ML20053C513
| ML20053C513 | |
| Person / Time | |
|---|---|
| Issue date: | 01/12/1982 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1941, NUDOCS 8206020264 | |
| Download: ML20053C513 (48) | |
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The ACRS Subcommittee on Nuclear Safety Research Program he ing on 1
January 6,1982 at 1717 H Street, N.W., Washington, D.C.
The entire meeting was closed to public attendance since it involved discussion of oreliminary budget information.
Mr. Sam Duraiswamy was the Designated Federal Employee for the meeting.
A list of documents submitted to the Subcommittee is included in Attachment A.
ATTENDEES ACRS:
C. P. Siess (Subcommittee Chairman), D. Okrent, J. J. Raj, W. M. Mathis, M. W. Carbon, D. A. Ward, J. C. Mark. W. Kerr, D. W. Moeller, S. Duraiswany (Designated Federal Employee).
Principal I
NRC Speakers:
D. Ross, R. Scroggins, 0. Bassett, C. Kelber, F. Arsenault, L. Shao, G. Arlotto, F. Rowsome, K. Goller, L. Shotkin, G. Knighton, R. Bernero, M. Silberberg.
EXECUTIVE SESSION Dr. Siess, the Subcommittee Chairman, convened the meeting at 8:30 a.m. and indicated that the entire meeting was closed to public attendance, as requested by the Office of Nuclear Regulatory Research (RES) Staff, to discuss the FY 1983 budget for the NRC Safety Research programs.
He said that the main purpose of the meeting was to discuss:
'The Office of Management and Budget (OMB) final mark on the NRC FY 1983 Safety Research Proaran budget.
' Draft 2 of the ACRS report (NUREG-0864) to Congress on the NRC FY 1983 Safety Research Program.
Dr. Siess said that Part I of the ACRS report is a compilation of ACRS general reconmendations and comments and it is intended to serve as an Executive Summary of the report.
He suggested that the Subconnittee try to take a m m c m r. 7, 43 8206020264 820112 PDR ACRS 1941 pop
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Nuclear Safoty Research Program January 6,1982 position on the items identified in the proposed Part I during the subject meeting.
Part II of the ACRS report contains nine chapters, each of which re-presents a Decision Unit of the NRC Safety Research Program.
He suggested that the appropriate Subcommittee Chairmen try to make sure that Part II reflects the recommendations in Part I.
Mentioning that the RES Staff has submitted its responses, in writing, to the ACRS comments and recommendations delineated in NUREG-0795, ACR5 Report to the Commission on the NRC FY 1983 Safety Research Program Budget, he suggested that the appropriate Subcommittee Chairmen go through RES responses to see whether ACRS reconmendations have been responded adequately.
PRESENTATION BY RES - DR. D. ROSS, MR. R. SCROGGINS OttB Final Mark on FY 1983 NRC Safety Research Program Budget Mr. Scroggins reviewed briefly the FY 1983 budget for the NRC Safety Research Programs, indicating that, in its initial budget request to OMB, the Commission had included a budget of $256.5 million.
However, the OMB Target, that was developed in accordance with the Adninistration's budget reduction program, included a budget of $194.2 million.
Subsequent to receiving the OMB Target, the Commission had submitted a Reclama to OMB for a budget of $213.1 million.
But, the OMB final mark includes a budget of $195.2 million for the FY 1983 NRC Safety Research programs. He said that the 01B has specified funding level for each Decision Unit as shown below:
- $ 31.0 M LOFT
- $ 15.0 M Accident Evaluation & Mitigation - $ 47.2 M l
Advanced Reactors
- $ 13.0 M Reactor & Facility Engineering
- $ 38.0 M i
Facility Operations & Safeguards - $ 13.5 M Waste Management
- $ 13.6 H l
Siting & Environmental Research 9.0 M Systems & Reliability Analysis
- $ 14.9 M l
Total
$195.2 M l
l l
Nuclear Safety R0 search Program January 6,1982
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The allocation of the budget for various Subelements is included in Attachment B (pages 2-9).
Mr. Scroggins said that, subsequent to receiving ~the OMB final mark, the Commission tried to get an additional $20 million for the LOFT test program fron OMB that would be needed if they have to conduct all the tests identified in the RES testing matrix as directed by the Congress.
However, the OMB did not peovide any additional funds for the LOFT program other than the 515 million that is included in the G12 final mark.
DISCUSSION OF DECISION UNITS LOCA and Transient Research Mr. Scroggins said that, in it its Reclama budget request to OMB, the Commission included a total budget of $31 million for this Decision Unit; OMB final mark also includes the same budget for this Program.
Mr. Bassett discussed briefly some of the planned programs in this Decision Unit ( Attachment C, pages 1-6).
With reference to a memo from H. Denton to R. Minogue, dated December 30, 1981, regarding " Request for the Conceptual Design of a Facility for the Study of B&W and CE Integral System Characteristics" (Attachment D), Dr.
Okrent asked about the necessary funding for this program in FY 1983 and the source of fundino.
I Mr. Bassett responded that if this program is agreed to be initiated, they may need about $4 million in FY 1983 which, he believes, may have to be obtained through reprogramming of funds from the Subelements on Semiscale, Separate Effects Experiments and Model Development, and 3-D Program. He added that since they first have to formulate the program in collaboration with the Elec-tric Power Research Institute, it may be well into FY 1983 before they start spending money on this program.
U Nuclear Safety Research Progran January 5,1982 Dr. Okrent asked whether the Office of Nuclear Reactor Regulation (NRR) of the NRC has assessed the capability of the Plant Analyzer Program. Mr.
Bassett responded that RES does not have a user endorsement for this Program and they plan to request one from NRR.
He added that the industry seems to have some interest in the Plant Analyzer Progran, and he believes that Combus-tion Engineering and Westinghouse have been doing some work in this area.
Dr. Jkrent asked whether the NRC Staff has any fast-running codes to analyze B&W plants. The Staff responded that they have close to real time capability to analyze B&W plants.
')r. Ross said that sometime ago RES had sent a proposel for developing fast-running codes to NRR for endorsement. However, NRR did not endorse that proposal, indicating that it would not work.
RES now l
plans to make certain changes ~ to that proposal and resubmit it to NRR for endorsement.
Mr. Shotkin added that proposals have been requested from Los Alamos Scientific Laboratory and Idat snal Engineering Laboratory a
for developing fast-running codes.
Dr. Okrent commented that close to real-time capability is different from faster than real-time capability.
He believes that the NRC Staff should first write down their needs in this area and then develop a program to complement those needs.
He added that although it has been about three years since the Three Mile Island, Unit 2 (TMI-2) accident, the Staff does not seem to approach this issue more vignrously.
Considering that the de-monstration of Plant Analyzer Program feasibility for PWR plants is scheduled for FY 1985, which is about seven years since the TMI-2 accident, he does not believe that the Staff's efforts in this area are adequate.
Indicating that the behavior of spray distribution in BWRs is being looked at in Japan, Dr. Okrent asked whether the NRC Staff has any efforts planned to stt.dy, in general, the BWR spray distribution issue.
The RES Staff responded that they have been analyzing the test data obtained from the Japanese on this matter, and results of the analysis have been given to NRR.
Based on the available information, RES does not believe that the
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- Nuclear Safety R2s2 arch Program January 6,1982 BWR spray distribution issue is a major concern for certain plants. However,
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NRR has to make a decision on the significance of and need for further study on this issue.
Dr. OkrEnt Commented that things of this sort substantiate that overemphasis in research has been placed on PWR issues, and BWR issues have been neglected on a continuing basis.
Dr. Siess asked about the status of the FLECHT-SEASET program in FY 1983.
The NRC Staff responded that this program will be completed ir. FY 1983.
LOFT Mr. Scroggins said that, although in its revised FY 1982 budget request to OflB the NRC has requested a total budget of $29.5 million for the LOFT program with the intention of terminatina the test program by the end of FY 1982, the Congress, in its appropriation, specified that the NRC should carry out all of the tests identified in the RES testing matrix.
As a result, RES has allocated
$42 million for the LOFT Program in FY 1982.
However, if they had to conduct the full complement of tests in LOFT as directed by the Congress, the $15 million budget included in the Ot1B final mark for LOFT in FY 1983 would not be sufficient; they may need at least an additional $20 million. Consequently, subsequent to receiving the OM3 final mark, the Commission tried to get additional funding for LOFT in FY 1983. But OMB did not provide any more funding for LOFT, indicating that the LOFT funding issue has to be dealt with by the Congress itself.
Mr. Bassett said that they are proceeding to meet the Congressional objectives with a total funding of 557 million ($42 million in FY 1982 and $15 million in FY 1983) for the LOFT program. However, if they have to operate with a total budget of $15 million for LOFT in FY 1983, they have to:
' Eliminate two contingency tests identified in the RES testing matrix.
l L3 Small-Break test (Hog Leg)
LS Intermediate Break (pressurizer surge line) test
' Minimize instrumentation for the L2-6 test
- Reduce management reserve i
a I
Nuclear Safety Research Program January 6,1982 He said that RES has cone up with a modified LOFT test progran and schedule based on a total budget of $57 million ( Attachment C, pages 7 and 8) and it is awaiting NRR concurrence.
If it is approved by NRR, the final test (L2-f, large-Break LOCA test with pressurized fuel) will be conducted in January 15, 1983.
They are also discussing with Idaho Engineering Laboratory means to run tests in LOFT effectively with less budget. At present, they are not sure whether they will be able to fulfill the Con-gressional objectives with a total budget of $57 million for LOFT for FY 1982 and FY 1983; he believes that they may have a better idea a few months from now.
Mr. Bassett said that subsequent to the completion of planned tests in LOFT, the fuel will be removed from the reactor and it will be placed in a cold standby condition and maintained at that statLs with a minimum maintenance support of about 30 staff.
He mentioned that the industry and certain foreign countries continue to support the LOFT Program. Some people in the Department of Energy (DOE) are thinking about running LOFT as an international facility; if it is decided by June of 1983, then the NRC will place the facility in a hot standby status af ter completion of the planned tests.
Further, if DOE runs this as an international facility, NRC may participate to run some important tests, as necessary.
Dr. Siess asked about the funding necessary for standby activities.
Mr. Bassett responded that they may require about $3.7 million for decommissioning, and about $3 million per year for standby activities.
Accident Evaluation and Mitigation Mr. Scroggins said that the 01B has included a budget of $50 million for this Decision Unit in its final mark.
However, since OMB has drastically cut the budget for equipment, the NRC has taken about $2.8 million from this Program to provide the minimum funding necessary for equipment. As a result, the budget for this Decision Unit in FY 1983 is $47.2 million.
Mr. Bassett discussed briefly some of the planned programs in this Decision Unit ( Attachment C, pages 9-14).
l January 6,1982 Nuclear Safety Research Program Dr. Siess asked which Subelement provides support for the Power Bu Mr. Bassett responded that the PBF activities are supported (PBF) programs.
He added that this Subelement under the Subelement on Sebavior of Damaged Fuel.The allocation of this bud has a total budget of $25.7 million in FY 1983.
for various programs is as follows:
m
. $18.5 M Integral tests (PBF)
- $ 1.5 M Integral tests (NRU)
- $ 4.0 M Phenorenological Separate Ef fects, experiments ( ACRR + Lab)
Model code development, assessment, maintenance and analysis - $ 1.7
$25.7 M Total Dr. Kerr asked what types of questions that are expected to be answ The NRC Staff responded programs in the Behavior of Damaged Fuel Subelement.
d tand-that its objective is to develop data base and analytical methods for un ers It is ing and predicting core behavior over a range of severe accident sequen intended also to provide information to meet regulatory needs in the for accident management, improved risk assessment, and policy determinat It is also to help examine the TMI-2 core.
severe accidents.
Dr. Siess asked about the objectives of the PBF programs and the ty Mr. Silberberg responded questions that are expected to be answered by them.
t ding that PBF programs are intended to provide information for the unders an of damaged core configuration, core coolability characteristics, and the He added that hyorogen generation rate under the severe accident conditions.
the first series of tests in PBF are scoping tests to:
have a better understanding of what to look for in the TMI-2 1.
core examination, understanding the core configuration, and 2.
obtain a benchmark for analytical capability.
3.
definitive Dr. Okrent commented that the NRC Staff and the Comm es of criteria to review new plants in terms of features to mitigate the conse He does not believe that this program is focused on providin He does not severe accidents.
information necessary for regulatory decision-making in this area.
o Nuclear Safety Research Program January 6,1982
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believe that the PBF, Annular Core Research Reactor ( ACRR), and NRU programs will provide necessary information to meet the regulatory needs of the NRC.
In view of the lack of proper focus and clear relevence to important questions, he believes that the whole program needs to be restructured.
Dr. Ross said that a document that includes Staff's position in the severe accident area has been prepared by NRR and it is expected to be discussed by the Commission within a few days.
According to this document, operating plants in the next few years would not be required to do additional things in the severe accidant area other than those delineated already in certain Regulatory Guides, Regul ations, etc.
He mentioned that RES is preparing its plans for producing the research information needed to confirm regulatory decisions in the severe accident area and it is expected to be completed in the near future.
Dr. Okrent commented further that although ACRS has repeatedly asked the NRC Staff to look at and compare with events in foreign reactors, the NRC Staff has not done so; this kind of attitude does not seem to be appropriate.
He believes that the NRC Staff should look at foreign country approaches in certain areas, such as decay heat removal, to see why they are following that approach, and how it compares with what NRC is doing in this country.
Dr. Ross responded that they do not have a specific task to look at what other countries are doing.
Dr. Kelber added that although they do not have a specific task to look and compare the foreign country approaches in certain areas, they do exchange technical information on an informal basis.
Advanced Reactors Mr. Scroggins said that the OMB final mark includes a total budget of $13 million for this Decision Unit. Out of this $13 million, $10.5 million has been allocated to Fast-Breeder Reactor research and $2.5 million for the Gas-Cooled Reactor (GCR) Research.
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Nuclear Safety Research Program January 6,1982 Dr. Ross mentioned that RES and NRR have been directed by the Executive Director for Operations (EDO) to develop Clinch River Breeder Reactor (CRBR) licensing needs and the $10.5 million allocated for Fast-Breeder Reactors will be devoted to that purpose.
Dr. Siess said that the ACRS has been recommending that the Congress provide separate funding for the Liquid Metal Fast Breeder Reactor (LMFBR) resea rch.
However, Congress continues to appropriate money for developmental effort without apprcpriating for research to keep up with the developmental work.
Dr. Siess asked whether the $2.5 million budget for the GCR is intended for research related to Fort St. Vrain reactor.
Mr. Scroggins said that most of the budget for GCR is not in support of the Fort St. Vrain reactor research; it is directed toward licensing needs for GCRs expected to be proposed for con-struction in the future and is responsive to Congressional directives.
Mr. Bassett discussed briefly the ongoing and planned efforts in this Decision Unit ( Attachment C, pages 15-17).
Dr. Mark asked who is going to do the research related to sodium fires. The NRC Staff responded that the Department of Energy (DOE) has an extensive research program in this area.
In addition, research in this area is being done in France.
The NRC Staff believes that there is no need for them to have a separate progran to deal with this issue because they believe that they can get data from DOE and France. However, since they have limited access to French data, coupled with the possibility that they may encounter some unanticipated issues during the course of the licensing review, they may have to initiate a separate program in this area.
Dr. Carbon asked whether the NRC Staff has a program planned to examine the core. Dr. Kelber responded that DOE has completed a series of tests on the control characteristics of a heterogeneous core. He believes that the NRC Staff could use that information for the examination of the core, i
e January 6,1982 Nucleer Safety Research Pr: gram Dr. Carbon asked w' ether the NRC Staff has a program to provide information p
cith regard to the type of questions to be asked of DOE during the evalua-Dr. Ross responded that they do not have such tion of a heterogeneous core.
a program.
m Dr. Carbon commented that it may be very difficult to ask the right questions He believes that the NRC Staff should have a on the spur of the momant.
program to provide information to esk the right questions of DOE and to evaluate the core.
Dr. Siess commented that if the NRC Staff has to depend on DOE researen in this area, they should have a separate program to enable them to ask the right questions of DOE, and also to make sure that they are getting the right answers.
Dr. Siess asked that if the NRC Staff had additional money for the Advanced Mr. Bassett Reactor research, do they have plans developed to spend it.
l responded that they have ott.er programs that they could conduct with addi funding.
Reactor And Facility Engineering Mr. Scroggins reviewed the budget, indicating that the Ot1B allowance for th Decisior Unit is $38 million, which is about the same as the budget requested by the NRC.
With regard to NRC Staff's efforts to study aging effects, Mr. Arlotto said that, in response to the ACRS recommendation on this issue, the NRC Staff has developed a scoping study to examine the aging degradation of They plan to spend about $160,000 in f Y 1982 and about safety equipment.
He said that they want
$350,000 in FY 1983 for this scoping study on aging.
to find out whether this is a major safety problem prior to getting involved Based on the results of this scoping study, they plan deeply in this area.
to develop a detailed research program, m
Dr. Okrent asked about the basis for allocating such small funding for the Dr. Ross responded that they aging study as compared to the funding for PBF.
l Nuclear Safety R2 search Program January 6,1982 4
are not sure about how much money will be spent in this area since they do not have a detailed research program.
Btsad on the resuits of the scoping study, he believes that ttay would be able to develop a detailed program and the necessary funding therefor. Dr. Ok ent cwnented that what is being done in the aging area is P.ot systemat'ic; further, the scoping study approach to
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q define a research program on aging may be tin.e consuming. He suggested that the Staff invite some experts in this area and cxoduct a workshop to identify important issues that need to be looked at and thereby develop a research program.
Mr. Scroggins responded thit he agre'.'s with Dr. 0krent that aging is an.
important issue that needs to be looked at more carefully. They will take a look at the adequacy of the ongoing efforts in this area, and also try to conduct a workshop as suggested by Dr. Okrent to obtain some expert opinions to deal with this issue.
t Dr. Siess askad about the NRC Staff's efforts to study the issues associated with a water-hammer event.
Mr. Shao responded that some work is being done under the Technical Assistence Program. Thit. issue is also listed as an Unre-solved Safety Issue. Until NRR defines this problem better, RES may not be able to do research in thfs area.
Mr. Knighton, from NRR, said that water hammer, which occurs ' periodically in nuclear plants,-has not been considered as an overwhelming problem.
Dr. Siess commented that water hammer may cause a double-ended pipe break, because it may develop a longitudinal force on the pipe.
He suggested that RES and the group that handles the Unresolved Safety Issues look at this problem carfefully to find out different mea?s for fixing it.
l Dr. Okrent pointed out that, in its report to the Nuclear Safety Oversight Committee, the Reactor Safety Resarch Review Grcup identified water hammer as I
an important item for research.
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f Nuclear Safety Research Program January 6,1982 Dr. Ross said that the NRC Staff does not have an official position on the recommendations included in the Reactor Safety Research Group Report to the Nuclear Safety Oversight Committee. Further, he believes that water hammer is a design problem and he is not sure how research can prevent this problem. fir. Knighton added that, based on operating experience and on what industry has done in this area, the NRC Staff is in the process of developing a NUREG document; he does not believe that water-hammer is a serious problem that warrants research.
l Dr. Okrent commented that water hammer under certain transients may lead to loss of sone safety systems.
He believes that there is a need for research in this area to look at systems that are prone to water hamer, the consequences of water hanner, etc.
He believes that the Systems and Reliability Analysis (SARA) Decision Unit should include a research program to study water-hammer problems.
Dr. Ross said that he agrees that water-hamer issue should be looked at and he believes that it coulc be included under the SARA Decision Unit.
Facility Operations And Safeguards Mr. Goller said that OPIB has provided a total budget of $13.5 million for this Decision Unit, which is much less than that requested by the NRC in its Reclama budget request.
As a result, they have to cut or delay several programs in this Decision Unit in FY 1983 (Attachment C, page 18).
With reference to the ACRS recommendation in NUREG-0795 that this Decision Unit should include taose studies necessary for the NRC to be able to develop requirements with regard to design to protect against sabotage by an insider, Dr. Okrent asked whether the NRC Staff believes that the research being done is adequate to prevent sabotage by an insider in new plants.
The NRC Staff responded that it is very hard to define how much is " adequate". However, they believe that the current research is sufficient.
Dr. Siess asked whether, based on the available information, the NRC Staff will be able to specify criteria to prevent a saboteur from doing something that
t Nuclear Safety Researen Program January 6,1982 could lead to a core melt.
Mr. Goller responded that with the available information, they could get a good start in specifying criteria and then they could improve that later on.
He said that he does not believe that the Commission wants to go any further in this area in view of the budget con-straints.
Dr. Okrent asked whether the NRC Staff has any plans to look at the reliability of non-Class IE information in the control roon.
Mr. Wenzinger responded that they have a contract with Idaho National Engineering Laboratory (INEL) to assess the instruments needed to diagnose plant status in both normal and abnormal conditions.
However, INEL has been asked to look at the perfonnance require-ments of instruments and not the reliability requirements.
Dr. Okrent commented that the NRC Staff does not seem to have a cohesive progran to look at the relf 'ility of instruments.
Dr. Siess asked uat if the NRC Staff gets more money for this Decision Unit, what is their priority in spending that money.
Mr. Goller responded that, in his opinion, they may spend it on the Occupational Protection Program.
Indicating that in its report to the Nuclear Safety Oversight Committee the Reactor Research Review Group has identified Human Engineering as a high priority research item, Dr. Okrent wondered how the NRC Staff intends to spend more money on the Occupational Protection Progran instead of the Human Engineer-ing Program.
Mr. Ward asked why so much emphasis is being placed in the Human Engineering Program on research to support improvements in control room design. The Staff responded that the level of emphasis on control room design improve-ments is based on the TMI-2 accident experience.
Waste Management Mr. Arsenault discussed briefly various programs in this Decision Unit and the associated funding levels for FY 1983 (Attachment C, pages 19-21). He said that the Ot1B final mark includes a budget of $13.6 million for this Decision Unit which is $4.4 million less than that requested by the NRC in its Reclama budget request to OMB.
Nuclear Safety Research Pr: gram January 6,1982 Dr. Moeller asked about the impact of the budget reduction on the site characterization research program that is included under the High Level Waste Subelement. Mr. Arsenault responded that he believes that the curtailment of the budget may not slow the research efforts in this area; however, it may reduce the NRC Staff's level of confidence in the criteria they develop in this area.
Dr. Moeller asked whether the NRC Staff has a program to look at the possible impact of glaciation on the high level waste respository.
Mr. Arsenaul t re-sponded that they do not have any program on glaciation.
Siting And Environmental Research Mr. Arsenault said that the OPIB final mark includes a budget of $9 million for this Decisinn Unit for FY 1983, which is $3.5 million less than that requested by the NRC in its Reclama budget request to 0'iB.
He discussed briefly the Pro-grams in this Decision Unit and the associated funding levels ( Attachment C, paga 22).
Dr. Okrent asked about the purpose of and the need for the research program on Meteorology.
Mr. Arsenault responded that it is to:
' Conduct field tests to validate the meteorological models.
' Improve the meteorological models for application to nuclear power pl ants.
' Evaluate the meteorological dispersion during an accident.
Dr. Okrent commented that it is not clear to him how the infomation obtained from this research program will be used. He suggested that it would be helpful if the NRC Staff provided additional information on their research efforts in the meteorological area.
Mr. Arsenault agreed to provide such information at a later date.
Dr. Okrent suggested that the NRC consider doing some research in the area of Offsite Decontamination following radioactive releases fron nuclear power plant accidents.
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Nuclear Safety Research Program January 6,1982 Systems And Reliability Analysis The Subcommittee did not discuss this Decision Unit since it was discussed in detail on January 5,1982 at the ACRS Subcommittee meeting on Reliability and Probabilistic Assessment.
DISCUSSION OF THE DRAFT 2 0F THE ACRS REPORT (NUREG-0864) TO CONGRESS Owing to lack of time, the Subcommittee did not discuss the ACRS Report to Congress on the NRC FY 1983 Safety Research Program. Dr. Siess suggested that the Subcommittee members take a lock at the proposed recommendations included in Part I of the ACRS Report and bring their comments for discussion at the full Committee meeting on Januarty 7,1982.
FUTURE MEETING No further Nuclear Safety Research Program Subconnittee meeting is scheduled at this time to discuss the ACRS Report to Congress, However, it is scheduled to be discussed at the January 7-9, 1982, and February 4-6, 1982 full Committee meetings.
Dr. Siess thanked all participants and adjourned the meeting at 6:30 p.m.
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Nuclear Safety Research Program Meeting January 6,1982 LIST OF DOCUMENTS SUBMITTED TO THE SUBCOMMITTEE i
1.
OMB final mark on the NRC FY 1983 Safety Research Program Budget.
2.
Draft 2 of the ACRS Repnet to Congress on the NRC FY 1983 Safety Research Program.
3.
Presentation materials by RES.
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ATTACHMENT A
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1/6/82 FY 1983 CONGRESSIONAL NUCLEAR REGilLATORY RESEARCH (DOLLARS IN MILLIONS) d FY 83 COL. 5 FY 82 FY 83 ACRS FY 83 vEnsus FY 81 APPR.
REO. 9/81_
RE_Cm CONG.
00ld
$ 16.2
$ 30.9
$ 31.0
$ 31.0
$ 31.0 1
LOCA & TRANSIENT LOFT 41.6 42.0
'10.6 14.5 15.0 0.5 ACCIDENT EVALUATION &
MITIGATION 26.3 33.1 50.4 50.4 47.2
-3.2 ADVANCED REACTORS 9.7 7.5 23.5 22.5 13.0
-9.5 REACTOR & FACILITY ENG.
28.4 33.8 38.2 40.2 38.0
-2.2 FACILITY OPER. & SAFE-GUARDS 12.6 13.0 16.8 19.8 13.5
-6.3 WASTE MANAGEMENT 10.2 12.2 19.6 19.6 13.6
-6.0 SITING & ENVIRONMENT 12.9 9.0 14.7 14.7 9.0
-5.7 SYSTEMS & RELIABILITY ANALYSIS 13.9
_15.1 21.7 25.7 14.9
-10.8 TOTAL PROG. SUPPORT
$201.8
$196.6
$256.5
$238.4
$195.2
$-43.2 1
RTincNN6NT b
1/6/82 FY 1983 CONGRESSIONAL NilCLEAR RE6tiLATURY RI.SEARCil (DOLLARS IN MILLIGNS)
FY 82 FY 83 FY 33 FY 81 APPR.
REO. 9/81 CONG.
$ 3.1 5 7.5
$ 7.5
$ 8.1 SEPARATE EFFECTS EXP.
& MODEL DEV.
10.5 6.0 6.6 6.1
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3-D PROGRAM 9.fi 6.0 6.5 6.5 CODE IMPROVEMENT & MAINT.
6.2 3.li 2.6 2.5 CODE ASSESSMENT & APPL.
fl.2 5.5 6.9 6.9 FUEL BEHAVIOR UNDER OPER.
TRANSIENTS 7.8 2.5 0.9 0.9 TOTAL PROG. SUPPORT
$'16.2
$30.9
$31.0
$31.0 82
1/6/82 FY 1983 CONGRESSIONAL NilCLEAR REGULATORY RESEARCH (DOLLARS iPI fillL10NS)
FY 82 FY 83 FY 83 FY 81 APPR.
REO. 9/81 CONG.
ACCIDENT EVALUATION _4 MITIGATION BEHAVIOR OF DAMAGED FUEL
$13.7
$18.0
$25.7
$25.7 FUEL MELT 5.8 7.1
' 13.7 10.5 FISSION PRODUCT RELEASE TRANSPORT 3.6 4.1 5.4 5.4
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ACCIDENT MITIGATI0ff 3.2 3.9 5.6 5.6 TOTAL PROGRAM SUPPORT
$26.3
$7& 1 6 0. 11 Wa 8-3
1/6/82 FY 1983 CONGRESSIONAL NUCLEAR REGULATORY RESEARCH (DOLLARS IN MILLIONS)
FY 82 FY 83 FY 83 FY 81 APPR.
REQ.9/81 CONG.
ADVANCED REACTORS FAST REACTORS
$ 6.9
$ 5.0
$21.0
$10.5 l
GAS-COOLED REACTORS 2.8 2.5 2.5 2.5 TOTAL PROGRAM SUPPORT
$ 9.7 5 7.5
$23.5
$13.0 e
4 8-4
1/6/82 4
1 FY 1983 CONGRESSIONAL NUCLEAR REGULATORY RESEARCH (DOLLARS IN MILLIONS) l FY 82 FY 83 FY 83 FY 81 APPR.
REQ. 9/81 CONG.
REACTOR & FACILITY ENGINEERING MECHANICAL / STRUCTURAL ENG.
$ 9.6
$11.0
$12.7
$12.5 PRIMARY SYSTEM INT.
12.2 15.3 17.4 17.4 ELECT. EQUIP. QUAL.
2.5 3.7 3.8 3.8
[
FUEL CYCLE FAC. SAFETY 1.9 1.1 1.2 1.2 EFFLUENT CONTROL & CHEM.
SYS.
0.6 1.0 1.4 1.4 DECOMMISSIONING 1.6 1.7 1.7 1.7 TOTAL PROGRAM SUPPORT
$28.4
$33.8
$38.2
$38.0 8-5 9
1/6/82 FY 1983 CONGRESSIONAL NUCLEAR REGULAT9RY RESEARCH (DOLLARS IN MILLIONS)
FY 82 FY 83 FY 83 FY 81 APPR.
REO. 9/81 CONG.
FACILITY OPER. & SAFE-GUARDS HUMAN ENG./ MAN-MACHINE
$ 3.9
$ 4.2
$ 6.3
$ 5.1 PLANT INST. & CONTROL 1.5 2.7 4.7 3.6 0CCUPATIONAL PROTECTION 2.0 2.0 2.6 2.1 EMERGENCY PREP.
0.2 0.5 0.5 0.5 SAFEGUARDS 5.0 3.6 2.7 2.2 TOTAL PROGRAM SUPPORT
$12.6 113.0
$16.8
$13.5 1
86
(-
1/6/32 FY 1983 CONGR.ESSIONAL NUCLEAR REGULATORY RESEARCH (DOLLARS IN MILLIONS)
FY 82 FY 83 FY 83 FY 81 APPR.
REO. 9/81 CONG.
WASTE MANAGEMENT HIGH LEVEL WASTE
$ 5.2
$ 5.6
$ 9.6
$ 6.1 LOW LEVEL WASTE 3.2 4.0 5.5 4.3 URAN. REC 0VERY 1.8
- 2. 6 4.5 3.2 TOTAL PROGRAM SUPPORT
$10.2
$12.2
$19.6
$13.6 9
l
- g. 7 j
1/6/82 FY 19P,3 CONGRESS 10f!AL NUCLEAR REGlil.ATORY RESEARCH (DOLLARS IN MILLIONS)
FY 82 FY 83 FY 83 FY 31 APPR.
RE4. 9/81 CONG.
SITING & ENVIRONMENT EARTH SCIENCES
$ 6.6
$ 5.1
$ 7.1
$ 5.5 SITING 1.4 1.2 2.8 1.5 l
HEALTH EFFECTS 3.2 1.5 3.0 2.0 ENVIRONMENTAL IMPACTS 1.7 1.2 1.8 0
TOTAL PROGRAM SUPPORT
$12.9
$ 9.0
$14.7
$ 9.0 B-8
1/6/82 FY 1983 CONGRESSIONAL NUCLEAR REGtlLATORY RESEARCH (DOLLAltS IN MILLIONS)
FY 82 FY 83 FY 83 FY 81 APPi!.
REQ. 9/81 CONG.
SYSTEMS & RELIABILITY ANALYSIS RISK METHODS & DATA EVAL.
$ 7.2 5 6.5
$ 8.0
$ 4.9 REACTOR RISK & REL. ANAL.
4.8 7.1 11.0 8.5 TRANSPORTATION & MATL.
RISK 1.9 1.5 2.7 1.5 TOTAL PROGRAM SUPPORT
$13.9
$15.1
$21.7
$14.9 i
3 39
SEMISCALE SUBELEMENT o
C0flDUCT A LOSS OF ALL EXTERNAL POWER TEST SERIES BY 3/83 o
COMPLETE A STEAM GENERATOR TUBE BREAK (AND ASSOCIATED COMPLICATIONS) TEST SERIES IN FY 83 o
COMPLETE REPORTING OF THE SIEAM GENERATOR SECONDARY SIDE UPSET TEST SERIES o
COMPLETE RELAP 5 LWR TRANSIENT CODE o
PERFORM PRE AND POST TEST CALCULATIONS o
PERFORM COMPLETE DATA ANALYSES o
CONDUCT SPECIAL STUDIES FOR NRC AND, IF NECESSARY, PERFORM TESTS PRESENTLY NOT SCHEDULED TO SUPPORT NRC ACTIVITIES o
PLAN TEST SERIES FOR FY 8I4 COVERING UPSETS NOT PREVIOUSLY COVERED, INCLUDING TESTS NOT CONDUCTED IN LOFT o
CONTINUE ACTIVITIES IN SUPPORT OF A CONVERSION TO THE B&W CONFIGURATION (MOD 5)
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SEPARATE EFFECTS EXPERIMENTS Af1D MODEL DEVELOPMENT SUBELEMENT FLECHT-SEASET PROGRAM WILL COMPLETE THE ANALYSIS AND ISSUE FINAL REPORTS FOR THE NATURAL CIRCULATION AND BLOCKED BUNDLE TEST.
FIST PROGRAM WILL COMPLETE CONSTRUCTION AND INITIATE SHAKEDOWN TESTING PRIOR TO PHASE I TESTING THERMAL MIXING EXPERIMENTS AND DVERC00 LING SUPPORT EXPERIMENTALs PROGRAM TO BE CONDUCTED AT INEL WILL BE COMPLETED TWO PHASE FLOW MODELING AND HEAT TRANSFER C00 RELATION DEVELOPMENT AND DATA COMPARISON FOR THE ADVANCED CODES (TRAC AND RELAP 5)
WILL BE CONTINUED TWO PHASE FLOW INSTRUMENTATION TO f1EASURE VOID PROFILES AND FLOW WILL BE IMPROVED i
c-2,
3D-PROGRAM SUBELEMENT CYLINDRICAL CORE TEST FACILITY (CCTF) W o
USING CORE 11 BY THE END OF FY 1983 SLAB CORE TEST FACILITY (SCTF) WILL C0fiPLET o
CORE I BY MID FY 83 AND START MODIFICATION CORE 11 WHICH WILL BE COMPLETED BY EARLY 198f4 j
NRC WILL DEllVER INSTRUMENTATION NECESSA o
IN MID FY 83 TWENTY TWO TRAC CALCULATIONS FOR SCTF, CCTF AND U o
COMPLETED l
C-3
F
~
CODE IMPROVEMENT AND MAINTENANCE o
RELEASE TRAC-PF1/ MODI, TRAC-BDl/ MODI AND RELAP5/ MOD 2 TO THE NESC.
EMPHASIS ON MODELLING BOP COMPONENTS AND CONTROL FUNCTIONS o
RELEASE COBRA-TF TO THE NESC, SUB-CHANN L AND CONTAINMENT MODELLING o
RELEASE NUFRED BWR STABILITY CODE TO THE NESC INDEPENDENT CODE ASSESSMENT o
COMPLETE ASSESSMENT OF TRAC-PF1 AND RELAP5/ MODI, FOR FAST ANALYSIS OF SBLOCA IN PWR'S o
COMPLETE ASSESSMENT OF TRAC-BDI, FOR DETAILED ANALYSIS OF LOCA IN BWR'S CODE APPLICATIONS l
o COMPLETE ANALYSIS OF PRESSUR12ED THERMAL SHOCK IN ll AND CE PLANTS o
COMPLETE ANALYSIS OF PARTIAL ATWS IN BWR'S o
RESPOND TO ANALYSIS REQUESTS FROM NRR AND OTHER USERS o
DE50NSTRATE CAPABILITY OF BWR PLANT ANALYZER BA l
RAMONA-3B c-9
PLANT ANALYZER SCOPE AND OBJECTIVES o
A PRODUCT OF NRC CODE DEVELOPMENT EFFORT o
USING EXISTING NRC DEVELOPED AND ASSESSED SYSTEM CODES FOR LWR SAFETY ANALYSIS '. TRAC-PWR, TRAC-BWR, RAMONA-3B AND/0R RELAP 5) e o
CONVERT TO EXISTING COMPUTER AND DISPLAY HARDWARE o
ACHIEVE FASTER THAN REAL-TIME COMPUTING SPEED o
PROVIDE USER-0RIENTED INPUT /0UTPUT INTERFACES AND DISPLAYS, INCLUDING INTERACTIVE CAPABILITY 9
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PLANTANALY2ERPLANS o
CONVERT THE RAMONA-3B BWR SYSTEM CODE WITH KINETICS TO THE AD-10 MULTIPROCESSOR COMPUTER IN FY 82 DEMONSTRATE FAST RUNNING TIME IN FY 83 o
RECElVE PROPOSALS BY 4/82 FROM INEL AND LANL TO PERFORM COMPARABLE CONVERSIONS FOR THE CODES THEY HAVE DEVELOPED (TRAC-PWR, TRAC-BWR AND/0R RELAP 5) o DEMONSTRATE PWR PLANT ANALYZER FEASIBILITY IN FY 85 o
DEMONSTRATE BWR PLANT ANALYZER FEASIBILITY IN FY 85 c-6
FY 1982 - 1983 LOFT PROGRAM FY 1983 d
FY 1982--
=
($45M)
OCT JAN APR JUL OCT JAN APR JUL OCT
+
l l
1 I
l i
I I
I I
I I
I I
I I
I
($38M)
A2/F1 L9-3 L6-6 L2-5 L6-8 LS-4*
L3-8' L5-2*
F1/F2 L2-6 L8-1A
($30M)
+
1 l
l 1
I I
I I
($15M)
I L6-6/L9-3 FACILITY IN hot STANDBY.
($42M)
+
i i
i i
l i
i I
i i
i i
i i
I
($15M)
A2/F1 L9-3L6-6 L2-5 L6-8L9-4* F1/F2 L2-6 t
- CONTINGENCY IESTS t HIGH-PRESSURE CENTER BUNDLE w/ MIN INSTR.
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LOFT TEST FROGRN1 mlE TEST DESCRIPTION mRCH1982 l.9-3 Al}S INITIATED BY A LOSS OF FEED mTER.
APRIL 1982 L6-6 BORON DIllITION OPERATI0mL TPANSIENT.
MY 1982 L2-5
" WORST-CONDITIONS" LARGE-BEAK LOCA.
AUGUST 1982 L6-8 CONTROL-RDD WITIMAPAL SET OF OPERATIOML TRNSIENTS.
stratrBER1982 L911 AINS INITIATED BY A LOSS OF 0FFSITE POWER.
NORBER1982 Pull 350 PSI PRESSURIZED CONTROL BUNDLE.
INSERT 600 PSI PRESSURIZED CONTROL BlFRE.
JANUARY 1983 12-6 LARGE-BREAK LOCA; ECC DELAYED UNTIL FLEL CLAD FAILURE IN HIGi ALPM-PHASE.
ex
l l
l l
BEHAVIOR OF DAMAGED FUEL SUBELEMENT l
o 0BJECTIVE l
DEVELOP DATA BfSE AND ANALYT! CAL METil0DS FOR UNDERSTANDING AND PREDICTING CORE BEHAVIOR OVER A RANGE OF SEVERE ACCIDErlT SEQUENCE o
REGULATORY NEEDS ACCIDENT MANAGEMENT IMPROVED RISK ASSESSMENT POLICY DETERMINATIONS FOR SEVERE ACCIDENTS o
KEY ELEMENTS INTEGRAL, MULTI-EFFECT, SCOPING AND PROOF TESTS (PBF, NRU)
PHENOMEN0 LOGICAL, SEPARATE EFFECT EXPERIMENTS (ACRR + LAB.)
SEVERE CORE DAMAGE ANALYSIS PACKAGE (SCDAP)
INFORMATION FROM TMI-2 CORE EXAMINATION
-a
BEHAVIOR OF DAMfGED FUEL SUBELEMENT o
COMPLETE THREE PHASE I SFD TESTS IN PBF o
PREPARE FINAL PHASE I SFD TEST FOR PBF o
PLAN FOR PBF PHASE II TESTS o
DOCUMENT NRU LOCA SIMULATION TESTS COMPLETED IN FY 82 o
EVALUATE /PLAtl NRU-SFD FOLLOW-ON TESTS o
COMPLETE DAMAGED FUEL CONFIGURATION DRY-Ol!T MODEL VERIFICATION EXPERIMENTS IN ACRR FOR HIGH PRESSURE o
START DRY-0UT MODEL VERIFICATION EXPERIMENTS FOR OTHER PARAMETERS (FUEL SHAPES, MIXTURES) o COMPLETE FIRST EXPERIMEflTS IN ACRR ON DAMAGED Ft'EL CONFIGURATION FORMATION, PROGRESSION, AND RELOCATION o
COMPLETE FIRST MODIFICATIONS TO SCDAP AND ISSUE MOD I o
START LABORATORY EXPERIMENTS ON HIGH TEMPERATURE MATERIAL INTERACTIONS AND PROPERTIES TO COMPLEMENT C0flTINUING WORK IN FRG (KFK) o CONTINUE PARTICIPATION Ifl THE PLANNING / IMPLEMENTATION OF TMI-2 CORE EXAMINATION ooO
FISSION PRODUCT RELEASE AND TRANSPOP.T SURELEMENT o
COMPLETE UPDATED SOURCE TERM ESTIMATES OmREG-0772 FOLLOW-0N)
FOR REASSESSMENT (PPG) o COMPLETE PHASE I TESTS (2000 C) ON F.P. RELEASE FROM IRRADIATED FUEL AND START HIGHER TEMPERATURE TESTS.
o CONTINUE AEROSOL RELEASE TESTS FROM FUEL o
COMPLETE AND ISSUE TRAP-MELT (VERSION 3) o INITIATE TRAP MELT VERIFICATION TESTS FOR REACTOR COOLANT SYSTEM o
COMPLETE AEROSOL TRANSPORT TESTS IN STEAM (NSPP) o COMPLETE A00E0VS AND VAPOR PHASE CHEMISTRY OF TE
CORE MELT INTERACTION FY 83 o
CONDUCT LARGE SCALE FUEL MELT / CONCRETE TEST o
HOT SOLID DEBRIS / CEMENT INTERACTION TESTS (CEMENT TO INCLUDE BASALT, LIMESTONE AND HAC o
DESIGN FOR SUSTAINED HEATING COOLANT / CORE DEBRIS INTERACTIONS (INT 0 WATER AND WATER ONTO M0LTEN CORE DEBRIS) o o
STEAM EXPLOSION PHENOMENA (FCI RELATED TO THREAT TO RPV WILL SHIFT TO ACCIDENT PROGRESSION AND STEAM SPIKE)
HYDROGEN RESEARCH PROGRAMS FY 83 ACCOMPLISHMENTS EXPERIMENTAL ASSESSMENT OF FLAME ACCELERATION EFFECTS o
IMPROVED HYDROGEN TRANSPORI :0DE (COVERING VARIOUS RELEASE RATES) o o
ASSESSMENT OF PRE-INERTING, 02 DEPLETION AND POST-ACCIDENT C0INERTING 2
o INITIAL ASSESSMENT OF HYDROGEN AUTOIGNITION PROOF 0F CONCEPT FOR DELIBERATE FLARING IN CONJUNCTION WITH HIGH POINT VE o
COMPLETE STUDIES OF HYDROGEN GENERATION RATES AND PRODUCT MORPHOLOGY FOR ZN, AL o
AND ORGANIC C0ATED MATERIALS IN CONTAINMENTS o
DEVELOP MODELS FOR ASSESSING THE EFFECTS OF H 2 BURNS ON EQUIPMENT SURVIVAL -
COMPARE WITH EXPERIMENTS
-)
CORE RETENTION STUDIES o
ANALYSIS OF FIRST LARGE SCALE TEST RESULTS (BRICK CONSTRUCTIONS) o RETROFIT RETAINMENT CONCEPTS AND ANALYSIS CASTABLE CERAMICS (VIZ. HAC) AND RESISTANT AGGREGATES o
INITIATE HEAT LOSS STUDIES TO STRUCTURE (FOR OFFSHORE PLANTS AS WELL AS FIXED B o
SEVERE ACCIDENT SEQUENCE ANALYSIS (SASA)
EVALUATE PROCEDURAL AND PLANT CHANGES SUPPORTING SEVERE ACCIDENT POLICY o
o EVALUATE SAFETY IMPLICATIONS OF CONTROL SYSTEMS EVALUATE PLANT ABNORMAL AND EMERGENCY OPERATING PROCEDURES o
CHARACTERIZE PLANT BEHAVIOR UNDER COMPLEX TRANSIENT WITH MULTIPLE FAILURES o
o TRANSPORT OF RADIONUCLIDES IN SEVERE ACCIDENTS r.,,,
FAST REACTORS FY 83 CONTINUED EFFORTS IN SUPPORT OF CRBR LICENSING COMPLETE BALANCE OF PLANT MODELING IN SSC AND ANALYSE DESIGN BASIS ACCIDENTS AND o
TRANSIENTS ANALYSE TEMPERATURES AND FLOWS DURING NORMAL OPERATION AND NATURAL CONVECTION WIT o
SSC AND COMMIX o
ANALYSE KEY PROBLEMS IN ACCIDENT ENERGETICS FROM CDA'S CARRY OUT TRANSITION PHASE MODELING TESTS AND OTHER TESTS FOR SIMMER VERIFICAT o
o COMPLETE S0DIUM - CONCRETE INTERACTION TEST AND MODELING COMPLETE AND APPLY THE CONTAIN CODE TO ACCIDENTS THAT THREATEN CONTAINMENT o
o CARRY OUT AUDIT OF CRBR PROBABILISTIC RISK ASSESSMENT o
COMPLETE LMFBR AEROSAL RELEASE AND TRANSPORT PROGRAM
FAST REACTOR FY 83 PROPOSED NEW PROGRAMS DEVELOP BASIS FOR REGULATIONS ON OCCUPATIONAL RADIATION PROTECTION AT LMFBR'S o
DEVELOP A BASIS FOR HUMAN FACTORS REVIEW 0F LMFBR'S (CONTROL ROOMS, STAFFING, ETC.)
o ASSESSMENT OF CONTROL, PROTECTION, AND INSTRUMENTATION SYSTEMS TO IDENTIFY AREAS o
NEEDING CONFIRMATORY RESEARCH o
EVALUATE THE POTENTIAL OF MECHANICAL DEFECTS AS ACCIDENT INITIATION o
EVALUATE DETECTION AND MITIGATION OF MECHANICAL DEFECTS o
DEVELOP SITE SUITABILITY SOURCE TERMS, t
c /6
a ADVANCED REACTORS FY 83 HIGH TEMPERATURE GAS-COOLED REACTOR SAFETY RESEARCH ASSUMPTION:
HTGR - SC/C LEAD PLANT UTILITY AND SITE TO BE IDENTIFIED IN FY 82-83, PSAR TO BE SUBMITTED IN 1985 AND CONSTRUCTION AND O. L. TO BE COMPLETE IN 1993.
O COMPLETE DATA BASE FOR LEAD STEAM CYCLE /C0 GENERATION PLANT PRE-APPLICATION REVIEW o
COMPLETE EVALUATION OF EXISTING AIPA AND ASSESSMENT OF RESEARCH PRIORITIES o
DEVELOP HTGR - SPECIFIC SITE SUITABILITY SOURCE TERM, SITING CRITERIA AND POSSIBLE SITING RULE AMENDMENT FOR HTGR'S o
FURTHER DEVELOPMENT OF HTGR FISSION PRODUCT BEHAVIOR, HTGR COMPONENT STRUCTURAL TECHNOLOGY AND SYSTEM ANALYSES i
e
.co
IMPACT OF FY 83 OMB ON FACILITY OPERATIONS AND SAFEGUARDS DECISION UNIT PROGRAMS THAT WILL BE SUBSTANTIALLY DELAYED AND/OR CUT BACK FAC OPER & SAFEGUARDS RESEARCH FOR ADVANCED REACTORS REG GUIDES ON HUMAN FACTORS CRITERIA FOR EQUIPMENT DESIGN EVALUATION OF PROBLEMS IN WORKER EXTREMITY DOSIMETRY OPTIMIZATION STUDIES FOR OCCUPATIONAL ALARA LWR DOSE REDUCTION: LOW-MAINTENANCE EQUIPMENT, WASTE HANDLING, COLLECTIVE DOSE INCENTIVES EVALUATION OF THE SAFETY IMPLICATIONS OF CONTROL SYSTEMS AND ASSOCIATED PLANT ELECTRIC SYSTEMS t
EVALUATION OF INSTRUMENTATION FOR ACCIDENT MONITORING EVALUATION OF ON-LINE RADIATION MONITORS FOR REACTOR COOLANT RESEARCH ON APPLICATION OF STRATEGIC ANALYSIS TO SAFEGUARDS
@-C.
~
1/6/82 1
I FY 1983 CONGRESSIONAL NUCLEAR REGULATORY RESEARCH (DOLLARS IN MILLIONS)
)
FY 83 WASTE MANAGEMENT CONG.
HIGH LEVEL WASTE WiSTE FORM ANDPACKAGE PERF.
$ 1.11 NEAR-FIELD MIGRATION AND GE0CH.
1.7 REPOSITORY DESIGN AND ENG.
0.7 SITE SUITABILITY 1.1 PERFORMANCE ASSESSMENT 1.1 l
l STANDARDS AND TECH. ASST.
0.1
$ 6.1 O-I n
1/6/82 2
FY 1983 CONGRESSIONAL NUCLEAR REGULATORY RESEARCH (DOLLARS IN MILLIONS)
FY 83 WASTE MANAGEMENT CONG.
LOW-LEVEL WASTE WASTE FORM AND PACKAGE PERF.
$ 1.0 DESIGN, OPER. AND MONITORING 0.8 SITE SUITABILITY 1.3 PERFORMANCE ASSESSMENT 0.7 STANDARDS AND TECH. ASST.
0.5
$ 14.3 O
1/6/82 3
~
FY 1983 CONGRESSIONAL NUCLEAR REGULATORY RESEARCH (DOLLARS IN f11LLIONS) 4 FY 83 WASTE MANAGEMENT CONG.
URANIUM REC 0VERY WASTE CHAR. AND OPERATION MGMT.
$ 1.0 SITING, PATHWAYS, AND IMPACTS 1.2 DECOMMISSIONING
.6 STANDARDS AND TECH. ASST.
.4
$ 3.2 l
9 C -2. i
1/6/82 4
l FY 1983 CONGRESSIONAL i
NUCLEAR REGULATORY RESEARCH (DOLLARS IN MILLIONS)
FY 83 SITING & ENVIRONMENT CONG.
l EARTH SCIENCES l
GEOLOGY AND SEISM 0 LOGY
$ 3.1 METEOROLOGY AND HYDROLOGY 2.4 SITING 1.5 HEALTH EFFECTS 2.0 ENVIRONMENTAL IMPACTS 0
$ 9.0 O.7h
5
[. msg'g UNITED STATES NUCLEAR REGULATORY COMMISSION e
0 tvASHIfeGTON, D. C. 30665
(
Decenber 30, 1981 MEMORANDUM FOR: Robert B. Minogue, Director Of fice of Nuclear Regulatory Research FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
REQUEST FOR THE CONCEPTUAL DESIGN OF A FACILITY FOR THE STUDY OF L&W AND CE INTEGRAL SYSTEM CHARACTERISTICS
REFERENCE:
(1)
G. Willeutt, "Small Break and Transient Audit Program Technical Note "LA-SBTA-TN-81-3, October 1981 (2) " Generic Evaluation of Small Break'LOCA Behavior in B&W Designed 177-FA Operating Plants," NUREG-0565, January 1980 (3) " Clarification of TMI Action Plan Requirements,"
NUREG-0737, November 1980 (4) Memo, Levine to Denton, " Endorsement of the
{g Redirected Semiscale Program and Facility Upgrade,"
k October 19, 1979 (5) Memo, Ross to Shao, "NRR Proposed Prioritization of Semiscale Facility Mc I 2A Testing," January 29, 1981 (6) " Transient Response of Babcock and Wilcox - Designed Reactors," NUREG-0667, May 1980 Purpose The purpose of this user need letter is to identify NRR experimental data needed to confirm the thennal-hydraulic behavior under transient and acci-dent conditions of plants with a 2 x 4 loop configuration (B&W and CE).
Background and Statement of Problem The prototype PWR used for the design and scaling of both the Semiscale and LOFT facilities has been a Westinghouse 4-loop plant. For the euluation of large break LOCAs, design differences among the three PWR NSSS vendors were not considered to have a significant offect on the overall thermal-hydraulic phenomena and behavior. Recent comparisons of small break and plant transient behavior among the three NSSS vendors have identified several aspects of the thermal-hydraulic behavior unique to each vendor's design:
(1) B8W plants are calculated to behave in a cyclic manner during some j(
portion of some transients with a two-phase mixture in the primary ATTMcNu&ui- )
000002 e'^ymniir
{
Robert B. Minogue December 30, 1981 (2) B&W plants are equipped with vent valves which are calculated to be of significant benefit to primary system recovery from a two-phase condition.
(3) The B&W NSSS exhibits significant sensitivity to secondary side upsets.
(4) CE plant calculations have led CE to recommend that one pump be left running in each of the loops in the event of a small break.
There are no thermal-hydraulic integral system data for the plants with a 2 x 4 loop configuration that are comparable to those that have been obtained for the Westinghouse NSSS.
In addition to these examples, there exists a need to complete the data base required for model verification for B&W and CE plants.
These items are discussed in more detail in the following sections of this user request.
B&W Design Characteristics Natural circulation is important to heat removal during small breaks, and
(
operator instructions are designed to initiate and/or continue this process.
Recent B&W analyses have shown the system pressure to behave in a cyclic manner that could be confusing to the operator during certain small break conditions.
Specifically, during certain small breaks in plants of the lowered loop design, B&W calculations indicate that once steam being gener-ated in the core fills the upper part of the reactor vessel, it will then collect in the top of the hot-leg U-bend. This is predicted to interrupt natural circulation once the bend is filled.
Natural circulation is cal-culated by B&W to be restored once the liquid level in the steam generator primary drops to expose a condensing surface.
For the raised-loop plant design, B&W analyses have predicted this interruption to be cyclic.
(Three significant interruptions and restorations of natural circulation prior to draining below the condensing surface were predicted by B&W.)
prior to TMI-2, the repressurization phenomenon had not been identified by B&W and was only predicted to occur after a close moci.1 examination and modeling change was made.
We do not know if the unique oscillatory phenomenon is real or an artificiality of the analyses. Recent analyses of small breaks in B&W plants by LANL (Ref.1) do not show the repressurization phenomenon calculated by B&W. We believe the predicted phenomenon could produce false symptoms of other events, such as loss of heat sink, and lead to incorrect operator actions which could result in more severe consequences than now predicted for the SBLOCA.
The phenomenon described is considered unique to B&W-designed NSSSs because
{
of the hot leg and Once-Through-Steam-Generator (OTSG) configuration.
Neither LOFT nor Semiscale as presently configured are applicable for verifying this phenomenon experimentally.
)- 2.-
's
?
Robert B. Minogue December 30, 1981 1
The need for experimental verification of heat removal capability during small breaks in B&W plants is discussed in Reference (2). The need for i
experimental model verification is part of TMI Action Plan, Item II.K.3.30 (Reference 3).
A second phenomenon which needs to be considered is the apparent benefit of vent valves in B&W piants. Vent valves are predicted to aid in main-taining core cooling during many small breaks.
In PWRs designed by Westinghouse and CE, small breaks in which the break flow exceeds the HPI flow require the loop seal at the pump inlet to " clear" (allow steam to pass to the break) prior to inventory recovery. Due to the hydraulic coupling of the liquid in the core to that of the steam generator prior to this clearing process, the liquid level in the vessel must drop to approxi-mately the elevation of the bottom of the loop seal. This results in a predicted uncovering of a significant portion of the core for a relatively short time.
In B&W plants of the lowered-loop design, the bottom of the cold leg loop seal extends to an elevation near the bottom of the vessel.
Without the aid of the vessel vent valves to pass steam into the upper downtomer annulus and hydraulically disconnect the vessel liquid from the steam generator liquid, a complete uncovering of the core for B&W plants would probably be predicted prior to loop seal clearing.
With the aid of the vent valves, however, this clearing process is not predicted to occur, and in fact, the vent valves operation is predicted to keep the B&W reactor core covered for a wider spectrum of small break sizes than either Westinghouse or CE predict for their plants.
A final aspect of our data needs regarding the capability of B&W plants to accommodate small breaks is (a) the potential benefits of high point vents restoring natural circulation and (b) the advantages of incorporating sprays at the top of the hot legs (and vessel head) to condense trapped steam.
We believe that the benefits of these modifications need to be explored and confirmed experimentally.
l Another aspect of the B&W-designed NSSS is its relatively rapid response to secondary side upsets, particularly in the feedwater system. Methods are being studied which could alleviate this behavior. A more complete assessment of this concern is documented in reference (4).
Presently, any design changes or modifications proposed to reduce this sensitivity can be justified by analysis only, without the benefit of experimental verification (unless it can be demonstrated with full scale startup testing.)
An experimental facility would be of significant benefit in this area of concern.
At present, we have no confimatory integral systems data with which to
[
verify the acceptability of the predicted behavior of transients and acci-dents including small break LOCAs in B&W designed reactors. Al so, the long-term hydraulic stability of the plant following a SBLOCA has nevar (r
been analytically or experimentally confimed.
so
(
Robert B. Minogue 4
December 30, 1981 CE Design Due to the similarity in steam generator designs (U-tube type), plants with CE-designed NSSSs are expected to behave similarily to Westinghouse-designed NSSSs for small break LOCAs. However, there are differences.
One aspect of the CE 2 x 4 design as it affects small break behavior is pump operation during small breaks. Specifically, CE has recommended that one pump in each of the two loops be tripped and one pump in each of the two loops remain running. The basis for this recommendation is that it provides forced convection flow during small break LOCAs, yet was not calculated to result in unacceptable consequences if the punps were tripped at any time during the event.. Although we intend for LOFT test L3-6 to provide data for code verification, any decision to support the CE partial pump operation recommendation would be based on analysis of the system for a mode of operation not directly comparable to the L3-6 test. Moreover, a facility in the CE configuration would complete the available data base for all three PWR NSSS vendors' designs.
Information Needs
(
The concerns identified previously in the " Background" section of this memo indicate the inadequacy of the present Semiscale configuration to properly characterize the key themal-hydraulic phenomena predicted to occur during small breaks and selected transients in B&W NSSSs. The importance of these phenomena in detemining the course of small-break LOCAs leads us to seek additional integral systems data from a test facility more representative of the 2 x 4 NSSS design.
The need for such a facility was previously documented in References (5 & 6).
We believe that at this stage of planning for such a facility, it is appro-priate to identify to you the basic criteria which such a facility should meet. These criteria are identified below:
(1) The system should scale, to the extent practicable, the power to volume ratio of B&W-designed NSSSs, including the steam generator secondary.
(2) The system should preserve prototypic elevations.
(3) Major design features of B&W plants should be represented, including vent valves, hot-leg U-bend, etc.
(4) Sufficient flexibility should be incorporated into the design such that:
(a) the steam generators could be configured to represent either a lowered-loop or raised-loop B&W design; and 2 - 9.
Robert B. Minogue December 30, 1981 (b) the once through steam generator design could be replaced with a U-tube steam generator design to represent the CE 2 x 4 loop configuration.
(5) The system should exhibit to the maximum extent practicable the key thermal-hydraulic phenonena predicted to occur in B&W plants during small breaks as previously discussed.
Use in Licensing Section 1.C.1 of the TMI Action Plan (NUREG-0660) is to provide " Guidance for the Evaluation and Development of Procedures for Transients and Accidents,"Section II.K.3.30 addresses " Revised Small-Break Loss-of-Coolant-Accident Methods" to show compliance with 10 CFR Part 50, Appendix K.
Programs consisting of reanalysis, comparisons with experimental data, and sensitivity studies are being conducted by the industry in response to these items.
Several industry studies will be reported by June 1982. There will remain substantial uncertainty in any B&W analysis of SBLOCA pending experimental verification. While we realize such a facility would not be available to produce useful data on a time scale necessary to complete efforts on these tasks, we would require that licensees confirm their models against
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the facility data once the data became available. A completed conceptual study including cost data will be essential on or before the end of June 1982.
Cost and Schedule Preliminary assessments by RES indicate the cost of the base facility in the B&W configuration to be approximately $18-20M, including contingency and inflation.
This is based on constructing a facility similar to, but separate from, the present Semiscale MOD 2A facility.
We believe your office should identify various options available for satisfying the identi-fied data needs and prepare more detailed cost / schedule information to serve as a basis for final decisions on such a program. We also do not believe that NRC should be solely responsible for the funding of the facility to obtain the needed data. We request that RES investigate means in which the regulated industry shares an appropriate position of the financial costs. We could best use this information in July 1982, about the time we expect to make decisions regarding the long-term adequacy of analytical models for improved accident procedure guidelines.
Expected Benefits l
In addition to integral system data on small break and transient behavior expected in B&W plants, a program of the sort propased here could serve to assess the appropriateness of emergency operator guidelines and pro-r I
cedures. Additional uses of this facility would be to provide NRC with k
data applicable to specific problems as they arise and to aid management b5 l
- g. e e Robert B. Minogue Decenber 30, 1981 decisions during and following plant emergency situations (similar to the role of'Semiscale M003 during the TMI-2 accident).
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Harold R. Denton, Director i
Office of Nuclear Reactor Regulation l
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