ML20042E096
| ML20042E096 | |
| Person / Time | |
|---|---|
| Issue date: | 10/10/1989 |
| From: | Beckjord E NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Todreas N MASSACHUSETTS INSTITUTE OF TECHNOLOGY, CAMBRIDGE |
| Shared Package | |
| ML20042D143 | List: |
| References | |
| NACNSRRC, SECY-89-225, NUDOCS 9004200098 | |
| Download: ML20042E096 (8) | |
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October 10. 1989 Professor Neil E. Todress, Chairman Nuclear Safety Research Review Comittee Massachusetts Institute of Technology Nuclear Engineering Department Building 24-219 77 Massachusetts Avenue Cambridge, MA 02139
Dear Professor Todreas:
I am replying to the third report of-the Nuclear Safety Research Review-Comittee (N3RRC) in your letter dated August 11, 1989. As noted, the NSRRC haloped its views in the course of subcomittee meetings in the fall of 1988 on sr vere accidents, aging and reactor pressure vessels, human factors, and.
wasce disposal; and at the full Comittee meeting on May 23 and 24,1989.
In this letter I respond to the 12 topics listed in your letter, in the same order as given.
1.
Interaction of RES and NRR I agree with NSRRC's assessment of the utility of user need letters from the Program Offices, and I intend to continue to invite the Office Directors to meet with the full Comittee at its meetings.
2.
Maintaining Staff and Contractor Expertise Through Research The Comittee is concerned about the eroding U.S. capability in research areas important to regulatory needs.
I recognize tY: r.ead to maintain expertise in those areas important to regulatory requirements.
Thermal-hydraulics is.a prime example, because of the need to. maintain core cooling and remove heat from reactor 1
primary and secondary systems. The research task in this area is to maintain the j
capability to address the regulatory iswes.that arisa from operating events, and 4
at the same time to provide technical challenge in research projects that will attract the interest of the most competent people.
I enclose a copy of the Comission paper (SECY-89-219, 7/24/89)(Enclosure A)
' for the Comittee's information. The concept of-the paper was presented to NSRRC at the May meeting, and the paper itself has been reviewed by the ACRS and the Comission. The paper proposes to maintain essential capability in T/H. The. definition of future research in T/H is given on pages 6-12.
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Professor Neil E. Todreas 2
October 10, 1989 3.
Introduction of Advanced Technology Into U.S. Plants The Comittee stated that utility control rooms were yer.rs behind the chemical I think the main processing industry in terms of using electronic technology (.NPP)areagingand factor in this regard is the fact the nuclear power plants there are not any new commercial plants in the pipeline.
It is true that regulation can inhibit introduction of new. control technology in operating i
plants, because of possible impact on technical specifications or Itcensing i -
L,ases. At the same time digital control systems are replacing the original i
analogue systems in several plants with the likelihood that more will follow.
s Also, nuclear utilities are introducing advanced technologies, including expert systems, into their routine and planned operations where it is not adverse to i
safe operations. For instance, EPRI is developing over twenty different expert systems for both nuclear and fossil applications. The NRC is working jointly with EPRI in order to verify and validate expert systems.
Furthermore.
General Electric is incorporating advanced electronics technology in the Advanced Boiling Water Reat. tor (ABWR), now under review for certification.
i l
The intent of the RES presentation at the NSRRC meeting was to sunnarize the l
motivation for NRC research into improved regulatory review criteria for advanced technologies in nuclear facilities. The initial phase of the research is to determine the types of advanced technologies and the anticipated schedules as l
utilities introduce advanced technologies into nuclear facilities. The Human g
Factors Branch (HFB) is completing a survey oi, this topic, I suggest that HFB present its curtsnt findings to the NSRRC during the November Human Factors and Reliability Subcommittee meeting. The regulatory research aims to provide review criteria which remain consistent with the new technologies in a manner that assures public health and safety. This work will help to accelerate regulatory acceptance of advanced technologies.
In addition, both new and old technologies are subjected to rigorous reviews before nuclest power plant control rooms may be changed. These reviews provide assurances that any change to the control rooms, from either an old or a new technology, does not introduce new or more likely problems. Each change is rigorously reviewed in the context of plant performance over the full spectrumofnormalandabnornalgiantstates.
It might appear to the NSRRC that these rigorous reviews are severe restrictions regarding introducing l
advanced technoir.gy in our nuclear plant control rooms." However, regulatory l
reviews are intended to assure the public health and safety rather than to hinder the advancement of technology.
The NRC is now spending high-priority review time in ongoing reviews of pro)osed changes to control rooms motivated by both advances and obsolescence in tecinology.
The current reviews are performed in NRR by the Instrumentation and Control Systems Branch and portions of other branches. The largest program activity in the HFB/RES budget is directed toward research on the human-systems interface, the largest portion of which is on advanced technologies. To rny knowledge no " higher priority issue" is interfering with the research into the issue of advanced technologies associated with the human-systems interface at nuclear facilities.
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Professor Neil E. Todreas 3
October 10, 1989 The NRC research itself is recognized internationally as advanced. The NRC participates with other technologically advanced countries in the Halden Project. NUREG/CR-5348, " Man-Machine Interface Issues in Nuclear Power Plants," is attached which reports the man-machine interface criteria and guidelines, as well as verification and validation experiments resulting from the international workshop held by HFB/RES during January 1989.
l When the NRC m established the Congress intended to make a clear distinction i
between regulatory responsibl11 ties and technological development or promotion.
By law the NRC may not develop or promote plant technology. The responsibility of the NRC was and continues to be the assurance of public health and safety regardless of the newness of electronic technology.
i I am ready to discuss the role of the NRC with the NSRRC Subcommittee on Human Factors and Reliability at greater length during the November 1989 meeting.
4.
Attracting Outside Researchers to Bolster Management The Comittee emphasized the need for a program to enhance the management capabilities of the current people in RES.
I agree with this point and take your suggestion.
The management training program for the Office of Nuclear Regulatory Research consists of a supervisory developmnt program that includes courses in supervising human resources, performance management, performance appraisal, managing change, and workshops in s>ecial topic areas. Emphasis is also me d towards expanding scientific and tecinical perspectives through participation in the Brookings educational program as well as the Federal Executive Institute and Execative Seminar Center. More work needs to be done to provide additional outsido exposure i
t to RES management from industry, national laboratories, and universities.
I will look into this carefully to see how to pursue this idea vigorously in the coming year.
5.
Annual Program Reviews with the Principal Performers of Research I think you make an important point, because performance feedback has the
)otential for contractor performance improvement, increased productivity, and 1ence better use of research funding. RES holds a number of specific program reviews during the year; the Division of Engineering and Division of Systems Research reviews in the spring of 1989 were extensive. The marketplace also provides such a mechanism indirectly: inadequate performance by a contractor
' generally means reduction or loss of future research assignments, but res)onse by this route is neither very fast nor efficient.
I will give more thoug1t to effective use of performance feedback in ways that avoid excessive paperwork.
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li, Professor Neil E. Todreas 4
October 10, 1989 h46.
Peer Review of Research Products 1 agree with.the Comittee's point that RES products should be peer reviewed.
To that end, RES encourages all contractors to publish their technical findings in high caliber, refereed journals. RES has in the past, and will continue in the fu'ure, to stress the need for exposing our technical work to careful peer reviews. The peer review of NUREG-1150, chaired by Dr. Herbert Kouts, began in June of this year, and is expected to be finished in March of-1990.
7.-
RES Grant / Contract Procurement
'I strongly endorse the Comittee's recomendations that more be done to implement plans to attract industrial and university participation and to stimulate new approaches to solving the difficult safety problems. The Divisions within the RES organintion are continuing to make effective use of the Broad Agency Announcement (BAA) as one mechanism to get institutions outside of the National Laboratories involved in nuclear safety research. As suggested by the Comittee, RES presented its overall grant program in Atlantu, Georgia on June 5,'1989 to the Nuclear Engineeri.79 Department Heads Organization '(NEDHO).
In addition to Nuclear Engineering De netments, BAAS will be mailed out to other engineering departments (e.g., mectanical) as a means to increase participation by the university comunity. RES has developed an Historical Black Colleges and o
Universities (HBCU) Research Initiative to improve minority access to the RES nuclear safety research activities.
8.
Establishment of a Research Planning Process I appreciate your cunclusion that RES is using the prioritization procedure appropriately at this time.
In August the research budget took a $20M cut (19%), as a result of the Appropriation Comittee actions on the NRC budget as a whole. The consequent deferrals and cancellations are a serious blow and j
impairment to the ability of RES to perform according-to-the 5 Year Plan. These are detailed in Enclosure B.
A furtier cut, in the' event of a triggering of i
Gram-Rudman during this month could be devastating to the~ research programs.
I propose to discuss prioritization again with the Comittee, by which time we will know the full impact of budget actions. Frankly, the budget outlook is bleak.
9.
Severe Accident / Accident Management
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The Comittee stated that the revised SARP plan lacks specificity with respect l
to deliverables and milestones. Although this is true of the schedules and milestones, the SARP goa?s and objectives are'quite specific. Detailed schedules, success criteria, and milestones are the main substance of the statements of work that have been prepared for FY 1990 in order to implement I the revised SARP plan.
I suggest a review at the earliest opportunity. The i RES staff will be pleased to meet with the Comittee to discuss how the goals of the revised SARP are being met through these work statements.
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L' a,o, Professor Neil E. Todreas 5
October 10, 1989 Human Factors 10.
u I am pleased that the letters of June 8,1989, and June 23, 1989, are expected to be useful to the full NSRRC concerning the selection of organizations to perform research on the topic of Organization and Management. The Comittee also requested that certain additional issues be addressed at the next Subcomittee meeting. The best means of pursuing your questions would be for the HFB/RES to make a presentation to the Human Factors and Reliability Subcomittee at its November meeting on this subject. The HF8/RES presentation would cover (1) the processes used to select the research organizations for the topic of Organization ind Managessent, (2) the status of'our collaborative endeavors with other agencies doing Organization and Management research, and (3) a sumary of the progress to date. A highlight of the progress to date has been the review of the entire human factors research program by the ACRS. The ACRS concluded in a memorandum to the Chairman (May' 9,1989) that'"The utilization of a number of diverse institutions and organizations as human factors research providers is comendable. This is-particularly noteworthy in the organization and management and in the reliability assessment program elements of the research plan. The use of diverse research providers has already generated new input to, as well as interest in, the human factors research program."
j The second topic listed under the " Human Factors" portion of the letter on page 5 deals with *Delphi and other methods for elicitation and refinement of subjective probabilities from human experts." This was used in the NUREC-1150 (Severe Accidents Risks) research effort rather than human factors research. Subsequent to the NSRRC meeting, information on this subject was provided to Dr. T. Sheridan at his request.
I suggest that the effort would best be discussed in'the context of a further discussion of NUREG-1150 with the Committee.
- 11. Waste Disposal, Regarding user need input from the States, I agree that the Statesneeds should be considered in developing the LLW research program. There is now a process in place to do so. However, the Comittee may not be aware that the NRC and EPA have been charged by Congress to develop' standards and rules for LLW burial facilities in support of the States and compacts of States that will be developing those facilities. The States' needs are addressed through a program sponsored by) DOE under the Low-Level Rauicactive Waste Policy Amendment Act (LLRWPAA. DOE prepares and publishes an annual report which includes needs identified by tie States. This report is a key document used by NMSS in preparing the user need request forwarded to RES each year.
In addition, RES maintains direct contact with the States through attendance at various LLW coordinating groups and periodic meetings held by HRC (GPA) for State regulators. Since the sources for State needs for DOE's report are understood to be almost entirely from LLW site developers and operators, State regulatory needs may not be adequately represented. To correct that possible deficiency,1 intend to invite comments directly from the State regulatory bodies on.the first edition of the LLW Research Plan (NUREG-1380), naw scheduled to be i
published in October 1989.
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Professor Neil E. Todreas 6
October 10, 1989 I agree with the Comittee's view that realistic planning and priorities should f
be establiched for LLW and HLW research, given that the limited resources budgeted 1
l and planned will only support a portion of the identified user needs at any given time.
In the LLW area RES has made considerable progress through the development of a draft Research Plan that realistically spells out and prioritizes the work to be done in considerable detail. The latest published draft of the HLW Research Plan is less specific and RES and NMSS are working to prepare a more detailed plan that will identify and prioritize the work sufficiently to develop a HLW research program reflecting realistic budget constraints.
In the Spring 1990 update of the HLW research plan, RES and NMSS will attempt further to prioritize needs in view of identified available resources.
The policy issue raised by the NSRRC regarding the criteria for placing work at the Center for Nuclear Waste Regulatory Analyses (CNWRA) at the Southwest Research Institute was fully addressed by the NRC's management in a recent Comission paper (SECY-89-225,7/28/89)(EnclosureC). The Comission's conclusion was that the existing criteria for placing work at the CNWRA should not be changed.
Essentially, these criteria require the placement or HLW technical assistance and research at the CNWRA, absent compelling reasons not to ) lace it there. The basis for this approach is to concentrate the NRC work so t1at the Center can become an effective center of expertise, dedicated on-a continuing basis to the NRC program, without conflict of interest. The Comission's conclusion also recognized the likely continuation of limitation on NRC resources for HLW-related work, and therefore the NRC made the CNWRA contract show a comitted level of funding to the CNWRA.
As stated in the Comission paper, RES and HMSS are comitted to the principle of establishing the CNWRA as the long-term body of expertise in HLW matters.
Further the staff is aware of short-term concerns that will require extra management attention during the transition from existing contractors to the CNWRA. To this end senior RES, NMSS and CNWRA managers met in August in order to review research needs in FY 1990. The CNWRA has responded with their plan for personnel recruitment. Through continued management interactions, RES believes the CNWRA can provide the engineering and scientific excellence that the Comittee recomends.
i In the same Commission paper, the staff noted that there were compelling reasons for continuing research with at least one other contractor because of demonstrated capability. This was the theoretical modeling and field studies in hydrogeology, and the studies on borehole and shaft sealing on rock, all at the University of Arizona.
In the materials degradation area HMSS and RES have just recently agreed not to terminate work at Cortest Columbus and NIST in FY90 (as originally plannned in the Comission paper) when it became evident that the CNWRA was not expected to i
reach desired staffing levels in FY90.
In FY91 and beyond, there may be selected research areas that will require use of demonstrated capability outside the CNWRA-(within conflict of interest limitations) such as hot cell studies on spent fuel, i
source term, and transuranic chemistry research.
Although I would prefer more flexible criteria, I think we will be able to assign work selectively in special areas to other qualified contractors in the spirit of the NSRRC's recomendations, provided the HLW Five-Year Plan resource projections are realized.
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Professor Neil E. Todreas 7
October 10, 1989
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i The CNWRA has been directed by RES to develop expertise in a number of key, relevant, disciplines, in accordance with criteria, provided by RES, aimed at assuring " center of excellence" level capability as noted in the Research Philosophy.
12.
Committee Review of Full Scope of RES Activities'- (Page 6):
The Comittee identified the issue of the credibility of RES research results, using as an example the apparent discrepancy between the results of NUREG-1150 and the Mark I improvement proposal. During the May briefing, RES presented 0
the results of the NUREG-1150 risk analyses to the Committee, including results for the Peach Bottom (Mark I) plant indicating that the plant's risk.was beneath the NRC's safety goals. This led to the question of why the Mark I improvements were being pursued.
I would like to make several points on this issue. First, the showing in NUREG-1150 that the Peach Bottom plant meets the NRC's safety goals clearly does not also provide a demonstration that all 24 Mark I plants meet these r
goals, because the results of NUREG-1150 are primarily plant specific to Peach i
Bottom. Rather, the extensive risk analyses available for Peach Bottom for the past 15 years tend to support the view that this plant is at the low end of risk with respect to plant design, systems, and equipment.
PRA studies of other Mark I plants, showing higher core damage frequencies, also support this statement. Peach Bottom has been studied intensely, and the dominant sequences have already been fixed by hardware and procedural changes, and this is an important factor in its low risk. Second, the core damage results reported in NUREG-1150 for Peach Bottom were based on the presumption that venting, one of the key staff recommendations, was already being used to essentially eliminate the long term loss heat sink (TW) sequence.
In fact, the finding from NUREG-1150 that venting could essentially eliminate the (TW) sequence, and thus significantly reduce total core damage frequency, served as a major basis for staff recommendations However,)NUREG-1150 did not include several factors for venting of Mark I plants.
a a concern for vent rupture, in the event relevant to venting Mark Is, namely:
that a low pressure ductwork path were used during a high pressure scenario, coupled with b) a staff survey which indicated that vent capability, procedures, and operator training relative to venting may be inadequate and varied greatly amor.g Mark I plants.
Third, the Mark I improvement program was initiated based on concerns about the conditional probability of early containment failure if a core damage accident were to occur. Neither the first nor second draft of NUREG-1150 provide a basis to discount this concern. The staff recommendations taken together form a balanced approach, addressing both core damage prevention and containment failure probability given a core damage.
For these reasons, I do not believe that the NUREG-1150 results are inconsistent with the direction of the Mark I improvements program.
Rather, the Peach Bottom results tend to support the staff recommendations relative to Mark Is.
In your letter the Mark I analysis and evaluation is a specific example of a more general question of what are the criteria that research results must meet.
I think you.have answered the question in general: research must be useful to the staff and Commission in making decisions. The Research Philosophy elaborates on this idea
.a p.
1 Professor Neil E. Todreas 8
October 10, 1989 l
and provides additional criteria. However, I think a complete answer calls for RES to show how research findings have been and are being applied in specific j
I can recall a number of research presentations to the Comittee from cases.
which you could infer that research is used in regulatory decisions.
I would i
like to address this question directly in our next meeting and show how results l @ $ are being used in specific cases.
Every program presentation to the NSRRC should include a showing of how expected results of research will be used in the regulatory I
l process. Beyond this, I accept your suggested list of areas for discussion in the l
spring.
l In closing, I wish to say that I appreciate the dedication and effort.of the l
Comittee in carrying out its charter. You have pointed out key research issues, and helped RES management to focus on them. We have a better and mo e effective research program for your efforts.
Sincerely,
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Eric S. Beckjord, irector Office of Nuclear Regulatory Research
Enclosures:
A.
SECY-89-219 B.
Deferrals / cancellations C.
SECY-89-225 D.
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j ENCLOSURE A
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POLICY ISSUE 1
(Commission Meeting) shsly 24.1989 SECY-89-219 ESI:
The Commissioners Egga:
Victor Stello, Jr.
Executive Director for Operations subiect:
STATUS AND PLANS FOR THERMAL HYDRAULIC RESEARCH CONDUCTED BY THE OFFICE OF NUCLEAR REGULATORY RESEARCH
Purpose:
The purposes of this paper are to:
1.
Brief the Commission on the current status of this research.
3 2.
Inform the-Commission of future goals and directions of this research.
Backarcund:
TherR11 hydraulic research conducted-since i
the laception of the NRC in 1975 and its predecessor agency, the AEC, provided the basis for the recent' revision of 10 CFR 50.46 and Appendix K, issued in October 1988.
This.
rule' revision reflects the results of the large amount of work performed during the late 1970's and early 1980's, summarized in t
Reference 1.
The revised rule permits i
realistic analysis of loss-of-coolant-accidents (LOCA), while retaining the option to use the former prescriptive, artificially conservative approach.
Rescinding.the requirement to use artificial,foyerly; conservative analysis methods.will allow licensees more operational flexibility,,such as extending useful life by lowering neutron flux exposure of-reactor vessels and concomitant rate of vessel steel i
embrittlement, and-increasing-plant capacities.
Eventual revision of the LOCA Rule was a goal set forth by the Atomic I
d Contacts:
i D. Bessette, RES, 49-23572 L
L. Shotkin, RES 49-23530 d
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l Energy commission when it adopted the rule in L
1973 (Reference 2), and this has'now been achieved.
In addition, research has studied i
a spectrum of small break locAs and transients, and computer codes have: boon developed and assessed for such events.
As a result of these research accomplishments, our.
j understanding of and confidence in predicting-LWR ther1 sal' hydraulic performance during transient and accident. conditions has been l
greatly improved.
commensurate with this improved level of understanding, our research' i
efforts in this area have decreased significantly over the last several years and i
are scheduled to decrease further.
At this r
juncture, therefore, it is appropriate to inform the Commission of the current research-status and the future goals and directions for thermal hydraulic research'(Reference 3).
Appendix 1 reviews the history of this I
research.
Discussion:
Thermal hydraulic research is. intended to I
support the Staff in the following areast o
Understanding: reactor transient events and their broad implications for operational safety; i
o Detection of previously unrecognized i
issues important to safety;-
o Investigating and resolving apacific issuer >,
e.g.,
the effectiveness of_ decay heat removal via feed and bleed; o
Evaluating the-effect of design'and operatiens-related changes, including operating procedures and changes to technical specifications and setpoints; Analyzing the early_phasea of risk o
dominant accident sequences and other postulated severe accident scenarios; o
Evaluating: strategies and procedures for accident management; o
Evaluating the reactor and plant systems designs of new standardized LWRs; and I
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Confirming safety margins in licensee analyses by performing audit analyses.
Thermal-hydraulic research is conducted under
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the plant performance program element, and includes three activities.
The status and r
schedule of each activity is summarized briefly in the following.
1.
Babcock and Wilcox Testing The basic objective of this work, endorsed by
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the ACRS, is to provide a data base for B&W designs that is comparable to that which
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exists for the other NSSS vendors. Test facilities such as LOFT, SEMISCA12 and FIST provided integral facility data for Wustinghouse, Combustion Engineering and General Electric designs.
RES developed and is carrying out a plan to provide comparable data for B&W designs (Reference 4).
Following the TMI-2 accident, the best-estimate codas TRAC and RELAP were used to analyze Babcock and Wilcox (B&W) designs for small break LOCAs.
Similar analysis was performed by B&W.
Significant discrepancies in calculated plant response were noted which could not be resolved due to insufficient experimental data.
The lack of data to validate the calculated results led to the establishment of the Integral System Test (IST) program in 1983.
This program included the construction of the Multi-loop Integral System Test ' MIST) facility and the performance of a small break LOCA test series under a cooperative, cost-sharing arrangement among the B&W Owners Group, B&W, EPRI, and NRC.
The successful completion of this program has provided a small break LOCA data base, and the codes are currently being validated against these data.
Several transients that occurred in B&W plants (e.g., 1985 Davis Besse and Rancho Seco events) since the establishment of the IST program indicated that the unique thermal hydraulic behavior of B&W plants resulting from steam generator design was not limited to small break LOCAs, but included non-LOCA
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i transients as well.
This resulted in a follow-on test series in MIST to investigate transient behavior, as well as initiating discussion on the need for erperiments on once-through steam generator (OTSC}
performance.
Recently, agreement was obtained with the B&M.Cwners Group to conduct a cooperative' experimental program on OTSG l
- performance.
This experimental program, expected to be completed in FY 1992, will make the. experimental thermal hydraulic data
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- base for B&W plants comparable.to that which-exists ' for the 3 other vendors.
2.
Experiments and Analysis i
I It became apparent around 1975 that the cost for NRC to unilaterally obtain large scale experimental data necessary to resolve-the LOCA/ECCS issue was prohibitive, so RES began discussions with Japan and the Federal-1 Republic of Germany on the conduct of a' joint program.
Several years of planning and i
negotiation led to the formation of the 2D/3D program, which began in 1980 and will be completed in'1990.
Three large facilities were constructed, two in Japan (Cvlindrical Core Test Facility, Slab Core-Test facility) and one in Faderal Republic of Germany.(Upper Planum Test. Facility).
The RES contribution included advanced' instrumentation for these
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facilities, and the development of advancer; i
analytical tools needed to :nodel the - compl*w phenomena being studied.
Experimentation.
will be completed:in the current fiscal year, analysis in FY 1990, and final reporting in FY 1991.
Following the THI-2 accident, Japan decided.
. to build a large scale (1:50) integral testi facility to investigate small break LOCAs in PWRs.
A number of ancillary facilities also-maka.up the program imown as ROSA-IV.
RES interacted with Japan from the start of the program and provided advanced-instrumentation
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to the facility.
.A bilateralEagreement was signed in 1984, experimentation began.in 1985,'and the cooperation currently extends to 1992.
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5-Several past. experimental programs such as 14FT, SEMISCALE,-.and FIST successfully completed their mission to provide thermal hydraulic data for Westinghouse, Combustion Engineering.and General Electric designs.
These facilities-were decommissioned since the cost to maintain them in standby. mode could not be justified in the absence of an immediate demonstrated need.
The closure of these' facilities left the United States with no' domestic experimental facilities to provide information on small breaks or transients in non-B&W designs..' ROSA-IV helps fulfill this role, and RES periodically makes requests to the Japanese to perform experiments relating to different issues.
3.
Modeling This activity includes code development, code assessment, and code applications.
The International Code Assessment-Program (ICAP), a' cooperative effort among 14 countries, is the principal source of independent assessment of the RELAP and TRAC codes.
Additional input is received from the experimental programs described above.
The assessment results and discussions held under the ICAP program provided a common understanding of the performance of the RELAP and TRAC' codes for.nodeln.g PWRs.. RES developed a code improvement plan to guide the code development during the FY. 1988-1989 time period.
This will culminate in the last planned versions of these codes, which will be released by the end of the current year.
Following their release, the codes will be ansessed under the ICAP program throdgh 1991.
No new versions of these codes are planned unless assessment results show significant deficiencies in the ability to predict nuclear plar.t performance.
In this code development activity starting with the effort to revise the ECCS rule,.RES' has emphasized code documentation and software quality assurance.
Documents were issued to describe all models and correla-tions contained in the-codes, their source, I
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data base, and range of applicability.
User guidelines for nodalization_and applying the code wore improved and documented.
An independent review was conducted of software quality assurance procedures in use at INEL and:IANL to assure.confornity with ANSI standards.and prevailing industry practices.
i ruture Renamreh plana l
The governing objective for' future thermal 1
hydraulic-research was developed as a joint, j
staff position.
RES and the principal user office, NRR, interact on a routine basis, both formally and informally (Reference 5).-
4 AEOD also participates in'this process.
The formal mechanisms ares (1) the Reactor Systems Safety Senior Research Program Steering Group, headed by.the Director, Division of Systems Research, RES and-l constituted at the Division Director level; and (2) a Regulatory Research Review Group-constituted at the Branch Chief level.
During 1988, these groups reviewed future thermal. hydraulic needs.
This review-concluded that L
o The NRC staff will continue to need independent expertise and analytical' capability for addressing transients in PWRs and BWRs; and o
The principal RES analysis codes (TRAC-PWR, TRAC-BWR, REIAP5, and RAMONA) should be maintained for active use, The recent revision to 10 CFR 50.46 and-Appendix K does not terminate the need to 4
maintain expertise in the thermal hydraulic area.
Nuclear safety-encompasses a number of-engineering disciplines; one of the more important being the. field of thermo-fluid
[
mechanics.
Issues requiring thermal' hydraulic analysis arise periodically and the-Staff is called upon to address them.
Such-t issuasitypically require coda calculations.-
The codes are large, complex and require experienced users.
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7 Code development has proceeded in an interrelated fashion with experimental programs.
Most of this work is completer however, certain projects remain.
The 2D/$D program will end.in 1990.. The experimental prgrams associated with B&W designs will-finish in 1992.
The Japanese ROSA-IV progre.a of integral and separate effects experiments on small breaks and transients will continue-until-1992.
The ICAP program continues until 1991.
RES will complete these efforts, to the extent. appropriate, and incorporate important information into the codes.
Several policy issues are present with respect to the near term and longer term future.
These are described in the subse-quant bullets, followed by the Staff's plans.
o Thermal hydraulic research has been oriented towards resolution of specific issues such as ECCS performance and pressurized thermal shock. -These and other issues.have been resolved, and the current program is aimed at resolution of other remaining issues.
Absent specific new issues,:what level of thermal hydraulic research should be pursued and'to what end?-
Future projects must be technically chal-longing to attract and retain researchers and must be consistent with the NRC's mission.
We plan tot 1.
Complete the current projects within the three activities described previously, which should occur by:1992.
2.
Continue code maintenance for the RES.
PWR and BWR systems codes, and keep the software in pace with advances in i
computers.
3.
Perform research as appropriate, directed at improving the accuracy of the codes in areas determined.to warrant improvement.
1 il a
i I
8 4.
Interact with domestic and foreign research progrens in the subject area.
1 E.
continue and improve our utilization of university expertise.
l We have made a determination of the minimum research support level needed for the long i
term to assure that adequate expertise is available to the NRC.
In doing so, our contractors have.been consulted.
This determination considers the specialties and numbers-required for group dynamics and research. peer review.. The staff b111 eves that, in the long term, the budget for contract research in this program element I
should be maintained at about $3M.
The q
. current fiscal year budget for plant 4
performance research is $8.1M and declines to
$5.SM in FY-1990.
For reference, Figure 1 shows the history of RES thermal hydraulic i
funding.
Planned RES professional staffing for the plant performance program element is three FTEs.
It is expected that a significant fraction
~
(i.e., approximately 25%) of the work will be university projects, or joint national j
laboratory / university projects.
University involvement will be sought-to utilize academic expertise.
With the success of the B&W test loop at the University of Maryland, plans are being formulated to initiate other test loops at universities.to provide
.I experimental data and models in-the future, j
L
. I o
In the past when a large effort was-underway,-research was conducted at most
{
of the national laboratories, as well as with other contractors.
Now, the level of effort is greatly diminished and is limited primarily to three national laboratories.
To what extent should the-a research be further consolidated?
Several years ago as part of its plan for technical integration of thermal hydraulics (Reference 6), the N R C established a-policy of concentrating its remaining thermal j
J hydraulic research at INEL.
This' step was 1
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teken because research reductions had raised a concern that an effort dispersed among several national laborateries would result in groups too small to have effective. group dynamics and internal peer review.
INEL was chosen since.it had the largest remaining thermal hydraulics grcup and has performed well.
Certain work continues to be performed at IANL and SNL.
At LANL, the-research centers i
around the TRAC-PWR code, which was developed at that laboratory.
The BNL effort is i
associated with the RANONA code, used for:
analysis of reactivity-transients in BWRs.
i To date and for the near future, this arrangement is the most efficient, fWe'would.
of course continue support for university and international test programs, as requirs4.
)
l The possible negative aspects of consolida-tion were considered, three of which can be identified.
First is the effect of com-1 petition, or_ lack thereof, of different laboratories submitting proposals on a given j
- issue, second, is the question of assuring sufficient external input and review of the l-research..Here, interactions with universities and domestic and international l
organizations can be utilized to overcome l
this difficulty.
In terms of project =
j 1
management, a benefit is ehtained from consolidaticn'in that the tasks of i
j coordination.and integration become less-important-compared to planning and review of l
i the research.
The third possible l
detriment of consolidation is the question of i
whether unique expertise becomes lost if j
laboratory programs are terminatec
{
arbitrarily.
These factors mitigate against immediate, complete consolidation.
We are-proceeding.with these factors in mind.
o.
Industrial and international cooperation have played an important role in the research effort.
The intent was to i
pursue. cooperative programs with-industry and international organizations.
when6ver subjects were identified which 1
would be mutually beneficial to resolve S'
I N
4 10 via cooperative research.
To what
~
extent should industrial and international cooperation be continued?
Our current domestic industrial thermal hydraulic cooperation consists of the B&W Testing Program, while our international cooperation includes:
four agreement based programs (2D/3D, ROSA-IV, BETHSY, and ICAP);
participation in CSNIr and temporary assignments et technical personnel.
The NRC has, in the poet,-played a leadership rnie in fostering international cooperation.
- Also, because of the extensive past research, the RES codes REIAP and TRAC are currently the world standard for safety analysis of reactor transients.
This technological leadership may be difficult to maintain in the future due to strong research programo underway in certain countries, notably France, Japan, and Federal Republic of Germany.
We propose tot (1) continus a multilateral prgram based on RF. LAP and TRAC folicwing expiration.of the current ICAP program in 19918 (2) continue to-explore bilateral lesearch agreements of benefit to the NRCI (3) continue to explore cooperative programs with domestic industry (including EPRI); (4) 5 continue to' participate in CSNI activitiest (5) continue to accept temporary assignments of foreign technical personnel (although few of these are--anticipated in the future); and (6) consider temporary assignments of U.S.
technical personnel to leading foreign research programs.
The multilateral program (Item 1) will be developed with the following objectivest (1) share the effort and resources required for code maintenance; (2) retain access to foreign experimental-facilities and expertisst and (3) continue international cooperation concerning improved analysis of reactor transients.
o Considering that'a large amount of information and considerable expertise has been developed from past research, what future applications of such-knowledge to ass!.st the Staff are
'j envisaged?
~
w
l
'8 i
l 11 A new prograu element, Reactor Applications, was initiated last year to apply.the research results from the Plant Performance element.
I Reactor Applications research consists of performing-LWR systems studies for both operating reactors and those of advanced design.
Thus, Reactor Applications is 1
concerned with resolving issues by 1
application of information and computer codes l
developed through Plant Performance research.
l This element includes three activitiest (1) Analysis of Operating Reactor. Events; (2) Light Water Reactor (LWR) Systems studiest and (3) Thermal Hydraulic Technical f
support Center.
The first activity: relates to issues highlighted by operating experience.
For i
example, a subject of recent interest, as a result of the LaSalle event of March 1988, is-BWR stability.
The purpose of the research i
in progress is to determine the extent of safety concerns associated with BWR oscillations and whether any unacceptable conditions exist that might warrant 3
regulatory action.
To accomplish this, we 4
ares (1) performing code validation for such applications; (2) performing additional ATWS f
analysis to determine whether oscillations would be expected and.their consequences; and (3) determining the key parameters affecting onset and amplitude of esci11ations and their 1
mode (uniform or nonuniform).
The instability research is the most recent t
example of-research needs identified as a i
result of operating experience, either i
directly or indirectly.
Issues continue to arise requiring analytical assessment in the area of plant response to off-normal I
The second activity (LWR Systems Studies) is associated with advanced LWR designs.. The first task is to determine.whether, and in what ways, the transient response of advanced t
LWRs differs from current designs.
From.
l this, a determination will be made of whether i
the thermal hydraulic c?9ee are validated for such applications.
The seceta task includes b
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er.,.. - -
-.,-v--w
---,,,w-...
r
/
e 12 code validation and model development, as required, to establish or improve the applicability of. current analytical tools to transient analysis of advanced LMRs.
The goal of this effort is to establish the accuracy and reliability of the current RES LWR systems codes for the new plant designs.
Since these new designs have not been fixed, the long-term funding level for~this activity may require further definition.
The third activity cited (Technical support center) includes three tasks.
First is to help resolve regulatory issues.
Two examples of the types of issues that are addressed are long-term cooling following a LOCA, and the consequences of an interfacing systems (high-low pressure) LOCA.
The second task is the preparation of plant input decks.
Plant designs are generally unique, and'the particular design is important to the outcome of a given transient scenario.
This task is concerned with extending the library of plant.models available for analytical studies.
The third task within this activity is concerned with the synthesis and integration of information on given subjects.
Information normally exists in the form of a multitude of reports on experimentri end code calculations.
The task of synthesis and integration is intended to distill this information into a form that can be more easily utilized.
The Reactor Applications program element is funded at $5.3M for FY 1990 and' planned at i
about $3 to $4 million in later years.
This assumes that there will be no major thermal hydraulic test facility.for advanced LWRs.
The RES staffing level is three FTEs.
coordination:
The ACRS has reviewed this approach for q
future thermal hydraulic research and they agree with the. general objective of the research program to maintain, within the NRC and its contractors, a capability for thermal hydraulic analysis sufficient to deal with-safety and regulatory concerns that might-
E 13 arise in the future.
Also, they agree with the general level of funding projected for the next several years.
In addition, tho' ACRS provided several relatively specific I
recommendations which were, for the most part, consistent with our plans.
We intend to continue to interact routinely with the ACRS to ensure they are kept informed and to
-obtain their recommendations.
See for the ACRS letter and Staff response.
Pecommendation:
That the Commission take note of the approach proposed herein for the future goals and directions'of thermal hydraulic research.
Schedulingt This paper is scheduled to be considered at an open meeting on August.3, 1989.
i ter S 1
, Jr.
E cutive rector for operations Attachments:
1.
Appendix 1:
Historical Perspective of Thermal Hydraulic Research 2.
Figure 1:
Funding for Thermal Hydraulic Research 3.
ACRS letter of June 15, 1989 and Staff response This paper is tentatively scheduled for discussion at an Open Meeting during the Week of July 31, '989.
Please refer to the appropriate Weekly Commission Schc6if., when published, for a specific date and time.
DISTRIBUTION:
Commissioners OGC OIG LSS GPA REGIONAL OFFICES EDO ACRS ASLBP ASLAP SECY
^
+
4 References 1.
Compendium of ECCS Research for Realistic 'IcCA Analysis, NUREG-1230, December, 1988.
2.
Acceptance criteria for Emergency Core. Cooling Systems for Light-Water-Cooled-Nuclear Power. Plants, United States Atomic Energy Commission, Docket No. RM-50-1, December, 1973.
i 3.
Nuclear Power Plant Thermal-Hydraulic Performance'Research Program Plan, NUREG-1252, July, 1988 4.
Thermal-Hydraulic Research Plan for Babcock and Wilcox i
Plants, NUREG-1236, January, 1988.
5.
Plan for Integrating Technical Activities Within the U.S.
NRC and Its Contractors in the Area of Thermal Hydraulics, NUREG-1244, March, 1987.
6.
Reviews of Modern Physics, Volume 47, Supplement 1, 1975.
(
l l
L u
[
ATIACHPEtG 1 L
APPENDIX 1 i
Historical Perspective of Thermal Hydraulic Research For many years following its emergence in the mid-19(Os, the issue of Emergency Core Cooling System (ECCS) performance was dominant in reactor safety.
The need to standardize ECCS analyses for licensing proceedings eventually led to the adoption of 10 CFR 50.46 and Appendix K to 10 CFR 50.
Upon promulgation i
of the original ECCS rule in 1973 the Commission mandated that a research program be carried out to develop a better understanding of ICCA related phenomena.- A subsequent review of the research
~
p n y m by the American Physical Society highlighted phont senological and modeling issues which needed to be addressed i
to develop a realistic treatment of LOCA/ECCS (Reference 6).
This review noted the importance of scaling and systems modeling.
The NRC, as well as other organizations, planned and carried out a large program of thermal-hydraulic experimentation and model development.
Puel behavior experiments were also performed to study LOCA-related fuel issues.
This work concentrated initially on large break LOCA since this was believed to be the most i
limiting event from the standpoint of ECCS effectiveness.
The advent of WASH-1400 began to focus attention additionally on small break LOCAs and transients, a process which was greatly accelerates and enhanced by the occurrence of the TMI-2 accident.
Emphasis on thermal hydraulic research likewise shifted to small break LOCAS and transients.
The NRC successfully accomplished the original mission.of the LOCA/ECCS research.
Questions and concerns associated.with ECCS performance are no longer important issues..When questions arise regarding LOCAs or transients the experimental data base, analytical tools, and expertise are available to be applied to resolve the issue.
This being so, 10 CFR 50.46 and Appendix K were revised in 1988 to permit the use of best-estimate analyses of LOCA for licensing purposes, with appropriata accounting for uncertainties.
The principal products of thermal hydraulic research were j
computer codes that can be applied to understand and predict plant response to deviations from normal operating conditions.
The codes model the plant behavior by describing the processes of 4
heat transfer and fluid' flow.
Code development and experimental programs proceeded according to a feedback process.
As different i
scenarios were encountered or postulated or potential modeling deficiencies identified, particular experiments were run to obtain data necessary to establish the code-accuracy or to improve the code.
The interlinkage of code development and i
experimental programs is such that one cannot exist without the i
other.
Events involving new phenomena were encountered periodically in operating reactors for which code applicability i.
Y S
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I
..'a l
I 2
had not been verified.
This necessitated a dual analytical and experimental approach to help resolve such issues.
The attached i
table provides examples of issues resolved using this approach.
Since the ability to determine the uncertainty of a large complex
- i systems code was subject to some debate, RES undertook to develop a suitable methodology.
The approach is known as Code i
Scalability, Applicability and Uncertainty (CSAU) methodology.
The method is general in concept and was. applied to the TRAC-PWR code modeling a large break LOCA in a Westinghouse 4-loop plant.
-Briefly, CSAU determines whether the code, which is developed and assessed against scaled experimental facilities, is appropriate
]
for full scale applications.
It also includes a determination of whether the code has appropriate models for the important phenomena.
Finally, CSAU incorporates the ranging of important i
parameters over their uncertainty ranges, as determined from separate effects experiments.
RES believes that the CSAU application produced final closure on the issue of large break l
LOCA.
It provided a best-estimate core peak temperature history for the event, along with a statement of uncertainty.
This was done in a scrutable, traceable, and auditable manner.
{
e The process by which the CSAU methodology was developed was as essential as the product.
It was a cooperative effort among l
personnel from three national laboratories, Idaho National Engineering Laboratory (INEL), Brookhaven National Laboratory (BNL), and Los Alamos National Laboratory (LANL).
It also utilised input from university professors.and consultants.
The personnel had extensive experience and expertise, developed largely through their participation in past NRC thermal hydraulic research.. The participants proceeded under a technical program group structure headed and closely coordinated by a RES Project Manager.
This successful method'is now being applied by RES to address scaling issues associated with severe accident research.
i
{
While not relevant to the mission of the NRC, it is noteworthy that past RES thermal hydraulic research has produced other i
unintended benefits.
Currently, RES codes are being utilized for l
safety analysis of production and research reactors.
Research into two-phase flow has also been applied to petroleum and t
chenical industries and to space applications.
1 l
Role of International Cooeeration i
i International cooperation has played an important role in thermal-hydraulic research.
The goals RES pursued were to share safety technology in the interest of international nuclear l
safety; obtain access to foreign experimental results and expertise; and save resources by carrying out large programs with other countrics on a shared basis.
The principal means of international cooperation RES has undcrtaken are described in the following, U
'l 3
It became apparent around 1975 that the cost to obtain large scale experimental data necessary to resolve the IOCA/ECCS i'ssue unilaterally was prohibitive, so RES began discussions with Japan and the Federal Republic of Germany on the conduct of a joint program.
Several years of planning and negotiation led to the formation of the 2D/3D program, which began in 1980 and will completed in 1990.
Three large facilities were constructed, two in Japan (Cylindrical core Test Facility, Slab Core Test Facility) and one in Federal Republic of Germany (Upper Plenum Test Facility).
The RES contribution included advanced instrumentation for~these facilities and the development of advanced analytical tools need to model the complex phenomena being studied in these facilities.
Following the TNI-2 accident, Japan decided to build a large scale (1:50) integral test facility to investigate small break L
A number of ancillary facilities also make up the program known as ROSA-IV.
RES interacted with_ Japan from the start of the program and provided advanced instrumentation to the facility.
A bilateral agreement was signed in 1984,.
experimentation began in 1985, and the cooperation currently extends to 1992.
Until 1984, RES made its thernal hydraulic codes openly available through the National Energy Software Center.
Different countries obtained the codes and modified versions proliferated.
RES did not attempt to obtain feedback from foreign users nor were.
foreign code assessment results particularly useful, being performed as they were with various unique code versions.. Until that time RES had sponsored its own independent code assessment program.. Budget constraints made this no longer possible, therefore, RES organized the International Code Assessment and Applications Program (ICAP) through a series of bilateral agreements with foreign safety authorities.' The ICAP duration is from 1986-91.
The intent was tot o
Develop a common understanding of the ability of the code to appropriately represent important physical phenomenn; o
Share user experience on code assessment and present a well documented assessment data baser o
Share experience on code errors and inadequacies and to cooperate in removing the deficiencies to maintain a single, internationally recognized code version; and l
Establish and improve user guidelines for applying the.
o code.
'l v
o 4
4 The ICAP program is successfully accomplishing these goals.
Participating organisations have benefited from the establishment of a code users group.
ICAP provides the NRC access to foreign facilities, experimental results, and expertise.
An additional role was played by the Committee on the Safety of Nuclear Installations (CSNI) thermal-hydraulics working group.
This provided a forum for multilateral discussion and information exchange.
It also provided a structure for the conduct of international standard problem exercises to evaluate how well codes and analysts could calculate thermal hydraulic experiments.
Mention should be made of the role of temporary assignments of foreign technical personnel to national laboratories, principally INEL, where about 70 such assignments have occurred since 1974.
The purposes of these assignments included contributing to collaborative programs, training, and liaison.
On the U.S. side, a limited number of assignments to collaborative programs with Japan and Federal Republic of Germany have taken place, the last of which is ending this fiscal year.
Cooneration with Domestic Industry Where appropriate, RES has cooperated with U.S. industry in carrying out research programs.
The establishment of joint programs is normally sought as a means of sharing the costs involved and in broadening the technical input to given projects.
The industry groups most often involved are the Electric Power Research Institute (EPRI), the NSSS vendors, and utility owners groups.
Past examples include cooperation with:
General Electric and EPRI on the development of the. TRAC-BWR code and the conduct of the FIST and BWR-FIECHT experimental programs; and' Westinghouse and EPRI on the-conduct of the FLECHT, FLECHT-SEASET and MB-2 experimental programs.
Currently, the only cooperative program with industry is the Integral System Test program with the B&W owners Group, B&W, and EPRI.
)
___-______m
L O
TABLE 1 EXAMPLES OF DUAL EXPERIMENTAL /
ANALYTICAL APPROACH TO RESOLVE ISSUES 4
no.,s Mergin of Conservatlem in LOFT, Somlocale UPTF, CCTF Analyses of Test Apoendia K; Revision to SCTF, TLTA, SSTF, FAST, LOBI,-
Foollity Data and Appendia K (LOCA)
Marviken, PKL, Create, BCL Full-Scale,LWRs Creare, Purdue, Semlocale, TRAC-PWR and RELAPS Pressurized Thermal Shock UPTF, HDR, Finland, and Analysis of Oconee H. B. Robinson and Calvert Cilite Small Break LOCA in W PWRs Semiscale, LOFT, ROSA-IV, LOBI TRAC-PWR and RELAP5 Small-Break LOCA and Natural TRAC-PWR and RELAPS Circulation in B&W Reactors Analysis of Data l,.
^"* '#
Feed-and-Bleed Procedures g,gp for Decay Heat Removal
^ ^ ^" Y"'
S-PL-3. LOFT LP-FW-1.
in PWRs Performance of Upper Head
~
and Upper Plenum injection COBRA / TRAC TRAC-PWR 3D, CTF S F In W Reactors Sme!1-Break LOCA with loss of ROSA-IV, LOFT, LP-SB-3, RELAP5, TRAC-PWR, High-Pressure injection Semiscale S-NH series SASA Analyses Semiscale S-UT-6, al-UT-8, RELAP5, TRAC-PWR, Gen a ors Our ng S 11 S-LH-1. S-LH-2, ROSA-IV NOTRUMP Break LOCAs ROSA-IV, Semiscale TRAC-PWR, RELAP5 Rupt te (SGT i
-3, W, REW5, RAWA-3B, Wt ut ta T S)in Semiscale S-PL-T, FIST TRAC-BWR l
PWRs and BWRs lodine Behavior Following MB-2, ORNL, Northwestern CITADEL, TRAC-PWR, I
SGTR University RELAP5 Natural Circulation TRAC-PWR and RELAP5 3 LE M
$fQ
L TABLE 1 (CONTINUED)
EXAMPLES OF DUAL EXPERIMENTAL / ANALYTICAL-APPROACH TO RESOLVE ISSUES w ee Empewneau Anse Fluid Structure interaction on Reactor Core Barrel and HDR, sal K-FIX (FLX)
Vessel Internais after LOCA BWR Containment Pr>ssure MIT, GE, Uvormore PELE, '80LA 8.ppression Pool Loads Marviken Stability Margins for BWRs DRESDEN, FRIGO NUFREO
~IMI "'
M{S TRAC-PWR, RELAPS TMI-2 Accident TlS Plant Transients ANO-il LOFT L6-7 Crystal River Semiscale S-PL-3 TRAC-PWR, RELAP5 Ginna LOFT L6-6C St. Lucio Semiscale S-FS series Davis-Besse MIST, OTIS
~
~
Effects of Reactor Coolant
~80~2
~ ~
Pump Operation During PWR TRAC-PWR, RELAPS g
g,gg,p '
Small-Brn'.; LCCA
-P2,-P7 e-e
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AWACHMEt@ 2; Figure.I-FUNDING FOR T/H RESEARCH 100 m
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0 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89. 901 91 FISCR WAR
- Dollar Amounts not Adjusted for Inflation a
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I
j.' -.
i ATTACHMEtR 3 4
Ma
- aes,q%',.
n UNITED sT ATEs NUCLEAR REGULATORY COMMISSION s
. [
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS waggggeGTON, D. C,30646 -
- o
- e.... /
June 15, 1989'
- L J
The Honorable Lando W.: Zech, Jr.
Chairman U.S. Nuclear Regulatory Connission i
-Washington,D.C. 20555
Dear Chairman Zech:
-l i
SUBJECT:
NRC THERMAL-HYDRAULIC RESEARCH PROGRAM-During the 350th meetingL of the Advisory Connittes' on JReactor : Safeguards,.
+
we - reviewed the NRC's. plan for. continuing themal-hydraulic d
June.8-10, 1989, research as-relatedi to'-the design and operation of ' nuclear power ' plants.-
This-matter was also considered - by our Subcommittee en Thernal Hydrau'Jic 4
Phenomena at.a meeting on May 23, 1989.
During these meetings, we :had-the-benefit of presentations by representatives of the Office ofiNuclear Pegula-
'm tory Research '(RES).
We also had the benefit of the documents referenced.
The Committee last commented to you on this subject in our, report of June 7.-
g 1988..
Thermal-hydraulic research has, always been 'a central ' and major -part of1the d
NRC's research program.: Much ofc the work wasiinspired by thel perceived need J
better understand. hypothetical large-break -loss-of-coolant-accidents to(LB-LOCAs) and - the performance - cf emergency core cooling systems dECCSh
-l Experiments and - analytical' models such as. the ~ RELAPiand TRAC codes,. have j
confirmed compliance with the ECCS. rule.
Continuing research - on ' LB-LOCAs -
1 culminated with a 1988 revision to the ECCS rule which permitsLlicensees:to-i use more accurate means ~of analysis and.makes possible certain_ safety -and l
operationhl improvements in existing plants.
NRC contractors:.have demon-'
strated a methodology that. can-be used to estimate? the magnitude of uncer-tainty associated with~ code-predictions, j
In addition, the experimental.information base.and the codes have been:found j
'useful in aspssin<3 andL predicting the consequences of transients ~and small-
]
break loss-of-coolant-eccidents- ~(SB-LOCAs) which are.now recognized.to be much more risk significant than the LB LOCAs. The' codes are also being used j
to analyze the early stages of severe accident scenarios.
1 Proposed Research Program i
We understend the continuingENRC1 program; in-thermal-hydraulic research to
.have two. principal purposes:-
Bring development of the major' computer codes -to a successful 4 comple-0
- tion, a
j q
a w
4.,1 L
E E
The Honorable Lando W. Zech, Jr. June 15, 1989-
=
' Maintain, within the NRC and its contractors, a capability for thermal-hydraulic' analysts sufficient to deal with safety and: regulatory' con.
L 4
- cerns that might arise in the future.
This includes -the continuing y
availability of a cadre of experts.
J
~
.RES' representatives indicated. these L general purposes would be realized i
[
through achievement of several specific objectives:
- The major-codes will be maintained indefinitely and some further devel-
)
r 5:
opment will be -carried out. ' The scope and depth of further development.
seems not to have been decided. Apparently, it will includ.e appropriate reactions to new data from foreign experimental 1 programs and assessments which are expected : to continue for some time.
It may also include.a T
review and redevelopment of the important constitutive equations in the-R codes.
=
' The current experimental. programs related -to: specifics of the Babcock' l
F i
and Wilcox (B&W) nuclear steam supply (NSS) system will be completed.
Beyond this,' any further experimental programs will-be carried out at
{s univortities, rather ' then-by the creation - or operation of any major
. facilities at national laboratories.
Relatively inexpensive " integral" facilities, of scope similar to the ' facility now operating: at the=
a University - of Maryland, are being considered as contrasted with what -
b have been called " separate effects" facilities.-- These would;be mockups y
of specific NSS systems and of.an' advanced LWR (600 MWe' size) design.
An expanced program of sapplications; research.is-planned.
Apparently, j
T much of this activity is expected to be in response to issues that'arise.
H L
from experiences with operating plants.
But, it -will' include prepara--
tion of input data for several mort plant types-than arei now available
'l to the NRC. This will pennit more rapid ' analysis Lthan 'would otherwise be possible in response. to : future safety or regulatory issues.
This
. program c" also include exploratory,J iri-depth studies of a range of :
possible transients for a variety of plants.
In addition, two other specific program elements were mentioned:
c A further demonstration of the " Code Scaling, Applicability, and Uncer-tainty" methodology will be carried out for Kn 58-LOCA with:RELAP5/M002, similar to that recently completed for an LB 40CA.
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- Improvements wili te made to the NSS system process models now incor-i porated in training simulators at.the HRC Technical Training Center.
1 This will permit more. accurate simulationi of-off-normal scenarios for
.the' study of emergency.and accident management procedures.
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J. The Honorable Lando W. Zech, Jr. June 15, 1989 l Before commenting on these research proposals, it is pertinent to consider two statements 'made by the NRC staff at the May 23, 1989 Thermal Hydraulic Phenomena Subcommittee meeting, because the ideas expressed have an influence on our recomendations: A representative of L the Office of Nuclear Reactor Regulation said, "NRR is not-relying extensively on the codes to address current licensing issues." 1 A representative of RES said, " Codes have now reached an accept-able level of accuracy-and maturity... further development-is not -likely to produce major changes in our understanding of 1 [ plant]performanceor-[ accident] consequences." l - ACRS Recomendations We agree with the general. objective of the research, program to maintain. within the NRC and its contractors, a' capability for -thermal-hydraulic ] analysis sufficient to deal. with safety and regulatory concerns that might arise in the future.,Also, we agree with the -general-level of funding pro-jected for the next - several ' years. However, we believe there1 is too much l emphasis on further development of the-existing codes in the-planned program.. 4 Maintenance of the needed NRC. capability is more at matter of. ensuring.- the availability of a cadre of experienced and expert analysts and access to the' general body of exparimental data,- than it is of improving or even ensuring the : availability of large systems codes. The Comittee: reiterates lits' comments in the report of June-7,1988, that " marginal improvements :that -f could-be made [in the codes) over' the next few years -by extrapolating-the recent' levels of development work will not be sufficient to attain a signifi-cantly higher plateau of code accuracy and validation."' 1 To accomplish this general purpose, we recommend a: program of -four primary j elements: 1 (1) Code Development Maintain the - present large system codes. TRAC-PF1/ MODI, RELAP5/M002, 1 TRAC-BWR, and RAMONA-3B, for an indefinite period. ~ Limit improvements i only:to those required by: (a) the discovery of.important errors or (b) crucial new information from the foreign experimental and assessment programs or the B&W testing program. Do not undertake major new re-structuring or. "zero-based" improvements - to. the constitutive ' equations - l -or numerical algorithms. in these codes. We are not convinced - by the. l arguments given-for the need to develop TRAC-PF1/ MOD 2 and RELAPS/M003.- It is our view a ' the proposed modifications will not substantially-improve the codes. i y
q a 4 '/ %,, 9 ~ t 9 The Honorable Lando W. Zech, Jr. June _15, 1989 L Instead, consideration should be given to the development of a new type L of systems. code that will be more: useful for egiysis of, extended plant l L transients involving interactions of plant systems. The Comittee also i made this recomendation in its < June 7,1988 report.- TRAC and RELAP were originally 1 designed to analyze the LB-LOCA,_ a rapid and severe z [k", reactor transient, in great detail. There is a: need for a' more empir- . ical and' efficient analytica1' tool.. 'We envision a code that would be able, for example, to make a rapid and sufficiently accurate analysis of: the power oscillations observed last year.at the LaSalle County Station.- Unit. 2-plant. Such a' code would - be more akin to ' advanced simulator codes than to TRAC and RELAP. The BWR. code -(HIPA) now in use-at Brook-' l haven National. Laboratory is -an1 example of the. type of j code fwe are j suggesting. y 1 (2) Experimentation-L The staff proposal to develop: relatively inexpensive' " integral" = test j facilities at universities is sound. -We see this as consistent with uur-i recomendation for a_ new: type of systems code. We agree that it wuld i be s inappropriate to J build 'several such facilities att one - tim? A'
- j gradual approach is warranted. -- The first such,new facility might, ce q
that would incorporate features.'of the advanced LWR designs. Ah a. *: Jd will be better to: completely assess the' benefit. that has'heen obtained. from tests with' the University of Maryland facility mentioned above. 1 In-addition, a small program to' deal with more : fundamental. research. should be maintained. These are experiments of the sort that'have been j m previously called " separate effects" tests. An effort should be made to develop a consensus among experts as to which particular -phenomenon. should be investigated. 'At this time, we suggest consideration be given E ~' to.the investigation of: l L ' fluid-elastic instability related to vibration ~ of; tubes in U-tube-steam generators,- . j departure from nucleate boiling with esci11ating flow and power in; [
- BWRs, l
dynamic instabilities and loads on valves. (3) Data Analysis A major effort is - needed to organize data - fromL test -programs - into a l useful fonn other than the -large systems codes. In particular, with the, 20/30 ROSA-IV, and the B&W - test programs all; coming to _ closure, mea-il sures are needed to ensure that -_these expensive and valuable bodies of data are preserved and used. In addition, older data from, for example, = { l i y r i l 4
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+ The Honorable Lando W. Zech, Jr. -5 Juns 15. 1989 1 the. FIST. and FLECHT programs can be of greater -value if they are effectively organized into more useful forms. 1 (4)_-. Applications Research l A program in this area'should include three elements:
- Analysis of-transients. indicated to be of interest as' a result of plant operating experience.
01 Preparation-of input data decks for several classes of plants so ,i that turnaround time for analyses in response to experience -is - shortened. 1 O I Analysis of transients that are indicated by PRA' or other. sources of -information to '.be of particular interest. but which are not presently well understood.- We suggest'the following fer'considera-tion: j - feed and bhed scenarios - secondary depressurization scenarios. y Finally. we suggest that RES broaden its perspective as --to what otherJ re-search in the-thermal sciences should be included in.its program, rather than being limited to-.the traditional scope of concerns. in. thermal-hydraulic areas. 'We suggest that it include studies of' a broad range -of thermal-and: fluid transport issues related to reactor safety. ACRS Members William Kerr and Forrest Remick did-- not participate',-in' the review of this matter. Sincerely, David A. Ward Acting Chairman
References:
1. U.S.- Nuclear Regulatory Comission. draft SECY Paper:. Status and Plans for ~ - Thermal Hydraulic Research - Conducted by the Office of ~ Nuclear-Regulatory Research,"lprovided to the'ACRS in May 1989.' 2.. U.S.-- Nuclear Regulatory Comission, NUREG-1252:: " Nuclear Power Plant Thermal-Hydraulic. Performance REsearch Program Plan," Office of Nuclear-Regulatory Research,' July 1988 i .t [ f m o i1 l
..'l g p**C* %. i J#- k, UNITED STATES - 8 p - . NUCLEAR REGULATORY COMMISSION : 1.,,..)f. I i wasHWGTON, D. C. 20665 Jul. 2 0989 3 1 Mr. David A. Nard,, Acting Chairman -1 Advisory Committee On Reactor Safeguards U.S.: Nuclear Regulatory Commission Washington, D.C.. 20555 7 7,
Dear Mr. Ward:
SUBJECT:
NRC THERMAL-HYDRAULIC RESEARCH PROGRAM The' Advisory-Committee on Reactor Safeguards (ACRS)- lettier of ' June;15,z 1989,.provided comments on the sub act program. The 3 ACRS= agreed;with the general-objective to maintain expertise in i thermal-hydraulics to meet future agency needs in this field. The committee-alsocagreed with the general level of-funding 1 projected for the next several' years.- Parallel to the. staff's' e plans,:n program ccmprised-of four elements was recommended. ~ l statf review provides the.following responses:! ( (1) Code Development The ACRS may have misinterpreted'a staff" statement-to l, mean that older versions of computer' codes..were already-nature. This'would have ledLthe ACRS to recommend-that I no further work'is needed on'the final code versions,- .J l-TRAC-PFl/ MOD 2 and - REMP5/ MOD 3. ,s The staff-believes that the. final' versions of TRAC-PWR (TRAC-PF1/ MOD 2) and REMP (REMP5/ MOD 3),i as completed r
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'in 1989, have an acceptable' level of-accuracy-and that [ .further development is not likely.to produce.. major: L changes in our-understanding of plant performance and-accident consequences.- Previous versions.of:theseltwo I codes (TRAC-PF1/MODl' and REMP5/ MOD 2) were released in ' December 1984.' Since then, peer review and code.. applications studisea identified a number _of.modeling t deficiencies which we determined must be resolved to provide a reasonable, sufficiently accurate. representation of physical phenomena. The. knowledge-and. experience gained over the last five. years is reflected:in the~ final code. versions.- The codes are. being provided to'the International Code. Assessment j Program,(ICAP) members for assessment to be completed- -in December 1991, when.the ICAP program ends.- 1 We-believe that these. final code versions, plus the other NRL codes'altc 1 m aM m ed~(RAMONA-3B, HIPA, TRAC-BWR, at ' COB'. Y pc- ..e NRC with; sufficient i analytical ccpubi n - .w et us future needs, 1 4 4 i r I ~ if L m i, a- . a
.,n ?; t se 1 Mr.= David A. Ward 2 + i .i including-any neededl analyses of extended plant-transients involving-interaction of plant: systems. 1 This has been confirmed at several meetings among the .RES staff and-the user offices of NRR and AEOD that . ere convened'for-this purpose. The' staff thus wconcludes that no new-systems code development is q required:at thic time. .(2) Experimentation We noteuthat~the ACRS endorses our_ plan to develop . university experimentalDracilities. We intend to, I clearly-define what we expect from1any new test ~ facilities at' universities beforeLany proposals to perform: this-research > are issued.: Concerning:the suggestion to maintain a small program to deal with more fundamental research,_we' intend to include this in. 1 a Broad Agency Announcement fer' proposals similar:to l the ones you hava suggested.=-In-addition, we will' l continue our grants progtam which has: supported.such research efforts for several_ years. (3) Data Analysis You have' raised anconcern'about how beststo preserve. our research results:for use by;" future-generations" of-reactor safety experts. ~You are aware that we-have an experimental data bank, code maintenance programs, and gi publish-researchisynthesis reportsion special1 topics.1 j However, these existing programs may.not-fully _ address. your concern. The staff will" review.our current ~ capabilities and identify any, ways we can improveLon putting completed'research into-the'most useable form. We-vould appreciate'further interactions;with'the Committee regarding where.they think-our-current
- programs are deficient or'could be' improved.
(4) Applications Research. We agree with the'three-elements'you have suggested,- l and would add a-fourth, code applicability studies. 4 -Such' studies:better define the' classes of reactor-i georotries and accident: scenarios for which: specific l comy ,r: codes are presently applicable. -This-would . understood. 'j help to assure that code performance is well f i f . ) -
n ..~ .,.,,g. Mr. David A. Ward . 3 -- i i i 1 With regard to.the suggestion of including other research in the i thermal sciences, we propose to discuss this.further at a subcommittee meeting-in order to understand what the ACRS has in-nind. j sincerely,. l ./ -f. i or el , Jr. ecutive rector-for Operations. -ect chairman'Carr Commissioner Roberts -l Commissioner Rogers i Commissioner Curtiss OGC .l SECY L ? i 4 . I-1 8 ...1
r, .'a .f, I T ED NTATES Jf NUCLEAR r'E'.ULATORY COMMISSION c- [l g: Yv A?HNC !oN. O.C. 20666 June 26, 1989 oFFRE oF THE - COMul8840NER MEMORANDUM' PORT. Victor Stello, Jr.: Executive Director for Operations Kenneth. C. - Rogers.. PRON:
SUBJECT:
ACRS LETTER OF JUNE 15, 1989 ON NRC THERMAL-HYDRAULIC RESEARCH PROGRAM. I. consider the subject ACRS letter excellent:and lucid advice ~ I on the office of. Nuclear Regulatory Research' (RES) ; thermal-hydraulics.research program. I' support'the ACRS:recommenda-tions in the letter and tho'four primary elements of this' ' Program. Please inform me of any RES comments on the'ACRS recommendations or their. plans for implementation,. f 1 1, 4 Kenneth C. Rogers Commissioner cc: Chairman Zach Commissioner Roberts Commissioner Carr Commissioner.Curtiss OGC SECY 'l 4 q AdG:i,Eic De:a __ (. M t') q q -.1 . i 'i 100 --004 5 E1 i 3}}