 l the more likely six years or more, causes the staff to over- l l estimate the releases of radionuclides with half liv'as I l shorter than a few years. For example, the longer cooling would reduce the releases by factors of about 1.4 for l tritium, 1.5 for krypton-85 and 64 for ruthenium-106 63/ 1 Iodine-131 would have decayed away. l l E. Reprocessina Experience and Projected Improvements , i I The record of the proceeding contains much testimony
- about 1 the. experience at NSF and the basis for the staff's O
g
'62/ BG&E/FRA, pp. 123-132.
63,/.BG'&E/IWS, pp..II-34 thru II-35. - L
i I assessment of the model facility. That part of the record is summarized in this section.
- 1. Systems and Equipment
- a. Shearing of Fuel Elements The " head end" preparation of the fuel for the l Purex process is the major difference between the reprocess-1 ing of metal fuels in U. S. government plants and oxide fuels in commercial reprocessing plants. Although there are processes for dissolving the cladding, as in the gov-ernment plants, shearing is the preferred method of head-end treatment of commercial fuels.5A!
At NSF the ends were sawed off the fuel elements and the fuel-containing tubes were sheared into small pieces before being fed into the dissolver.55! These operations caused major problems during the entire period of NFS opera-tion.$5! At Barnwell only the shear will be used.52! The design of the NFRRC includes shears; whether end pieces will be removed by a saw or other means has not yet been decided. 63/ 64/ Tr. 172. 65/ Tr. 851; Pohl and Resnikoff, Appendix B, p. 81. 66/ Pohl and Resnikoff, Appendix B, pp. 93-94. 67/ Tr. 856. G 6_8/ 8 Tr. 853-855.
r3 'b/ Difficulties were experienced with both the saw and the shear at NFS. Saw blades dulled or broke and parts of the saw mechanisms failed on several occasions.59/ Shear blades dulled or broke. The shear lubrication system failed. NFS reprocessed several types of fuels. The fixtures that held the fuel elements in place during saw-ing and. shearing had to be changed for each type of fuel. On one or more occasions the fixtures broke and they did not function properly at other times.1S/ Despite the many difficulties, the shear at NFS was capable of sustained operation at significantly greater than the design rate of 1 ton / day.71/-- The sh' earing at NFS was carried out in the Process Mechanical Cell (PMC). This cell was provided with two
' overhead cranes, a power manipulator, and four pairs of through-the-vall manipulators to operate'and do remote maintenance on the equipment in the cell.12/ The manipu-lators were required to carry out many operations in order to change saw and shear blades and to change the holding fixtures in the shear.13/ The manipulators and cranes 69/ Tr. 851, 852; Pohl and Resnikoff, Appendix A, pp. 93-94.
70/' Tr. 852; Pohl and Resnikoff, Appendix A, pp. 93-94. 7_1_/ Tr . 853. 72/ Pohl and Resnikof f, Appendix A, pp. 80, 92. 73 __ 3 / Tr. 852, 853, 854. l
were responsible for many of the difficulties in the PMC. Electrical cables to the cranes were pinched, snagged and cut. Crane hoists and_ parts broke or malfunctioned. The manipulators required frequent repair or replacement and . were too weak for some of the jobs, some unanticipated, that they were called upon to do. 74/ Although the cranes could be withdrawn from the PMC into another room for main-l tenance, they could not be decontaminated before maintenance personnel entered the room. 75/ i l The shear installed at Barnwell is a French design l built by St. Govain. This shear was chosen because the French company had built and cold-tested several shears. The shear at Barnwell is designed so that the blade and other in-cell parts can be removed as modules by use of an impact wrench on an overhead crane. Also, the entire shear can be removed and replaced by a spare. The bearings and associated mechanical parts are outside the cell where
/
they can be maintained directly. 74/ Tr. '852-853; Pohl and Resnikoff, Appendix A, pp. 96-97. 75/ Tr. 862. 76/ Tr. 856.
-77/_Tr. 856-858.
u
Two overhead cranes are provided to do most of the maintenance operations in the shear cell and they are assisted by a manipulator. The cranes are similar. to., but not the same as, cranes that are used for remote main-tenance at the Hanford and Savannah River plants. The cranes and power manipulator can be withdrawn into a shielded cell for maintenance. The equipment can be par-tially decontaminated before maintenance personnel enter the cell to make repairs. Although the maintenance con-cept resembles that at NFS, the designers believe that improvements incorporated at Barnwell will overcome the problems encountered at NFS.1 ! The extent to which the problems of shearing fuel at NFS have been overcome at Barnwell can only be learned if and when that facility is permitted to operate. Apparently, the equipment has functioned well in cold tests. 9! The French shear is installed in the La Hague plant in France and has been used on irradiated fuel. According to the Sierra Club, that shear has broken down after shear-ing less than 100 MT: of high burnup oxide fuel, the pro-visions for remote maintenance proved inadequate, and the g~.3 '78/ Tr. 859-862. V 7 9/ Tr. 857.
j(}
' shear must be repaired by direct contact. The Sierra Club does not indicate that the provisions for maintenance at La Hague include the improvements claimed for Barnwell.
The NFRRC design differs from Barnwell. It has two shears in the mechanical cell. One is to be capable of shearing 1500 MT/yr but two are required for the expanded 1 1 capacity. The maintenance concept is similar to that at '
.Barnwell. A shear is being designed and developed for the Exxon facility'and the entire head end process is to be tested on full scale before being installed in the plant. 81/ /
I Also the DOE has begun a program at the Oak Ridge National l Laboratory to develop a much improved shear system for future reprocessing facilities. 82/
- b. Solvent Extraction In the Purex solvent extraction process that follows the head-end treatment ~, Barnwell, NFRRC and the model facility have been changed from NFS in several ways. A centrifuge j has'been provided between the dissolver and the solvent 80/ Sierra Club Position on the Environmental Effects of I the. Uranium = Fuel Cycle, June 26, 1978, p. 7. The 1 statement indicated that the manipulators were not adequate to'do.the repairs.
81/,Tr. 853-856, 898-899..
-82/ NRC/FRA, p. 1-40; Tr. 175-176.
.]
' 1
<I E .. ..
t
/S g extraction equipment to remove suspended solids to reduce the frequency of plugging of transfer lines that caused problems at.NFS. Centrifuges are used for removal of solids in the Purex plant at Hanford and Savannah River.S1!
Barnwell, NFRRC and the model facility will use cen-trifugal extractors in the first stage of solvent extrac-tion in place of'the columns used at the government plants and at NFS. The radiation level in the first stage of solvent extraction will be much higher in plants that reprocess high burnup fuels. The solvent is irradiated for a much shorter time and degrades less in a centrifugal 84/
~~
contactor. The Barnwell contactors are of a French I design that is used in reprocessing plants in France. The Savannah River has used centrifugal contactors of a some- . l what different design with good results.g5,/
- c. Acid Recovery Evaporators Acid recovery evaporators are essential components of Purex plants. Difficulties with the acid recovery evaporator at NFS caused. higher occupational exposures and 1
83_/ Tr. 886; NRC/FRA, p. 1-50. 84,/ NRC/FRA, pp. 1-54, 1-55; Tr. 210.
. --85/ Tr. 210-212; NRC/FRA, p. 1-49.
7-s d 0 .
'~') 41 -
(J lower decontamination f actors than the designers had anti-- cipated. The radioactive liquid in the evaporator was often entrained in the vapor stream, contaminating equip-ment in unshielded areas and liquids that were to be released to the environment. 86/ The staff and the utilities believe that NSF's prob-lems were caused by design deficiencies. They believe that the experience at NFS and the government plants has been used to eliminate the deficiencies from the evapor-ator designs for Barnwell and NFRRC. Also, the equipment at Barnwell, as in the government plants, is installed in shielded areas to protect the operators if excessive amounts of liquid are entrained occasionally. At Barnwell excess water will be vaporized into the atmosphere. This extra evaporation step is to provide an additional decon-tamination f actor. 87/
- d. Conversion of Plutonium and Uranium In Barnwell, NFRRC and the model facility, the plutonium will be purified entirely by solvent extraction, thus eliminating an ion exchange process that caused some 86 / Pohl and Resnikof f , Appendix A, pp. 106-108; Tr. 204.
,3 87 / Tr. 204, 212; BG&E/ FRA, pp. 7-8, 89.
(_) e4 m e
^
j's} operational problems at NFS and safety concerns at govern-
! Because the plutonium from.the model ment plants.
facility is to be treated as a waste, about five percent of the fission products will be'left in the plutonium in order to make theft difficult. The plutonium nitrate will be converted to the oxide by a process similar to that used in the government plants. However, the processing j must be done remotely rather than in glove boxes. 89/ l l Uranyl nitrate will be converted to uranium hexafluoride by conventional processes. .
- d. Storage of HLLW_ l The HLLW from the first stage of sclvent extraction and the concentrated aqueous wastes from subsequent process-ing will be stored in large underground stainless steel tanks, similar to the Barnwell design, for up to five years before being solidified. The waste solution is acidic and will not contain enough solids to complicate removal from the tanks and further processing. The tanks will be equipped to cool the waste. The off gases will be treated before being released to the atmosphere. '! At Barnwell the HLLW and intermediate level liquid waste tanks are installed in 88/ BG&E/FRA, p. 117; BG&E/IWS, II-20.
(~)N 89/ Tr. 338-345. 90/ NUREG-0116, pp. 4-12 thru 4-17. ,
1:
"* . t a +
i (O/ J v concrete vaults lined with stainless steel. The vaults
~
are designed-to hold the entire contents of the tanks. Pipes connected-to the waste tanks are double contained. Tanks have leaked at two different reprocessing facil-ities. Although these leaks'have not had any known effect i off-site, '! their occurrence led some parties to question the staff's assumption that leaks would not occur at the model facility. l I Twenty or.more HLLW tanks have leaked at Hanford and released thousands of gallons of radioactive liquid into the" soil. 93/- The record shows that the tanks that leaked into the soil at Hanford were not double contained. Eight l tanks at Savannah River leaked, and in one case a few tens
.of gallons of liquid overflowed the outer containment, into n
the soil. In this case the outer containment was not designed to hold the contents of a full tank.
- No leaks occurred in the waste tank at NFS. None have occurred in 15 stainless steel tanks that have been used to hold acidic wastes at the Idaho Chemical Processing
' 91/ ,Id . , Tr . 371.-
9 2/ BG&E/IWS , p. III-15, III-16 ; Tr. 3 61-3 6 2, 365-367. 9 3/ BG&E/IWS , _ pp. III-15. ON 9 4/ BG&E/IWS , pp. - III-14_ thru III-18. _
-------___.m_ _ _ _ _ . , _ . , _ _ _ _ .
\
44 - [Q l Plant for more than 20 years. 95/ Since present NRC regu- ' lations require HLLW to be solidified it is doubtful that 1 HLLW will be stored as a liquid for even as long as five years. 96/ , I
- e. Solidification of HLLW NRC regulation 10 CFR 50, Appendix F, requires 1
that the HLLW in a fuel reprocessing facility be converted i to a stable dry solid. Several solidification processes have been under developnent. In the process chosen by the l l staff, the liquid is atomized into small droplets which are j 1 dried and calcined as they fall through a vertical, heated-wall chamber. The calcine drops directly into a canister l l in a multizone furnace below the calciner. Glass frit is 4 added to the canister along with the waste to produce a l l massive glass. After the canister is filled, it is cooled l and sealed, the surface is decontaminated, and it is removed to storage. The entire process is operated remotely within l shielded cells. '/ l 95/ Id., pp. III-19, III-21, III-22. 96/ Modifications at NFS and the design of NFRRC were based on prompt solidification of HLLW. The volume of tanks at Barnwell was to be the minimum necessary to store the waste until a solidification facility could be provided. NUREG-0116, p. 4-17; Tr. 931. () 97/ NUREG-Oll6, p. 4-19; BG&E/IWS, pp. III-25 thru III-27. e e -
-/7 LJ After the waste canisters have been sealed and the outer surface has been decontaminated they will be stored in a water basin for up to five years. The water basin would be similar in construction and operation to those presently in use for storing spent fuel elements at nuclear power plants.EE/
The spray calciner/in-can melter has been under development for more than 15 years at the Pacific Northwest Laboratories. Radioactive waste with the full level of radioactive heat and the full spectrum of fission products has been calcined and solidified on a-1 MTU/ day scale in a Waste Solidification Engineering Prototype (WSEP). Within the past year cold tests have been run at rates up to 9 MTU/ day in a recently completed engineering test
' apparatus. Although commercial reactor waste has not been used in the tests, the composition of the waste was adjusted to simulate commercial material.S1!
The utilities state that alternatives to the spray calciner/in-can melter are presently available or are being developed for solidifying ELLW. A fluidized bed calciner has been used to solidify HLLW at the Idaho Chemical Processing Plant since 1963. A pot calcination and l 7-w . A)s 98/ NUREG-Oll6, pp. 4-24, 4-25.- 99/ NRC/FRA, pp. 1-98 thru l-105; Tr. 287-304.
v
~~}
vitrification process was developed at ORNL and demonstrated with HLLW from Hanford. A continuous ceramic melter based on technology of the glass industry and a continuous metallic melter have been tested on an engineering scale. A fluidized bed / continuous ceramic melter system is pro-posed for the NFRRC. In France a production scale pot vitrification unit was placed in operation at the Marcoule
~
facility in 1969. Processes similar to those being developed in the U. S. have been operated on pilot plant scale, in some cases with radioactive waste, in France, England, and Germany. Industrial scale operations are scheduled to bsgin within the next 10 years.100/ The Si'erra Club was concerned because the staff expects less than 0.5 percent of the plutonium to be in the HLLW in the model facility, whereas, 1.5 to 2 percent was lost to the HLLW at NFS. ! Testimony that includes information on the plutonium losses to the HLLW in the large government plants was introduced in support of the staff's predictions.102/ For the purposes of this proceeding, it is assumed that all the plutonium will be disposed of as waste. Staff 100/ BG&E/IWS, pp. III-25 thru'III-36: Tr. 287-289, 892-893.
-[ )
' ~ '
101/ BG&E/FRA, p. 91; Pohl and Resnikoff, p.
- p. 73.
8, Appendix A, 102/ BG&E/FRA, pp. 90-91; NRC/FRA, pp. 5-3, 7-88, 7-89.
- ) ..
f) V testimony indicates that whether the HLLW contained 0.5 percent or 2 percent of the plutonium would not signifi-cantly affect.the waste processing or disposal. The. testi-mony of the utilities suggests that all the plutonium be left in the HLLW stream to be solidified and disposed of with the fission products. This possibility was rejected by the staf f because of potential criticality problems.'103/ l
- f. TRU Contaminated Waste The remainder of the radioactive wastes generated at a reprocessing facility are mostly TRU wastes. They are to be disposed of in a geologic repository. These wastes are cladding hulls and hardware, degraded solvent, various aqueous wastes, failed equipment, filters from the various air and gas handling systems, and general trash.
For the model facility, the staff has specified that the l cladding hulls and hardware are to be packed in dry sand l in drums. The sand is.used to reduce the potential for combustion of the zircaloy. Non-combustible liquids and dispersible solids are to be mixed with cement and packaged. Failed equipment is to be decontaminated and packaged with other non-combustible trash. Except for incineration of the combustible wastes, the processes for handling the TRU JN _) 10y BG&E/IWS, IV-1; Tr. 294-297, 888. i
vO wastes are. essentially those presently in use at the gov-ernment plants. 104/ Combustible solids and organic solvent are to be incinerated and the ash'is to be mixed with cement and
/ The staff chose incineration as packaged in drums.
the method for treating combustible waste to reduce volume, eliminate combustible material from waste sent to the Federal repository and, to ensure a conservative estimate 1 of the environmental impacts of treating waste.106/ I Combustible nuclear wastes have been incinerated in 1 J the U. S. and foreign countries for more than 25 years. l Because of mechanical and operational problems none of the incinerators is in use in the U. S. to process TRU waste
,at the present time. 07/ However, a production-scale, controlled-air incinerator for use with TRU contaminated waste has been constructed and is being tested at Los Alamos. The unit is a commercial incinerator, developed for disposal of industrial waste. It has been equipped to test off-gas cleanup systems designed to overcome past deficiencies. 08/ Small controlled-air incinerators are 104/ Tr. 447-448, 451-458.
,y
, 105/. H., pp. 4-43 thru 4-45.
'q_) '106/ Tr. 449.
107/ Tr. 441; ERDA 76-43, pp. 9.7 - 9.9.
'.108/ Tr. 443-445; NRC/FRA. 7-42; NUREG-0216, pp. 3-49, 3-50.
-, - l A 49..
\- ,! i used to burn TRU waste at Windscale in England and Marcoule in France. The unit at Windscale has burned 15,000 kg of l waste containing 17 kg of plutonium. 09/ Several alter-natives to the controlled air incineration process are
-used or are being developed to handle combustible TRU wastes. They include: simple compaction and baling, shred-ding and packaging (possibly mixed with. cement), several other types of incineration, molten salt combustion and l 110 acid digestion. /
The staff assumes that the packaged TRU waste will be stored at the fuel reprocessing facility for 15 years or so. Welded. stainless steel containers of cladding hulls and hardware will be stored in the fuel storage basin where the heat from the radioactivity can be dissdoated in the circulating pool water. Other waste in drums will be stored in an earth covered concrete vault. Similar mis-cellaneous TRU wastes are stored, covered with salt, on asphalt pads at the Idaho National Engineering Laboratory. ! r 109/ Tr. 442. 110/.BG&E.IWS, pp. V-6 thru V-12. 111/ NUREG-0116, pp. 4-63, 4-64. ! t l% / 1 e ? + , , e
.i J\s, 2. Radioactivity in Plant Effluents
- a. Sources The only releases of radioactivity to the biosphere from Barnwell, NFRRC, and the model facility will be in the gaseous effluent. Liquids will be vaporized or solid-ified and shipped with the solid wastes as indicated at
- p. 42 , supra. Most of the gaseous radioactivity will be in the off-gases from the shearing operation, the dissolver, ,
and other process vessels.
- b. Tritium The staff assumes that all the tritium in fuel
/
cooled for '160 days will be released to the atmosphere. No process for containing the tritium is presently avail-
'able.
! Factors which the staff and the utilities believe have resulted in the value in Table S-3 being an overesti-a mate of the amount of tritium released were discussed at
- p. 34, supra. Also, testimony of the utilities indicates that 15 to 50 percent of the tritium may be in the cladding and would not be released.114/
112/ BG&E/IWS, pp. 2-29, 2-35; Tr. 199, 205. 113/ NRC/FRA, pp. 1064 through 1-65; BG&E/FRA, pp. 126-127. 114/ BG&E/IWS, p. 2-35. (~} LJ
i ,. ',
-- M w) l I
- c. Krypton-85 l The staff assumes that all the krypton-85 in the feel will be released to the atmosphere. This amounts to l l
. 400,000 ci/RRY. EPA objects to the use of this value in i Table S-3 because it. exceeds the 50,000 ci/ gigawatt yr.
that will be permissible after 1983 under the EPA regula-tions in 40 CFR 190. The EPA believes that the technology is presently available to remove most of the' krypton from ,
/ Although not in the orig-the reprocessing off-gases. l l
inal design and construction, a krypton removal system will be required at Barnwell. A pryogenic system, which i l the designers believe will remove up to 95 percent of the krypton, is included in the design of the NFRRC.
-l The staff justifies the use of its value for krypton i release in Table S-3 on the basis that the regulation does not apply to fuel burned before 1983 and it does not apply to waste management. The value in Table S-3 is intended to apply to all phases of the fuel cycle and to apply to
~
spent fuel as a waste as well as to reprocessing. The staff assumes that all the krypton-85 would be released from the spent fuel after it is emplaced in a repository. A 115/ NUREG-0216, p. 3-122.
-- - 116/ Tr. 87 6-877, 880-881; BG&E/FRA, pp. 124-125.
= .-
("\ Q) The staff proposes.to change the values for krypton release when the EPA. requirements become effective.
- d. Carbon-14 l The staff assumes that all the carbon-14 in the fuel will be released to the atmosphere. There is no i regulatory requirement for containing and disposing of carbon-14. The model facility has no provisions for con-taining it. However, in the design of the NFRRC equipment I is provided for removing carbon monoxide and carbon diox-ide from the off-gas before it enters the krypton recovery system. This equipment would likely remove much of the carbon-14. The NFRRC plans to release the carbon-14 to the atmosphere but methods of confining it could be developed. !
- e. Iodine-129 i According to Sierra Club testimony, practically all the iodine-129 was released in the effluents at NSF.
The Sierra Club believes the staff should assume that a large fraction of the iodine-129 would be released in effluents from the model facility. '! 117/ NUREG-0216, p. 3-122; NRC/FRA, pp. 3-1 thru 303; Tr. 205-206. (O,) 118/ Tr. 882-884; BG&E/FRA, pp. 125-126. * ! 119 / P&R, p. 6, App. A, p. 25.
+ .. .
(^h, \~A , The staff calculates that 2.5 percent of the iodine in the' fuel will be'rel' eased to the en~vironment.from the model facility. The staff estimates that 90 percent of the iodine will be volatilized into the off-gas from the dis-solver and vessels ahead of the extraction process. This iodine will be removed from the off-gas by scrubbers and absorbers and will not contribute significantly to the release. Ten percent of the iodine in the fuel will remain in-the dissolver solution and will eventually enter the vaporizer where excess process water is evaporated into the atmosphere. Mercuric nitrate will be added to suppress iodine volatility and hold 75 percent of the iodine in the liquid. This will limit the loss of iodine-129 in the effluent to 2.5 percent of that which entered
'the process in the fuel. Of l l
l The staff believes that its estimate of the release I of iodine in the effluent from the model facility is con-servative. Recent work at ORNL indicates that less than one percent of the iodine will remain in the dissolver solution after it has been sparged with air in the 120/ BG&E/FRA, pp. 128-129. 0) [ . L
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dissolver and other head-end vessels. 21/ At Barnwell, iodine is to be removed from the off-gas from those vessels by passing it through mercuric nitrate scrubbers - followed by adsorbers packed with silver on zeolite. Parts of this system have been developed and tested with radioactive material at several locations, but the overall system has not been operated on a commercial scale.122/ The combination of scrubbers and absorbers is expected to give a decontamination factor of 1000. ! Mercuric nitrate will be used to reduce the release of iodine from the vaporizer at Barnwell. Considering the lower fraction of the iodine likely to reach the vaporizer and the retention there, the staff calculates that the iodine release from Barnwell could be as little as 0.05 percent -- a factor of 50 below the release from the model facility. The NFRRC propose'to use an iodine stripper and ion exchange column to remove iodine from the vaporizer feed water. 24 / The utilities estimate that , the release rate at NFRRC will be 150 times lower than the model plant for iodine-129 and 250 times lower for iodine-131. ! 121/ Id.; Tr. 885. () 122/ Tr. 199-203, 878; NRC/FRA, p. 1-70. 123/ Tr. 224; NRC/FRA, p. 1-70. 124/ BG&E/FRA, p.:129.
- 125/ NRC/FRA, p. 1-71.
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-Although the' staff has calculated that iodine-129 will be' removed in'the1 reprocessing plant, the release rate included in Table S-3 is for discharge of all the-iodine- :
129.to the biosphere. The value in Table S-3 applies both . to spent fuel as a waste and to uranium recycle. The staff assumes that.all.the iodine-129 will be released from the spent fuel after emplacement in a repository.1 !
-EPA considers the; assumed release of all the iodine-129 to be unacceptable because it does not comply with EPA requirements. The EPA regulations under 40 CFR 190 specify that the release of iodine-129 from the uranium fuel cycle shall not exceed 0.005 ci/ gigawatt' year which corresponds to 0.004'ci/RRY. These compare with the value 1.3 ci/RRY
! EPA has concluded that processes are in Table S-3.
'available forolimiting the iodine released in the effluents from. reprocessing facilities and can be developed for isolating the iodine containing wastes. EPA staff finds l
inconsistent the staff assumption that iodin, will be i released from a geologic repository but that other long-lived nuclides will be confined. 126/ NUREG-0116, .pp. 4-114, 115 ; Tr. 601. 127/ NUREG-0216, p. 2-6; EPA 520/4-7 6-016, p. A-27 ; Tr. 1027, 1029. 128/ Tr. 1022-1034. ! l
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(} The staff justifies continued use of the value for total release of iodine-129 on the basis that the EPA regu-lation does not apply to waste disposal or to fuel irradiated before 1983 and that there is no proven technology to meet the standards.129/ Also, iodine is more mobile than most of the other nuclides in the waste.
- f. Ruthenium-106 Ruthenium-106 was a major radioactive contaminant of the effluents from NFS. According to Dr. Resnikoff the 5
overall decontamination factor at NFS was 10 and he ques-8 tions the staff's use of a factor of 10 for the model facility. The staff expects the decontamination factors in the model facility to be higher because of improvements in the handling of process off-gases and plant ventilating air, 1 improvements in acid concentrator design, measures taken l 1 to suppress the volatility of ruthenium, and use of a l vaporizer system to further decontaminate water that is l released from the plant. The staff expects that less than i 10 5 of the ruthenium will be vaporized into the off-gas from the dissolver and other vessels in the model l l 129/ NUREG-0216, p. 3-122. (~)
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l 130/ Tr. 605. i 1 131/ Pchl and Resnikoff, p. 6. l
57 - (^)'). facility. The elaborate off-gas treatment systems are 4 5 expected to provide a decontamination factor of 10 to 10 for the off-gas. At NFS the difficulties with the acid concentrator were the cause for much of the ruthenium getting into the liquid waste that was discharged from the plant. Improve-ments in the concentrators are expected to eliminate most of the difficulties,and sugar is to be added to suppress the volatility at some stages of the process. Sugar has been shown to be more effective in suppressing ruthenium volatility than the sodium nitrite used at NFS. NFRRC proposes to use ruthenium absorbers wherever volatile ruthenium is likely to be produced. Barnwell, the model facility,and NFRRC will use high efficiency filters in the plant ventilation systems to remove ruthenium that escapes from the process equipment.132/ Water discharged from the facilities will be vaporized into the atmosphere, and the vaporizer bottoms will be recycled to the waste processing systems for additional treatment. 1 The above processes concentrate the ruthenium in the HLLW. Ruthenium is volatilized in the solidification of ELLW. Much of-the most relevant information on the behavior l'~\ l l 132 / BG&E/FRA, pp. 129-130; Tr. 203, 204, 213-216. I
/, i -(_/ ,
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- of ruthenium in the solidification process has come from the WSEP tests performed on a 1 MTU/ day scale. 33/
In the WSEP tests ruthenium was volatilized during the solidification process; up to 75 percent under some conditions, but, for some unknown reason, less than two percent'under conditions similar to those of the model facility. However, after the vapors from the solidifica-tion process had passed through the acid recovery, off-gas scrubbing, and final filtration equipment, the overall 8 decontamination factor was greater than the 10 used by the staff.134/
- g. Technetium As part of its contention that the staff has not i
'l adequately evaluated the impacts of the uranium fuel cycle, the Sierra Club' raised the issue that the staff had neglected to consider the impacts of the release of the long-lived fission product, technetium-99. Technetium-99 is present in uranium that is recycled to enrichment plants.
According to the Sierra Club the presence of the technetium in the uranium would make the uranium-only recycle option 133/.Tr. 291-294. See also p. 45 Infra, r~5 134/ BNWL-1667, Waste Solidification Program Summary Report,
-( ) Vol.. 11.
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() - 591-infeasible. The Sierra Club contended further that practi-r cally all the technetium, because of'its volatility and ' 213,000 year half. life, would be released from wastes to the biosphere where.its impact would be significant.135/ The participants agree that the technetium present in'the solution from the dissolver136/ will be partitioned between the HLLW and the uranium product stream. The Sierra Club estimates that 25 percent will accompany the uranium;137/ in NFRRC about 8 percent is expected-to be with the uranium. In NFRRC, the uranium hexafluoride will be passed through absorbers containing magnesium fluoride to remove technetium.138/ The magnesium fluoride, loaded with. technetium, will be packaged and. stored on site until it can be delivered to a-Federal repository.139/ l l 135/ Sierra Club Response' to Question 7 of Commonwealth l Edison,.et al.,' February 20, 1978. 136/.Irt thistproceeding it has been assumed that all the I technetium-99 dissolves. The Final Written Statement on Behalf of Commonwealth Edison, et al., p. II-38, R refers to laboratory data that indicates that a large
. fraction will not dissolve and would pass directly to the high level waste.
137/-Sierra Club Response to Question 7, Commonwealth Edison, et'al.,'pp. 2-4. 138/ Tr. 298, 898-900.
'139/'NFRRC PSAR, pp. 7.5-30 thru 7.5-32.
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The uranium hexafluoride product is expected to meet the requirements for gross beta activity in feed to the enrichment plants. Initially, Dr. Resnikoff agreed that the requirements would be met at NFRRC, but, after further study, he concluded that they probably would not be met.14qf' In the enrichment plant techt um deposits in equip-ment in the diffusion process. Special precautions must be taken to control the spread of contamination during maintenance. This has not impeded the operation or main-tenance of the plants.141/ Some of the technetium that enters the enrichment plant in the uranium feed is dis-charged in liquid and gaseous effluents.142/ The' remainder is disposed of by shallow burial. Technetium in the HLLW is to be solidified along with the other fission products. Although HLLW that was solidi-fied in the laboratory, WSEP, and other tests would have contained technetium, ! no data were obtained on the 140/ Tr. 946, Sierra Club Response, pp. 7-8. 141/ Sierra Club Exhibits and Responses to Oral Requests by Board, Exhibit 2.a., February 17, 1978; Tr. 966-971. 142/ Sierra Club Exhibit 2.e. 143 / The Sierra Club contends that no technetium was used in any waste feed for the WSEP tests (Sierra Club
,s Response to Question 7, p. 5) . BNWL-1667 states that i
i .Hanford IWW waste was a constitutent of the feed to k- give-an entire fission product spectrum (p. 3.19). j Apparently, molybdenum was added to simulate technetium L when raising the concentration to t,he required level.
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behavior of technetium. In'the~WSEP tests the behavior of: rutheniumJ and cerium was considered to represent the'
. behavior'of high' volatility and' low volatility nuclider, respectively.- 'The staff and the- ' utilities indicated r that they-expect the decontamination factor for technetium 1
to be greater'than'that for ruthenium. Technetium, being 1
'less volatile than. ruthenium, less is expected to be released.from the calciner and melter. Also, removal from the off-gas by sorption, condensation, and plate-out is t
expected-to be equal to or greater than that for ruthen-ium.144/ The Sierra Club argues that the staff and industry have no reasonable basis for their estimates that the volatility of technetium in large scale apparatus will be small, because there are no data for technetium, and the volatility data for the elements said to be similar to technetium were obtained with very small samples.- Further-more, if the equipment on which the technetium plates is ultimately disposed of by shallow burial, the Sierra Club' expects that the technetium would migrate to the biosphere in times much ' shorter than the half life.145/ 144/ BG&E/FRA,- 106-107;'NRC/FRA, pp. 5-7, 5-8.
.14 5/ . pohl~ and L Resnikof f, ' Appendix B; Sierra Club Response to. Question 7, pp.' 5-7.
/.a * * - -
, s ( ) V The lack of information on the behavior of technetium is attributed to its small contribution to the radioactivity of the waste, its low radiotoxicity, and the expectation that it would remain in the HLW. Under the conditions foreseen by the staff and industry, the impact of the tech-netium released would be small in comparison with the impacts from tritium and carbon-14. However, if a large fraction of the technetium were released to the biosphere, the short term impacts would be of the same magnitude and the very long term impacts could be much greater. /
- h. Other Nuclides In the staff's model the radiological impacts are dominated by the complete release of tritium, krypton-85 _
and carbon-14. The staff calculated release rates for the other radionuclides present in the spent fuel, but the ! l resulting health effects were found to be less than one ' l percent of those attributable to the three nuclides listed above.147/ The large decontamination factors for the remainder of the radioactive nuclides are attributed to the elab-orate off-gas and plant-ventilating-air cleaning systems
-146/ BG&E/FRA, 107-109; Sierra Club Response to Question 7, pp. 10-11; Tr. 435-436.
~
< '3 147/ NUREG-0216, p. C-29; Tr. 902-903.
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. (F and to the additional decontamination obtained by vapor-izing water from the liquid waste into the atmosphere.
Test data are cited in support of the decontamination factors assumed for some of the systems. ! l The utilities expect the decontamination factors in l NFRRC to be much greater and the releases much less than were estimated by the staff for the model facility. ! The Sierra Club argues that the decontamination factors used by the staff are much too large. In support of their 1 contention, they cited the difficulties experienced with l 1 the air handling and filtration systems at NSF, and the measured release rates for strontium-90 and cesium-137 at l NSF and La Hague. They contend, based on experience at those facilities, that the lower decontamination factors , l l should be used in arriving at release values for Table s-3.150/ 14 8/ NUREG-0216, pp. 3-12 thru 3-15, 3-47; NRC/FRA, pp. 1-96, l 7-119, 7-120. l 149/ BG&E/FRA, pp. 130-132. 150/ Pohl and Resnikoff, Appendix A, pp. 19-32, 113-114; Sierra Club Exhibits.and Responses to Oral Requests by Board, pp. 3-7. l l
v) [ 3. Occupational Exposure A major contention of the Sierra Club is that the staff's estimate of the occupational exposure to workers - at'a mo' del reprocessing facility is much too low. The staff estimates that the exposure will be 22 person rem /. RRY which corresponds to a total of 1250 person rem /yr for the workers. This value was obtained from the GESMO report and is distributed as follows: Facility Person-rem /yr Separations 400 UF conversion 150 6 Pu conversion 400 Waste solidification and packaging 300 1250 The estimate was based on exposure conditions set by design bases, personnel needed to carry out various operations, and the types of activity and associated occupancy time in various work zones. The estimates were for normal operations and include minor incidents. They do not include exposures resulting from activities associated with recover-ing from major accidents or. unexpected equipment failures.151/ l 151/_NUREG-0216, App. I; NUREG-0002, pp. IV E-35 and 36. () 4
f) () The Sierra Club discussed the occupational exposures
! The record, at NFS and how they increased with time.
as taken from testimony by the Sierra Club and the util-ities follows:153/ Fuel Occupational Exposure Processed No. of person rem Year MT Campaigns NFS Outside Total 1968 150 4 723 127 850 1969 136 5 859 109 968 1970 37 3 1098 433 1531 1971 69 9 1172 1194 2366 In view of this experience, the Sierra Club contends that the staff's projection of the occupational exposure is not based on real time studies and bears no relation to the experience at NFS. At the request of the Board the Sierra Club provided information about the occupational exposure at the La Hague plant. A paper by the employees' union at La Hague indi-cates that the individual dose to about 473 persons who worked directly with radiation was 650 mrem (310 person ram /yr) in 1974 when 650 MT of metal fuel was reprocessed.155/ 152/ Pohl and Resnikoff, Appendix A, pp. 48-66. 153/ BG&E Exhibit 1 at Tr. 870; Pohl and Resnikoff, Appendix fN A, pp. 61-66. l t 154/ Sierra Club Position on Environmental Effects of the Uranium Fuel Cycle, pp. 2-3. 155/ Sierra Club Exhibits and Responses to Oral. Requests by Board, pp. 4-5 and Exhibit 4a.
- . . . . , . ~ . . . ~ - .- -. .. . . . - ~ . /)L/ The union projects the individual exposure to rise by a .
1 factor of 12 when larger ~ amounts of more highly irradiated l oxide fuels are reprocessed. l Utilities witness, Dr. Rodger, testified that abnormal events were responsible for the high occupational exposure t at NFS in 1970 and 1971. Numerous small batches of fuel were reprocessed and this required frequent changes of fixtures in the shear. Extensive clean-up of the fuel pool was required as a result of fires in a uranium metal zirconium clad fuel, a type that would not be reprocessed in the model facility. Also, clean-up of parts of the plant was beginning in preparation for making the proposed modifications. Improvements-in the design of cells and equipment, in the shielding, and in provisions for maintenance are expected to reduce the exposures in large commercial reprocessing facilities over those at NFS. Dr. Rodger indicated that occupatione.1 exposures were dependent on several factors and could not be sealed in proportion to the plant capacity. He considered the exposures in 1968 and 1969 to represent a reasonable expectation of what could be achieved in a plant of the vintage of NFS. He l V
C\ V conclude that the improvements incorporated in Barnwell and the design of the NFRRC should make those plants capable of achieving the staff's ectimate of 400 person-rem /yr occupational exposure in the separations facility.156/ s 1
- 156/ Tr. 864-873, BG&E/FRA, pp. 8-12, 95.
U 1 1
=- -
a _ , p ,. , f O , 68 - II. Hich Level Waste Disposal - 'A.-. Staff Tsstimony. -
- 1. General Criteria ;
' Testimony on high level waste (HLW) disposal
_ covered both thefno-recycle.and the uranium-only recycle options. In.the no-recycle option the high level waste is composed of complete fuel elements (which includes the normally' defined high level waste (10 C.F.R. 850 App. F) plus plutonium and other transuranium products). In a uran- l ium-only recycle the uranium is -returned to the fuel cycle leaving plutonium and other long lived radioactive isotopes as waste material. At the end of the normal reprocessing procedure the high level wastes are in liquid form but 10 C;F.R. s50 Appendix F_ requires these wastes to be solidified before shipment. Therefore, for the purposes of the discussion on waste disposal, it is assumed that the waste has been solidified at the reprocessing plant prior to disposal'. See infra p. 44. In its assessment of the environmental impacts of permanent disposal of high level wastes the staff utilized t
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a model based on geologic disposal.157/ (Generally appli- : cable to'either salt bed or hard rock. disposal.) In a
' discussion with the Hearing Board, the staff witnesses pointed out that theLDepartment of Energy (DOE ) is respon-sible forLselecting the method of disposal and the disposal site. Tr. 573-586'. While DOE has made it clear that they presently favor geologic disposal, no' site has yet been selected.- When-such sites are selected they are subject to licensing review by NRC and their in'dividual characteristics must be such that they meet the NRC licensing' criteria. The staff under present organizational responsibilities must not be in the position of prejudging any particular site before DOE presents its application.
- (M1 this basis staff witness Hewitt stated that the staff
~
' site suitability criteria !
are really a series of factors that we think-have to be considered in determining whether. i 1 or not a site is suitable. There aren't any 157/ In answers to questions by the States of Ohio, Wisconsin and New York the staff pointed out that in the past , ELW have been disposed of by various countries in- ! deep ocean' water or inactive wells. This practice is no. longer used. Several countries (e.g., Canada, j Netherlands,nBelgium, France, Germany) have programs l to develop either salt dome or crystalline rock depos- 1 itories.- u e
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exact numerical-type parameters.in the criteria,- 1 for the simple reason that a balance.has to be i smade on each individual site, in order to see- l t that- the: environmental and radiological objec - d
- tives are met. In other words, it wouldn't be possible or it wouldn't be' reasonable to say-that a' salt bed has to be X meters thick, or i so many kilometers away from a body'of water, because those factors are'really determined on ,
how the hydrogeological system acts, what the l gradients are, what the permeabilities are, what the actinide absorption factors are. So'what we have tried to do here is to give,a series of factors that have to be' con - sidered, so we can have reasonable assurance y that the radiological performance objectives , we establish for a repository would be met by 1 the repository. This series of criteria may be summarized as follows:
- a. The repository site must be owned by the United States Government (including all mineral and exploration rights so that the
- site can be isolated from other human activity),
~
There must be "a multiple set of barriers to l b. waste migration." Salt of course would be one-barrier if a salt bed is chosen. The uniformity of geologic formations and the hydrology of the area would be of extreme importance as well as any chemical reactions which would effect isotope migration. > (- I:
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- c. The site should be in a seismically stable I area.
I 1 L d. The likelihood that subterranean construction and the heat produced by the stored materials will cause major disturbances of the over-I burden must be small. This is necessary to assure that migration of waste from the repository into ground water that might be used for drinking or food production would be precluded.
- e. The site should not unnecessarily deprive society of valuable resources.
- f. If several sites are available the one chosen should be based on minimization of risk to the public not only from the site itself but transportation to the site.
l
- g. Preferably the site should be such that long l term monitoring and surveillance is not nec-essary.
These general criteria appear to be compatible with what is known about the present DOE site selection criteria. t Na 1 l i I
./ W j x_e Before selecting its disposal site model the staff considered such alternatives as rocket ejection of the wastes, transmutation of the long life radioactive isotopes j to other nuclei by nuclear bombardment, ice sheet isolation, I seabed isolation and various forms of geologic disposal. j l
There was no serious suggestion by any party that more l than salt bed or hard rock ceological disposal should be considered. I
- 2. The Model HLW Disposal Facility The staff's model HLW disposal facility is dis- l 1
cussed in the staff's initial testimony. NUREG-O ll6, pp. 4-71 to.4-95. It consists of a surface land area l l totaling some 18,000 acres of which the central 2000 acres would be over the actual burial zone. The waste material would be buried at depths of from 1000 to 4000 feet. Such a repository is estimated to be capable of handling the waste from 4000 RRY/30 yr or for the life-time of 133 reference reactors. The surface structures, shaft to the mine level and transportation equipment in the mine could be designed to handle either the solidified material from the uranium reprocessing plants (if such plants are in operation) or
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l packaged. fuel elements directly from reactor or regional l 1 storage pools. The mine itself would use the conventional 1 room and pillar excavation. High level waste material l will be embedded in glass cylinders at the reprocessing l plant and placed in sealed stainless steel canisters for insertion into holes drilled in the floor of the mine. Plutonium will be reduced to powder form before encasement and burial. TRU wastes will normally be place.d in sand or cement filled containers. See Part 2, I.A., infra. l Spent fuel assemblies would also be enclosed in larger sealed canisters, probably helium-filled (for l l better cooling) , which would also be inserted in the floor of the mine. In either case, the placement of the waste is governed by the amount of heat given off by the can-isters. The staff's present placement criteria limit heat loading to a maximum of 150 kw/ acre. NUREG-Oll6, p. 4-114. In its discussion of the salt mine model the staff recognizes that digging a cavern into a salt mine and placing heavy, heat producing materials on and into the salt floor may cause some of the following interactions:
- a. Flow of brine pockets toward the heat source, eventually possibly producing a briny slush L %l t
(') surrounding some of the hot canisters,
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b' . - Gamma l radiation of the brine may produce corrosive chemicals which may add to the brine corrosion of the' stainless steel can-isters and eventually lead to leaching of the" glass in which reprocessing' wastes are buried.
- c. Similar corrosion will occur in the canisters around the fuel assemblies and thus expose j the' cladding of fuel elements and eventually l the fuel pellets to the brine.
.l
- d. Heated salt,tends to become fluid because of c!
brine migration toward the heat source and may result in'the fuel assembly packages } sinking into the floor or tipping over, so that care must be taken in placement to avoid { i any possibility of creating criticality by having several assemblies come together. i l i l The~ staff points out, however, that many of these 1 problems are~ site specific. Some salt mixtures are better than others. Obviously the' lower the brine content of. d 1 the, salt the better. In this connection the staff points
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in the Lyons, Kansas, salt mine as part of Project Salt Vault. In those experiments packaged spent fuel assemblies were successfully emplaced in the mine. See NUREG-Oll6,
- p. 4-113 This experiment lasted about 19 months.
Emplacement sleeves showed no sign of damage at the end of that period. Tr. 572. The staff's calculation of the environmental impacts of the geologic repository was based on the conclusion that radiological releases would occur only during the time that waste was being placed in the repository. The staff assumed that the only significant radioactive releases t from the repository would be in the gaseous effluent. This effluent would contain all the gaseous fission products from the spent fuel and any gases released by other wastes. , i It would also contain small amounts of fission products i and TRO nuclides that are released into the air and pass through the HEPA filters in the ventilation system. i The staff assumed that no radioactivity would be ! I released after the repository was filled and sealed. Although the canisters would corrode and release their 1 s contents to the salt, the salt and the surrounding media would isolate the radioactivity. Studies were made of [ l
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The results of the study showed that for reasonable storage conditions the potential incremeatal radiation doses would be of the Fame order as, or less than, doses from natural sources ...." NUREG-Oll6, pp. 4-80, 4-89, 4-93; Tr. 600-604.
- 3. Variations in Salt Mine Model with Tvoe of Fuel Cycle (No recycle or uranium recycle)
In its testimony the staff discussed in some detail the variations in their model which must be assumed depending o'n whether no recycle or uranium only recycle is utilized. This testimony was amplified in the staff's
- answers to the questions of other parties and the Hearing Board.
In the uranium-only recycle the HLW, TRU wastes and plutonium must be.placed in' the disposal facility. Until ecently it had been assumed that the bulk of the plutonium (98.5 to 99%) would be recycled as useful fuel with only the chemical process losses (1.0 to 1.5%) being included in either the HLW or the TRU wastes.
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Plutonium as a waste can be handled in either of two ways: (a) as part of the HLW or (b) processed one more step and separated from the HLW in the ' arm of highly radioactive oxide power. Under the fir.:' method the plutonium would be a portion of the liquid HLW which would be initially stored in tanks.then solidified in glass and placed in stainless steel canisters for burial. The staff admits that this method has had very little study since the possibility of such disposal had not been considered until recently. Using the second method the plutonium oxide, powder would be encased in heavy stainless steel cylinders for insertion in the mine floor. In either case, the disposal facility must be designed to hold approximately 100 times more plutonium than had previously been considered
'on the basis of the complete reprocessing option. 11RC/FRA Questions p. 1-140.
Under the no-recycle option used fuel assemblies, l after cooling for 10 to 25358/ years in water storage pools, I would be enclosed in large canisters for placement in the i salt mine floor. In response to a question from CERCDC , 1 i the staff listed the following items which must be considered l l ('] 158/ Staff model actually assumes at least a 10-year cool-(/ ing period. NUREG-0116, p. 4-107.
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when burying spent fuel in a salt mine initially designed for solidified HLW.
- a. In the high level wastes most of the uranium, plutonium and fission product gases present in spent fuel are removed during reprocessing;
- b. the volume of spent fuel (per metric ton of heavy metal of reactor fuel) is larger than high-level waste;
- c. the total time integrated heat output of spent fuel (per metric ton of heasy metal) k is larger than that of high level waste;
- d. the leach rate of fission products and acti-nides from spent fuel may be different than high level wastes, and
- e. the possibility of forming a critical assembly in the repository must be considered because of the presence of plutonium and uranium in in spent fuel.
All'of-the above items must be considered in the
-design.-- The staff notes that one important effect is that N a " larger repository area would be required for spent fuel
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than for the high level wastes * *
- because of the increased heat output of the spent fuel and to allow for movement of the different. size waste canisters." Id. at 1-140, 141.
Additionally, due to the longer (15' vs. 10' -- See Tr. 554) canisters required for the spent fuel elements the mine must be higher to accommodate these canisters. In answer to questions by Wisconsin and Ohio, the i staff also pointed out that there will be ten times as many canisters required for spent fuel disposal per RRY than would be required for disposal of HLW after reprocess-ing. Id. at p. 2-24. B. Comments on Staff Model Comments on the staff's model geologic disposal
. facility were received from several of the parties. The most detailed of these comments came from the United States Geologic Survey (USGS) both in written testimony ! and l answers to questions during the proceeding. .
1 l
- 1. The thrust of the USGS initial testimony was that while it was working with the Department of Energy (. DOE) to 1
establish detailed geologic criteria for repository sites 1 it questions "whether in_the absence of final criteria a l 1 (~N ' (_) 159/-Statement of George D. DeBuchananne, Chief, Office of Radiohydrology, L ter Resources Division, U. S. Geological Survey, October.3, 1977. ,! l 1 1
I is definitive assessment of such (environmental) impacts can be made." In their oral testimony, USGS witnesses George D. DeBuchananne and Noel J. Trask make it clear that their primary concerns are (1) the continuing stability of the salt bed after intrusion by man and (2) placement of heat emitting wastes in the salt. Tr. 702-708. With respect to man's intrusion into the salt mine they noted that the Lyons, Kansas mine had been abandoned as a long tarm repository because its integrity was doubt-ful due to an unknown number of bore holes which earlier had been drilled into it in scarch of oil. Tr. 704-05. Even with a controlled number of shafts they felt that sealing of such openings against water intrusion for a long period of time was an unknown technique. This may be particularly true for access shafts large enough for the elevators to place equipment and waste canisters into the mine. The most serious problem foreseen by USGS was the
- stability of the salt floor and the sait backfill around the canisters. They stated that brine in the salt would tend to migrate "from a relative small area, 30 centimeters perhaps" toward the source of heat. (Tr. 724). It would rN U
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1 i O g therefore be necessary tofhave a low level of brine occlu-
'sionssin the salt lto'begin with. However, long term venti-lation of.the mine might supply water.to the salt (the salt acting'as a,dessicant) producing a continuous supply of brine. Tr. 724. Although the. witness did not believe that any weakening of the salt support columns would occur 1
(Tr. 726), they envisioned the possibility of weakening ] salt around the canisters-which could allow them to shift , or tip over. In any' case, the witness concluded that no problem should occur "if the temperature of the waste.were 100*C cur. less" above the ambient temperature of the mine. , He suggested that this be achieved by storing the waste above groun'd for 30 or'40 years. Tr. 732-734. In connection with the heating problem in salt , Dr. DeBuchananne mentioned the use of anhydrite beds.160/ j In this. material, according to'the witness, brine tends to migrate away from a heat source and would therefore not ; produce the weakening effect around the canisters that j occurs in salt beds. He noted that at the New Mexico site { which USGS ~is investigating.for DOE,'there appears to be L 1 a few hundred feet of anhydrite on top of the salt bed i i 160/ Anhydrites (CaSO 4 ) is of ten deposited on top of normal j salt beds to-depths of up to a few hundred feet. l
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which is several thousand feet in depth. Tr. 711-12. Dr. DeBuchananne emphasized that use of such beds is a recent concept and there may be other stability problems such as brittle cracking that may prohibit their use. The USGS witnesses were also concerned that the ambient mine temperature might rise enough to cause suffi-cient expansion of the overlying rock to crack it. This possibility is being looked at by other DOE contractors, but not by USGS. 61/ The USGS investigation is still underway. No specific site has yet been investigated sufficiently to be sure that the witnesses questions are answered. They also recognize that their research is concerned only with the salt and that the waste containment canisters, the rock formations surrounding the salt mine and the hydrology of those forma-tions form a multiple barrier situation independent of the salt mine itself. Tr. 171. Despite their "research" concerns the USGS witnesses, in answer to Hearing Board questions, stated that "we would l agree that the concept of geologic disposal is a valid I l concept" (Tr. 728) and that: finding repositories that will I l IT ,g,/ This issue was admittedly outside the witnesses' l
's_/ '
expertise. l l l
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give the low release rates estimated by the staff is-possible. In response to Board questions the staff stated that during the 19 month Lyon experiments (Project Salt Vaultl the heat density in Project Salt Vault was higher than one would anticipate for a half i metric ton of spent fuel irradiated to 33,000 megawatt days * * *. So Project Salt Vault densities were more than would be -- than to be expected with spent fuel disposal, somewhat lower than that anticipated for high level waste disposal, but in the same ball park, certainly transferable data." Tr. 571-572. The fuel elements in the Salt Vault experiment were stored for some 18 months and were retrievable at the end of the period with no apparent damage to either the can-ister or the salt bed surrounding the containment hole. In other words there was no sign of brine intrusion.
- 2. In general the comments of CERCDC, Sierra Club, NRDC and Wisconsin and Ohio empnasize the uncertainties of the staff model concerning the actual release of radio-active isotopes from the salt mine disposal site. However, as discussed below, witnesses had varying points of view 1
on the staff's model. l
'l l I
. t p
h a. CERCDC witness Anderson 162/ believes that the staff model should have included an analysis of other geologic disposal methods than the salt beds and that the results of the disposal method with the most severe impact should have been used for conservatism. Additionally, the witness believes that it is a serious error to assume that plutonium must be disposed of irretrievably. At the very least, in his opinion, another analysis should have been made on the assumption that plutonium might be needed in the future. Similarly, Mr. Anderson believes that the staff should have considered "the option of storing spent fuel in a retrievable manner, so that at a later date, if policy or c'conomic considerations changed, it could be reprocessed for its uranium and/or plutonium content. If necessary such temporary retrievable storage can be accom-plished at a spent unreprocessed fuel f acility (SURFF) which is being considered by DOE.
- b. On the other hand, witness David A. Deese 6!
for the 2tates of Ohio and Wisconsin referred to the tech-nical problems encountered at the West German ASSE salt 16%/ Statement of Robert N. Anderson on Behalf of the California Energy Resources Conservation and Develop-ment Commission, Oct. 3, 1977, pp. 27-32. () l 163/ Statement of David A. Deese on behalf of the States of Wisconsin and Ohio, Oct. 3, 1977, pp. 13-17.
V('3 mine repository. These were mainly brine seepage into the
'mine and creep of the salt dome. He believes that there is insufficient test data to conclude that the Carlsbad, New Mexico, mine will be completely satisfactory and, like the USGS, is concerned about the effects of human intrusion, brine and heat on the salt.
- c. Sierra Club witness Marvin Resnikoff l64/ does not agree with the staff and industry that reprocessing will reduce the volume of waste material that muut be managed. He claims that the addition of chemicals in reprocessing will increase volume. It is unclear in his statement just how much of this additional volume is low level waste. While the witness agrees that heat output is the largest problem for the disposal, his analysis treats
'no-recycle and plutonium-plus-uranium recycle but does not treat the situation of uranium-only recycle that is of interest in this proceeding. Id. Appendix C, Tables 1, 2, 3.
- d. While NRDC did not present direct evidence on the salt bed disposal their concluding statement appears to indicate agreement with the USGS concerns.
l l 1 164 / Appendix C to Statement of Marvin Resnikof f , Sep. 30, , j 1977. l (
, )
(q 3. Testimony by Baltimore Gas and Electric, Common-wealth Edison, et al., Tennessee Valley Authority and Pacific Legal Foundation all expressed general approval of the staff's geologic disposal model.165/ All expressed the opinion that the staff's assumptions were overly con-servative. In particular, it was noted that geologic disposal creates a series of barriers to the escape of radioactive isotopes that are underestimated by the staff. In summary,the following points were made by witnesses for these parties concerning the staff's over-conservatism.
- a. For spent fuel buried in stainless steel canisters, the staff assumed the integrity of the canisters, fuel elements and fuel pellets would only last long enoug.i to back fill the areas of the mine in which they were located. This means that the stainless steel canisters, fuel cladding and pellets would corrode in less than 25 years and under the staff assumption would release all the gaseous radioactivity to the atmosphere before the mine is sealed. See p. 75 infra.
- b. For the HLW enclosed in glass, the staff l I
f ' assumed that the stainless steel canisters would corrode l l-("N- 165/ See Initial Testimonies presented by BG&E, et al.,
%- Commonwealth Edison, et al., TVA; PLF. i l
l
L . . 9 ( rapidly and the glass would be leached away with a half-life of approximately 200 years. Witnesses for PLF mentioned several experiments which indicate that the glass cylinders have a half-life of at least one million years (Tr . 767-769). The staff itself in answer to a question raised by NRDC stated: All materials have some solubility or leach-ability in water. The rate of leaching of a glass is dependent on many things, including: glass composition, particle state, temperature of the water, pH, chemical contents, flow rate of water, length of recction time, and element measured. With the wide range possible, waste glasses can have reported leach rates from 10-3 to 10-8 grams /cm2 / day. . Since leaching is a surface area dependent value, large scale tests are not normally used to determine leach rates. However, one semi-large test has been conducted in Canada with blocks of HLW in ground waters. Leach rates after 15 years burial are continuing to decrease and are about 105 times lower than leach rates determined in the laboratory. NRC/FRA, p. 7-93.
- c. The staff and USGS assumed that any water leakage into the mine would result in instantaneous removal of radioactive material from the mine. Witnesses for the utilities and PLF note that that assumption ignores the tremendous volumes of salt involved and points out that the staff itself stated in answer to a Board question:
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88 - l 7 Q. Well, the USGS says that one has to assume I that the salt mine is.' going to get flooded at some time. Other people have expressed concerns about water getting access to the salt and dis-solving out fission products or dissolving out plutonium. So one would like to know, are there places in nature where salt deposits have existed and where water has gotten access to the salt deposits and dissolved out the material, or are there not situations like this? A. (Witness Bishop) There certainly are such situations. If we can use a concrete example, in southeast New Mexico, it is quite well known that the salt fronts at the edges of that major salt bed in the Permian Basin are being dissolved away at a rate which, although it is not in my head, it is in a number of reports, from Sandia Laboratory and the old Oak Ridge National Labora-tory reports -- I am reminded that the units are sort of a mile or so per million years. Tr': 610-11. PLF witness, Dr. Cohen noted that 1
* *
- if all the ground water.now flowing through I all the aquifers above the salt formation (at the l New Mexico site) were diverted to flow through I the salt formation, it would take 50,000 years l to dissolve away the salt enclosing one year's i waste. l Tr. 819. l
- 4. The staff has assumed that once out of the salt mine itself the radioactive isotopes will be transported through the surrounding media by underground water.
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( Witnesses for PLF called attention to the OKLO phenomenon in Africa to show that heavy elements are not transported long distances by water. The OKLO phenomenon was discovered in a uranium mine near Gabon, Africa, where a natural chain reaction was found to have occurred approximately 1.8 billion years ago. The reaction required a unique set of circumstances to supply the right combination of uranium isotopes and water and it apparently lasted somewhat intermittently for thousands of years. The reaction on the basis of tests at the mine generated several tons of fission products and plutonium. Tests at the mine indicate that the fission products moved very little and none of the plutonium moved any detectable amount. PLF Initial Statement by Marc'W. Goldsmith, October 3, 1977 PLF makes the point that even if the fission products and plutonium are washed out of the salt mine the OKLO i data indicates their movement in the geologic formations around the salt bed would be minimal even with high water flow. / Tr. 772-774. l l i 166/ The State of Wisconsin and NRDC question the applica-bility of the OKLO phenomenon to waste disposal in salt formations. e
/'N V III. Low Level Waste Disposal A. Staff Testimony In its testimony (NUREG-0116, pp. 4-117 to 4-128) the staff defines the term low level waste as all solid radioactive wastes !other than high level and transuranic (TRU) wastes. Such wastes are large in volume but contain a relatively small portion of the total radioactivity. The principal radioactive isotopes involved have maximum half lives of approximately 30 years. %8/ They may require isolation for a maximum of a few hundred years. Under the staff's present criteria low level wastes will include no more than 10 nanocuries of TRU wastes per gram (NUREG-0 216 ,
- p. H-16).
l The staff's testimony, NUREG-Oll6, pp. 4-118, concludes that the preferable method of disposal for these radioactive wastes would be in shallow burial grounds. It bases its calculations of the environmental effects per reference reactor year (RRY) on the assumption that any burial ground 167/ The term solid wastes here includes wet solid wastes
~~
from the treatment of the various process streams, ) non-compactable small items such as tools and com-pactable' solid items such as rags, clothing and filters. 168/ This maximum half life does not include trace amounts l
/~ of such long life nuclides such as iodine-129, nickel-5 ^ ,
i technetium-99 and plutonium-239, which ray occur in scre !- ~ of~the LLW wastes.
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k_ , would meet a proposed set of criteria. The staff's major interim criteria for a satisfactory burial ground may be briefly summarized as follows:
- 1. The model facility would be located in a rural, sparse 7.y settled area, with annual precipitation of about 40 inches and low seismic activity.
- 2. A central portion of the area (about 100 acres) would be used for waste handling and burial. This area would be fenced to prevent human and wildlife access.
- 3. A surrounding buffer area would be required to provide physical isolation.
- 4. The site must be an upland area with no permanent surface water onsite, no nearby use of groundwater down-stream from the site an'd no major zones of perched water between the surface and shallowest regional water table.
- 5. Surface drainage must be good with no potential for abnormal erosion.
- 6. The regional water table must be about 50 feet below the surface with minor annual fluctuations.
- 7. The unconsolidated material of the burial site I
should have an adsorptive capacity such that most
.(
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/3f i v radionuclides which might be leached from the waste by percolating water will move slower than the water.
- 8. Trenches should be excavated to about 30 feet in ,
unconsolidated silty soil, i e., excavated to no nearer than 20 feet above groundwater. Trenches are'to be filled in such a way as to allow at least five feet of uncontam-inated backfill above the radioactive material. l l l Using these interim criteria and reasonable manage-ment procedures, the staff has' estimated the radioactive releases.from their model low-level waste burial sire. The staff admits that the composite characteristics of the , model site do not directly match any of the existing low-level waste burial sites and that some of those sites have had unexpected releases of radioactivity. The staff notes l (NUREG-0216, p. H-14) that there have been no releases from the low-level waste burial sites that presented a hazard to public health.169/ It is the staff's position that many of the problems with the present sites have been due to improper management. 1 1 Furthermore, the knowledge gained from the problems that I 169/ This appears to be acknowledged by NRDC witness Lash in his reply to a question (from BG&E) on the effect l of releases at Maxey Flats. l l m
n . . . . . . - . have occurred in the past has been included in developing the' interim' criteria. Tr. 382-385; 418-20. In other words,? future sites and future. operation._of present sites should be superior to what has occurred in the past. B. Testimony of Other Parties
- 1. The U. S. Geological Survey _(USGS) has been working on a.
Congressionally approved program for the last three years n . investigating low-leval~ waste.~ disposal. This progran has approximately another two years'to completion. Dr. DeBuchananne,171/ witness.for USGS, discussed the type of 7 criteria which he felt are necessary for LLW sites in his I; answers to NRDC questions and to the Hearing Board's follow-up questions. USGS Responses to Questions from NRDC and Tr. 738 to 760. These criteria are essentially the same as those listed by the staff for its model site. Dr. DeBuchananne noted that the present Beatty site and probably the Barnwell site meet these criteria, but
'170/ In answer to' questions from the Hearing Board the staff pointed out that a plan to develop regulations on low
~
~
-leve1 waste burial has be'en published.as the NRC Low Level Waste Management Program - NUREG-240. This -docu- -
ment contains-a' statement _ofithe NRC program aimed at-development of the desired site and management regula-Ltions.
;q. .171/ Dr. George DeBuchananne, Chief, Office of Radiohydrology, Q Water' Resources. Division, U.LS. Geological Survey.
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()/ - 94 - West Valley and Maxey Flats do not. However, only the Beatty site has been fully studied. The Barnwell studies are nearing completion. He concluded that suitable burial sites can be found for various isotopes, but no one site can be considered ideal for all isotopes because of the wide range of half-lives and toxicities involved. Tr. 750,
- 2. The Sierra Club, NRDC, and CERCDC al} object to the staff's model site as unduly optimistic in view of past experiences at some of the existing sites (partic-ularly Maxey Flats and West Valley). They clain that there is no proof that radioactivity found outside some of the burial grounds has come from surface spills or improper use of concentrators at the site. They believe that some of the radioactivity could have come from through-ground transmission which would indicate a greater permeability of the soil to radioactive isotopes than assumed by the staff. Sierra Club witness Resnikoff discussed the West Valley burial problems in detail.
- 3. On the other hand, witnesses for BG&E and PLF opined that the staff's assumptions and resulting interim i 172/ Testimony of Marvin Resnikof f . This testimony was
-~
first given as Sierra Club Testimony Related to i
,s Section IV.E. Reprocessing - GESMO 1. It was placed l l ( ') in evidence in the proceeding. The remarks on low N/ level burial are in Section II.B.1,.p. 17; II.B.3, pp. 30-33.
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(_)- i criteria for their model low-level waste storage are overly conservative. In their initial testimony on presently existing LLW sites.BG&E witnesses claim that only Maxey Flats and West Valley have shown any radioactivity off-site.173/. These witnesses believe the problem at both 1 locations was one of management, not of technical feasi-
.bility. BG&E testimony references experiments by New York' State and the operator of West Valley to show that the radioactive migration was'due to surface water filling the trenches through the overburden then carrying some y radioactivity across the run-off surfaces beyond the site limits. Maxey Flats' build-up of radioactivity at the site edge appears to have been caused by material which was spilled at the time of receipt, then carried across l the surface, and to the' emission of tritium by operation of an evaporator on the site. This conclusion is supported by measurements which show that the radioactivity levels i
offsite are dropping with time. No e'vidence was found of l a subsurface migration at either site.174/ l The Pacific Legal Foundation's witness Cohen (Initial Test. p. 2 and Attachment C) pointed out that, even under q 17?/ West Valley is no longer operating and Maxey Flats is running on a limited basis.
~
.,O 174/ BG&E Initial Test.,pp. VIII-9 to VIII-18.
r . .
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( conservative assumptions, the release and distribution through the soil of the LLW waste at Maxey Flats -- which the exception of I-129 and possibly Pu,239 -- would have an insignificant effect on human safety. He recommends
-that strict limits be placed on the amount of I-129 which may be placed in shallow burial.
However BG&E witnesses maintain the I-129 cannot be a health hazard becauce the maximum amount of the isotope that can be expected to reach the low level waste depository will be no more than 0.1% of the total I-129 produced. On this basis they calculate that a maximum of 2.5 curies would be handled at any low level disposal site. Because 7 of the long half-life (1.6 x 10 yrs) and low concentra-tion, they conclude that I-129 could not result in unacceptable exposures to either individuals or populations. BG&E Initial Testimony, pp. VIII-25, 26. They further conclude that the TRU limitation could be as much as 500 nanocuries/ gram -- in contrast to the staff proposed limit of 10 nanocuries/ gram -- for either humid or dry burial sites. I.
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(I IV. Interim Storage A. Spent Fuel'
- l. Staff Testimony The. staff's testimony assumed a model system in which spent. fuel would be sent without inte*rmediate storage to a reprocessing facility after 160 days.in the reactor ;
storage pool. This assumption allows for cooling of the spent fuel elements while nuclides with short half lives, iodine in particular, decay. .However, it transfers the major problems of containing radioactive materials to tia reprocessing plant (Tr. 169). According to the staff this procedure tends to maximize the environmental effects of reprocessing, therefore, it results in a conservative value for such impacts. For the throwaway (i.e. , no reprocessing) cycle the . i staff assumes that spent fuel will be stored in water 4 basins for 10 years. It will then be placed in canisters and shipped to a Federal repository for disposal. The environmental impact of the normal operation of an interim ! storage' facility is given in Table 4.25 on p. 4-110 of ; 1 NURE G-0116. The staff assumes that the packaging operation -l l I will add an impact about'. equal to.that of the extended storage.- According to the staff, discharged fuel is stored JO.
Le ^ Q ,]' mostly in water basins'at the reactor sites. However, additional facilities will be required, and a generic study by the staff of the environmental impacts of an off-site. interim storage facility is still in process.
- 2. Comments on Staff Assumption of No Interim Storage before Reprocessing Several of the parties criticized the staff's assum-ption of short cooling before reprocessing. All of them stated that in present circumstances some intermediate storage facilities must be utilized. However, none of them appeared to question the technical feasibility of away-fron-reactor interim pool storage.
- a. NRDC comments.
In answer to NRDC questions concerning the availability of fuel cycle facilities the staff stated that (a) "[R]epro-175/ ; cessing is now indefinitely deferred", ~~ (b) "Therefore, I we speculate, a high level waste repository will be 'com-176/ j mercially' in operation in 1985 or a few years thereafter", ~~ ) and (c) " DOE plans to place waste in a repository only after 177/
~~
it ' cools' for 5 years". 1 l 175' URC Responses to Questions Submitted by NRDC, Dec.
~~
2, 1977 at p. 7-19. 17 9 Id . U,,- l l 7' Ibid. at_p. 7-87a.
yg i ,
) b. CERCDC Comments.
178/ The CERCDC witness -- on this subject stated I believe that it is error to assume that spent fuel will be reprocessed within 160 days of discharge from the reactor and that this error wi11' result in a misstatement of'the environmental impacts from spent fuel reprocessing. The witness points out that reprocessing after longer cooling will reduce the impacts of reprocessing but will add-to the radioactivity released in storage. The witness strongly objects to " conservatism which nistates one rela-tively certain fact (here, that spent fuel will be stored for a long time) in order to be conservative about other impacts.
- c. Utility comments.
The utility groups (Commonwealth Edison, et al., Baltimore Gas and Electric Co., et al. and the Tennessee Valley Authority) commented on the need for interim storage 179/ in a combined statement See pp. VII-4 to VII-12. 1 They agree'.that the original plans for spent fuel manage- l 1 ment called for holding spent fuel at the reactor site for only a period of months. However, " [t]he residence l time of spent fuel in reactor site storage pools is now l \ measured in years rather than months. Ibid., p. VII-6. i
,f h i '
t' ,/ 17S/ Statement of Robert N. Anderson on Behalf of the
~~
California Energy Resources Conservation and Devel-cpment Commission, Oct. 3, 1977 at p. 20-22. ly Combined Initial Statement dated October 3, 1977.
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- 100 -
)
1 , Although during the past two years the NRC has received 1 applications for licenses to amend storage pools at 38 reactors, by 1982 at least four will lose their reload discharge capacity. While DOE is still considering alternatives to away-from-reactor unreprocessed spent fuel l i facilities ( SURFF 's) the most recent goal for completion 180/ of any such facility is placed at 1985. The utilities indicate that while underwater or in-air storage facilities appear feasible, more is known about underwater storage. They also state: " [C] orrosion rates of both stainless steel and zirconium alloys have 181/ been extremely low in experience to date". ,
- d. Pacific Legal Foundation (PLF) comments.
182/ PLF's witness considered the staff's choice of 150 (160) days for ia-pool storage one of the many conserv-atisms in its consideration of the environmental impacts. He asserted that "A more realistic way of evaluating the 180/ NRC points out in its answers to NRDC questions at p. 7-87a that by 1980 there will be 23,700 fuel assemblies accumulated, and some 58,000 assemblies discharged from reactors by 1985. 18f Combined Statement of October 3,1977 at p. VII-9. 187 Statement of Marc W. Goldsmith Prepared for Pacific
~~
Legal Foundation, Oct. 3, 1977 at p. 2. p
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44 - 101 - w/ Tl effect of cooling times and impacts might be a step function acknowledging the current situation" ( i_ . e . , long time interim storage). He considered that this conservative estimate, added to other "much larger philosophical" conservative estimates, made for an overly conservative rule. .i l l O. 1
s .
- 102 -
B. Mich Level Waste The staff's model provides for storage of HLW, as liquid and solid, at the reprocessing facility for ten years (see suora, pp. 42, 44). If a geologic repository is available, the HLW, after ten years decay, would be shipped directly to the repository for disposal. However, as a contingency measure, the s'taf f has assumed that the HLW would have to be stored for up to twenty years at a . Retrievable Surface Storage Facility (RSSF) until a facil-ity for geologic disposal is available. The sealed storage cask concept (SSCC) was selected for storing the canisters of waste at the RSSF cn the basis of a recommendation by a special panel of the National Academy of Sciences. In the SSCC, each canister of waste is contained in a massive concrete cask on a concrete pad. Cooling is by natural convection of air through the cask. In its analysis of the impacts of operation of the RSSF, the staff assumed that the waste would be in the form of calcine rather than glass. Releases of radioactivity would be greater from 183/ calcine than from glass. -- Questions concerning the impacts of alternatives to the SSCC and of longer interim 18 7 MUREG-0116, pp. 4-29 thru 4-37. O
103 ~ storage of HLW were asked by participants and answered by the staff. "o issues were raised concerning the interim 184/
~
storage at the RSSF. r}b 1 i i 1_8_4' NUREG-0 216, p. 3-62; NRC/FRA, pp. 1-114 thru l-116. O. 9 1
. . ~ . . ~ - - . _ . . . . . . . - - . - . . . - - - ~ - - -.
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C. Plutonium
' Provision is made for storage of plutonium at the reprocessing facility for up to five years. If a geologic repository.is available, plutonium, treated as a waste, would be shipped directly to the repository. Assuming i
that ar. geologic repository might not be available when needed, .the staff's model includes an additional twenty 185/ years of storage cefore geologic disposal. ~~ The interim storage facility selected, stores the containers of plu-tonium oxide in pressure vessels in stainless steel lined holes in a large, thick concrete slab. Cooling is by induced flow of air over the pressure vessels. The facility had not been specified when NUREG-Oll6 was prepared and
~ the impacts a re not included in Table S-3. Based on exper-ience with plutonium the staff judged that the special
,- from disposal of separated plutonium would be sma: lj[f Staff Responses to Questions on Economic Da' 18 F NUREG-Oll6, pp. 2-18, 4-102 thru 4-104; Staf. on Economic Data to Support the Feasibility c.
.Model, Enclosure 1; Capital Cost Estimate and 1 tion Impacts - Plutonium Storage Reference Facj 32-79c ' Interim Plutonium Oxide Stcrage 240 Met. t.
Tons Independent prepared by Bechtel Incorporated. Draft report provided by staff in support of cost analysis.
' /N V
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- LJ V. Transportation The staff described the transportation methods and environmental impacts in section 4.9 of NURIG-Oll6. Its
. consideration did not cover the impacts of transpor'ation c of fuel to and from a reactor 187/ since they are included in Table S-4 of 10 CFR 51.20. Transportation in the rest of the fuel cycle was based in large part on track ship-ment with some rail and barge usage.
The Staff points out that shipments of radioactive wastes related to the nuclear' fuel cycle move in routine commerce and on conventional transportation equipment, although normally on an exclusive-use basis. Primary - reliance for safety in transport is placed on packaging. The packagisg must meet regulatory standards established by F the U. S. Department of Transportatio- the NRC ar.d the states. .slthough HLLW has not been shipped, the solidified HLW in canisters would be shipped in casks similar to those j used for shipping spent fuel. NRDC and the State of Delaware asked some questions about the assumptions made in the staff's consideration 197/ The impacts of transportation of spent fuel are des-
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cribed in Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants, c WASH 1238, December, 1972. 1-u a
't
- 106 -
it. . of.'the environmental impacts of transportation, including the possibility of accidents and sabotage, but no party presented any serious challenge to the staff's discussion. However, the question did arise as to the' availability of sufficient shipping containers for both spent fuel and reprocessed material. 6 i 1 { i r~s . N
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1
/- - 107 - i VI. Accidents'
. 1 The staff testimony.(NUREG-Oll6, e.g., pp. 4-51, 4-52, '
4-102 to 4-104) and its answers to comments (NUREG-0216, j i e.g., 3-94, 3-95) and questions (NRC/FRA, e.g., pp. 1-124 thru l-126, 7-39 thru 7-41) discuss various types of acci-dents. Covered particularly are those which might occur during reprocessing and the handling and transporting of radioactive materials. In answer to a comment on the accidents reviewed in NUREG-Oll6, the staff stated that "not all accident scenarios are calculated or. evaluated in the available literature. Accidents that have been calculated range frc= very low-consequence to very high-consequence events. * ** In most cases, the one calculated was an event which was judged to be both of reasonable probability and of signifi-cant consequences." (NUREG-0216, p. 95). In following up on the questionc en this subject pro-posed by the participants, the Board questioned staff witnesses at some length on several types of accidents (Tr. 316-32; 341-42; 365-77; 452-54; 654-660):
- a. Accidents which might destroy the effectiveness 4 l of the reprocessing plant air filtering systers; l
- b. Accidents involving the spray calciner cr waste ;
solidification system;
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- c. Situations in the reprocessing which.might cause criticality events;
- d. Frequency of accidents in past operation of the military plants, e.g., Hanford;
- e. Possible movement in a salt mine of high level waste or spent fuel canisters that night produce criticality.
The Staff witness testified that the frequency of cccurrence is important in determining the overall impacts from accidents as compared with the impact of normal opera-tion. According to the testimony, a study was made of the impact of accidents that might occur in eight reprocessing plarts during the 40 year terms of their licenses. The consequence.of each accident was multiplied by the probability to get a risk figure that was compared with the normal releases. The witness stated that the conclusion was that the impact from accidents would be on the order of 1 or 2 pe: cent of the impact from normal operation. (Tr. 329,332). In other testimony the Staff witnesses asserted that ELU or spent fuel canisters vould have to be spaced so-widely in a salt mine repository, in order to comply with heat load ' 1 imitations, that movement of the canisters sufficient to l ,, produce a critical assembly would be extremely unlikely. p
- 109 -
l,b x/ . Mitne. eses'for the utilities testified to the conserva- t tism of the Staff's accident analysis for the nodel reproces-sing facilit.y. Doses calculated for a variety of accidents by use of assumptions that the witnesses believed to be r conservative'were shown to be much' lower than doses calculated by use of the staff's assumptions. (Tr. 295-898; BG&E/Ih'S Table II) . No party raised issues concerning accidents in its final stateraent. 1 1 J l i ! l l l ' /"s . ( sms
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L . , f#% i, j - 110 - VII. Sabotage The. staff discusses various aspects of sabotage in section 4.10 of UUREG-0016. It points out that the threat of sabotage can be related to either the release of radio-active materials or diversion of fissionable material. ' Motivation can range from political terrorism to a dis-gruntled employee. However, while various hoaxes, threats of violence, break-ins and arson have occurred at domestic reactor' sites, no bombings or seizures have been attempted. Nevertheless, the staff points out that the threat of sabotage must be assumed and guarded against. A. Three primary points are made by the staff:
- 1. A report by Sandia Corporation entitled " Safety and Security at Nuclear Power Reactors to Acts of Sabotage" (Sand 75-954, Albuquerque, January 1976) describes
" ... characteristics of commercial nuclear. power plants which greatly increase the difficulty of releasing radioactivity by sabotage:
The ' defense-in-depth' concept of reactor plant designs;
- The massive structure of the plant, which protects critical components from external attack;
- The' safety design basis of the plant, which i L emphasizes system reliability, flexibility, redundancy, and protection against common-mode failures; and
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- Engineered safety features, which are added i to the basic system to cope with abnormal j operations or accidents." l NUREG-Oll6, p. 4-158.
2.- Systems designed to limit the consequences of l accidents at every step of the fuel cycle also tend to l
.red uce the consequences of sabotage. l 1
- 3. In addition to the built-in safeguards, "regu-lations have been implemented.to prescribe a range of physical security measures that a licensee must follow.
The intent of the present regulations is for the licensee to. safeguard against the acts of a single insider and attc.cks by a small armed group." Id., pp. 4-155, 4-156. B. Questions by NRDC on the sabotage discussion were
.a n s w e r e d b y t h e s t a f f ,1 8 W and follow-up discussions between the staff and the Hearing Board ware held during the hear-ings (Tr. 333 to 357, 418-240 and 459-480).
In answer to questions from the Board, staff witness Page stated: f Essentially the elements for protecting reactorc against radiological sabotage are the same as
- protecting a reprocessing plant against radio-logical sabotage er theft of plutonium.
18W Staff Response 'to Questions submitted by NRDC, ;~ December-2,.1977 at pp. 162-190 and 202-207. i. 4 e
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;,-sg I V - 112 -
'Now there has been during the past year upgrading actions by NRC to elevate the level of protection for power reactors.
So I would say that the level of protection that is being applied under these new regulations, the so-called 73.55 amendments is a substantial upgrade from what existed before this time. Tr. 472. NRDC offered no comments on this subject in its Final Statement. C. Pacific Legal Foundation was the only other party that commented on sabotage. The Foundation's witness Goldsmith in his initial statement 0- L supported the staff's view that while there is little quantification of the potential risk due to sabotage at a nuclear facility, (t]he consequences of an act of sabotage appear to be small and the probability or likelihood i of the act occurring are also small.190/ l
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'189/ Statement Prepared for the Pacific Legal Foundation l lL by. Marc W. Goldsmith, October 3, 1977 at p. 5.
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VIII. Decommissioning There has been scant experience with decommissioning commercial fuel cycle. facilities. Most of the experience has come from decommissioning several demonstration reactors and small nuclear power plants and a large number of research reactors and criticality facilities. Studies have been completed or are in progress on decommissioning lar;e nuclear power plants and reprocessing f acilities.19Y The staff's estimate of the impact of decommissioning fuel cycle facilities was based on decontaminating the facilities to unrestricted use levels shortly after shut-down. This option was chosen because it would result in the most conservative evaluation of radiological impacts. Radiation levels would be higher and the amount of radio-active materials disposed of would be greater than if the decontamination were delayed for many years. During the exchange of questions and answers, some participants questioned the feasibility of immediate 19L/ NUREG-Oll6, pp. 4-129 thru 4-143; An Engineering Eval-u5 tion of Nuclear Power Reactor Decommissioning Alternatives, AIF/NESP-009; Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station, NUREG/CR-0130 (also referred to l as NUREG-0395 Working Paper); Technology, Safety and l ('N Costs of Decommissioning a Reference Nuclear Fuel [. () - Reprocessing Plant, NUREG-0278. F
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- 114 -
u decontamination. Although they did not disagree with the , i staff's conclusion that immediate decontamination would l have the greatest radiological impact, they questioned the staff's ability to estimate the impacts in view of scant experience with large-scale decontamination. Sev- I eral participants questioned the realism of the staff's model since entombment or mothballing were the methods chosen for decommissioning large nuclear power plants in j l current licensing proceedings. Several participants con- i tended that the staff had substantially underestimated the volume of radioactive waste from decommissioning.19F-- The staff had assumed that the reactor vessel and ) internal parts could be disposed of by shallow burial. i But testimony by the Sierra-Club indicated that disposing of the reactor parts could be affected by niobium-94.1$$/ . Niobium-94 has a half-life of 20,000 years and emits gamma rays during decay. The staff did not include niobium-94 in its analysis. The staff agreed that the-presence of the niobium-94 and long-lived nickel-59 and -63 might make it necessary to dispose of some reactor parts in the Federal repository.199-l 1 l 73 l_9 7 NUREG-0216, pp. 3-33 thru 3-43; NRC/FRA, 1-197 thru j w- ) 1-203i.Tr. 1224-1227.
, l[f Pohl- and Resnikof f, pp. 3-4, Appendix D. '
lj9;f NRC/FRA, pp. 5-9, 511-; NUREG-0216, pp. 3-36, 3-37.
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- 115 -
(v) After the rounds of questions and responses, the principal issue that remained was the cost of decommission-ing a nuclear power plant. In NUREG-0116 the staff indi-cated that the cost of immediately decontaminating and dismantling a reactor or reprocessing plant has been esti-mated to be 10 to 20 percent of the original facility cost or $80 to S160 million for an $800 million plant.195/ -- In its testimony on economic feasibility, the staff used
$30 million, based on a study ~for the Atomic Industrial Forum (the AIF study) obi' The staff had the AIF study reviewed by Battelle Pacific Northwest Laboratories (BPNL),
where a study of the decommissioning of the light water reactors was in progress. Representatives of BPNL generally concurred in the AIF study and cost estimates.02! State.of New York witness Skinner contended that specific costs in the AIF study had been grossly underesti-mated [bbE He estimated a total uninflated dismantling 19 5' NUREG-Oll6,, p. 4-131; NRC/FRA, p. 1-197. 196/ Staf f Testimony on Economic Data to Support the Feasi-
-- bility of the S-3 Model, Enclosure 4, p. 2.
I 197/ Staf f - Responses to Questions on Economic Data, pp. 1-29 thru 1-32; Tr. 1164-117 8 ; NUREG-03 95 Working Paper l ! (since published as NUREG/CR-0130) a copy of which was placed in the public document room by the staff. i , 19 8/' ' StatementL of Peter N. Skinner, P.E., on behalf of the
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.( T State of New York, pp. 28-39; Answers of Peter N. Skinner
! \%l to Suggested Questions.of Full Participants on Behalf (FOOTNOTE CONTINUED ON NEXT PAGE) . O e
i l r l U/ - 116 - l cost'of about 25 percent of the original facility cost or I about $300 million for a $1.2 billion nuclear power plant. This estimate was not based on a detailed analysis of the dismantling operations but on a comparison of the costs of dismantling the Elk River Reactor and the Sodium Reactor t Experiment. Mr. Skinner also supported his estimate with various tests, including comparison with an estimate for decommissioning the Oyster Creek plant.lg Mr. Skinner also questioned the use of the AIF and BMWL studies in this proceeding because the State was unable to get detailed figures and calculations for either report to independently verify their conclusions.2g 19F (FOOTNOTE CONTINUED FROM PREVIOUS PAGE)
~~ of the State of New York, pp. 5-6, 21-32; Responses'to Hearing Board's Oral Questions by Peter N. Skinner, P.E. on behalf of the State of New York, pp. 3 '
Attachment 3; Rebuttal Statement on behalf of the State of New York, pp. 1-13; Tr. 1223-1253. 199 Answers of Peter k. Skinner, pp. 21-32; Statement of
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Peter N. Skinner, pp. 28-39; Responses to Hearing Board, i pp. 3-6. ; 1 2;@f Statement of Peter N. Skinner, pp. 32-39; Rebuttcl l Statement, pp. 2-13; Tr. 1227, 1242. ' \ 73 I \~J ; l
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.IX. Socioeconomic ~ Effects ;
- 5. ' A. ' Staff's-Testimony-4 The staff considered all the~socioec'onomic effects of
-the fuel cycle together since, in its opinion,.the impacts I of. establishing reprocessing plants and waste-related j
g facilities."areinot exp'ected to differ in quantity or quality.from those associated with any' commercial nuclear l I power plant." The staff summarizes these issues as follows: Impacts that can be expected are comparable to , or less than those caused by LWR construction l activities and'could include noise and dust , around'the site; disruptions or dislocations
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of residences or businesses; physical or ! public-access impacts on historic, cultural, and-natural' features; impacts on public serv-ices such as education, utilities, the road system, recreation,'public health, and safety; increased. tax revenues in jurisdictions where d facilities are located; increased local expend-iteres for services'and materials, and social s' tresses. NUREG-0116, Section 4.11.1, p. 4-168. The staff emphasizes that the degrea of socioeconomic l effect of any'part of the fuel cycle will be site specific but' based on recent site-specific environmental analysis -
'and'on TVA's experiences in construction of nuclear reactors, ;
evenLpotentially severe ~ impacts are manageable. In the' <
- staff's opinion,:when the impacts of reprocessing and waste
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. ! ) - 118 - MJ burial facilities are spread over many power reactors, they add an insignificant amount to the environmental C impacts of an individual reactor. During the hearing the staff presented Dr. John. Bartlettb1 to testify on the socioeconomic effects of large nuclear fuel facilities near Hanford, Washington. Dr. Bartlett stated that Richland was one of several com-munities affected by the building of various nuclear facilities in the region. He noted that the impacts varied from community to community but the problems were handled by cooperation among them all. In general, the communities view the socioeconomic impacts as favorable to the residents. Tr. 684-88. Another staff witness, Frank J. Miraglia, stated that the residents of the areas surrounding the' Oak Ridge, l Savannah River, Rocky Flats or Hanford facilities had . 1 expressed no concern about the security measures required at any of-tne nuclear plants. Tr. 682-83, f 20Y Dr. Bartlett 'is the mayor pro tem of Richland,
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Washington, the site of the Hanford nuclear facilities. f l e X.,1
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.B . -Testimony of Other Parties J . . .
- 1. States.of Ohio & Wisconsin Witnesses' John Kelly and David Deese202/ asserted that the staff had given socioeconomic impacts only cursory
. treatment which was not adequate to meet the requi rements of NEPA. .These witnesses claim that "in terms of social and psychological stress and economic benefits -- two primary socioeconomic impacts.-- the effects of reprocess-ing plants and waste-related facilities differ substantially .
from those of commercial nuclear power plants." L Dr. Kelly asserts that for two reasons the social l t and psychological stresses on the1 population near a separa-tion plant or an HLW disposal facility are much greater
.than the stresses felt by the people near a nuclear reactor.
First, in the opinion of the. witness the level of security around the waste related plants is greater and more obvious than-that used for reactors. Second, the public connects the waste related' plants with available plutonium, an element which the public perceives as particularly toxic. q L l
~207 See - Statements . of Drs. John E. Kelly and David Z.
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Deese
'On Behalf.of the States of Wisconsin anc Ohio, dated 1 l: October 13,.1977.- See also' Testimony of Dr. John E.
Kelly in Lient of Oral-Testimony,. dated April 5, 1978. i . .:- A - ll ud w ! s
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In regard to the socioeconomic benefits Dr. Kelly sees the increased tax revenues as a positive factor for the waste connected facilities as well as reactors. But he believes that the public is able to visualize the benefits of reliable electricity.from a power reactor but not from the other facilities. Dr. Felly also emphasized l l that the long delay in selecting waste management technology ; and disposal sites has emphasized the problems of such i sites and caused many states to object to any waste dis-posal system being located within their boundaries. On a somewhat different tack, Dr. Deese claims that as a leader in nuclear technology the United Sta'tes is being looked to by many other countries to decide on the best waste management methods. Therefore, the Commission's decisions on waste management will have a socioeconomic impact throughout the world. In his opinion, the staff has completely ignored such far reaching effects. I
- 2. ' Natural Resources Defense Council !
NRDC witness Dr. Todd R. LaPorte 2
- 0y testified that ,
1 l the staff's review of socioeconomic impacts focuses on 207 Statement of Todd R. LaPorte entitled "On Increasing
,_s ~~ ' Scale and System' Reliability in the Management of j l Radioactive Wastes: Requisites for Credible Environ-L} ment and Cost-Benefit Analysis," October 3, 1977.
-l
[D - 121 - U the effects of one-reactor year and thus gives no concept of the total scale of operations or of the magnitude of zransportation to and from the operating and waste disposal facilities. The witness asserts that (i]n the absence of information about the scale of operations, the public and its leaders perceive that a number of critical issues have been left in a state of great uncertainty. The public, he asserts, is therefore uncertain as to the true socioeconomic impacts of reprocessing and waste dis-posal.
- 3. Commonwealth Edison, et al.
In the.ir concluding statementd I BG&E generally l supports the staff's position on socioeconomic effec's,e ; citing in particular the lack of evidence of negative l l socioeconomic impacts from nuclear f$cilities at Hanford l (Richland), Washington; Rocky Flats, Colorado; and Oak Ridge, Tennessee. They specifically disagree with i Dr. Kelly's assertions that the socioeconomic impacts at waste related facilities will be greater 'than suen impacts 20# Concluding Statement of Position on Behalf of Baltimore Gas and Electric Co., et al., June 26, 1978, SectionII-H. e v
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l L- - 122 - from reactors. lBG&E points out that " upgraded security / I safeguards $ provisions at' power reactors will be comparable to those'at' fuel cycle facilities." BG&E further claims that witnesses Kelly and Deese presented no factual basis for their arguments that the socioeconomic impacts of stress or those related'to international affairs actually j exist. Id. at pp. II-H-7 ' through II-E-10,
- 4. Tennessee Valley Authority. l In its own statement', ! supplemental to the'.BG&E. group of which it is a part, TVA supported the staff's consider-
.i i
ation of socioeconomic impacts. However, TVA objects strongly'to.what they.see as an implication arising from th,eHearingBoard'squestioninhbbbofthestaff'thatgov- O ernment-owned facilities, because they pay no taxes, would have great'er socioeconomic impacts than privately owned facilities. TVA~ points out that as a Federal resource agency'it considers the socioeconomic impacts a of its projects and it engages "in extensive programs to mitigate thesc. impacts.'"- In this-connection, TVA discusses its Hartsville nuclear project and asserts that its mitigation 1 i l h . .
~209 Concluding. Statement of the Tennessee Valley Authority, 1 '
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W dated June = 26, '1978. 3'A - 209 Tr'. 163-64. l J i u l;
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4 123 - projects " include millions of dollars in financial assis-tance to balance local. budgets, obtain professional plan-ning advice, buy temporary classrooms, contribute to the construction of new schools, buy supplementary water treat-ment facilities and to pay for added police personnel and equipment." . 1 l 1 l l l 4/ l 'A
- 124 -
d.m X. Economic Feasibility A. Staff Evidence Reasoning that such factors were outside of the scope of this proceeding, the staff did not consider the economic costs of the fuel cycle in its original testimony. Nevertheless, the States of Ohio, Wisconsin, and New York included economics in their initial written statements and in questions to the staff and utilities. Following a series of motions and rulings 207/ the staff submitted its " Staff Testimony on Economic Data to Support the Feasibility of the S-3.Model" (dated February 3, 1978). 207/ Wisconsin's questions were rejected by the Hearing Board (State of Wisconsin's-Proposed Questions, dated October 31, 1977). Wisconsin requested clarification . and reconsideration of the issue (State of Wisconsin's Objections to Board's Order of November 18, 1977 and Request for Clarification of Board's Position on Scope of Oral Testimony). By order of December 23,-1977 the Board rejected the Wisconsin _ request for full economic discussion as being beyond the scope of this proceeding, but it did agree that "a limited amount of cost data should be made available to allow the parties to test economic feasibility" (at p. 3). To accomplish this the Board directed the Commission's staff "to make i available appropriate economic data respecting the I reprocessing and waste disposal facilities from which l the values shown in Table S-3 were derived, * *
- in !
order that'the parties hereto may have an opportunity to test in this remanded proceeding the economic feas- , ibility of these facilities." This procedure was ( i approved by Commission Order dated February 9, 197'8.
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The staff's basic data were set forth in four enclosures: (1) Tabulation of estimates of facility costs; (2) Discussion of operating costs based on GESP.O economic data; (3) Supplemental operating cost information for those portions of the S-3 model not included in the GESMO data; (4) Discussion of decontamination and decommis-sioning costs. The staff points out that its tabulation of facility cost estimates is based on a 1977 study by Bechtel which was prepared for ERDA (now DOE). The decontamination and 20y decommissioning data were based on a Battelle study-- cnd 20S' an AIF study.--
- 1. Facility and Operating Costs of Reprocessing and Waste Management The staff had to apportion all of the costs for reproces-sing, solidification, high and low level waste disposal to a single reacter over its lifetime. Following the procedure j 20S'
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NUREG-02.73, Technology, Safety and Costs of Decommis- l sioning a Reference Nuclear Reprocessing Plant (pre- ' pared for URC by Battelle Northwest Laboratory). 20V Atomic Industrial Forum, AIF/NESP-009, "An Engineering
' Evaluation of Nuclear Power Reactor Decommissioning
,_ Alternatives", November 1976, f
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- 126 - /s '(,)a used in GESMO, the staff took the data from GESMO or Bechtel reports and' corrected them to 1977 cos.ts based on inflation i between 1975 (the year in which costs were calculated by GESMO End Eechtel) and 1977. Since the charges for various steps of reprocessing, burial, etc., would be incurred by the reactor operations at varying times in the future, these charges were discounted back to 1977 using a discount rate of 10%.
Ucrever, the Board concluded from its study of the ' l evidence and review of the parties' proposed questions that l the staff had not completely justified its procedure for apportioning the cost per kilowatt hour' incurred by a reactor due to the spent fuel cycle operations. At the Board's request, the staff supplied further clarifying data and discussion during the hearing. Tr. 1979-1181, particularly data following Tr. 1084, 1257. In the additional data and discussion, the staff pointed 1 1 out that the cost of all important facilities in the back end ' of the fuel cycle had been estimated. But some facilities might not be used, depending on the method and extent of interim storage, type of solidification, transportation l and disposal methods used, etc. Sj~sg
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The; staff explained that their-method of discounting was' based on the one used in GESMO. The discount. rate was recommended.by the Office of Management and Sudget (0:13) . t Tr. 1129. The staff stated that their calculations showed that a: discount rate of 4% would approximately double the costtof electricity per "<hr incurred because of fuel cycle charges. Tr.- 1130. The staff asserts that, assuming , a 10% discount rate,.the cost:of' waste management and re- , processing for uranium-only-recycle will amount to approxi-mately 3% of the total generating cost. Staff's Testimony
~
en Economic. Data to Support Feasibility of the S-3 !!odel, enclosure:2, p. 4. B. Evidence of.Other Parties,
- 1. Direct evidence on economic feasibility was presented r
, by vitnesses-representing New York State, and witnesses ' cc-sponsored.by the. States of Wisconsin and Ohio. La. The initial testimony from the' State of New York j 210/- < was supplied by witness Peter N. Skinner.-- This testimony c was supported byl rebuttal evidence supplied by Mr. Skinner 2 and Dr. Smernoff.1f 2Jp' Statement of Peter N. Skinner, P.E., on Behalf of the State of New York, September 30, 1977. gly / Rebuttal Statement on Behalf of the~ State of New York
- (Parts'Inand II by Peter N. Skinner, Part III by
~
. Dr'. Barry J. Smernoff), May.29, 1978.
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, 4 + ', _ 12g - Mr. Skinner's objection is that the staff's use of "the Discount Rate for Radioactive Waste Management" is incorrect. Mr. Skinner sums up his argument as follows: This economic strategem seems plausible, but is in fact totally inappropriate for radioactive waste management charges because:
- 1. discounting unjustifiably encourages long lag times between waste production and final safe disposal.
- 2. the funds, generated on paper by discounting are not liquid when needed or else lost in the accounting system.
- 3. discounting increases the cost of electricity for ratepayers who never generated the original wt. sten.
- 4. the funds generated on paper by discouncing assume a stable and growing economy and demand for the utility's product. i (Reb'chtal Statement, pp. 15-l'6.)
He justifias the four statements as follows: l (1) By discounting, the U.S. NRC staff justifies doing nothing about waste disposal now. The longer tha discounting period lasts under their scenario, the longer money nominally earmarked i for waste management charges will be making money at the assumed discount rate. Therefore if.we wait, say 100 years before paying $100,000 worth of repository charges, we will have,to only put away $7.26 now. In other words, the longer we wait, the cheaper the cost now seems to,be,. This assumption, implicit in discounting, is
,. not only morally wrong, but also is inconsistent with the policy *of Congress and the protestations
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.of the U.S. D.O.E. in theirfefforts to locate
.andiconstruct adequate waste management facilities to deal with the wasta presently being generated.
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[ Footnotes ' omitted . ] ' Id.,.p.-16. l l (2) The. funds generated by discounting will not be liquid when needed unless strict financial mechanisms (are required) to safeguard.the liquidity and accountability of such investment and the interest it generates. Since such-tight controls as. proposed by NUREG-02 7 8, page 10-12, have not yet been adopted, there can be no confidence that any money will be available to' pay the eventual reporting charges etc. Id., p. 17. (3) Discounting should not'be used because the interest supposedly earned on the deferred waste charges is money that the utility would ! have' earned.on that money anyway irrespective of.whether it was based on deferred charges. The utility. customer could have benefitted-
'from~this investment by'getting lower electricity
. charges. i Rather,'sidce the interest will go toward i repository charges,. future rate-payers will not
- get any benefit from'this investment and-there-
' fore be' helping to pay for waste generated by ,
earlier rate-payers. i Id., p.118. t,
- (A ) Finally;by using discounting the staff-(and the o
utilities) assume'that the utilities "will be able to continue V s l' i L
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to sell (their] product and that the econ round [them] is good." Mr. Skinner maintains that it i. .isky indeed" to assume that "the plant and equipment will be valuable . when the investment and its interest are needed for , repository charges." d I_d., p. 19. In addition, Mr. Skinner asserted that the staff's use of a 10% discount rate is unrealistic and that, conserv-atively, return on investment (ROI) should be expected to . be between 0 and 2%. .,
- b. In part III of the Rebuttal Statement entitled
" Economic Feasibility of ' Nuclear Waste Management,"
Dr. Smernoff asserts that on the basis of the report of the Nuclear Energy Policy Study Group of the Ford Founda-
, tion, "the comparative economics of the nuclear and coal -
options for generating bulk electricity is virtually a standoff and that the competitive advantage which nuclear , energy might have is marginal at best."212/ He further ' asserts that "[o]fficial projections for U. S. nuclear- .- electric capacity for the year 2000 have been sharply I reduced every few years" and that a de facto " Moratorium - 212/ Dr. Barry J. Smernoff, " Rebuttal Statement on Behalf -
'of the State of New York," Part III, p. 4, May 29, 1978. >
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M l - 131 - on contracts for new ' ear power planus is pt .ily due to the large f. c' ' risk involved." Id. at p. 5. On the basis - che above assertions Dr. Smernoff concludes that "the marginal economic advantage enjoyed by the ' barely alive' nuclear industry could well be erased by unforeseen cost escalation at the back end of the fuel cycle, suggesting that the economics of the tail may well wag the nuclear dog in this case." He believes that the public may insist on maximum protection from any possible environmental impacts of the fuel cycle thus forcing costs up sufficiently to make the entire cycle economically infeasible.
- c. On the other hand the Wisconsin and Ohio wit.ess, Dr. Irwin C. Bupp, stated in his initial testimony that there is a " complex argument about whether atomic energy is economically competitive with coal." This con-tinuing argument persists according to Dr. Bupp because the cost of building nuclear plants has not stabilized as expected but continues to rise. On the other hand, environmental costs are also increasing the costs of coal plants.
213/ Testimony of Dr. Irwin C. Bupp on Behalf of the States of Wisconsin and Ohio, October 3, 1977.
l: y ., c f ) f - 132 - However, Dr. Bupp states that the cost of the fuel cycle amounts to only 15% of the cost of producing nuclear power. Of this 15% only 1/4 is assignable to the back-end of the cycle. On this basis, he points out that even doubling the cost of the back-end of the fuel cycle would have an insignificant effect on the cost of power produced. In his supplemental testimony ! Dr. Bupp reiterated his concern about 'he uncertainty of all the costs in the fuel cycle. However, referring to the New York testimony, he asserted that: But, even if Skinner is correct, these costs do not in my opinion decisively alter the economics of nuclear pawer relative to realistic near-term alternatives. C. Comments by other Parties Only two other parties (Baltimore Gas and Electric, et al. and Commonwealth Edison, et al.) commented speci-fically on the economic feasibility issue.
- 1. BG&E, et al., in their final statement discuss the difficulty of considering the economics of the fuel cycle without balancing costs against environmental 214/ Supplemental Testimony of Dr. Irwin C. Bupp on Behalf of the States of Wisconsin and Ohio, received April 18, 1978.
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215/ Concluding Statement on Behalf of Baltimore Gas and Electric Co., et al., p. II G-1 to 8, June 26, 1978.
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consequences. They note that the staff uses a model system that tends to maximize the environmental effects (a con-servative approach for the environmental impact considera-tions). This, they state, does not necessarily result in conservative, i.e. high, economic cost estimates. How-ever, in the case of reprocessing, the staff presents alternate models with somewhat higher environmental impacts l and lower costs. BG&E considers "the basis for Mr. Skinner's estimates j questionable" and notes that he " conceded that there was not," contrary to his initial testimony," general agreement", that waste management cost estimates lacked foundation at least among government and industry participants. In this connection, they note the statement of Dr. Bupp, witness
'for Wisconsin and Ohio. See pp. 132, 133, supra.
The comments of Commonwealth Edison, et al. ! 2. supported the staff's estimates but indicated that the costs were conservative. They also point out that even with this conservatism the added cost of electricity due to the back-end of the fuel cycle would be less than one mill / kwhr using a 10% discount rate. Using a 4% discount rate, the added cost of electricity would be between 0.8 and 1.2 mill.
~216/Edison, Final Written Statement on Behalf of Commonwealth et al., p. XI - 1 through 13, June 26, 1978.
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l Commonwealth Edison, et al. are critical of New York witness Skinner for relying on the history of the Nuclear Fuel Systems experience at West Valley in both low level waste and reprocessing. They believe he ignores the improvements that can and are being made as the result of these early experiences. 4 _ - _ _ _ _ ____-__._-_____________2..__
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135 - PART 3 - OUTLINE OF SIGNIFICANT ISSUES Fuel Reprocessing
- 1. Whether the capacity and capacity factor specified by the staff in its model fuel reprocessing facilities are reasonably achievable.
- 2. Whether the staff's estimates of radioactive releases and occupational exposures are underestimated.
- 3. Whether the staff's model fuel reprocessing facilities ,
are economically feasible.
- 4. Whether the basis for the staff's choice of equipment and systems in its model fuel reprocessing facilities accords adequate consideration for past experiences in fuel reprocessing.
- 5. Whether the staff's design of equipment and systems in its model fuel reprocessing facilities places too much emphasis on projected improvements.
- 6. Whether the staff's assumptions on estimates of releases of long life radioactive elements in effluents from its model fuel reprocessing facilities are overly conservative or can be supported.
G e
, i
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]}
High Level Waste Disposal
- 1. Whether the staff's choice of salt mines for disposal of high level wastes is practicable and provides reasonable assurance for public health and safety.
Low Level Waste Disposal
- 1. Whether shallow land burial of low level wastes can be accomplished without endangering public health and safety.
Interim Storage
- 1. Whether the staff's model facilities accord adequate and practical considerations of interim storage of spent fuel before. reprocessing or disposal.
Transportation
- 1. Whether sufficient shipping containers are available for both spent fuel and reprocessed material.
Decommissioning
- 1. Whether decommissioning of a nuclear plant is economically feasible.
9
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Socioecor.pic Effects
- 1. Whether the staff gave adequate consideration to socioeconomic impacts of fuel reprocessing and waste n)anagement.
Economic Feasibility
- 1. Whether the staff's estimates of facility and oper-ating costs of fuel reprocessing and waste management are reasonable.
t 0
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l Li Attachment 2 l VIEWS OF THE HEARING BOARD ON WHETHER l THE RULEMAKING RECORD SUPPORTS AN EXTENSION OF THE .. l INTERIM RULE FOR A REASONABLE PERIOD PENDING '
,f COMPLETION OF OUR RECOMMENDATIONS AND COMMISSION ~
CONSIDERATION PRIOR TO ACTION ON A FINAL FUEL CYCLE RULE ,s e The Hearing Board has completed an in-depth review
; and analysis of the record develope,d in the reopened rule- , . ,.
making proceedings regarding the environmental effects of the reprocessing and waste management aspects of the . s Uranium Fuel Cycle. It is our view thdt the rulemaking
- record adequately supports an extension of the Interim ;
Rule for a reasonable period, pending completion of our recommendations and Commission consideration prior to action on a final rule, , As the Commission has been informed simultaneously with the submission of our report on the record, the H ' %s Hearing Board may take until October 31, 1978 to complete . the preparation of its recommendations regarding the reprocessing and waste management aspects of the fuel cycle. We have concluded that, as a result of the record , in the reopened proceedings, some additions in and mod-ifications to the fuel cycle rule as it affects reprocess- , ing and waste management are warranted and will, therefor,e, 1 be recommended. These additions and modifications are not, ,
wg ,
- 9l y . -
in our opinion, substantial enough to change the effects j of the present Interim Rule. Accordingly, an extension , of the Interim Rule for a reasonable period will not adversely affect the public interest. ( i . m
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