ML19211C748
| ML19211C748 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/1979 |
| From: | NRC COMMISSION (OCM) |
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| ML19211C733 | List: |
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| NUDOCS 8001140397 | |
| Download: ML19211C748 (37) | |
Text
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NRC REPORT TO CONGRESS ON THE STATUS OF DOMESTIC AND INTERNATIONAL EVALUATIONS OF NUCLEAR REACTOR AND FUEL CYCLE SYSTEMS 1742 136 November 1979 8001140 N
Executive Summary The U.S. Nuclear Regulatory Commission was directed by Congress, in the 1979 authorization legislation (P.L. 601), to submit periodic reports on NRC's assessment of the status of domestic and international evaluations of nuclear fuel cycle systems and a summary of all research activities relating to such matters. This is the first semi-annual report which covers NRC activities in alternative reactor and fuel cycle evaluations prior to June 30, 1979.
Events occurring after this date will be covered in subsequent reports to Congress in this series.
Because of the strain on available agency resources due to the Three Mile Island accident, NRC is able to provide only a limited analysis of the status of alternative reactor and fuel cycle systems evaluations.
Nonproliferation Alternative Systems Assessment Program (NASAP) and International Nuclear Fuel Cycle Evaluation (INFCE) final reports will not be available until late this year or early next year, and as of mid-1979 NRC had made only preliminary reviews of certain NASAP and INFCE documents.
NRC will provide additional comments on alternative reactor and fuel cycle systems in its subsequent reports to Congress.
In its analysis and review, NRC has examined information emanating primarily from four sources:
(1) DOE NASAP documentation, (2) INFCE Working Paper documentation, (3) the GAO, and (4) the Congressional Research Service (CRS).
NASAP was organized by DOE and is the technical program which provides U.S.
alternative systems concepts input to INFCE.
NRC staff members are inde-pendently reviewing DOE work related to the NASAP effort.
The majority of the NASAP material reviewed by the staff is contained in DOE's Preliminary Safety and Environmental Information Documents (PSEIDs).
These PSEIDs exist in seven volumes.
Six describe the mainline reactor types under consideration by DOE.
The seventh describes the associated fuel cycle facil-ities.
The six reactor types discussed are:
Light Water Reactor (LWR), Heavy Water Reactor (HWR), Light Water Breeder Reactor (LWBR), High Temperature Gas-Cooled Reactor (HTGR), Gas-Cooled Fast Breeder Reactor (GCFR), and Liquid Metal Fast Breeder Reactor (LMFBR).
The seventh PSEID provides information on candidate fuel cycle concepts for the reactor types.
Three fuel cycles are discussed:
Once Through Uranium Fuel Cycle, Uranium Fuel Cycle with Uranium and Plutonium Recycle, and Thorium Fuel Cycle with Thorium and Uranium-233 Recycle.
DOE has also provided several supporting NASAP documents discussing Nuclear Energy Characterization Data, Safeguards Considerations for the NASAP Alterna-tive Fuel Cycles, and the Denatured U-233 Fuel Cycle.
NRC has provided review comments on these documents as well.
INFCE began in October 1977.
It now has 53 participating countries and is organized around eight working groups.
The United States has offered about 50 papers to the working groups and has reviewed and commented on more than 200 papers contributed by other countries and organizations.
NRC participation 1742 137
has been in the U.S. organizational support structure for INFCE.
INFCE documents will not be released until the final reports are published and therefore NRC is limited in its discussion of them in this report.
NRC also has examined pertinent reports prepared by GA0 and CRS.
A brief discussion of these reports is also included in this report.
During FY 1979 Congress approved an $800,000 carry-over of FY 1978 funds for alternative reactor and fuel cycle research at NRC.
However, with Congressional approval NRC delayed all but $125,000 of this research effort in order to meet resource needs precipitated by the Three Mile Island accident.
NRC has implemented a modest technical assistance program directed to a better understanding of some of the licensing issues unique to the NASAP reactor conceptual designs.
NRC also has contracted with Brookhaven National Laboratory to provide background technical information for NRC's use in establishing future programs and policies relevant to the safeguarding of fuel cycles developed by the commercial nuclear industry.
During FY 1979, NRC will have devoted about four person years of staff time and expended about half a million dollars on technical assistance in support of the alternative reactor and fuel cycle effort.
NRC also maintains a signifi-cant research program in fast breeder reactors and a limited program in advanced converter reactors, funded at $14 million and $3.2 million respectively in FY 1979 and supported by about 14 person years of NRC staff effort.
This report provides limited preliminary comments related to the six reactor concepts proposed by DOE under NASAP.
In order to perform a systematic eval-uation of the safety and licensing issues for the six reactor concepts, addi-tional information and analysis in four general areas is needed:
(1) facility system descriptions, (2) safety approach, (3) accident analysis, and (4) research and development program plan.
For only the LWR do we have sufficient information in the four general areas.
However, each of the six concepts, given appropriate engineered safety features, could be potentially licensable.
The extended-life LWR may be the most readily licensable concept; also the U.S. has already had versions of the HTGR and LMFBR operating in this country.
The main proliferation concerns center on the back end of the fuel cycle, i.e., those operations associated with reprocessing and fabrication of recycle fuel, and waste management.
All of the fuel cycles considered, with the exception of the once-through fuel cycles, require reprocessing and recycling of fissile materials.
More information about the institutional and technological fixes that would be applied to these fuel cycles is needed in order for NRC to attempt an evaluation of the relative proliferation risks.
Presently, the NRC does not possess enough information to draw final conclusions regarding the relative safety, safeguardability, environmental impact, and licensability of the alternative reactor and fuel cycle concepts that have been reviewed.
However, it is doubtful that NRC would draw final conclusions on such matters until an actual application is received, so as to not prejudice the outcome of the licensing review process.
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TABLE OF CONTENTS Page I.
INTRODUCTION..
1 II.
SUMMARY
OF PERTINENT REFERENCES 3
III.
SUMMARY
OF NRC COMMENTS ON ALTERNATIVE REACTORS AND FUEL CYCLES 5
A.
Summary of NRC Comments to DOE on NASAP...
5 1.
Preliminary Safety and Environmental Information Documents 5
a.
Light Water Reactor (LWR).
6 b.
Heavy Water Reactor (HWR)...
7 c.
Light Water Breeder Reactor (LWBR) 9 d.
High Temperature Gas-Cooled Reactor (HTGR) 10 e.
Gas-Cooled Fast Breeder Reactor.
11 f.
Liquid Metal Fast Breeder Reactor (LMF8R).
12 g.
Fuel Cycle Facilities..............
14 2.
Nuclear Energy System Characterization Data 16 3.
Safeguards Considerations for the NASAP Alternative Fuel Cycles 18 a.
Material Control and Accountability.
18 b.
Physical Security.
21 c.
institutional and International Considerations 21 d.
Storage.
22 e.
Summary.
24 4.
The Denatured U-233 Fuel Cycle 24 B.
Summary of NRC Participation in INFCE...
26 C.
GA0 Reports.......
28 D.
Summary of NRC Comments on the Congressional Research Service Report: Alternative Breeding Cycles for Nuclear Power: An Analysis..
29 IV.
SUMMARY
OF APPLICABLE RESEARCH AND TECHNICAL ASSISTANCE 30 V. ADDITIONAL COMMENTS 32 VI. GLOSSARY.
33 1742 139 iii
NRC REPORT TO CONGRESS ON THE STATUS OF DOMESTIC AND INTERNATIONAL EVALUATIONS OF NUCLEAR REACTOR AND FUEL CYCLE SYSTEMS I.
INTRODUCTION The U. S. Nuclear Regulatory Commission was directed by Congress, in Section 9 of the 1979 NRC authorization legislation (PL-601), to submit a semi-annual report on NRC's assessment of the status of domestic and international eval-uations of nuclear fuel cycle systems.
The law directs the Commission to:
".. report to the Congress semiannually through calendar year 1980 and annually through calendar year 1982 on the status of domestic and international evaluations of nuclear fuel cycle systems.
This report shall include, but not be limited to, a summary of the information developed by and available to the Commission on the health, safety and safeguards implications of the leading fuel cycle technologies."
Section 9 of the Joint Explanatory Statement of PL-601 further states that, as part of the report to Congress, NRC snall provide:
"..a summary of all research activities relating to such matters, identification of areas of deficiency in such existing information and ongoing efforts, and recommendations regarding matters which should be undertaken to minimize any potential adverse impact on matters within the Commission's jurisdiction;"
This is the first semi-annual report and covers NRC activities in alternative reactor and fuel cycle evaluations prior to June 30, 1979.
Events occurring after this date will be covered in subsequent reports to Congress in this series.
Section II of this report references pertinent documentation.
Section III provides a summary of NRC comments on alternative reactor and fuel cycle systems described in reports that were reviewed.
Section IV summarizes applicable NRC research and technical assistance on alternative reactor and fuel cycle concepts.
Section V provides additional comments.
Identification and description of alternative reactors and fuel cycles by DOE are still being developed and the leading technologies are yet to be delineated.
Nonproliferation Alternative Systems Assessment Program (NASAP) and International Nuclear Fuel Cycle Evaluation (INFCE) final reports will not be available until late this year or early next year.
To this date NRC has made only preliminary reviews of draft NASAP and INFCE documents.
The Three Mile Island accident has placed a severe strain on available agency resources; thus, at this time the staff is able to provide only limited analysis of the status of alternative reactor and fuel cycle systems evaluations.
High priority demands on NRC resources will continue in FY 1980.
Nevertheless, 1742 140
2 after completion of the NASAP.nd INFCE reports, the staff will consider means to apply sufficient resources to the review and assessment of these documents so that in subsequent reports to Congress the relative merits of alternative reactor and fuel cycle systems may be further analyzed.*
NASAP was organized by DOE and is the technical program which would provide U.S. alternative systems input to INFCE.** NRC staff members are independently reviewing DOE work related to the NASAP effort.
The NASAP documents are com-prised of the Preliminary Safety and Environmental Information Documents (PSEIDs),
which provide information for each reactor and fuel cycle type, and several supporting documents. NRC staff has commented on these documents and the comments will be considered by DOE in the development of the NASAP final report.
Meetings were held with DOE contractors to discuss the reactor and fuel designs for several reactor types.
Our participation in INFCE originated as a result of informal con?.scts between NRC and the Department of State in September and October of 1977.
INFCE now has participation from 53 countries and is organized around eight working groups. Our staff members served as NRC contacts on the U.S. support groups for those INFCE working group areas in hich NRC has significant interests and expertise and have coordinated informal staff reviews of U.S. and foreign country analyses and studies.
NASAP assessments and conclusions are to be completed this year and will be presented by DOE in a report to Congress and the President by December 1979.
INFCE Working Group Reports are now being finalized.
The INFCE final report is scheduled to be presented to the Final Plenary Conference in February 1980.
- NRC has also agreed, in response to a GA0 recommendation, to provide a staff report to the President and Congress on our preliminary findings of known or suspected licensing issues and problems associated with alternative technologies under serious consideration by DOE.
This report could include a comparative evaluation of the alternative technologies studied from the safety, safeguards, environmental and licensing points of view.
NRC would publish this report subsequent to the com pletion of the NASAP program and the INFCE stucies.
- According to DOE, NASAP has the following objectives:
Assess the safety, environmental and safeguards issues that might impact a.
development and licensing of the principal alternative technologies; b.
Obtain inputs for estimating R&D costs ana times to accomplish; and c.
Compare alternative technologies for nonproliferation benefits and costs.
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3 II.
SUMMARY
OF PERTINENT REFERENCES In the analysis and review of alternative reactor and nuclear fuel cycle system concepts, NRC has examined documents emanating from four primary sources:
DOE NASAP documentation, INFCE Working Paper documentation, GA0 reports, and Congressional Research Service (CRS) Report.*
NASAP Reports The primary NASAP documentation is the Preliminary Safety and Environmental Information Documents (PSEIDs).
The PSEIDs are composed of seven volumes.
The first six volumes address six classes of alternative reactor and fuel cycle system concepts; these are Light-Water Reactors (LWRs), Heavy-Water Reactors (HWRs), Light-Water Breeder Reactors (LWBRs), High-Temperature Gas-Cooled Reactors (HTGRs), Gas-Cooled Fast Reactors (GCFRs), and Liquid-Metal Fast Breeder Reactors (LMFBRs).
The seventh volume addresses Fuel Cycle Facilities for all the reactor concepts discussed.
In addition, 00E provided supporting documents discussing Nuclear Energy System Characterization Data, Safeguards Considerations for the NASAP Alternative Fuel Cycles, and the Denatured U-233 Fuel Cycle.
NRC has provided review comments to DOE on all these documents. A summary of NRC comment on these documents is provided in Section III.A of this report.
INFCE Reports INFCE reports are papers contributed by the U.S. and other foreign organiza-tions.
With participation from 53 countries, INFCE is organized around eight working groups covering the topics of fuel and heavy water availability; enrich-ment availability; supply assurances; reprocessing, plutonium handling, and recycle; fast breeders; spent fuel management; waste management and disposal; and alternative fuel cycle and reactor concepts.
The United States has offered about 50 papers to the INFCE working groups and has reviewed and commented on more than 200 papers contributed by other countries and organizations.
The NRC has participated in the review of many of these papers.
Since all reports prepared for INFCE are " Official -- For INFCE Use Only" and no INFCE documents are to be released until the final reports are published, NRC's discussion of them in this report is limited. A summary of NRC's participation in INFCE is provided in Section III.B of this report.
GA0 Reports The GA0 report of primary concern is the March 7, 1978 GA0 Final Letter Report regarding NRC's role in selecting fission technologies which was prepared for the Joint Economic Committee of Congress.
A brief discussion of this report is provided in Section III.C of this report.
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e 4
II I
GA0 has also prepared other reports on topics related to Nonproliferation and Alternative Fuel Cycles.
NRC revievzd a d. aft version of such a report late in 1978.
A final version of this report was received by NRC too late for coverage in the discussions of this report.
CRS Report The Congressional Research Service (CRS) prepared a report entitled, "Alterna-tive Breeder Cycles for Nuclear Power: An Analysis," October 1978, for the Subcommittee on Fossil and Nuclear Energy Research, Development and Demonstra-tion of the Committee on Science and Technology, U.S. House of Representatives.
A brief summary of NRC comments on this report is provided in Section III.D of 4
this report.
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5 III.
SUMMARY
OF NRC COMMENTS ON ALTERNATIVE REACTORS AND FUEL CYCLES A.
Summary of NRC Comments to DOE on NASAP 1.
Preliminary Safety and Environmental Information Documents This portion of the report summarizes our review and assessment
- of the Prelim-inary Safety and Environmental Information Documents (PSEIDs) provided to us by DOE.
These PSEIDs are in seven volumes, with six describing the six mainline reactor types under serious consideration by DOE and one describing the associ-ated fuel cycle facilities.
This review and assessment is a first step in the direction of providing a comparative evaluation of the alternative fuel cycle and reactor types from the safety, safeguards, environmental and licensing points of view.
At this stage, it appears that each of the concepts is potentially licensable, given adequate engineered safety features would be provided.
Under an agreement with DOE, NRC provided to DOE on June 13, 1979 a set of comments and questions on tt.e PSEIDs.
The comments are summarized on the following pages.
In order to place this preliminary NRC review and assessment into perspective, it is important to understand what NRC would need in order to perform a systematic comparative evaluation of the safety, safeguards, environmental and licensing issues, and to compare these needs with what DOE has provided and what is generally known about these alternative reactors and fuel cycles.
Ideally, NRC would like to have consistent and well oefined (relative to the preliminary nature of a conceptual study) information in four general areas.
.The NRC commants on the six PSEID documents on reactor types are grouped in these four area, namely:
1)
Facility System Descriptions - For reactors:
a description of the mechan-ical, nuclear and thermal hydraulics aspects of the reactor core region; the heat transport system, including the decay heat removal system; the facility engineered safety features; instrumentation and control; contain-ment; and the balance of plant.
For fuel cycle facilities:
a description of process, chemical, mechanical, criticality, ventilation, scrap and waste handling systems; confinement and radiation cc.7 trol systems; effluent treatment and handling systems; engineered safety systems; safeguard and physical protection systems; and proliferation resistance features; 2)
Safety Approach, including a description of areas of conformance or departure from NRC's Code of Federal Regulations, the General Design Criteria, standards and regulatory guidelines; "At this stage NRC's preliminary " review and assessment" comprises: an updating of our present licensing position regarding these reactors and fuel cycles; comments on safety aspects; and questions directed to DOE that will aid NRC in its continuing review process.
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6 3)
Accident Analysis, including the low probability, potentially high conse-quence accident assessment; and 4)
Research and Development (R&D) Program Plan, in particular identifying the key safety areas with estimates of time and cost.
It is understandable that DOE could not provide all this information within their budget and schedule constraints, in particular for some of the less-developed reactor and fuel cycle concepts.
As NASAP evolves, NRC, both by independent investigation that makes use of past experience with similar reactor and fuel cycle concepts, and through evaluation of further analysis and assess-ment provided by DOE, will continue to review and assess the mainline reactor and fuel cycle types in these four informat'on categories.
The six reactor types and associated fuel cycle variations are:
a.
(PSEID Vol.I) Light Water Reactor (LWR) with (a) extended burnup; (b) denatured (U-233/U-238) fuel; (c) spiked plutonium recycle; and (d) a plutonium / thorium cycle; b.
(PSEID Vol. II) Heavy Water Reactor (HWR) of the Canadian Deuterium Uranium (CANDU) type; c.
(PSEID Vol. III) Light Water Breeder Reactor (LWBR) with four Prebreeder/
Breeder Pair Variations; d.
(PSEID Vol. IV) High Temperature Gas-Cooled Reactor (HTGR) with 2 fuel cycle variations; e.
(PSEID Vol. V) Gas-Cooled Fast Breeder Reactor (GCFR) with 2 fuel cycle variations; f.
(PSEID Vol. VI) Liquid Metal Fast Breeder Reactor (LMFBR) with 15 core type, fuel form, and fuel cycle variations; and g.
(PSEID Vol. VII) Fuel Cycle Facilities; for numerous variations of both uranium and thorium cycles.
The discussions of individual concepts that follow are intended to portray the present state of the dialogue between NRC and DOE relative to safety, safeguard-ability, and licensing issues.
These issues have been communicated to DOE in the staff letter of June 13, 1979.
For the most part, envirormental issues (and the related issues of economics and fuel utilization) have not been addressed at this time.
The Light Water Reactor (LWR) - Variants (PSEID Volume I) a.
Four variations on a standard LWR-design are being considered as part of this NASAP study.
These are:
extended burnup, from 30,000 Megawatt Days / Metric Ton (MWD /MT) to 50,000 MWD /MT; a denatured fuel (U-233 or U-235 mixed with U-238) cycle; a plutonium / thorium cycle; and spiked plutonium recycle.
For all four 1742 145
7 variations the deviation from a standard LWR design, in this case the Combus-tion Engineering System-80 PWR design, is slight and therefore the facility system description, safety approach, and accident analysis are all quite similar to the CE-System 80.
In fact, other than the fuel-rod design itself, the reactors are virtually identical.
From the view of reactor safety, these LWR variants have limited and well-defined problems.
(Before NRC can make a similar statement on environmental impact, more information must be supplied.)
Differences arise in fuel storage requirements and in fuel handling due to higher enrichments and higher radiation levels, in particular from the fresh spiked and fresh denatured fuels.
Innovations in the areas of fuel fabrication and fuel reprocessing for the spiked and denatured fuel cycles are considered in section III.A.1.g, " Fuel Cycle Facilities."
1)
Facility System Descriptions - Since they deviate only slightly from standard LWR designs, the descriptions are relatively complete.
Only specific fuel rod design information, and fuel storage and handling strategies, need further amplification.
For example it is not clear just how added fission gas inventory in the extended burnup pins will be accommodated.
2)
Safety Approach - Since these variants differ only slightly from licensed LWRs, this item is clearly defined.
An example of a potential change in regulatory requirements is in 10 CFR 51 (Licensing and Regulatory Policy and Procedures for Environmental Protection) where the applicability of Tables S-3 and S-4 is limited to 4 percent enrichment and 33,000 MWD /MT.
3)
Accident Analysis - This assessment depends considerably on the steady-state and transient behavior of fuel rods.
Comments have been made to DOE regarding the need for transient testing of extended burnup fuel pins and for a more comprehensive picture of fuel behavior in general for the denatured fuel cycle option.
4)
R&D Program Plan - NRC requested information on DOE's R&D plans in the safety area for all the LWR variants but in particular regarding fuels for extended burnep and the denatured fuel cycle.
- b. The Heavy Water Reactor (HWR) (PSEID Volume II)
The conceptual design for the HWR as submitted by 00E is a combination of an upgraded CANDU* core ind reactor coolant system (RCS) coupled with a CE-System 80 standardized LWR design for the steam generators, power conversion system, and balance of plant.
Thus, except for the system interfaces, the reactor core and the RCS, this conceptual design is well understood by the NRC.
A thorough understanding of this HWR then requires an understanding of the Canadian reactor core and RCS design, the interfacing of the RCS with the rest of the plant, and the upgrading features determined by DOE.
The upgrading basically is an increase
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8 in:
(1) size of core and RCS, (2) temperature and pressure; (3) enrichment (from natural uranium to 1.2% enriched), and (4) the burnup (from 8,000 MWD /MT for CANDU to 20,000 MWD /MT for the DOE design). Although not a very familiar reactor concept to the U.S. industry and U.S. licensing, the Canadian HWR has extensive operational experience and the benefit of extensive Canadian safety review. DOE gave NRC a draft report on the CE-HWR design which was helpful in the initial assessment.
(1) Facility Systems Descriptions - The discussions are fairly detailed and complete.
Either the system is part of the CE-System 80 standard design for a LWR plant or it is an upgrading of the existing CANDU design.
Questions were asked of DOE regarding systems descriptions which were unique to the upgrading.
For example, there were questions regarding the nuclear characteristics of the core for the extended burnup.
(2) Safety _ approach - There are intrinsic characteristics of this HWR concep-tual design which require a rethinking of traditional licensing regula-tions and practices.
The HWR pressure boundary (pressure tubes) is in a high radiation environment and has rolled joints (as opposed to welded);
these two features are at variance with LWR reactor coolant pressure boundary requirements.
The HWR has a positive power coefficient which may be at variance with the LWR General Design Criteria.
NRC has asked DOE a variety of questions regarding the conformance of the HWR to V. S. licens-ing practice for LWRs.
(3) Accident Analysis - The accident analysis area needs considerable attention for the HWR, as reflected in the NRC comments and questions.
Little is known to the NRC staff regarding a Loss-of-Coolant Accident (LOCA), a Loss of Regulation Accident (LORA), a Loss-of-Heat Sink Accident (LOHSA), and natural convection cooling for "his particular HWR design.
One of the most crucial considerations for the HWR is the leak-before-break concept for the pressure tubes.
If the NRC staff does not accept the leak-before-break concept, a number of accident scenarios involving rapid failure of pressure tubes will have to be considered.
(4) R&D Program Plan - Comments and questions on the safety R&D program for the HWR centered around the leak-before-break issue.
Even with access to valuable Canadian data and experience with pressure tubes and pressure tube failures, NRC still envisions a rather extensive R&D program directed to a U.S. version of the HWR.
Other R&D areas include seismic effects on pressure boundary including the fuel loading machine and fuel steady-state and transient performance.
Much of the effort to date has been to understand better the extrapolation of the Canadian a reactor concept that has had little licensing exposure in this country.
This has led to requests for more information on the CANDU design itself and for analyses and R&D assessments of those upgrade features that are unique to this HWR conceptual design.
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4 9
- c. The Light Water Breeder Reactor (LWBR) (PSEID Volume III)
The LWBR conceptual design is basically a pressurized water reactor (PWR) with a seed-blanket core composed of fissile U-233 oxide and fertile thorium oxide replacing the standard low enrichment uranium oxide core.
In a certain sense then the LWBR is similar to the LWR-variants, discussed above, where modified cores are essentially installed in existing PWRs.
For the LWBR, however, the deviation from the standard PWR is more substantial -- involving not only fuel pins but also the overall core power rating, the reactivity control mechanisms, the core thermal hydraulics and the introduction of a seed-blanket concept.
An additional aspect of the LWBR that must be kept 1-mind is that for every LWBR there must be a "prebreeder" reactor (i.e., prebreeder fuel cycle) designed to convert sufficient Thorium into U-233 in order to provide an initial fissile inventory to start up a LWBR (U-233 is not found in nature).
This is because a LWBR only breeds enough fuel to compensate for the fuel it burns; it provides none for additional LWBRs (i.e., it is not a net breeder).
Thus, in consider-ing safety and licensing aspects of the LWBR, the prebreeder conceptual designs must be taken into account.
This is done by always considering the LWBR in Prebreeder/ Breeder pairs.
Past licensing history for the LWBR can be considered in two parts:
(1) a considerable fraction of the reactor plant, excluding the core, is a stand-ardized PWR which has had considerable licensing exposure; (2) the Division c?
Naval Reactors (DNR) of DOE presently has operating a small (60 MW(e)) experi-mental LWBR at Shippingport, Pennsylvania, which has had a favorable NRC safety review.
(The Shippingport LWBR is not licensed since it is a government owned reactor, and therefore it has not had the benefit of a formal licensing review.)
00E has submitted a LWBR PSEID that presents conceptual designs for four Prebreeder/
Breeder pairs.
NRC concerns have centered on the core design, the reactivity control system design (the Shippingport LWBR has moveable fuel, and advanced concept designs employ moveable fertile pins), and on the scaling up of the nuclear steam supply system (NSSS) and balance of plant (BOP), if undertaken, to match the total electrical power output of a standard PWR (alternatively, the standard PWR could be used, resulting in a reduced plant power output).
(1) Facility Systems Description - The only relatively complete facility systems description presented was for the Shippingport reactor.
Although this has been very helpful in better understanding some of the LWBR con-cepts, there is considerable need for more detailed descriptions of the fuel pin designs, the reactor's mechanical, nuclear, thermal hydraulic features and performance, and the reactivity control system's features and performance.
This is reflected in the questions asked of 00E.
(2) Safety Approach - The licensing positicn taken by DOE for all these Prebreeder/
Breeder pairs is that ther e is c)mplete coi./ormance to the spirit of the General Design Criteria (GDL) established by NRC and that the features of these reactors that are unique and distinct from licensed PWR's are not major obstacles to licensing.
NRC needs considerably more information before it can draw similar conclusions in areas reflected in (3) and (4) below.
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4 10 (3) Accident Analysis - To date, there has been little accident analysis for any of the prebreeder or breeder types.
Because there is little or no experience with a number of the proposed fuel pin designs, the operational performance and transient behavior of these fuel pins is not well known.
Closely coupled to this is the response of the whole core and enlarged NSSS to such severe accidents as a LOCA.
In the unlikely event of a LWBR core meltdown, there is a possibility of recriticality for some of the fuel types proposed.
This matter should receive careful consideration.
Questions regarding accident scenarios have been sent to 00E.
(4) R&D Program Plan - The safety R&D program for the LWBR should concentrate on fuels qualification and transient behavior, thermal performance, and accident analysis.
In addition, questions regarding the hydraulic pressure equilization scheme for the control rods will have to be addressed.
A more comprehensive and better specified approach to a safety R&D program will be needed from 00E.
It would assist NRC's assessment of the LWBR considerably if the number of options (four) presently being considered by DOE could be reduced to one or two.
- d. The High Temoerature Gas-Cooled Reactor (HTGR) (PSEID Volume IV)
The original PSEID submittal for the HTGR was the standardized "GASSAR" HTGR design modified slightly to accommodate the medium enrichment fuel required so as to meet the non proliferation fuel enrichment requirements (e.g., up to 20%
enrichment for U-235).
Thus, with the exception of any safety or licensing implications from the different fuel enrichment, the large NASAP HTGR has had the benefit of rather extensive licensing review both in GASSAR proceedings and the reviews of the Summit and Fulton reactors before these reactor applications were cancelled.
Again, except for the change in fuel enrichment, we find that the facility systems description, the safety approach, the accident analysis, and the safety R&D program all are documented elsewhere and are familiar to the NRC.
For this relatively standard HTGR with a steam (turbine) cycle, as summa-rized in the HTGR-PSEID, the initial informational needs are basically met.
However, a major design modification was introduced in Addendum II to the HTGR PSEID, namely the direct cycle gas turbine. With this scheme the same helium gas that is heated in the HTGR core drives the gas turbines directly.
The gas turbines have shafts coupled directly to the main electric generators, which produce the electric power.
The emphasis of NRC's questions to DOE were on the gas turbine (GT), and the thrust of NRC's continuing review of the NASAP-HTGR will be on the safety implications of a GT design.
(1) Facility Systems Description - In the referenced Addendum and in a meeting with General Atomic (the DOE contractor for the NASAP HTGR) the plant layout for and unique features of a HTGR with a gas turbine was covered only in very general terms.
As this GT modification is in its very early stages of development, more detailed systems descriptions associated with the GT components are not available.
Questions to DOE reflect these concerns.
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s 11 (2) Safety Approach - It will be necessary to establish licensing criteria for the HTGR-GT, in particular to meet the unique aspects of the gas turbine design.
The objective of these criteria will be to assure that at least a comparable level of safety is achieved in comparison with other commercial reactors.
For example, NRC has asked DOE to provide its initial evaluation of differences in safety approach and licensing criteria between the Federal Republic of Germany's HTGR-GT and the U. S. version.
(3) Accident Analysis - The preliminary accident analysis for the HTGR-GT conceptual design will be important in sorting out problem areas and safety R&D emphasis.
Accident analyses specifically related to the GT modification, such as failure of the turbomachinery housed in the reactor vessel, were requested, as well as those " generic" accidents common to all HTGR's such as adiabatic core heat-up caused by sustained loss of forced convection cooling, internal pressure equalization accidents, control rod ejection, and core support structure failures.
(4) R&D Program Plan - The safety R&D program before the introduction of the GT was, although extensive, well understood.
With the GT modification there will have to be a considerable extension of the R&D directed to safety issues, into areas not so well understood today.
NRC thus requested additional information on the HTGR-GT research, development and testing program, including the depth of involvement of foreign participants.
NRC also emphasized to DOE that it would be very helpful to have a firm commit-ment from DOE for either the GT or the conventional steam cycle.
- e. Gas Cooled Fast Breeder Reactor (GCFR) (PSEID Volume V)
Of all the six mainline reactors under consideration by DOE and being reviewed by NRC, the GCFR is the only one for which there is no test, prototype, or comercial reactor experience.
Yet, for th ee reasons NRC is acquainted with the GCFR concept and its licensing issues:
(1) the AEC completed a conceptual design study review of a GCFR in 1974; (2) the GCFR, because it also is a fast reactor, has many core-region similarities with the LMFBR; and (3) because it is gas cooled, much of the HTGR technology is applicable.
Basically the GCFR PSEID describes a reactor system quite similar to the one
~
studied by NRC's predecessor, the AEC, in 1974.
At that time AEC considered the GCFR licensable but listed a number of major concerns, one of which was the reliability of core cooling which was dependent on a forced convection system.
DOE has recently introduced an upflow* design, as part of the NASAP-GCFR study, in order to provide a natural circulation cooling feature to the design. While NRC welcomes these types of design modifications, it should be noted that any "In the previous GCFR design, helium is forced down through the hot core in order to remove the core gene ated heat.
In the new GCFR design, helium is forced (pumped) up through the hot core.
This latter arrangement is referred to as "up-flow" and has the potential advantage of shutdown heat removal by natural circulation (no pumps (circulators) running).
1742 150
12 major design change introduces new unknowns with potential new safety and licensing issues.
(1) Facility Systems Description - This item is not as well established as for some of the more highly-developed reactors.
Up to the point of the "upflow" modifications, however, there was adequate information to proceed with this stage of the NRC review.
It is in the area of upflow modifications that considerably more information will be needed.
It is also important to understand the impact of scaling up from the 300 MW(e) plant, for which there is considerable information, to the 1200 MW(e) plant proposed in the NASAP study.
(2) Safety Approach - Assuming that upflow is adopted, it will be important for DOE to redescribe the safety approach, specific safety characteristics and design features present for the prevention and mitigation of postulated accidents.
In terms of licensing potential, NRC views the containment design criteria as important.
Thus, we noted to DOE that similar criteria to those we summarized for the NASAP LMFBR (" Bases for Containment Design in large LMFBRs") would be appropriate for a large GCFR.
(3) Accident Analysis - As with other reactor types, accidents which have the potential for adversely affecting the health and safety of the public usually involve core disruption and core melting.
Thus accident analyses must eventually include the impact of core melt and disruption on the reactor vessel and containment integrity as well as thermal margins for natural convection cooling.
In particular the accident analyses should take into account the effects of upflow.
.(4) R&D Program Plan - There are a number of points which need clarification and development in the area of safety R&D.
As noted above, the specific plan for the impact of upflow needs greater definition, as do some of the more generic, in place programs, such as reliability of vented fuel pin designs.
Over the past several years the GCFR program has relied on a combination of U. S. LMFBR and "TGR and foreign HTGR and GCFR R&D programs to considerably complement il own.
NRC needs to understand the relative importance of these programs and the impact of their termination on the GCFR program itself.
- f. The Liouid Metal Fast Breeder Reactor (LMFBR) (PSEID Volume VI)
The PSEID for the LMFBR is a description of core characteristics for 15 varia-tions on a large 1000 MW(e) sodium-ccoled fast breeder reactor, These 15 concepts include variations in core design (a homogeneous core vs a hetero-geneous core); fuel form (carbide fuel vs. oxide fuel); and fuel and blanket type (Pu/U fuel, Pu/Th fuel, U-233/U fuel; U Blanket, Th Blanket).
But, other than a description of the core characteristics for these 15 variants, there is virtually nothing presented in the PSEID that further describes the LMFBR.
It will be very difficult to factor the LMFBR into a comparative evaluation of NASAP reactor types without specific information on the NSSS and B0P and an 1742 151
13 understanding of the basic safety approach.
For example, there is no indica-tion whether the LMFBR will be a loop-type
- olant or a pot-type
- plant - a decision which will have considerable safety implications.
NRC believes that it will be unfair to the NASAP-LMFBR review for NRC to assume and use an extrapo-lation of the Clinch River Breeder Reactor (CRBR) design for making safety and licensing assessments.
The CRBR, being a loop design of an early 70's vintage does not reflect recent design innovations and improvements.
Also, a number of key safety issues associated with CRBR remained outstanding at the time of suspension of the safety review (spring 1977).
Nevertheless, NRC has gained valuable LMFBR experience in the licensing pro-ceedings for CRBR as well as for the Fast Flux Test Facility (FFTF) which were completed, for the most part, in November 1978.
Based on this experience and on the information provided in the PSEID, NRC asked questions and made comments to DOE in two broad areas:
the general design and safety approach area; and the specific core and fuel design area.
1)
Facility Systems Description - For the LMFBR descriptions were virtually non-existent.
In order to complete an adequate review we need descript'./e information on such systems as the primary heat transfer system (PHTS),
the decay heat removal system, the plant protection system and the contain-ment system.
2)
Safety Approach - Other than a statement that the LMFBR will conform to the licensing requirements established for CRBR, there was little safety approach guidance from DOE.
It is important for DOE to recognize that any one of the LMFBR-variant conceptual designs must be consistent with and conform to the spirit and intent of the staff licensing positions as reflected in the regulatory guides, criteria in the Standard Review Plan, the GDC, and other licensing regulations.
In order to aid DOE in con-sidering a safety and licensing approach for the NASAP LMFBRs, NRC provided, along with questions, a recapitulation of recent licensing staff positions in containment design entitled " Bases for Containment Design in large LMFBRs."
3)
Accident Analysis - The physics data presented in the PSEID for the variants is important, necessary input to an accident analysis and evaluation.
For example, it is valuable to know how the sodium-void reactivity coefficient "An LMFBR has three heat transport loops:
a primary sodium loop which remo"es heat from the reactor core, a secondary sodium loop, and a water / steam loop which drives the turbine generator.
In the got (or pool) arrangement all primary reactor components (core, pumps, intermediate heat exchangers (IHX's)) are immersed in the sodium coolant and located inside the reactor vessel.
In the loop design, the primary loop pumps and IHX's are located outside the reactor vessel and connected to the core region by piping.
1742 152
14 varies from one core design or fuel form to another.
However, this is only a starting poir.t for key accident sequence analyses.
It was requested that DOE provide more information on accident analyses, especially the low probability -- high consequence core disruption and meltdown analyses which are key in establishing containment requirements.
4)
R&D Program Plan - The NRC has been closely following the DOE safety R&D program for LMF8Rs over the past 10 years and has a considerable confirm-atory research program of its own.
For the Most part these programs have addressed safety issues for the conventional LMFBR design (homogeneous core, loop plant, Pu/U oxide fuel, U-blanket).
Although many of the problem areas are generic enough that they are appropriate areas for these NASAP LMFBR types, there are safety problems unique to these variants.
It will be important in the future to understand the DOE R&D program plan that will address these problems.
Areas of importance include the steady-state and transient behavior of carbide fuels, the core disruption and meltdown sequence for a heterogeneous core, and the material interaction effects between molten thorium carbide or uranium carbide and sodium or concrete.
In order to facilitate the continuing review of the NASAP LMFBR, it would be helpful for DOE to reduce the options presently in place (15) to two or three.
- g. Fuel Cycle Facilities (PSEID Volume VII)
This 00E document was written to provide information on the safety and environ-mental issues and permit URC to, identify licensing issues for all those fuel cycle facilities that would be required to service any of the reactor concepts
. described above.
These fuel cycles include:
Once through uranium fuel cycle - this is the extant fuel cycle for LWR's.
Uranium fuel cycle with (uranium and) plutonium recycle.
Thorium fuel cycle with (thorium and) uranium-233 recycle.
All of the fuel cycles considered in the NASAP program, with the exception of the once through fuel cycles, require reprocessing and recycling of fissile materials.
Operations for the front end of the uranium or therium fuel cycles are similar to one anothet and involve mining and milling of ore, production of a pure uranium (or thorium) compound, enrichment of uranium in the U-235 isotope (enrichment of thorium is not required), and use of the uranium or thorium in fuel assemblies.
In addition, thorium compounds have been produced in the past for limited use in reactors or other articles of commerce.
Because of the similarities between the front ends of the uranium and tnorium cycles, and the fact that thorium compounds have been manufactured in the past, and since the main proliferation concerns center on the back end of the fuel cycle, NRC has focused the greater amount of attention on the back end of the fuel cycle, i.e., those operations associated with recovery (reprocessing),
recycle (fabrication of recycle fuel) of fissile materials, and waste management.
1742 153
15 Reprocessing Chapter 5 of this PSEID, Reprocessing, contains a description of fuel reproces-
^
sing for five variations of uranium fuel reprocessing and seven variations of thorium fuel reprecessing.
The descriptions are almost totally qualitative, with little detail given to many important subjects, such as effluent treatment, safety features, safeguards systems, etc.
DOE notes correctly that:
There is a requirement for additional development work on technology for reducing effluents from uranium fuel reprocessing to meet current health and safety regulations.
There is little actual experience relevant to commercial reprocessing of thorium based fuel assemblies.
NRC's reviews of potential licensability of the reprocessing variations could not be detailed or in depth because the PSEID did not contain enough quantita-tive data to permit definition of appropriate licensing criteria.
Recycle Fuel Fabrication In the PSEID, DOE aggregated the discussion of recycle fuel fabrication of twelve different fuel forms into four classes of fuel, losing much information at this level of aggregation.
The data provided to NRC, to be useful, should adequately describe the differences between the processing of the fuel forms considered covering such items as emissions, radiological doses, etc.
For this operation also, NRC was unable to develop licensability criteria from the small amount of data given.
Waste Handling and Treatment The PSEID contains a statement that " current DOE National Waste Terminal Storage (NWTS) program calls for the selection by 1979 of two sites overlying suitable salt formations, followed by the construction and startup in 1985 of one NRC-licensed repository..." It appears that the program cannot be implemented on the schedule proposed.
Current NRC estimates for the time required for preparation of a license application, processing a repository license applica-tion, and reaching a licensing decision are such that the startup of a licensed repository would not occur before 1992.
Hence, DOE planning should be realistically based on a repository startup in that time frame.
International Fuel Service Centers (IFSC)
The documentation prepared for DOE on the bases for and potential problems in the design, construction, and operation of an IFSC are not available to NRC at this time.
The description of tne IFSC contained in the PSEID is too brief for NRC to comment knowledgeably on it.
1742 154-
16 Summary In general, the PSEID on fuel cycle f acilities touched on the principal safety and environmental and licensing issues that would be associated with various alternative fuel cycle systems.
In addition, it also suggested some areas where more information would have to be developed through further studies or future research.
It did not, however, make a quantitative comparison of safety or environmental trade-offs in areas such as occupational exposure, regional exposure, accident risk, or environmental, safeguards, or proliferation impacts.
NRC believes that such comparisons, even though not available at this time, will have to be developed as part of the NASAP final evaluations if any conclu-sions or recommendations are to be reached about promising proliferation-resistant alternatives.
While the DOE PSEID identified systems and principal issues related to nonpro-liferation alternatives, it did not assess the proposed systems or facilities in sufficient depth to permit definition of appropriate licensing criteria or the potential difficulty of meeting such criteria.
However, there is no reason to assume that the systems or proposed facilities could not be designed to meet the applicable local, state and Federal criteria at the time they may be proposed.
Although many of the fuel cycle systems might never reach the demonstration phase -- since they may not have enough promise to warrant appropriate funding --
the proposed facilities (within the extent of present definition) all appear to be fundamentally licensable on a qualitative basis.
2.
Nuclear Energy System Characterization Data The draft DOE document, " Nuclear Energy System Characterization Data," presents
. voluminous numerical data, together with a limited amount of analysis.
The data is primarily to be used as source material in the NASAP analyses.
General Three overall uranium fuel cycles are delineated in the document:
The once through fuel cycle is properly identified as being limited by uranium availability.
An improvement of enrichment technology is noted as a potential enhancement for the once through fuel cycle. While this may be true from a resource availability standpoint, no mention is made of potentially significant considerations with regard to the important pro-liferation and safeguards espects of such technological advances.
The second fuel cycle outlined is the uranium recycle option, which might possibly involve the LWBR. While uranium recycle is theoretically a potential alternative, in other countries it is only being actively con-sidered by those that plan large and aggressive breeder programs.
Either in the U.S. or abroad, this cycle will require facilities for the long term storage of plutonium, which are not mentioned in the DOE document.
The economic and proliferation-resistance implications of such require-ments have not been covered in any documents reviewed to date.
1742 155
9 17 The third fuel cycle considered is the uranium and plutonium recycle case.
All modes of recycle in this case seem to be based on two theoretical safeguards regimes.
One regime assumes universal worldwide safeguards; the second regime assumes two safeguards regions, one region with no restrictions and one region where only denatured, spiked, or diluted fuel can be utilized.
Thus, in this document, the concept of uranium and plutonium recycle seems to depend upon the acceptability of technologi:al measures (denaturing, spiking, or dilution) as deterrents to proliferation resistance.
Based upon progress to date in the INFCE work, NRC notes that it appears such technological measures alone are judged to have little potential for preventing national prolife ation and are only considered helpful for preventing sub-national diversion.
Reactor Data and System Analyses No new reactor types are projected to be introduced before erriv in the next century.
On this basis, the effects of new types on proliferacion resistance, resource utilization, and new facility requirements will be minimal for the next 20-25 years.
The report forecast that new reactor types could have a potential of capturing a maximum of 50% of the market for new and replacement reactors in about 11 years and 100% in about 30 years, based upon market penetration begining at the time of first commercial deployment.
The current nuclear situation adds uncer-tainty to both the penetration rate and the time of new system introduction.
Further, the introduction of any new reactor technology into the market will probably require substantial government support and resolution of institutional impediments.
Analytical results indicate that improved LWR fuel can be introduced into all reactors, both existing and projected, within a five year period from 1985-1990.
This would appear to imply complete industry penetration within five years.
In view of development, demonstration and regulatory requirements, it appears that such a fast penetration rate is overly optimistic.
The data presented in this section of the DOE document include cumulative spent fuel storage requirements through the year 2025.
From this presentation it appears that no Federal waste repository would be available to accept spent fuel until that year.
This basic assumption on respository availability does not appear to be consistent with Federal policy nor with the assumptions in PSEID, Vol. VII, Waste Treatment and Handling.
Summary The three basic LWR uranium fuel cycle options are discussed in this DOE docu-ment.
However, in the case of each option, the discussion is flawed by the lack of consideration of key proliferation aspects which would be of signifi-cant importance in evaluation and selection of improved modes of operation.
The discussion of reactor data and systems seems to take an overly optimistic view of the commercialization timing of improved LWR fuels and new reactor 1742 156
18 types in view of performance to date and the climate of increased design, development and regulatory requirements.
3.
Safeguards Considerations for the NASAP Alternative Fuel Cycles This section focuses on the potential safeguards issues that distinguish the 00E NASAP candidate fuel cycles from each other and from the reference uranium-plutonium fuel cycle.
The accompanying table on page 23 presents an overview of the 17 candidate fuel cycles and the proposed means each would employ for safeguardino against proliferation -- namely, spiking, use of U-233/ Thorium fuel, denaturing, and coprocessing.
The chart also presents a summary of safeguards issues (e.g., material control and accounting, physical security, institutional / international considerations, and storage) associated with each of these 17 candidate fuel cycles.
Safeguards issues are those which might have an impact on licensability, fuel cycle research and development, commercial implementation or operations.
These findings are preliminary and should not be interpreted as commiting the NRC to specific positions in future licensing actions.
a.
Material Control and Accountability (MCA)
The following preliminary comments summarized the major NRC concerns with regard to potential MCA safeguards issues and problems.
1)
Dynamic Real-Time Accounting The NASAP reports assume that dynamic real-time accounting processes will be used in conjunction with periodic clean-out physical inven-tories and that automated MCA systems will be available.
NRC has not yet determined the degree to which dynamic real-time accounting systems can supplement periodic clean out physical inventories.
Further, research and development will be needed to develop ways of assessing the adequacy of automated MCA systems before the NRC is prepared to make such a determination.
Accountability problems of hold-up and clean-out in high throughput facilities, particularly in radioactively hot processes, need to be examined in detail before dynamic real-time accounting or automated MCA systems can be considered ready for licensing.
2)
Spiking Four methods for spiking are proposed by NASAP.
These include the addition of radioactive material to nuclear fuel; partial reprocess-ing of spent fuel, retention of some portion of the fission product in the recovered p'utonium; pre-irradiation of fuel assemblies prior to shipment to a reactor; and mechanically attached radioactive sources.
Proposed minimum radiation levels (rem per hour at one meter), at a time two years after fabrication, range from 10 rem per fuel assembly to 1,000 rem per kilogram of heavy me-tal for Pu02 '"d high enriched UO powder or pellets.
2 1742 157
19 For the first two concepts, the radioactive spikants would greatly complicate the taking of samples for chemical analysis and somewhat complicate the analysis itself.
Most nondestructive assay (NDA) techniques would be rendered inoperable.
L3cking such techniques, it would be very difficult to perform real-time or near-real-time account-ability in bulk processing plants.
In addition, the physical inventory process would be slowed, material balances would be less accurate than in the absence of spiking, and inspection activities, both domestic and international, would be hampered.
The timeliness of both NRC and IAEA verification of material balances would be reduced.
A decision to proceed with significant research and development and commercialization of the NASAP alternate fuel cycles which require spiking would necessitate definition of spiking levels and threshold concentrations of special nuclear material (SNM) useful for producing a nuclear device.
If spiking were adopted, NRC would have to develop regulations governing its use.
3)
U-233/ Thorium Cycles There is little experience with the commercial reprocessing of highly irradiated thorium fuels.
It is therefore impossible to judge whether present NRC material accountability regulations can be met in commercial size reprocessing and fabrication plants for U-233/Th fuels.
A unique characteristic of U-233 fuels is the high radiation levels associated with U-232 and its daughters.
These levels are high enough to require remote fabrication, and non-destructive analysis techniques will have to be developed for any fuel cycle using U-233.
The feasibility of performing real-time accountability in U-233 fabrication plants will depend on the successful outcome of such efforts. As with spiked fuels, the taking of samples will be laborious and time consuming.
4)
Denatured
- U-233 Fuel For some of the reactor concepts involving denatured U-233 fuels, such as the. LWR, substantial quantities of plutonium appear in the spent fuel.
The fuel will therefore have to be reprocessed by a combination of the Purex** and Thorex** processes.
Very little, if any, experience in reprocessing denatured spent fuels exists, and
- Denaturing is defined as the addition of a non-fissile isotope to a fissile isotope in such proportions as to make the fast critical mass of the mixture impracticably large for a nuclear explosive weapon.
The Thorex process is the conventional solvent extraction process for recovering fissionable materials from thorium based fuels.
1742 158
20 therefore it is impossible to project how well NRC's accountability requirements can be met in such a reprocessing plant without detailed study.
For denatured fuels, NRC would have to review the composition data for U-233 fuels before selecting enrichment thresholds for safeguards requirements for uranium contai ing this isotope and for uranium n
containing both U-233 and U-235.
Another consideration that may enter into setting threshold enrichments for uranium containing U-233 is the greater ease of separating this isotope from U-238, compared with that of separating U-235.
5)
Heavy Water Reactors Two safeguards problems associated with the use of heavy-water reactors (HWRs) are the increased availability of heavy water in large quanti-ties and on-line refueling.
Heavy wat," can be used to moderate reactors fueled with natural uranium, and these can be used to produce plutonium.
A substantial commitment to the heavy water reactor fuel cycle in the U.S. might require, therefore, the imposition of safeguards on heavy water, not now required by NRC regulations.
Safeguards would likely consist mainly of material accounting, surveillance and containment.
The technology of safeguarding heavy water is not well developed.
Current safeguards practices place considerable reliance upon optical surveillance, item counts, and periodic checks of operating records to detect any undeclared activities.
For heavy-water production facilities of commercial size (at least 200 metric tons (Te) of 0 0 7
per year), present accountability techniques appear to be too inaccu-rate to detect the diversion frc' the extraction process of the minimum quantity (* 10 Te) of D,) required to supply the initial inventory of a small plutonium production reactor (annual production rate 8 kg Pu).
It appears that applying safeguards to a heavy-water production facility would involve problems different from those NRC has encountered in other types of safeguarded plants to date.
Con-siderable development of criteria and methods for safeguarding large plants of this type would have to be undertaken if they were to become a reality in the U.S. licensed industry.
For the heavy water reactor concept, NRC would have to determine the minimum amount of heavy water of safeguards significance, and the threshold concentration of deuterium in water for safeguards to apply.
With respect to on-line refueling, the frequency of this operation, the large inventory, and its rapid turnover impose a need for more safeguarda attention than required by current fuel cycles for light water reactors.
I742 159
21 6)
Coprocessed Fuels The safeguards issues relative to coprocessing appear similar to those for the uranium / plutonium cycle, with the exception that pure plutonium would not be stored.
The presence of uranium with plutonium would slightly complicate measurement of plutomium within coprocessing fuel cycles.
For coprocessed fuels, the concentration of plutonium in mixed oxides for thermal recycle fuels would probably be too low for direct use in a nuclear explosive. This may not be true for fast breeder reactor mixed oxide fuel, with its much higher concentration of plutonium.
The percentage of plutonium would be based on considerations of the practical needs of the fabrication plants and reactor neutronics, b.
Physical Security The present criteria defining "self protecting" SNM (i.e., not readily separable from other radioactive materials and with a radiation dose rate of more than 100 Rem per hour at 3 feet) are under study within NRC.
Based on these study results, NRC may need to consider any revisions to the existing protection requirements needed for fixed sites and transporta-tion from the standpoints of both theft and sabotage for licensed SNM, spikants, spiked SNM, and high level wastes.
DOE should coordinate with NRC to assure that NASAP review criteria remain consistent with future study results.
If the plutonium recovered from spent denatured fuel is recycled in energy centers, the safeguards technical problems are essentially the same as those discussed in the NRC GESMO proceeding, with the modifications associated with the physical and administrative nature of energy centers.
NRC and DOE have ongoing research programs to characterize the potential radiological source that could arise from sabotage (explosive breaching) of a loaded LWR spent fuel shipment and to estimate potential health consequences.
Additional research may be required to characterize the source term and health consequences of explosive breaching of spiked fuel shipments, shipments to which a spikant sleeve has been mechanically attached, or fuel shipments associated with the remaining candidate fuel cycles.
c.
Institutional and International Considerations 1)
Spiking A significant safeguards issue presented for NRC consideration in NASAP is the effectiveness of spiking fuel as a safeguards measure against nuclear weapons proliferation.
It can be postulated that U.S. exports of unirradiated fuel that afford nuclear weapon potential could be spiked to emit radiation levels comparable to spent reactor fuel.
If the importing country plans to use this material for the 1742 160
22 development of nuclear weapons, the fuel would first require processing by remote means, imposing processing requirements equivalent to those for reprocessing of spent reactor fuel.
Remote processing appears to be a relatively minor hurdle to overcome for any technologically advanced nation that decides it needs nuclear weapons.
Thus, spiking does not appear to reduce the risk of national proliferation of nuclear materials for diversion to nuclear weapon development; it merely assures that such overt action is not made easier to accomplish than it would be from the spent fuel, which many nations already possess.
Spiking may have substantial adverse effects on the environment thggughtheproductionofincreasedamountsofradioactivity(e.g.,
Co ) beyond those normally required for the generation of electric power.
An environmental impact statement may therefore be required prior to a national decision to implement spiking and an environ-mental impact proceeding probably would involve appreciable delays.
2)
Heavy Water If heavy water were to be safeguarded in the interests of non-proliferation, the definition of a safeguards significant amount of heavy water should be consistent with international commitments.
Safeguards on heavy water are not required under the NPT-INFCIRC/153 system of the IAEA, but may be required under bilateral or trilateral agreements or voluntary submissions.
Historically, the IAEA has accepted the responsibility for applying safeguards to heavy water when so requested.
It is not regularly safeguarded under NPT because heavy water is not included in the definition of nuclear material by the IAEA.
d.
Stora,ge A once-through fuel cycle implies the indefinite, perhaps permanent, storage of spent reactor fuel in repositories.
NRC believes it already possesses legal authority to license the storage of spent fuel in Away-From-Reactor facilities because spent fuel is considered to be a high level waste for the purposes of Section 202(3) of th Energy Reorganization Act of 1974 (ERA).
Explicit legislative confirmation of this authority in the proposed H.R. 2586, the Spent Nuclear Fuel Act of 1979, would prec'ude confusion on this point.
Disposition of the plutonium separated from spent denatured U-233 fuel is also significant from the standpoint of proliferation.
If this plutonium is considered waste, NRC believes it would be subject to licensing authority under Section 202(3) of the ERA.
Plutonium recovered from spent denatured fuel and separated U-233, from U-233/ Thorium and, possibly, from denatured fuel cycles, may be either stored or recycled in " secure" energy centers.
If storage is in licensed facilities, the safeguards probleas will be, generally, those already considered in the GESMO proceeding.
2 lbf
I XISl!NG OR ',USPECllD SAllGUARDS kt t r.!!DNSHIPS IUM filf NASAP IUf t CYCLES EXAMINIO PROPUS10 AiltNN IVL ltOINUtOGICAL APPRUACHt5 10 NUNPROL Ill RAll0N PO!ENIIAL IMPACl5 ON SAILGUARDS IUtt CYCifS A.
SPIKihG
- b. U-231/th IUt t C. DENA!UNING D. CDPROCES$1NG A. MCA
- 8. PHYSICAL S[CUN!1Y C.
IN5111/INItt D. STORAGE CUN51DLRATIONS tight Water kcastoe>
- l. I PWH Uut e lia suyh (U))
l.2 PWR Modilled UI X
3 X
X7 X
X X
X I.1 PWR U/Pu Neayole 1.4 IWW Denatured U 231 Ret.ycle X
X X
X X
X I !'l _t Wa t e;r_ tire;edes Rede tury
- h 2.1 11as k r i t l'a e-us eedes U/U-233 Recycle X
X X
X
- 2. 2 Advanteit P e-Drecatei U-211-th/U 213 Recyc le X
X X
X X
2.1 Pee-Ureeder U l h/U-2 3 3 Rec yc le X
X X
X 2.4 Beceder UU -lh/U-233 Retycle X
X X
X X
j RC'yy W.it e;r Reagor
- 3. 0 llWN Denatured Ul X
X X
X I!i' 8. h ay G.9 {anf ed ke.jt ury l
t ev J3
'd 4.1 tilLN Denatur ed Ul X
X X
4.2 lil6N Den.st. R.:t yt. le U - 2 35-f li/U-23 3 X
X X
X X
G.ss luuled I.sst Neattur S I LLi k U-Pu/U.*13 Ret yt te X
X X
X X
X X
t lej Netal last tirseder He.Htors b I iMlUN O Pu/U Netytle X
X
- te. 2 L M8 DR tl-Pu ket yc le X
X X
X X
N b. 3 IMibN U-Pu/lh Retyt.le X
X X
X X
X gb4 I Mi t!N lla-Pu/ th Net yt le X
X X
X X
X K
6.5 LMittN Denat. U 21'c l h/lh ket yc le X
X X
X X
g
-1/ The potentaal s.s t egua r ds i mpat t s a s t eijos I d est lie s'e apgily to the d iles n.st ive tise t t yc les and ene ept for sturoge (uncerns du saut.spply to the ourrent t WR fi c l a yc le.
- /
Y NASAP Neattus/luel/Cy<le Systems for NRI' keview.
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- the X's put t r ay ex is t a ny or suspes. teil s e l.stlunship.
- etween tiee f uc t ty( les examined, t he-t es.hno log ic a l 4"pruat tees t o pueeps u l i f era t iosi, eint the potentsai lug at t s nu :..s f ego.o r.l,
a
24 e.
Summary As noted, the findings presented herein are preliminary in nature and should not be interpreted as committing the NRC to specific positions in future safeguards licersing actions.
More definitive statements on safe-guardability of alternative fuel cycles will be presented in future reports on the basis of r? view of the NASAP and INFCE final reports, further fuel cycle definition, research and development by 00E and NRC staff work.
4.
The Denatured U-233 Feel Cycle The report, " Interim Assessment of the Denatured U-233 Fuel Cycle:
Feasibility and Nonproliferation Characteristics," ORNL-5388, was reviewed as a NASAP support effort. While the impacts of U-233 usage on reactors were not con-sidered, the document is an exhaustive study of the denatured U-233 fuel cycle and contains analyses of many denatured fuel systems.
The analysis presents voluminous data on the uranium, uranium / plutonium and denatured uranium fuel cycles covering eight fuel cycle options with four cases being studied for each option, and two levels of uranium supply (high and intermediate) used in each case and option.
A total of 64 cases were studied.
Nuclear Power Demand The demand curve for nuclear power used in the analysis presented in this report required that nuclear power reactors be capable of supplying 1100 GWe in 2049.
Not unexpectedly, for the high-cost U 0 supply runs (constrained resource 3g availability), only those fuel cycle options that include breeders are able to meet the demand.
For the intermediate-cost U 0 runs (larger resource avaH-3g ability), options using heavy water moderated converters are also capable of meeting the demand.
Although NRC recognizes the need to consider the ultimate capability of any resource base in terms of GWe produced /MT U 0, it does not appear reasonable that a projection of installed nuclear poweh h0 years from now is a realistic decision criterion on which to base near-term (1980) decisions.
NRC believes that the near-term decisiens should include consideration of the transitional period (say 2000-2020) showing changes in the nuclear fuel cycle as new types of reactors, instead of LWR's, are built.
NRC recommends that some detailed evaluations of the differences among the options over the nearer term transition period be developed and factored into the decision process.
" Secure" Centers With the exception of the throwaway option, all other opti.ons considered in ORNL-5388 require fuel reprocessing of LWR e.nd other types of fuel.
- Further, all options except the Pu throwaway option recycle plutonium to reactors built on a " secure" center together with fuel reprocessing facilities.
The charac-teristics of the " secure" center are not completely delineated.
The restric-tion of certain reactors to secure centers can be a potentially significant factor in the growth of nuclear power.
The ratio of the installed nuclear capacity outside an energy center to the installed nuclear capacity inside the i742 l63
25 center, termed " energy support ratio," is a measure of the restrictions in nuclear power plant siting attributable to the requirement for secure centers.
NRC recognizes the desirability of obtaining the maximum amount of information from the complicated systems studies, and therefore considers the data on the energy support ratio valuable.
However, data must be available to consider the nuclear power fuel cycle and reactor complex in the early years (say 2000-2020) of the next century. Options that appear to be very different in the mid-21st century may be quite similiar in the early years of the next century, and very likely near term considerations may dominate the selection of alternative fuel cycles.
Hence, a discussion of the energy support ratio' over the early years of the next century as well as an evaluation of the limiting ratio should be included.
In addition, the requirement for a" secure" center is an assumption in Chapter 6 of the ORNL document.
Since this assumption does not affect the results of systems studies, in terms of numbers of reactors, development costs, etc., it would appear desirable to clearly indicate that this assumption does not impact on other study results.
(The energy support ratio is, of course, dependent on the assumption.)
Summary This support document projects the nuclear power industry 70 years in the future and considers over 60 alternative fuel cycles.
Thus it is both too diffuse and long range in its perspective to provide much insight into near-term decision making.
It is believed that the transitional phases of some long-term alternatives in the 20-30 year time frame would lixely be very significant in the decision making process and should be analyzed in the NASAP
. effort.
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26 III.B
SUMMARY
OF NRC PARTICIPATION IN INFCE In October 1977, forty countries and four international organizations met in Washington and agreed to undertake an International Nuclear Fuel Cy:le Evaluation (INFCE).
INFCE, which now has fifty-three participating countries and is scheduled to conclude in February 1980, is organized around eight working groups and a Technical Coordinating Committee made up of representa-tives of the twenty-two countries chairing the eight working groups.
The eight working groups and their "co-chair" countries are:
1.
Fuel and Heavy Water Availability (Canada, Egypt, India) 2.
Enrichment Availability (France, Federal Republic of Germaay, Iran) 3.
Supply Assurances (Australia, Philippines, Switzerland) 4.
Reprocessing, Plutonium Handling, Recycle (Japan, United Kingdom) 5.
Fast Breeders (Belgium, Italy, U.S.S.R.)
6.
Spent Fuel Management (Argentina, Spain) 7.
Waste Management and Disposal (Finland, Netherlands, Sweden) 8.
Alternative Fuel Cycle and Reactor Concepts (Republic of Korea, Romania, United States)
NRC's participation has been in the context of the U.S. organizational support structure for INFCE.
Individual staff members have been named as contacts to work with the U.S. support groups for six of the eight INFCE working groups (numbers 3-8) and three special U.S. cross-cut review groups (safeguards; nonproliferation; technical and economic aspects).
These personnel have reviewed, along with specialists from the Department of Energy, the Department of State, the Arms Control and Disarmament Agency, the Environmental Protection Agency, and other interested organi-zations, various U.S. and foreign papers circulated by the U.S. INFCE Coordinator's Office.
The U.S. has offered about 50 papers to the INFCE working groups and has reviewed and commented on more than 200 papers contributed by other countries and international organizations.
This combined total of 250 papers was preceded by several hundred preliminary drafts and discussion papers on sub-topics, most of which were also reviewed in whole or in part by members of the U.S. support structure.
NRC staff member comments have been of two types, those of a general technical nature and those concerning regulatory considerations.
The latter comments have included licensability questions:
when and under what circumstances might a given type of technology or control system be 1742 165
27 considered sufficiently well understood and supported by adequate data to permit it to be licensed and used in the U.S.?
Such comments on alterna-tive reactor and fuel cycle concepts have been provided by NRC in the NASAP context (see discussion above), but have been repeated in key areas as appropriate in INFCE.
In tnis regard, it is noted that one purpose of NASAP was to provide technical support to INFCE, especially to Working Group 8 of which the U.S. is a co-chair country.
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28 III.C.
GAO REPORTS In a letter report to the Joint Economic Committee of Congress, dated March 7, 1978, GA0 recommended that the NRC (1) establish a program to systematically and independently monitor the development of alternative fission reactor and fuel cycle technologies for the future, and (2) identify and report to the President and cognizant t.ongressional committees known or suspected licensing issues and problems associated with the reactor and fuel cycle technologies under serious consideration by DOE before any are scheduled to be selected for future development.
GA0 also recommended that, to the extent possible, the Chairman of the Commission should rank the reactor and fuel cycle technologies for desired development in the United States from a licensing point of view, and clearly identify the relative safety, safeguards, and environmental advantages and disadvan-tages of each.
In letters to GA0 and various Congressional Committees dated June 9, 1978 the Chairman of NRC responded to the GA0 recommendations.
NRC indicated that it has been monitoring the alternative fission reactor and fuel cycle technologies described in the Nonproliferation Alternative Systems Assess-ment Program (NASAP) as well as those in the International Fuel Cvele Evaluation (INFCE).
Comments have been and will continue to be sent to 00E on their NASAP documentation.
NRC has established a NASAP/INFCE working group to coordinate the Agency's activities in alternative fuel cycle assessments and research efforts.
NRC has agreed to provide a staff report to the President and Congress on preliminary findings of known or suspected licensing issues or problems with alternative technologies under serious consideration by 00E.
NRC plans to publish the report subsequent to the completion of the NASAP program and the INFCE study.
A future edition of these semi-annual reports to Congress may serve as the vehicle for reporting to the President and Congress on these matters.
In late 1978, NRC reviewed a draft of a report produced by GA0 entitled
" Nuclear Reactor Options to Reduce the Risk of Proliferation and to Succeed Current Light Water Reactor Technology." In January, 1979, NRC provided GA0 with comments on this draft report.
Although the report contains no recommendations or conclusions directed at NRC, the Agency did provide GA0 with comments to improve the accuracy and completeness of the report.
For example, NRC staff pointed out technical inaccuracies and conclusions requiring additional supporting verification and offered additional perti-nent information where appropriate.
A final version of this report was published by GA0 in May 1979.
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29 III.D.
SUMMARY
OF NRC COMMENTS ON THE CONGRESSIONAL RESEARCH SERVICE REPORT:
ALTERNATIVE BREEDING CYCLES FOR NUCLEAR POWER:
AN ANALYSIS NRC reviewers of this report found it to be of high quality and comprehen-sive in its treatment of the various alternative breeder concepts.
- Also, some of the views expressed in the report appear to be in general agreement with those of the NRC.
The NRC has reviewed this report in part to monitor the degree of technical accuracy in " companion" efforts to NASAP/INFCE and in part to provide perspective for NRC efforts as well as constructive comments to the authors of the CRS report.
A meeting was held at NRC April 11, 1979, with a CRS representative during which such comments were exchanged.
The report does not deal as well with the ex-reactor portion of the fuel cycle as it does with the reactor concepts.
Since it is the vulnerability of the special material while it is outside the reactor that gives rise to proliferation concerns, a more appropriate focus, therefore, would have been on alternative braeder fuel cycle systems.
It is noted in the report that there seems to be little "discernability" between the proliferation risks of one alternative breeder concept and another and that for all of the concepts the risks appear to be relatively low but not yet quantified or compared with other risks of public concern.
The report also noted that there is somewhat of a " gap" or lack of meaningful connection between nuclear power breeder reactors and nuclear weapons proliferation.
That is, the scenarios describing diversion of special material and the creatica of a threat have not been critically reviewed in depth or on a comparative basis.
Although the report notes this problem, it does not set forth criteria for determining the risk from such scenarios relative to others, including non-reactor scenarios, nor does it discuss how to judge whether disproportionate applications of safeguards may occur with one of several routes to proliferation.
These considerations still need to be given attention and review for adequate Congressional and public understandings.
To the extent allowed and useful, NRC will continue efforts to coordinate reviews and advice on safety concerns with the CRS as they progress in their efforts to clarify various proliferation scenario facts for Congressional decisions.
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30 IV.
SUMMARY
OF APPLICABLE RESEARCH AND TECHNICAL ASSISTANCE NRC has developed short-range research and technical assistance programs to help identify outstanding safety, safeguards and environmental issues associated with alternative reactor and fuel cycle concepts.
During FY 1979 approval was requested and obtained from Congress to apply $800,000 of FY 1978 carryover funds to this work as a part of ongoing programs at several National Labora-tories; however, with Congressional approval NRC delay this research effort because of resource needs precipitated by the Three Mile Island accident.*
NRC has implemented a modest (for FY 1979, 5300 K) technical assistance program directed to a better understanding of some of the licensing issues unique to the NASAP conceptual designs.
These areas include:
1.
Failure dynamics of HWR pressure tubes and implications of break before leak; 2.
The reactor physics characteristics of the HWR; 3.
Accident sequences peculiar to the HWR; 4.
The variation in core disruption consequences as a function of the varia-tions in the LMFBR designs; 5.
Accident sequences peculiar to the HTGR design with a gas turbine; and 6.
Core melt and core retention characteristics for fuel forms and fuel cycles different than plutonium and uranium mixed oxide.
In FY 1979, NRC also contracted with Brookhaven National Laboratory (BNL) to provide a preliminary analysis and assessment of the requirements for safeguards and tr-difficulty of safeguarding the various alternative nuclear fuel cycies being considered by 00E in NASAP.
The BNL study is intended to identify and evaluate safeguards regulatory issues and potential solutions which could occur given implementation of the NASAP alternative fuel cycles.
To date, the study has provided an overview and preliminary assessment of generic safeguards issues, potential safeguard's problems, and alternative solutions for each fuel cycle.
The study has explored the impact of the current regulatory structure upon the time required to implement each cycle.
Remaining tasks address ranking of the major alternative fuel cycles by difficulty of safeguards, impact of legal obstacles that could restrain implementation of the fuel cycle options, impact on implementation of the USfblEh safeguards agreement, and impact on operation of multinational fuel center' 1-the United States.
- Subsequent to the delay, Congress directed NRC to Y,ma:
'.]00 for fuel cycle studies.
This authorization was received in August 1S'79, after the end date of this Report (June 30, 1979).
I742 i69
31 When completed in early FY 1980, this $154 K study will provide background technical information for the NRC to use in establishing future progra.;A and policies relevant to the safeguarding of fuel cycles developed by the commercial nuclear industry.
NRC also maintains a significant research program in fast breeder reactors and a limited program in advanced converter reactors, funded at $14 million and
$3.2 million respectively in FY 1979 and supported by about 14 person years o' NRC staff etfort.
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32 V.
ADDITIONAL COMMENTS Section 9 of Public Law 601 (FY79 NRC authorization legislation) states that, as part of the Report to Congress, NRC shall provide:
"...a summary of all research activities relating to such matters, identification of areas of deficiency in such existing information and ongoing efforts, and recommendations regaroing matters wnich should be undertaken to minimize any potential adverse impact on matters within the Commission's jurisdiction;"
It has been noted elsewhere that this first semiannual report is based on only a preliminary review, over about a 6-month period, of the DOE PSEIDs.
As such, a definitive " summary of all research activities", including a listing of deficiencies in present programs and recommendations for future work, cannot be part of this first report.
A.
Reactor Types As a first step, in Section III. A, we have indicated the status of information provided to us by DOE for each reactor type in four general categories --
facility system doscriptions, safety approach, accident analysis, and the research and development program.
The reader is referred to these portions of the summaries for the status of information provided for the specific reactor types.
Until more information is provided by DOE in the first three information categories, the NRC can neither be more definitive in evaluating the merits of an in place plan nor be more responsive in estimating safety R&D costs and durations.
Nevertheless, we have provided estimates of trends, emphasis, and
.needs.
B.
Fuel Cycle Concepts As we noted above (Section III.A.), the information received from DOE in support of an NRC evaluation of the NASAP concepts did not assess the proposed systems in sufficient depth to permit definition of appropriate licensing criteria.
For NRC to provide meaningful evaluations related to health and safety, safe-guards, and environmental considerations, documents that approach the level of detail of an applicant's report would be required.
Since fuel cycle operations tend to te generic in nature, it is suggested that DOE choose a few typical fuel cycles that it would like NRC to comment on and define them in sufficient depth for elCLation purposes.
To permit NRC to provide definitive opinions on alternative fuel cycles, DOE should provide detailed analyses of the safety, safeguards, and environmental effects of typical fuel cycle facilities, and a comparison of the effects of such facili-ties with those of the once-through LWR fuel cycle.
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33 VI.
GLOSSARY ACRS
- Advisory Committee on Reactor Safeguards B0P
- Balance of plant CANDU Canadian Deuterium Uranium CE
- Combustion Engineering CFR Code of Federal Regulations CRBR Clinch River Breeder Reactor CRS Congressional Research Service D&D Decommissioning and Decontamination ERA
- Energy Reorganization Act FFTF
- Fast Flux Test Facility GASSAR - General Atomic Standard Safety Analysis Report GCFR Gas-cooled fast breeder reactor GDC
- General design criteria GESMO - Generic Environmental Statement on Mixed Oxides GT
- Gas turbine GWe
- Gigawatts electric HTGR
- High temperature gas reactor HWR
- Heavy water reactor IAEA International Atomic Energy Agency IFSC
- International Fuel Service Centers INFCE - International Nuclear Fuel Cycle Evaluation LMFBR - Liquid metal fast breeder reactor LOCA
- Loss-of-coolant accident LOHSA - Loss-of-heat-sink accident LOPA
- Loss-of-regulation accident LWBR Light water breeder reactor
. LWR
- Light water reactor MCA
- Material control and accountability MWD /MT - Megawatt days per metric ton MW(e) - Megawatts electric NASAP - Nonproliferation Alternative Systems Assessment Program NDA
- Nondestructive assay NNPA
- Nuclear Non-Proliferation Act NPT
- Nonproliferation Treaty NWTS
- National Waste Terminal Storage NSSS
- Nuclear steam supply system ORNL
- Oak Ridge National Laboratory PHTS
- Primary heat transfer system PSEIDs - Preliminary Safety and Environmental Information Documents PWR
- Pressurized water reactor RCS
- Reactor coolant system R&D
- Research and development SNM
- Special nuclear material i742 172
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