ML19343C378
| ML19343C378 | |
| Person / Time | |
|---|---|
| Issue date: | 02/28/1981 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | |
| Shared Package | |
| ML19343C377 | List: |
| References | |
| FACA, NUREG-0751, NUREG-751, NUDOCS 8103230599 | |
| Download: ML19343C378 (61) | |
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-M UA soucLEAR RESULATORY C0bastges0N NUREG-0751 BIBLIOGRAPHIC DATA SHEET 2.Eomeweef TITLE AND SueTITLE 64sw vebme Na. sf amprepaew1 4 Review and Evaluation of the Nuclear Regulatory Comission Safety Research Program for Fiscal Year 1982
- 2. RECIPIENTS AccEssice.No.
- 5. DATE REPORT COMeLETED
- 7. AUTHORISI woNTn lvsam February 1981 bMORMING ORGANIZATION N AME AND MAILING ADDRESS # ache le Com/
DATE REPORT ISSUED Advisory Committee on Reactor Safeguards F"ebrua ry l'1981 o*'"
'^a U.S. Nuclear Regulatory Comission s Ee.e==2s Washington, DC 20555
- 8. Eeme w ohl
- 12. SPONSORING ORGANIZATION NAME AND M AILING ADORESS #acke le Ceeiel
" ' * " ' " * * ' " ~
Same as 9, above.
- 13. TYPE OF REPORT pe n.oo cova nto #achsae asesJ Report to Congress Yy 1982
- 14. Eeme'Weet
- 15. EJPPLEMENTARY NOTES
- 16. ABSTRACT 200 werws or apast Public Law 95-209 includes a requirement that the Advisory Comittee on Reactor Safeguards submit an annual report to Congress on the safety research program of the Nuclear Regulatory Comission. This report presents the results of the ACRS review and evaluation of the NRC safety research program for Fiscal Year 1982.
The report contains a number of coments and recomendations.
- 18. AVAILAgILITY STATEMENT
- 19. MCURITY CLASS ITAs wertl
- 21. NO. OF PAGES unclassifisd l
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8103230599
NUREG-0751 Review and Evaluation of the l\\ uclear Regulatory Commission Safety Research Program for Fiscal Year 1982 A Report to the Congress of the United States of America ate u ished e u 1 1 Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555 p.~%,
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'o UNITED STATES
'n NUCLEAR REGULATORY COMMISSION
{
ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS 0,
g WASHINGTON. D. C. 20555
- '+,
,d February 12, 1981 The Honorable George H. W. Bush The President of the Senate The Honorable Thomas P. O'Neill, Jr.
The Speaker of the House Gentlemen:
I am pleased to transmit to the Congress the report of the Advisory Committee on Reactor Safeguards on the Nuclear Regulatory Commission's safety research program for fiscal year 1982.
This report is required by Section 29 'T the Atomic Energy Act of 1954 as amended by Section 5 of Public Law 95-209.
Part 1 of this report is intended to serve as the Executive Summary.
A copy of this report is being ser.c to the Chairman of the Nuclear Regulatory Commission.
Respectfully submitted, J. Carson Mark Chairman
r TABLE OF CONTENTS PAGE i
PREFACE.............................................................
vii PART 1 GENERAL RECOMMENDATIONS 1.
Introduction................................................
3 2.
Need for Research on Systems Reliability and Plant Behavior.
4 3.
Priorities..................................................
5 4.
Budget Recommendations......................................
7 5.
General Comments............................................
8 6.
Specific Comments and Recommendations.......................
9 TABLE 1 - Proposed FY 1982 Budget...............................
10 PART 2 SPECIFIC COMMENTS 1.
LOCA AND TRANSIENT RESEARCH.................................
13 2.
LOFT........................................................
19 3.
PLANT OPERATIONAL SAFETY....................................
20 4.
SEVERE ACCIDEMi PHENOMENA AND MITIGATION RESEARCH...........
25 5.
SITING AND ENVIRONMENTAL RESEARCH...........................
28 6.
WASTE MANAGEMENT............................................
32 7.
SAFEGUARDS AND FUEL CYCLE SAFETY............................
34 8.
SYSTEMS AND RELIABILITY ANALYSIS............................
38 APPENDIX A - REFERENCES.............................................
45 APPENDIX B - GLOSSARY...............................................
46 APPENDIX C - ACRS CHARTER AND MEMBERSHIP............................
48
.y
PREFACE This is the fourth report by the Advisory Committee on Reactor Safeguards (ACRS) 'that has been prepared in response to the Con-gressional requirement for an annual report on the Nuclear Regulatory Cormiission (NRC) reactor safety research program. As was requested by the Congress last year, this year's report has been delayed until the proposed budget for FY 1982 has been submitted to the Congress and reviewed by the ACRS.
As in previous reports, the ACRS has interpreted the words " reactor safety research" to include safety-related research in all phases of the nuclear fuel cycle and power plant operations, excluding only that having to do with nonsafety-related environmental concerns.
Part 1 is a compilation of our general recommendations regarding the NRC safety research program, and includes an identification of areas where higher priority of efforts should be directed and our budget recommendations.
It is intended to serve as an Executive Summary.
Part 2 is divided into eight chapters, aach of which represents a Decision Unit of the NRC research progra.
In each chapter, we have included specific comments on the resear ' involved in the Decision Unit, an assessment of priorities, and re enendations regarding new directions and levels of funding.
All references to funding in this report rel, to funds budgeted for program support.
Funds allocated for NRC p.
'onnel, administrative support, and eqeipment have not been included.
In the preparation of this report the ACRS did r., ' take into account the reallocation of responsibilities and resources which may occur as a result of Public Law 96-567, Nuclear Safety Reset 'h a'nd Demonstra-tion Act of 1980.
vii
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P!LU 1 GE2EPJL RECCt1EDATIOS 1
1.
Introduction In its report to the Congress on the Nuclear Regulatory Commission (NRC) Safety Research Program for FY 1981 (Ref.
1)*,
the Advisory Committee on Reactor Safeguards (ACRS) stated that greater emphasis than was then planned by the NRC should be placed on studies of se-rious accidents and on studies of plant operations and systems be-havior.
We a'so singled out two then existing budget Decision Units for substantial increases in funding beyond the level included in the proposed budget; namely, Risk Assessment and Research to Improve Re-actor Safety (which would have included studies of conceptual designs for improved containment systems for serious accidents and imp. oved shutdown heat removal systems).
We recommended that the lacter increases in funding be provided, even if it should require a can-comitant reduction in other portions of the overall safety reseat ch program.
Those recommendations on funding were not adopted by the Congres, in the appropriations for FY 1981, and the NRC has failed to give the emphasis and priorities recommended by the ACRS.
Although increases have been made in these areas in the proposed FY 1982 budget, they are not considered adequate.
We believe that these actions are to the disadvantage of improved reactor safety for several reasons including the following:
(a) The NRC Commissioners and Staff will continue to lack important information on design alternatives to mitigate the consegaences of degraded core and core melt accidents; infonnation important for reactors in operation and under construction as well as for the related rulemaking process which is underway.
(b) The enhancement of safety which might be achieved by improved shutdown heat removal systems will be delayed still further.
(c) The enhancement of safety which might have been achieved by greater efforts on reliability analysis of plants in operation and under construction will be unavailable.
(d) The greater depth of knowledge which should be achieved about plant operational behavior, including the impacts of control systems on safety, will be delayed.
- References appear in Appendix A.
3
2.
Need for Research on System Reliability and Plant Behavior Operational experiences since the Three Mile Island Unit 2 (TMI-2) accident sho,< a need for continuing investigation of means to improve nuclear plant system behavior and reliability.
Areas deserving consideration include:
(a) Establishing that core cooling continuity will not be threatened by circumstances similar to but more serious than the following incidents:
On May 19,1979, the Oyster Creek boiling water reactor (SWR) experienced a transient in which the operator incorrectly isolated the core from its source of water for shutdown heat renoval and an important set of water level instrumen-tation provided erroneous information, On June 11, 1980, the St. Lucie Unit 1 pres-surized water reactor (PWR) unexpectedly encountered steam bubble f mation in the reactor vessel head during a transient which involved natural circulation cooling, and On May 20, 1980 and on August 12, 1980, the Calvert Cliffs Unit 1 PWR essential service water system failed because of air binding due to an air compressor failure.
(b) Assuring that themal shock or overpressure avents, separately or in cocbination, cannot jeopardize pressure vessel integrity by an event similar to but more severa than the incident on October 17, 1980 at the Indian Point Unit 2 PWR when the reactor vessel cavity was flooded, subjecting the reactor vessel to the possibility of undesirably ow temperatures at elevated pressure due to a combination of failures.
(c) Providirg tolerance for and decreasing likelihood of accidents involvi.ia failure of reactor scram systems of types comparable to but oc 9 severe than the June 28, 1980 event i1 which the Browns Ferry Unit 3 BWR suffered a partial failure to scram due to a malfunction i.- the hyaraulic discharge system.
This partial listing of reliability matters deserving attention indicates an urgent need for an expanded research effort to improve plant reliability.
4 W
t 3.
Priorities It is disconcerting, to say the least, that despite a research program with funding levels which have ranged from $100 million to $200 million a year, the possib.lity had not been recognized that the i
hydraulic discharge of a BWR scram systc,a might be as vulnerable as it proved to be.
The suitability of priorities in the research program must be questioned.
Although the Risk Assessment Review Group (Ref.
- 2) and the ACRS have recommended that the NRC reevaluate its current and proposed research programs in terms of risk reduction potential and major regulatory needs, this is not reflected in the proposed budget.
As a consequence, about $80 million a year continues to be spent on loss-of-coolant ace.ident (LOCA) research and about $20 million on fuel behavior experiments, much of which represents neither significant risk reduction potential nor major regulatory needs, while a total of less than $2 million is devoted in FY 1981 to improved shutdown heat removal systems and conceptual systems for mitigation of serious accidents.
In addition, there is little current effort being directed to trying to anticipate and prevent events such as those listed in Section 2 above.
Furthermore, although it has been clear for some time that major changes are required in the single failure criterion and possibly in the safety role attributed to control systems, these are being treated as long range rather than priority efforts in the NRC research program.
Only a very modest beginning of work is scheduled for FY 1982.
All of the safety matters mentioned above should, in principle, be addressed in the design of any proposed new light-water-reactor power plant.
However, in the absence of general guidance from the NRC, the nuclear industry finds itself in a position where it is difficult to assess what the licensing requirements will be, although it could be forward-looking and take considerable initiative in this regard.
In previous reports to the Congress (Refs.1 & 3), we have taken the point of view that all of the safety research being proposed by the NRC was generally useful but have also recommended areas which should be given priority or increases in funding. This year, we are adopting the point of view that budgetary constraints will exist and that it is unlikely that an NRC safety research program budget significantly larger than that recommended in the proposed budget will be approved by the Congress, unless a specific addition is made to provide for safety research on liquid metal fast breeder reactors (LMFBRs).
Furthermore, a situation tnat has occurred repeatedly, and can be expected to occur again during FY 1982, is that, by then, many of the 5
major NRC regulatory needs will have changed, and different issues of considerable significance will warrant higher priority and more em-phasis than is now foreseen.
This is a hallmark of a safety research program which is responsive to changing needs.
Hence, we recommend that the Congress direct the NRC to place greater priority and in-creased emphasis, including funding support, on the matters identi-fied above and reiterated below.
We recommend also that the Congress provide the NRC considerably greater latitude in shifting funds between budgetary Decision Units.
In summary, we recommend that the following areas be given higher priority and greater funding in FY 1982:
(a) The role of control systems in safety; (b)
Plant operational safety, including system behavior as a func-tion of design; (c) Reliability analysis of existing plants, including emphasis on the more detailed understanding which might prevent many operating occurrences which have potentially serious implica-tions; (d)
Improved, more reliable shutdown heat removal systems, including dedicated and bunkered systems; (e)
Studies of degraded core and core melt accidents with emphasis on the conceptual design and evaluation of features to mitigate such accidents; (f) Studies of the physical and chemical behavior of fission products in post-accident environments; (g) The early development of an approach to supplement or replace the single failure criterion; and (h) Other matters relevant to setting the principal design bases for plants to be constructed.
In addition, we recommand that there be a viable NRC program in LMFBR safety research, unless the Congress judges that it is highly unlikely that the NRC will be involved in any new LMFBR licensing efforts in the next few years.
6
4.
Budget Recommendations 4.1 Base Budget In order to provide adequate funding for the high-priority matters identified above, within the level of total program support requested in the proposed budget, we recommend reallocation of funds among the Decision Units as shown in Table 1.
The amounts we recomend for the programs on Plant Operational Safety, Severe Accident Phenomena and Mitigation Research, and Systems and Re-liability Analysis are increased by a total of $18.5.aillion over the amounts proposed.
Offsetting reductions are recommended in Decision Units 1, 2, and 6.
We recommend a modest reduction of $1.5 millior, in the pr 79 ram on High Level Waste Management (Decision Unit 6). We believe that this reduc-tion will have little effect on the progress of this rapidly expanding program.
In Decision Unit 1, we recommend a reduction of $1.0 million in the program relating to Code Assessment and Application, and a reduction of $2.0 million in research on Fuel Behavior Under Operational Tran-sients.
The bases for these reductions are discussed in Sections 1.3.5 and 1.3.6 of Part 2 of this report.
The remaining $14.0 million can be taken from the LOFT program (Deci-sion Unit 2).
We have recommended to the Commission in July 1980 (Ref. 4) that the LOFT test program be terminated after FY 1982 and we repeat that recommendation here.
If this schedule is followed, the high operational and overhead costs associated with this test facility can be reduced significantly in FY 1982 and the sum of $30.0 million that we recommend in Table 1 should be sufficient to pcanit four to six tests. We believe that this number of tests, if carefully selected, will serve to bring this program to a satisfactory conclu-sion.
The foregoing recommendations, as summarized in T able 1, are based on the assumption that the NRC safety research program support budget will be funded at the full $213.2 million requested.
If a smaller sum is authorized or appropriated, we recommend that the NRC be given considerable latitude in judging where these reduct, ions should be made to have the least impact on the ability of the research program to improve nuclear safety and to meet the most important regulatory needs.
7
4.2 LMFBR Safety Research The proposed budget includes no funds for LMFBR safety research.
We believe that such research is needed and should be conducted by the NRC unless it is certain that the current development program will not continue or will not require licensing decisions within the next few years.
If the Congress continues to fund a Department of Energy (00E) development program in this area, it should provide funds for the appropriate and necessary NRC shiety research program, at the level discussed in Section 4.3.4 of Part 2 of this report.
This sum should be appropriated in addition to the funds shown in Table 1.
Otherwise, it will have to be obtained by reallocations within the existing budget, and will most likely come from the Decision Unit providing for high priority research on Severe Accident Phenomena and Mitigation, unless the Congress permits other steps to be taken.
Such realloca-tion from that Decision Unit, or from others designated as high priority areas in this report, would have a serious effect on the priorities and effectiveness of the program recommended herein.
If, however, the Congress should find it necessary to direct the NRC to conduct research on LMFBR safety without providing funds beyond those recommended in Table 1, we urge that the NRC be directed to obtain those funds primarily by reallocation from Decision Units other than Decision Units 3, 4 and 8, which are described herein as including high priority research programs.
Considerable latitude in reallocat-ing funds should be permitted.
5.
General Comments We reiterate several general comments provided to the NRC Commissioners in July 1980 (Ref. 4).
If those safety research areas which are judged to have the potential for greater impact in protecting the public health and safety are to receive the necessary priority, several steps will need to be taken, including the following:
(a) The NRC Commissioners will have to provide prcmpt policy guidance on the major open safety issues.
(c) The NRC Staff research user offices will have to reevaluate their approach to formulating requests for research and strive to consider these in some broad framework which takes into account the major issues confronting the agency.
(c) The Office of Nuclear Regulatory Research (RES) of NRC will have to reevaluate its current and proposed programs in terms of risk reduction potential and major regulatory needs.
8
(d)
The NRC will have to judge whether some research, particularly that which involves large scale component testing or the appli-cation of existing methodology, should be the responsibility of the industry or DOE rather than of the NRC.
(e)
The NRC should reduce sharply that research which is merely confirmatory in nature where there is good reason to believe that current regulatory requirements provide adequate protec-tion to the public.
(f)
The research program should be geared to providing that infor-mation which is most important to NRC decisionmaking on de-graded core and core melt accidents as expeditiously as possible.
6.
Specific Comments and Recommendations Specific comments and recommendations regarding the scope, nature, and funding levels of the various elements of the research program are presented in Part 2 of this report.
9
TABLE 1 PROPOSED FY 1982 BUDGET (Dollars in Millions)
ACRS CHAPTER DECISION UNIT PRO ~'0 SED RECOMMENDATIONS 1.
LOCA and Transient Research 51.0 43.0 2.
LOFT 44.0 30.0 3.
Plant Operational Safety 37.0 42.5 4.
Severe Accident Phenomena and Mitigation Research 20.2 24.2 5.
Siting and Environmental Research 14.4 14.4 6.
Waste Management 21.5 20.0 7.
Safeguards and Fuel Cycle Safety 10.2 10.2 8.
Systems and Reliability Analysis 14.9 23.9 Total Program Support 213.2 213.2 10
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11
1.
LOCA AND TRANSIENT RESEARCH 1.1 Scope Research in this area includes several programs which are directed toward improved understanding of reactor behavior during loss-of-coolant accidents 'LOCAs) and transients. There has been an extensive reorientation of the program since the Three Mile Island Unit 2 (TMI-2) accident and the emphasis is now on small breaks and tran-sients.
Included here are the improvement and assessment of computer codes which have as their objective an analytic description and understanding of light-water-reactor (LWR) transients. Other programs are directed toward the understanding of fuel and core behavior under conditions of inadequate cooling.
1.2 General The proposed budget includes items requiring a large financial commit-ment over a long period of time.
In many cases, costs for portions of the program begun in earlier years have peaked and are declining.
Other programs have tten reoriented as Nuclear Regulatory Commission (NRC) perceptions of needed research have changed since the TNI-2 accident.
Small breaks and transients which may lead to inadequate core cooling are now receiving the attention they need.
The planned program addresses these problems and should make important contribu-tions to their rc. solution.
1.3 Comments 1.3.1 Semiscale This facility has shown itself to be increasingly useful as an experi-mental tool for contributing to an understanding of pressurized water reactor (PWR) thermal-hydraulic phenomena under LOCA and transient conditions. The NRC has undertaken a serious study of the limitations and scaling questions which arise in applying the data to large scale facilities.
We approve and commend this effort.
A modification of i
Semiscale (M00-2A) has been recently completed.
This modification will provide improved represeatation of systems with U-tube steam generators.
Another modification of Semiscale, MOD-5, is under consideration.
This change would be directed toward the representation of a Babcock &
Wilcox (B&W) PWR with onco-;.hrough steam generators ard core vent 13
valves.
The cost for such a system has not been included in the budget.
We understand that the NRC is exploring with industry groups the possibility of funding support for such a facility.
We believe that the M00-5 program would make significant contributions to the understanding of the basic physical phenomena needed for improved code modeling, and strongly support its implementation.
1.3.2 Separate Effects Experiments and Model Development There are several programs grouped in this item which deserve separate discussion.
One of these is the Two Loop Test Apparatus (TLTA) which is intended to do for boiling water reactors (BWRs) what Semiscal e does for PWRs.
TLTA, however, is inadequate for the purpose and needs to be upgraded significantly to address the current problems with small breaks and transients.
We support the need for experiments in an improved TLTA type facility, preferably one involving two full-length fuel bundles.
Whether such tests are made in a new facility financed wholly or in part by the NRC or in a facility operated by a foreign agency is a matter to be determined by the NRC.
Another facility related to BWRs is the Steam Sector Test Facility which has the objective of studying BWR core spray behavior.
The program is directed toward large-break LOCAs and cannot readily be reoriented to small-break problems.
We agree with the NRC that the experimental program be phased out in FY 1982 as scheduled.
The FLECHT-SEASET program at Westinghouse has been reoriented to include an examination of natural circulation.
Even though the facility is limited to low pressure, it has good steam generator representation and will be useful for code assessment.
We believe that the present effort in code assessment is inadequate.
For an improved program, additional experiments on separate effects are required.
Many of these experiments would not require large facilities since the experimentation should be directed toward getting basic physical and engineering bases for the codes.
The NRC should extend its efforts in this direction.
Further comments on code assessment are given below.
The model development prograr consists largely of relatively small projects in various university laboratories.
We endorse this kind of program as being useful, productive, and cost effective.
A number of these programs can also provide useful information for code assessment.
At the same time, the program provides a helpful interaction with an important part of the engineering and scientific comunity and should be extended and increased.
14 L
1.3.3 3-D Program This is an international program involving Japan, the Federal Republic of Germany (FRG), and the United States.
It was begun when LWR safety research was preoccupied with large-break LOCAs, and it was with this in mind that two large facilities were designed and built in Japan.
One, the Cylindrical Core Test Facility (CCTF), has been completed.
The other, the Slab Core Test Facility (SCTF), is under
-construction.
Both are limited to low pressure.
A thoroughly in-strumented experimental program is planned in CCTF on natural circula-tion, and two-dimensional (2-D) effects in core refill following a large LOCA are planned in SCTF.
Both facilities, within their cap-abilities, are well constructed but suffer from an insufficient test engineering staff.
The NRC is planning to assign several engineers to th3 facilities in Japan and the FRG.
We believe that this will enhance the value of this international cooperative program.
There was early appreciation in Japan of the low-pressure limitations of CCTF and SCTF, and they have completed the design of a high-pressure facility, ROSA IV.
This facility should be of importance to the U.S. reactor safety program, and we urge that the NRC explore the i
possibility of participation in the program.
The FRG effort in the international 3-D program will consist primarily of the construction of the Upper Plenum Test Facility (UPTF).
This facility will presumably provide some information relating to special questions regarding large LOCAs.
The facility will also make some contributions to the interaction of hot leg injection with steam upflow through a core and other counter-current flow questions.
A large part of this program relates to a special feature of the FRG PWR design.
We suggest that the NRC attempt to secure a redirection of the FRG program to make it more relevant to present problems. The program in the FRG PK'. facility has been useful, but the major item in the FRG program is the UPTF which promises to be a vary costly facility with insufficient return in pertinent data.
The U.S. contribution consists of two programs.
One of these is the supply of experimental measuring devices for these large foreign research facilities.
We have for some time urged the development of new and improved instrumentation which could be installed in operating power reactors and would encourage some contributions frc1 the 3-D program in the direction of instruments for power reactors.
The second contribution from the U.S. consists in supplying a bridge between the various tests in Japan and FRG by means of the TRAC 15
1 computer code.
The NRC should carefully consider whether this com-putational effort contributes effectively to the basic requirement of a useful code description of nuclear power plant transients.
The TRAC bridge may prove to be of limited value in view of the present stage of development of this code.
1.3.4 Code Improvement and Maintenance The NRC proposes to complete in FY 1982 best estimate codes for PWR and BWR systems.
These codes will be adaptations of TRAC.
The NRC appears to believe that TRAC has been adequately developed and as-sessed so that these efforts will be meaningful. We believe that TRAC has been inadequately developed and assessed in spite of the large effort that has been expended.
We strongly support the RELAP-5 effort.
The version of TRAC which has just been developed by NRC for BWRs is totally inadequate. Among other deficiencies, we especially note that the present version of this code has no treatment of the upper plenum.
He also note that the NRC has not yet sponsored a BWR version of RELAP-5.
We believe that this omission is inexcusable and recommend the prompt development of such a code.
1.3.5 Code Assessment and Application It has been noted that the NRC code assessment is inadequate, particu-larly in the case of TRAC.
In spite of a large effort, TRAC is not yet a code that is adequately developed and assessed.
In some re-spects RELAP-5 has indicated greater promise with a smaller effort than TRAC has received.
We nevertheless recommend that both TRAC and RELAP-5 be continued since the NRC has already made a large investment in TRAC and this code may eventually give a 2-or 3-dimensional calculational capability.
We recommend that the NRC drastically reduce, or even terminate, the effort going into the development of component codes.
Further, the tendency toward a multiplicity of such codes should not continue.
These efforts should be replaced with attaining increased code cap-abilities for small-break LOCAs and other reactor transients with RELAP-5 and TRAC.
The program of code assessment is so arranged that it is difficult to incorporate changes and improvements into the codes as research and development proceeds. Such a bureaucratic approach is clearly undesir-a.le.
We also believe that the code assessment matrix is much too heavily skewed toward concerns with large-break LOCAs.
The code assessment matrix should be immediately revised to reflect the need for emphasis on small-break LOCAs and operational transients.
16
Before embarkirg on a massive assessment program, the NRC should undertake a limited and carefully monitored code assessment program on a modest scale.
The effective development of RELAP-5 shows the value of an integral facility, even one of modest scale like Semiscale. The close associa-tion with observation in such facilities contributes to better under-standing of what is needed to make a code an effective tool for the description of reactor transients.
As a general comment on the code development program, we approve the development of the so-called " advanced codes" like TRAC, but we consider it unfortunate that the PRC does not yet have a full and adequate audit calculation capability. This essential need of the NRC has not been met even though great attention has been given to these advanced codes.
We question whether the WRAP code is being pursued effectively so,as to fulfill the NRC audit needs in a timely manner.
1.3.6 Fuel Behavior Under Operational Transients The purpose of this program is to aid the NRC in the prediction of fuel element behavior under design-basis-accident conditions and operational transients. It consists of in-pile tests of fuel rods, development and assessment of fuel codes, and basic studies of clad-ding behavior in fuel bundles.
The future function of this program and especially that portion which is undertaken in Power Burst Facility (PBF) has been of concern to us.
- The particular matters currently being studied in PBF represent a negligible risk to the public and the information now available in this area seems adequate for any high priority regulatory need.
The full-length fuel bundle tests planned to start in FY 1982 at Oak Ridge National Laboratory and the in-pile experiments on pellet-clad interac-tion planned at the Canadian Test Reactor (NRU) represent other examples of-studies to which we would sssign a low priority.
1.3.7 Core Damage Beyond LOCA This program includes several in-pile experiments aimed primarily at studying the core degradation processes that occur in a:cidents that overheat the core.
The. tests of full-length fuel' bundles in NRU will look primarily at the question of clad ballooning and coolant blockage.
The PBF severe core damage tests will take a short fuel bundle up to 4000 F in steam.
The ESSOR tests in Europe, which NRC is supporting, will do similar tests on longer fuel bundles.
17
The PBF tests are a unique component of the NRC effort supporting the Degraded Core Cooling rulemaking now underway as a result of the TMI-2 accident.
It is important that they be carefully planned and have the benefit of a broadly based review and evaluation while they are still in a formative stage.
They should be closely coordinated with ongoing complementary research in this area, particularly with research per-formed under the " Severe Accident Phenomena and Mitigation" (Decision Unit 4).
We believe the related work in the areas of Post-Accident Coolant Chemistry and the Hydrogen Studies are important outgrowths of the TMI-2 accident lessons learned and deserve strong support.
1.3.8 PBF Operations The PBF is the only reactor in the U.S. that is dedicated to studying the behavior of fuel in operational and short-period transients.
While the PBF has provided useful licensing infomation, we believe its future usefulness lies in the study of severe fuel damage condi-tions.
1.4 Recommendations We recommend that the LOCA and Transient Research Program be funded at the level of $48.0 million, which represents a reduction of $3.0 million from the proposed budget.
We recomend a reduction of $1.0 million in the program on Code Assessment and Application through a reduction of the effort going into the development of component codes, and a reduction of $2.0 million in the program on Fuel Behavior Under Operational Transients.
18
f 2.
LOFT 2.1 Discussion The LOFT facility has been useful and has contributed to an understand-ing of the phenomena encountered in large LOCAs, and in demonstrating, within reasonable limits, the effectiveness of PWR emergency core cooling systems.
More recently, it has been useful in studies of the phenomena which may appear during transients and small LOCAs.
We believe, however, that these programs are not cost effective when compared to other programs.
We believe that LOFT can conc'ede its NRC mission with funding for tests terminated by the end of FY 1982.
This time period allows for an orderly completion of some tests of interest to the NRC.
Since the facility will be entering its final phase, many ancillary programs which have been added over the years can be eliminated.
Costs can thereby be reduced so that the few remaining tests of concern can be performed within a budget of $30.0 million.
2.2 Recommendations We recommend therefore that the NRC test program in LOFT be terminated by the end of FY 1982, and that the budget for' FY 1982 not exceed
$30.0 million.
This budget will, in our opinion, allow an orderly termination ~of the program and will at the same time make possible completion of the significant tests under consideration.
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3.
PLANT OPERATIONAL SAFETY 3.1 Scope Research in this area is directed toward improvements in the opera-tion and maintenance of reactors in a safe and reliable manner.
Specific areas covered by this program include the following:
Man-Machine Interface, Instrumentation and Electrical, Plant Systems Behavior, Mechanical Components Safety, Structural Safety, Fracture Mechanics, Operating Effects on Materials, and Nondestructive Exami-nation.
3.2 General We believe that plant operational safety is a high priority area and
- recommend that $5.5 million be added to the proposed budget for this Decision Unit to permit an increased level of effort in the following areas:
Man-Machine Interface (a) Augmentation of the program on man-machine interface, includ-ing the examination and evaluation of a broad range of possible applications of disturbance analysis systems.
(b) An-examination of interlocks that may involve operator action either by plan-or by error.
Instrumentation and Electrical (a)
Accelerated and broadened research on safety implications of 1
the use of computers in protection systems and process control systems, including hardware and software.
(b) -Augumented research intended to enable an improved evaluation and assessment of the design adequacy of instrumentation, electrical equipment, and power systems.
Plant Systems Behavior (a) Influence of control system design philosophy and reliability on safety.
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(b) Detailed studies of plant benavior as a function of time for a wide range of transients for the several generic plant designs in operation or under construction, including an examination of the sensitivity of such behavior to possible
)
variants in plant design and operator intervention encompass-ing multiple failures and related f ailures of redundant systems.
(c) A deterministic analysis of the course of events following a wide range of postulated accident precursors for a representa-tive group of specific plant designs in order to provide in-sight into possible steps which might be taken to reduce the likelihood of occurrence of those events which have the potential for. posing significant challenges to safety.
Seismic Safety Augmentation of the Seismic Safety Margins Research Program (SSMRP) to include a plant-specific, detailed seismic risk study for a t
BWR in FY 1982, and to provide for extensive general and detailed recocinendations as to possible changes in the current approach to seismic safety in FY 1982.
PWR Secondary System Integrity Initiation of a program to investigate the overall aspects of secondary system integrity and functional reliability insofar as nuclear safety is affected by failures in these systems.
3.3 Cocunents 3.3.1 Man-Machine Interface This has been an expanding area of research since the TMI-2 accident.
The research deals with a need for better operator response to plant transients.
The work on developing improvements in instrumentation and information display is expected to have progressed by FY 1982 to a point where firm recommendations can be made for status monitoring and diagnostic display arrangements.- A program on human factors measure-ments and improved instrumentation and control displays will be con-tinued through FY 1983/1984.
l Man-m6
'Cerface programs have been initiated as results aither of req from NRC research user offices, or Congressional recommen-dations,.r directly from the NRC Action Plan (Ref. 5).
Thr.;e pro-grams will provide improved safety and reduced risk.
We take note of the fonnation of the Operational Safety Research Branch and agree
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1
2 that efforts to consolidate the research in the field of the man-machine interface are warranted.
The Severe Accident Sequence Analysis (SASA) program whose objective is to improve the understanding of reactor accident phenomena and of the man-machine interface during a broad spectrum of accident se-quences is supported.
Emphasis will be placed on the operator, the operator need for information, the alternative actions the operator might take given various combinations of component failures, and the consequences of those actions. These programs demonstrate and develop diagnostic tools which will contribute to operator knowledge of plant conditions.
These programs are considered important to plant opera-tional safety and should be continued and expanded.
3.3.2 Instrumentation and Electrical Research in this area will include testing of advanced instrumentation to follow the two-phase liquid level in reactor vessels for possible use to alleviate TMI-2 type problems.
Work on fire protection re-mrt.h concerning fire suppression systems is scheduled for completion in FY 1982 and full scale replication tests of actual cable area configurations are in progress.
Tests will be completed in the qualification test program and work will be initiated to address safety concerns from the environment of small LOCAs.
Qualification testing and postmortem analysis will be performed on TMI-2 in conjunc-tion with Department of Energy (D0E) sponsored programs during plant decontamination and rec'overy scheduled for FY 1982/1983.
A system review of generic safety-related instrumentation and elec-trical equipment is scheduled for FY 1982 to identify basic design, fabrication, wear, aging, and other reliability problems, and to determ'ine the ability of such equipment to withstand post-accident temperature and steam conditions.
We note that the NRC., 0; memorandum dated May 23, 1980 (Ref. 6),
directed its Staff ta proceed with the replication tests for fire protection systems. Je have been informed that the acceptance test criteria have been structured to minimize expenditures as much as possible consistent-with the test objectives.
If such work is deemed necessary, it should be performed by industry, not NRC.
However, we continue to believe that the expense of fire replication tests is f ar too great for the information to be.obtained, and we do not support this particular part of the work.
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The FY 1982 budget should contain provisions for the initiation of new programs, including some that address problems with safety-related instrumentation and electrical equipment, software verification, and the study of nuclear plant electrical supply system design problems.
3.3.3 Plant Systems Behavior The demonstration phase of the test at the Sequoyah plant of a contin-uous on-line surveillance system to show how pattern recognition can be used to alert plant operators to anomalous conditions is expected to be completed in FY 1982.
A significantly increased effort on assessing nuclear plant operational behavior will be initiated in FY 1982.
This effort will include assessments of operational transients on system behavior, safety consequences of shared systems, and design requirements for plant systems and facilities to allow them to cope safely with accident conditions.
These programs also demonstrate and develop diagnostic tools which contribute to reducing the frequency of accidents.
3.3.4 Mechanical Components Safety The primary objective of this program is to provide the NRC licensing staff with methods for evaluating structural integrity in terms of margins of safety. The SSMRP is supported largely by this effort. We continue to support the SSMRP and recommend that it be structured to provide input as early as is feasible into the broad safety policy considerations, as well as practical applications,. concerning the seismic design bases of nuclear power plar.ts.
Programs in this area are being developed on the reliability of systems and components which have not been seismically qualified.
The results of this effort should influence the design of plant mechanical equipment for improved reliability and ease of maintenance and inspection.
~3.3.5 Structural Safety This program is well defined and well balanced among the several identified reed:
and is oriented. strongly toward questions relating to the safety of operating plants.
A significant portion of the research in this program is to be done by ' independent contractors rather than by National Laboratories.
The results of this action, in terms of cost, timing, and effectiveness should be evaluated as the program progresses..NRC should maintain cognizance of the structural 23
research being done by industry, both domestic and foreign, and should be in a position to utilize the results to the extent practicable.
We have commented on the SSMRP research in Section 3.3.4.
Those comments are applicable to the structural safety work under this program.
3.3.6 Fracture Mechanics This long-range program is providing a sound basis for decisions on the integrity of primary systems.
It should continue to address the question of thennal shNk in pressurized systems which represents an important uncertainty in the integrity of the older reactor pressure vessels.
This program should be actively pursued to provide a basis for decisions in this area.
In the piping area, work should continue to define programs which will provide an acceptable basis for reducing the number'of constraints on primary piping systems while maintaining adequate' safety margins.
3.3.7 Operating Effects on Materials The largest uncertainties in assuring the integrity of the primary system a'e attributed to operating environment, upecially radiation and water chemistry.
The program in this area addresses these issues in a sound, coherent manner.
That part of the study of the Surry steam generator which - determines the relation between eddy-current indications and actual defects in the tube assembly will be valuable to the NRC in its decisions on other operating ste;m generators.
We are not convinced of the merits of the other portions of the steam generator program, and recommend that funding previously allocated for this work be directed to other research programs in this Decision Unit having a greater risk reduction potential.
3.3.8 Nondestructive Examination Periodic inspection of reactor components is regularly carried out to assure that no dangerous flaws are present.
The NRC must be capable of judging how reliable these technique: are and must be able to develop _ criteria for the acceptability of new techniques.
Several programs' are planned or are in place to enhance this capability.
3.4 Recomendations We believe that a funding level of $42.5 million, an increase of $5.5 million over the proposed budget, is required for this Decision Unit which we.believe to be of high priority..
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4.
SEVERE ACCIDENT PHENOMENA AND MITIGATION RESEARCH 4.1 Scope This is a new Decision Unit providing redirection of some work previ-ously included in other Decision Units to support aspects of the Degraded Core Cooling, Siti ng, and Emergency Planning rulemaking procedures.
This Decision Unit provides only one part of the total research effort on these rulemaking procedures.
Other Decision Units include several projects that will be useful in these procedures.
This Decision Unit also covers the research related to the licensir9 of Fast Breeder Reactors and Advanced Conu -ter Reactors.
4.2 General The research'needed to sup) ort the Degraded Core Cooling, Siting, and Emergency Planning ruleuaking proceedings represents one of the highest priority areas in the research program.
We suggested in NUREG-0699 (Ref. 4), tha*. a high level task force be established with the charter of recommending the research needs and resources to support these rulemaking proceedings.
This recomendation has been implemented in part by the establishment of a Degraded Core Cooling Steering Group to coordinate rulemaking activities. We recomend that this Group act promptly in identifying the needed research, and that the resources which must be committed to accomplish rulemaking proceed-ings in a timely manner be made available.
~he approach suggested by the ACRS for (1) identifying major Class 9 recident categories and (2) identifying the information needs and time scale for each major category, should be followed.
Resource comitments should be kept flexible until the Steering Group has defined the needs for this work.
The Commissioners should devote the time needed to provide the safety philosophy and objectives that must guide this work.
We continue to recomend, as we have repeatedly done in the past, that a viable program in Fast Breeder Reactors be continued.
4.3 Coments 4.3.1 Fuel Melt Behavior This program provides for the analyses of' fuel melt behavior at a time -
after the molten core material falls to the core support plate and 25
through subsequent attack on the reactor cavity basemat.
Th i s re-search provides information needed for the development of fil*ered-vented containment and core retention systems.
Containment and core melt mitigation system response are also included.
Areas needing attention are the feasibility of cooling molten core debris in the reactor vessel and whether water should be introduced into the reactm cavity following an accident.
If the molten core is not coolable in the reactor vessel, a determination needs to be made as to whether water should be introduced or kept out of the reactor cavity once it has been cetermined that slumping of the core has occurred and that vessel melt-through is inevitable.
4.3.2 Fission Product Release and Transport This work provides the basis for defining the radiological source term for severe accidents.
The source term must be defined in order to determine the radiation environment and the quantity of radioactive material which must be dealt with by engineered safety features and by mitigation features to be considered in the Degraded Core Cooling rulemaking.
Large uncertainties exist in the amount and behavior of radiciodine and suspended aerosols of various fission products predicted to be released to the environment during the course of a number of postu-lated serious accidents. Recently, the amount of radiciodine likely to be released in certain accidents has been a matter of particular controversy. We believe the general subject of fission product source terms for serious accidents - is important and should be given high priority.
4.3.3 Severe Accident Mitigation This work provides for the determination of the feasibility, value, impact, and design of additional features which may be incorporated in nuclear plants for mitigation of severe accidents. Priority should be given to studies to develop detailed cenceptual designs of potential mitigating features for the various combinations of reactors, contain-ments, and sites. These studies should be carried out concurrently in order to determine the usefulness of such foreseeable severe accident mitigation features.
An area requiring near term decisions is that of hydrogen burn pres-sure mitigation features for containments designed for. low pressure.
Research in e area also requires high priority.
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4 This program should recef ve all needed resources before final commit-ments are made to the large experimental program envisaged in Section 4.3.1 on Fuel Melt Behavior.
4.3.4 Fast Breeder Reactors We repeat in this report bott the technical and fiscal recommendations on l' quid metal fast breeder eactor (LMFBR) research which were made in NUREG-0699 in July 1980 (Ref. 4).
If Congress expects the U.S. to move forward with either the Clinch River Breeder Reactor Project or a larger demonstration unit within the next few years, we believe that the NRC should be prepared to take the initiative and make a positive input to the safety decisions on such plants, and a research budget of approximately $20 million is needed.
Further, if early U.S. LMFBR action is likely, we believe the NRC should be directed to make a new, detailed study of research needs and costs on a priority basis and to submit a new budget to the Congress at the earliest opportunity.
4.4 Recommendations We recommend that the research programs on Severe Accident Phenomena and Mitigation, exclusive of research on Fast Breeder Reactors, be funded at a level of $24.2 million, an increase of $4.0 million over.
the proposed budget, for FY 1982.
This is a high priority area.
If research on Fast Breeder Reactors is to be conducted, funds should be provided in addition to the $24.2 million recommended above, as discussed in Section 4.3.4, in order not to impair the effectiveness of the other high priority programs in this Decision Unit which are important to LWR safety.
If additional funds are not provided, the funds needed should be obtained from those Decision Units not desig-nated herein as being of high priority.
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5.
SITING AND ENVIRONMENTAL RESEARCH 5.1 Scope Research in this program area includes studies of Seismology, Geology, Meteorology, Hydrology, Environmental Impacts of Airborne and Aquatic Effluents, Occupational Exposure and Health Effects, Siting Alterna-tives, Socioeconomic Impacts, and Planning for Emergencies.
5.2 General In our review of Siting and Environmental Research, we had an opportu-nity to discuss a number of individual projects with the NRC Staff.
In many cases, however, the importance of specific projects to the attainment of program goals and, indeed, the nature of the overall programs, themselves, are not clear.
In the interest of helping to clarify this situation, we have outlined below what we consider to be the more important programs in this subject area.
5.3 Comments 5.3.1 Development of Seismic and Geologic Behavior Data for Impar. tant Regions gf the U.S.
This program is devoted primarily to developing a better understanding of the seismic and geologic behavior of several important regions of the U.S., and is responsive to our recommendation of several years ago.
We believe that the studies are of considerable importance to the establishment of an improved seismic design basis for future nuclear facilities and to an improved assessment of the seismic safety of existing facilities.
5.3.2 Hea'th Effects When the l'RC sets or modifies 11; standards for radiation protection in 10 CFR Part 20, it is clearly obligated by law to follow Environ-mental Protection Agency (EPA) guidelines for environmental applica-tions.
While the legal requirements are not as clear in the case of standards for occupational radiation exposures, it is probable that the recommendations of EPA will essentially be followed.
Since the NRC is currently revising 10 CFR Part 20, only the results of health effects research currently nearing completion will be useful as input to this effort.
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A review of the NRC research on health effects shows that it is a collection of some 16 projects with no substantive programmatic theme.
Several are projects that were initiated by EPA or D0E. Upon learning that these agencies intended to discontinue support far these projects, the NRC decided to assume responsibility for their funding.
On the basis of our review, we make the following recommendations:
(a) That each of the projects taken over from EPA or DOE be care-fully reviewed to determine whether it should be continued.
Several might be referred to the relevant National Institutes of Health Research Committees for proper disposition.
(b) That the NRC continue its research on relevant health effects problems that are not being pursued by other Federal agencies.
Examples are studies on the effects of internally deposited uranium and thorium, and on the effects of neutron exposures.
(c) That NRC consider continuing a modest research effort on the offects of higher level radiation doses (with serious to termi-nal consequences in about one year) for input to evaluations of accident consequences.
(d) That the NRC develop strong contacts with groups, such as the National Council on Radiation Protection and Measurements and the National Academy of Sciences, which are conducting in-depth reviews of health effects data, and ultimately attain an inde-pendent capability for making assessments of the more contro-versial aspects of such data.
5.3.3 Emergency Preparedness Following the TMI-2 accident, various groups, including the NRC and Federal Emergency Management Agency, addressed the subject of emer-gency preparedness and all agreed that activities in this area needed to be upgraded.
Included in this effort was the need for research on developing a better description of the source term, on defining the consequences of radionuclide releases accompanying an accident, and on controlling and mitigating adverse effects. Although certain of these needs are being addressed, we believe the level of effort is not commensurate with the currently accepted role of emergency prepared-ness in the defense-in-depth concept.
Specific topics on which research is needed include:
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)
5.3.3.1
'-termination of the Nature at1 Path of Released Radionuclides Although progrecs is being made in the development of portable field equipment for mor.itoring airborne radiciodine, it is not clear whether such equipment can quantitatively collect and analyze iodine in its several possible forms. In addition, there is a need to improve capabilities for tracking plumes of contamination released under accident conditions. The reseasin programs should include a review of the specific kinds of information needed on a real time basis for immediate decisions, an analysis of various mechods by which this can be accomplished, a review of the adequacy of existing technologies for perfonning these functions, and the identification of any addi-tional methodology development and research needed to fully implement practical systems.
5.3.3.2 Meteorology and Patt.vay Analysis Analysis of the consequences of postulated releases involves having the necessary meteorological knowledge to trace plume pathways (in-cluding the effects of wet and dry deposition), understanding the hydrological movement of liquid releases including their effects on groundwater, having the necessary mathematical models (such as an improved Calculation of Reactor Accident Consequences (CRAC) code) to predict environmental transport, and having an ability ta predict the potential health effects of estimated exposures.
We are impressed with the thorough and systematic manner in which the meteorological aspects of these problems are oeing addressed. This program is moving along well, and the resultino data can strengthen the accuracy of the CRAC code, as well as contribute to the ongoing reassessment of any necessary backfits to the Indian Point and Zion nuclear power plants.
The program deserves str3ng support, especially in research on wet deposition.
The Radioiodine Pathway Analysis study may be an impor-tant part of this evaluation.
Although the efforts related to underground water pathways need further strengthening, we are encour-aged to see the close ties that have been established with related studies on the underground disposal of radioactive wastes.
5.3.3.3 Countermeasure Actions 1
We believe that the NRC should support research on a full range of countermeasures to protect the public in the event of a nuclear power plant accident.
Such actions include evacuation, sheltering (includ-ing the use of respiratory protection), and the administration of blocking agents.
The current level of research effort in this area appears to be inadequate. Funds for this purpose can probably be made available through discontinuation of support for several of the previously cited health effects studies inherited from EPA and 00E.
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5.3.3.4 Human Factors in Emergency Response As mentioned in our previous report to the Commission (Ref. 4), we believe that there is a need for better information on the anticipated behavior of population groups in emergency situations.
The program proposed by the NRC in this area is reasonable.
5.3.4 Research Related to Reactor Siting The NRC is currently engaged in rulemaking on siting for nuclear power plants.
We believe there is a need for a comprehensive review of any research that may be required to provide data in support of this effort.
Once this is completed, an appropriate research plan should
-be developed. Specific aspects that need attention include the estab-lishment. of criteria for determining the acceptability of sites for single as well as multiple-unit stations, and the bases for determin-ing that one site has clear and distinct advantages over another.
j.4 Recommendations We recommend that this Decision Unit be funded at the proposed level of $14.4 million.
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1
- 6. WASTE MANAGEMENT 6.1 Scope The Waste Management research progran is dirc:ted to the health and safety problems that result from the handling and ultimate disposal of high and low level radioactive wastes and uranium mill tailings.
The safe disposal of such wastes has been and continues to represent a major public concern.
6.2 General In previous reports to Congress (Refs.1 & 3), we have criticized the NRC management of research work in this program.
However, in the FY 1981 report (Ref.1), we noted that some positive steps had been taken to improve this situation and we again have observed continued improve-ment in this regard.
We encourage further efforts to coordinate this research work with other Federal agencies, particularly DOE.
We believe that the NRC should expedite its planned studies on the development of risk assessment methodology for potential early appli-cation of this technique to assist in the selection of research work to be undertaken on Waste Management, to set priorities for this work, and to enable risk assessment across the nuclear fuel cycle.
We recommend again that the NRC make more use of outside qualified people to assist it in deciding what research work is realistically needed, priorities among research areas, and whether the research should be supported by the NRC or by other organizations, e.g.,
D0E.
6.3 Comments 6.3.1 High Level Waste (HLW)
We agree with t'ie NRC that research work on handling and disposal of HLW should be given high priority.
The NRC has proposed a funding level of $13.0 million (compared to $3.4 million and $0.6 million in FY 1980 and FY 1981, respectively) for this area but we believe that this can be reduced to about $11.5 million and still provide timely information for the NRC to make its decisions on licensing and regu-latory actions for HLW management. We recommend that the $1.5 Pllion be shif ted to other Decision Units earlier identified in this report that we strongly believe should have substantially greater funding than is proposed.
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F 1
l 6.3.2 Low Level Waste In NUREG-0657 (Ref. 1), we emphasized the need for sufficient research work to expedite the licensing and regulation of the handling and disposal of low level radioactive wastes.
We reiterate that position for FY 1982.
The existing situation mandates the selection of new disposal sites within the near future.
Research related to the development of criteria for assessment of the acceptability and operating procedures for such sites should be expedited.
6.3.3 Uranium Recovery The disposal of uranium mill tailings which result from uranium recovery and concentration operations has long been a public concern.
We agree that the work to develop methods for dealing satisfactorily with the large number of existing uranium mill tailings piles and to provide early guidance for the licensing and regulation of new mills warrants the amount of funding requested and that it should be given high priority within this Decision Unit.
6.4 Recommendations We recommend that the budget for research on Waste Management be reduced to $20.0 million, that the reduction of $1.5 million be accomplished by a corresponding reduction in the program related to HLW.
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7.
SAFEGUARDS AND FUEL CYCLE SAFETY 7.1 Scope This program deals with Safeguards, including Physical Protection, Material Control and Accounting, and Threat and Strategy, and wi;h Fuel Cycle Safety, including Fuel Cycle Facility Safety, Decommis-sioning, Transportation, Effluent Control, By-Product Safety, and Occupational Protection.
7.2 General The situations and materials being studied are those associated with the operation of LWRs.
In a number of instances, the scope of the proposed studies ought to be reconsidered, and possibly broadened, particularly in view of the possibility that the country's present policy concerning reprocessing and breeder reactors may have been changed by the time the FY 1982 budget is in effect.
We recommend that consideration be given to the expansion of research on the reduction of occupational exposures associated with major maintenance and repair operations such as the replacement of steam generators.
This work could help provide the information necessary for making additional progress in controlling occupational exposures in LWRs.
7.3 Coments 7.3.1 Physical Protection A major fraction of this effort will be devoted to applying techniques already developed for use in the licensing and regulatory process.
Work will be' continued, and new work started on spent fuel storage problems. Studies will also be made of potential conflicts between safety and safeguards requirements for_ operating reactors.
7.3.2 Material Control and Accounting Here, also, a major fraction of the effort will be devoted to trans-ferring developed techniques for use in the licensing and regulatory process.
Increased attention will be given to determining the amount of enriched material held up in processing equipment.
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7.3.3 Threat and Strategy This is a small program to develop appropriate responses to threats or appropriate actions in the event of successful sabotage or theft.
In our view, work in this area would have a lower priority than work in Physical Protection, and Material Control and Accounting; however, we recommend the initiation of a program intended to provide insight into appropriate actions by response groups for a wide range of scenarios involving the postulated takeover of part or all of a nuclear power plant.
7.3.4 Fuel Cycle Facility Safety A major research effort in this area is devoted to analyses of acci-dent scenarios that might lead to aerosol generation in fuel cycle facilities and to the development af realistic models for assessing aerosol transport within, and releases from such facilities. We agree with the importance of this effort.
Another significant research effort in this program is that directed to the development and applica-tion of risk assessment methodology in the fuel cycle.
7.3.5 Decommissioning We have previously recommended a larger NRC research program on decommissioning. We continue to support this position.
7.3.6 Transportation We believe that the research studies related to safety in the trans-portation of radioactive materials are generally needed.
- 7. 3. 7-Effluent Control We agree with the need for this research program which is directed mainly at improving the accuracy in evaluating effluent control system performances in "WRs and fuel cycle facilities.
In order to help achieve this objective, a major research effort will be made to obtain more accurate radionuclide source tena data.
We believe that the study on " Advanced PWR Effluent Treatment Model" should be either combined with the one on " Source Term Measurements" or deleted.
7.3.8 By-Product Safety This is a new program.
A first step will include developing an inventory of the products requiring consideration and a scale of relative public risks associated with these products.
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l 7.3.9 Occupational Protection Included within this program area are exposure evaluation and control.
A key research need relative to exposure evaluation is the development of better techniques for monitoring and assessing neutron exposures.
Research needs relative to exposure control include the development of additional design and fabrication approaches for preventing the production, release, and buildup of radionuclides within reactor cooling systems, the application of risk / benefit assessments to procedures for the maintenance, repair, modification, replacemer.t, and disposal of major nuclear power plant components, such as steam generators, and the development of design features to facilitate the decommissioning of nuclear power plants.
A review of the NRC's Draft Long Range Research Plan (Ref. '),
indi-
/
cates that the NRC Staff has a firm grasp of the nature of the prob-lems in this area and that they have specified what needs to be done.
The necessary research is being implemented. We agree that there is a need for development of a better understanding of the " corrosion, erosion, transport, and deposition phenomena within the primary coolant system and the effects on these of changes in design, mate-rials of construction, quality control, housekeeping practices, and operations."
We agree also that "there are few methods or data for analyzing the performance and reliability of LWR decontamination syst' ems or for evaluating the net contribution (or reduction) they might make to occupational exposure."
The continued increase in collective occupational doses at operating nuclear power plants makes this an area of progressively greater interest.
We note that studies sponsored by DOE and the Electric Power Research Institute, on the relation of water chemistry to the behavior of radionuclides within reactor cooling systems and on in-reactor decontamination, are underway at several utilities.
The NRC should have a program to analyze the results of these efforts, and should consider the implementation of any necessary research to fill in the voids, as well as to review the general applicability of the work to commercial nuclear power plants.
The research programs proposed for initiation in FY 1982 to address these problems are appropriate.
7.4 Reconinendations -
For-Safeguards, there is difficulty in comparing the priority of work in this field with work aimed at improving the operational safety of 36
v reactors but, in view of the public interest and potential importance of possible acts of theft or sabotage, we believe this work should be continued at about the existing level.
For Fuel Cycle Safety, we recommend funding at the level proposed, but suggest that the amount allocated to Occupational Protection should be increased sufficiently to support a meaningful study of the control of crud buildup in LWRs.
In summary, we recommend funding for this Decision Unit at the proposed level of $10.2 million.
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v 8.
SYSTEMS AND RELIABILITY ANALYSIS 8.1 Scope This Decision Unit, Systems and Reliability Analysis (SARA), includes many but not all of the programs which previously were grouped under the former Decision Unit entitled Risk Assessment.
SARA includes Methodology Development, Reliability and Human Error Data Analysis, Systems Analysis, and Consequence Analysis.
8.2 General We have previously given strong support to this research program.
In NUREG-0657 (Ref. 1), we placed our highest research priorities on the FY 1981 Decision Units entitled Improved Reactor Safaty and Risk Assessment; we recommended increases in the President's Budget requests for these two Decision Units.
The forthcoming rulemaking on Degraded Core Cooling and the growing emphasis during recent months on the use of reliability and risk analyses and the development of quantitative risk criteria give strong support to that recommendation.
8.3 Comments 8.3.1 Recommendations for Expanded Program We recommend that this Decision Unit be allocated at least $23.9 million for FY 1982, an increase of $9.0 million over the proposed budget, and that this additional funding be allocated approximately as follows:
(a) A large increase in emphasis and resources for the task on alternate ' decay heat removal systems, including consideration of sabotage, enabling enough effort to provide a basis for regula-tory decisionmaking no later than the end of FY 1982 ($2.0 million).
(b) A very considerable acceleration in the development of informa-tion needed to estimate the likely effect on risk of various potential design changes intended to mitigate accidents leading to severe core damage or core melt in LWRs ($1.25 million).
(c) The development of accident precursor screening techniques and their extensive application to the existing operating plants
($1.0 million).
38
Y (d)
The early development of a focused, cohesive program to provide the information needed to determine the appropriate regulatory approach to control systems and to information needs of the reactor operator ($1.0 million).
(e) Critical review and evaluation of probabilistic analyses and risk studies performed by licensees and construction permit holders ($1.0 million).
(f) An examination of possible weaknesses in the current application of the single failure criterion, and the early development of an improved approach ($0.5 million).
(g) The development of a basis for an improved approach to minimiz-ing significant design errors (50.5 million).
(h) A systematic approach to possible design steps to reduce the potential for serious accidents which might be caused by sabo-tage by an insider (50.5 million).
(i) A program to better define property damage from accidents involving large releases of radioactive materials, including the effect on societal resources ($0.5 million).
(j) A critical evaluation of the merits of LWR regulatory require-ments in other countries which differ significantly from those of the NRC (50.5 million).
(k) The early development of quality assurance (QA) criteria for probabilistic analyses to be used in the regulatory process (50.25 million).
We recommend that the n itters listed above be given priority, even if it means reducing the funding for other programs, ongoing and proposed, in this Decision Unit.
Some jditional manpo':::r will be needed for the NRC Staff that manages this program.
8.a.2 Methodology Development The NRC justification and planned accomplishments are reasonable.
However, as discussed in Section 8.3.1, we believe that priority in this program should be given to the most pressing needs of the NRC.
These needs include the following:
the development of a methodology suitable for early use by the industry and the NRC in system and 39
accident probability evaluation; QA guidance and a peer review tech-nique for probabilistic analyses performed by the NRC and the industry; quantitative risk criteria; changes in the regulatory approach to control systems; screening techniques for precursors of potentially serious accidents; a methodology for reducing or catching design errors; and accidents involving common mode and multiple related failures.
The program of research on floods should receive adequate priority taless and until it can be shown that present methods of estimating the incidence and consequence of floods are adequate to reduce the risk from such events to a minor level.
8.3.3 P,eliability and Human Error Data Analysis We support the proposed research program on human error. We recommend that it have the benefit of considerable interaction with the Office of Inspection and Enforcement (IE).
Such interaction, if it includes working personnel from both the training and inspection sections of IE, could be useful to both groups.
While we agree that work on component failure rates and downtime is worthwhile, we recommend that this program be evaluated to see if the proper priority is being given to systematic and common cause failures of all kinds, including sabotage.
8.3.4 Systems Analysis Several of the specific recommendations made in Section 8.3.1 for high priority research and much more funding also apply to this program.
They include the following:
(a)
A reevaluation of the NRC safety research program in terms of risk-reduction potential.
(b)
The extensive application of accident screening precursor methods to operating reactors.
This should include the review in detail, of actual representative systems such as BWR hydrau-lic scram systems and LWR air systems, and potentially undesir-able scenarios such as those which might cause cold overpressuri-zation of the reactor pressure vessel.
(c)
A systems analysis of the contributions of risk arising from control systems and the potential for improvement.
(d)
Conceptual design studies of possible approaches to reduce the potential for serious effects arising due to sabotage by an insider.
40
(e) Much greater emphasis on alternate decay heat removal systems.
This emphasis should be sufficient to provide a basis for regulatory decisionmaking by the end of FY 1982.
(f) A major acceleration in conceptual design studies of systems to mitigate and reduce the consequences of degraded core and core melt accidents.
(g) Adequate review of the probabilistic analyses performed by industry.
8.3.5 Consequence Analysis The following should receive high priority in this program:
liquid pathways; power plant siting; emergency planning; and reexamination of nearby and distant effects of a large atmospheric release of radioac-tive material in relation to property damage and societal resources.
If the property damage and other impacts on societal resources from serious accidents are substantially larger than the effects. estimated in the Reactor Safety Study (Ref. 8), this could impact directly on siting and on the assessment of the importance of mitigation features for serious accidents.
8.4 Recommendations We recommend that research in the Systems and Reliability Analysis area be funded at a level of $23.9 million, an increase of $9.0 million over the proposed budget, for FY 1982.
41
APPENDLES 43 1
?
APPENDIX A REFERENCES 1.
Advisory Committee on Reactor Safeguards, U.S. Nuclear Regulatory Commission, " Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program for Fiscal Year 1981 - A Report to the Congress of the United States of America," NUREG-0657, February 1980.*
2.
Ad Hoc Review Group, U.S.
Nuclear Regulatory Commission, " Risk Assessment Review Group Report to the U.S.
Nuclear Regulatory Commission," NUREG/CR-0400, September 1978.**
3.
Advisory Committee on Reactor Safeguards, U.S. Nuclear Regulatory Commission, "1978 Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program - A Report to the Congress of the United States of America," NUREG-0496, December 1978.**
4.
Advisory Committee on Reactor Safeguards, U.S. Nuclear Regulatory Commission, " Comments on the NRC Safety Research Program Budget for Fiscal Year 1982," NUREG-0699, July 1980.*
G.
U.S.
Nuclear Regulatory Commission, "NRC Action Plan Developed as a Result of the TMI-2 Accident," NUREG-0660, Volumes 1 and 2, Revised August'1980.***
6.
U.S. Nuclear Regulatory Commission, " Memorandum and Order in the Matter of Petition for Emergency and Remedial Action," May 23, 1980.
7.
U.S.
Nuclear Regulatory Commissipn, "Long Range Research Plan - Fiscal Years 1983 throug51987," NUREG-0740, to be published March 1981.
8.
U.S. Nuclear Regulatory Commission, " Reactor Safety Study -
An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants," WASH-1400, October 1975.***
- Available for purchase from the NRC/GP0 Sales Program, U.S. Nuclear Regulatory Commission. Washington, DC 20555, and the National Technical Information Service, Springfield, VA 22161.
- Available for purchase from the National Technical Information Service.
- Available free upon written request to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
45
APPENDIX B GLOSSARY ACRS Advisory Committee on Reactor Safeguards B&W Babcock and Wilcox BWR Boiling Water Reactor CCTF Cylindrical Core Test Facility CRAC Calculation of Reactor Accident Consequences DOE Department of Energy EPA Environmental Protection Agency ESSOR Multi-National Research Reactor Complex at Ispra, Italy FRG Federal Republic of Germany FY Fiscal Year HLW High Level Waste
.IE Office of Inspection and Enforcement LMFBR Liquid Metal Fast Breeder Reactor LOCA Loss-of-Coolant Accident LOFT Loss of Fluid Test LWR Light-Water-Reactor NRC
. Nuclear Regulatory Commission NRU Atomic Energy of Canada Ltd., Test Reactor 46
'I PBF Power Burst Facility PKL Test Facility in Germany designed to model plant systems behavior during Loss-of-Coolant Accidents and Transients PWR Pressurized Water Reactor
~
Quality Assurance QA RELAP-5 Advanced System Code used to model Loss of-Coolant Accidents RES Office of Nuclear Regulatory Research SARA Systems and Relf ability Analysis SASA Severe Accident Sequence Analysis SCTF Slab Core Test Facility SSMRP Seismic Safety Margins Research Program TLTA Two Loop Test Apparatus TMI-2 Three Mile Island, Unit 2 TRAC Transient Reactor Analysis Code UPTF Upper Plenum Test Facility WRAP Water Reactor Analysis Package 1
47
APPENDIX C THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS The Advisory Committee on Reactor Safeguards was established as a statutory Committee in 1957 by revision of the Atomic Energy Act.
The ACRS was charged with the responsibility for review of safety studies and facility license applications submitted to it, and to make reports thereon, advising the Commission with regard to the hazards of proposed or existing reactor facilities and the adequacy of proposed reactor safety standards, and to perform such other duties as the Commission might request.
Section 182b of the Atomic Energy Act requires ACRS review of the construction permit and operating license applications for power and testing reactors and spent fuel reprocessing facilities licensed under Section 103, 104b or W4c of the Atomic Energy Act; any application for a research, developmental or medical facility 1icensed under Section 104a or c of the Act and which is specifically referred to it by the Commission; and any request for an amendment to a construction permit or operating license under Sections 103 or 104a, b, or c which is specifically referred to it by the Commission.
The Energy Reorganization Act of 1974 transferred' operation of the ACRS from the Atomic Energy Commis-sion to the Nuclear Regulatory Commission.
In 1977, Public Law 95-209 added to its other duties a requirement for the ACRS to undertake a study of reactor safety research and to prepare and submit annually to the United States Congress a report containi.79 the results of this study.
The first of these reports was submitted to the Congress in December of 1977.
48
ACRS MEMBERSHIP - January 14, 1981 CHAIRMAN:
Dr. J. Carson Mark, Division Leader, Los Alamos Scientific Laboratory, Los Alamos, New Mexico (retired)
VICE-CHAIRMAN:
Dr. Paul G. Shewmon, Professor and Chairman of Metal-lurgical Engineering Department, Ohio State University, Columbus, Ohio Mr. Myer Bender, Director of Engineering Division, lak Ridge National Laboratory, Oak Ridge, Tennessee (retired)
D r.
Max W.
Carbon, Professor and Chairman of Nuclear Engineering Department, University of Wisconsin, Madison, Wisconsin Mr. Jesse Ebersole, Head Nuclear Engineer, Division of Engineering Design, Tennessee Valley Authority, Knoxville, Tennessee (retired)
Dr. William Kerr, Professor of Nuclear Engineering and Director of the Office of Energy Research, University of Michigan, Ann Arbor, Michigan Dr. Stephen Lawroski, Senior Engineer, Chemical Engineering Divition, Argonne National Laboratory, Argonne, Illinois (retired)
Dr. Harold W.
Lewis, Professor of Physics, Department of Physics, U1iversity of California, Santa Barbara, California Mr. William M. Mathis, Director, Planning, United Nuclear Industries, Inc., Richland, Washington (retired)
Dr. Dade W.
Moeller, Chairman, Department of Environmental Health Sciences, School of Public Health, Harvard University, Boston, Mas-sachusetts Dr. David Okrent, Professor, School of Engineering and Applied Science, University of California, Los Angeles, California Dr. Milton S. Plesset, Professor of Engineering Science - Emeritus, California Institute of Technology, Pasadena, California Mr. Jeremiah J. Ray, Chief Electrical Engineer, Philadelphia Electric Company, Philadelphia, Pennsylvania (retired)
Dr. Chester P. Siess, Professor Emeritus of Civil Engineering, Univer-sity of Illinois, Urbana, Illinois Mr. David A. Ward, Research Manager of Nuclear Engineering, E. I. du Pont de Nemours & Company, Savannah River Laboratory, Aiken, South Carolina 49
ACRS SUBCOMMITTEE ACTIVITIES RESPONSIBLE ACRS SUBCOMMITTEES OVERALL REPORT NRC Safety Research Progran CHAPTER i
1.
LOCA and Transient Emergency Core Cooling System (ECCS)
Research Reactor Fuel 2.
LOFT ECCS 3.
Plant Operational Safety Reactor Operations Metal Components Structural Engineering Extreme External Phenomena 4.
Severe Accident Phenomena Class 9 Accidents and Mitigation Research Advanced Reactors 5.
Siting and Environmental Reactor Radiological Research Effects Extreme External Phenomena 6.
Waste Management Waste Management 7.
Safeguards and Fuel Safeguards and Security Cycle Safety Waste Management Reactor Radiological Effects 8.
Systems and Reliability Reliability and Probabilistic Analysis Assessment Reactor Radiological Effects 50
MEMBERSHIP 0F THE ACRS SUBCOMMITTEES f
1 NRC Safety Research Program C. P. Siess, Chairman M. W. Carbon W. Kerr S. Lawroski J. C. Micrk D. W. Moeller D. Okre'st M. S. Plesset P. G. Shewmon S. Duraiswamy, Staff Advanced Reactors M. W. Carbon, Chairman M. Bender W. Kerr J. C. Mark M. S. Plesset P. G. Shewmon R. P. Savio Staff Class 9 Accidents W. Kerr, Chairman M. Bender S. Lawroski D. W. Moeller D. Okrent P. G. Shewmon C. P. Siess D. A. Ward G. R. Quittschreiber, Staff Emergency Core Cooling System (ECCS)
M. S. Plesset, Chairman M. W. Carbon J. C. Ebersole H. Etherington*
H. W. Lewis D. Okrent D. A. Ward A. L. Bates, Staff P. A. Boehnert Staff 2
- Member Emeritus 51 i
l Extreme External Phenomena D. Okrent, Chairman M. Bender M. W. Carbon H. Etherington*
H. W. Lewis J. C. Mark D. W. Moeller C. P. Siess R. P. Savio, Staff Metsi Components P. G. Shewmon, Chairman M. Bender H. Etherington*
W. M. Mathis D. Okrent M. S. Plesset D. A. Wa rd E. G. Igne, Staff Reactor Fuel P. G. Shewmon, Chairman M. W. Carbon H. Etherington*
S. Lawroski J. C. Mark W. M. Mathis D. Okrent D. A. Ward P. A. Boehnert, Staff l
l Reactor Operations W. M. Mathis, Chai rman J. C. Ebersole H. Etherington*
D. W. Moeller D. ' 0k rent J. J. Ray D. A. Ward R. K. Major, Staff
- Member Emeritus 52 e-,
a m
- _ _ =
t-Reactor Radiological Effects D. W. Moeller, Chairman J. C. Coersole S. l_awroski D. Okrent J. J. Ray
- 3. G. Young, Staff Reliability and Probabilistic Assessment D. Okrent, Chairman M. Bender J. C. Ebersole W. Kerr H. W. Lewis J. C. Mark C. P. Siess G. R. Quittschreiber, Staff Safeguards and Security J. C. Mark, Chairman M. Bender M. W. Carbon S. Lawroski W. M. Mathis P. G. Shewmon C. P. Siess R. K. Major, Staff Structural Engineering C. P. Siess, Chairman M. Bender J. C. Ebersole D. Okrent P. G. Shewmon E. G. Igne, Staff Waste Management S. Lawroski, Chairman M. W. Carbon W. Kerr J. C. Ma rk W. M. Mathis D. W. Moeller M. S. Plesset J. J. Ray G. G.' Young, Staff 53
g 93 US NUCLEA3 RGOULATORY COMMISSION
- 1. REFO^.T NUMBER Hascyby coCJ 8IBLIOGRAPHIC DATA SHEET NUREG-0751 1 TITLE AND SU8 TITLE Mew Venume No o/ epornprunel 2- (Leave blast)
Review and Evaluation of the Nuclear Regulatory Comission Safety Research Program for Fiscal Year 1982 3 RsciPiENT S accession NO.
- 7. AUTHOR (S)
- 5. DATE REPORT COMPLE TED lVEen woNTH Februa ry 1981
- 9. PERFORMING ORGANIZATION N AVE AND MAILING ADDRESS (facevor le Coor)
DATE REPORT ISSUED Advisory Committee on Reactor Safeguards
%oNT" 5^"
U.S. Nuclear Regulatory Commission February l'1981 Washington, DC 20555 8'****'
8 (Leave bashi
- 12. SPONSORING ORGANIZATION N AME AND MAILING ADORESS Itacivde le Code /
Same as 9, above, n.cONTRAcTNO.
- 13. TYPE OF REPORT PE Rico cove mEO (lactus.ve Waws/
Report to Congress Fy 1982
- 15. SUFPLEMENTARY NOTES
- 14. (Leare 'weias
- 16. ASSTR ACT 200 words or ess)
Public Law 95-209 includes O requirement that the Advisory Committee on Reactor Safeguards submit an annual report to Congress on the safety research program of thm Nuclear Regulatory Comission. This report presents the results of the ACRS raview and evaluation of the NRC safety research program for Fiscal Year 1982.
The report contains a number of coments and recommendations.
- 17. KEY WORDS AND DOCUMENT AN ALYSIS 17a DESCRIPTORS 17b. IDENTIFIERS /OPEN.ENDE D TERMS l
- 18. AVAILAtlLITY STATEMENT
- 19. SECURITY CLASS (Th,s reporrl
- 21. NO. OF PAGES l
unclassified unlimited zogi,Tyg(ra,s,.Pri 22 price t mc Pomu 33s (7.??)
l l
m.