ML20247J871
| ML20247J871 | |
| Person / Time | |
|---|---|
| Issue date: | 07/24/1989 |
| From: | Stello V NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| TASK-PICM, TASK-SE SECY-89-219, NUDOCS 8907310357 | |
| Download: ML20247J871 (30) | |
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%..... +"g POLICY ISSUE July 24, 1989 SECY-89-219 For:
The Commissioners From:
Victor Stello, Jr.
Executive Director for Operations Subiect:
STATUS AND PLANS FOR THERHAL 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.
2.
Inform the Commission of future goals and directions of this research.
Backaround:
Thermal hydraulic research conducted since the inception of the NRC in 1975 and its l
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 Reference 1.
The revised rule permits 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, overly 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 embrittlement, and increasing plant capacities.
Eventual revision of the LOCA Rule was a goal set forth by the Atomic Contacts:
D. Bessette, RES, 49-23572 l
L. Shotkin, RES 49-23530
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Energy Commission when it adopted the rule in 1973 (Reference 2), and this has now been achieved.
In addition, research has studied a spectrum of small break LOCAs and transients, and computer codes have been developed and assessed for such events.
As a result of these research accomplishments, our understanding of and confidence in predicting LWR thermal hydraulic performance during transient and accident conditions has been greatly improved.
Commensurate with this improved level of understanding, our research efforts in this area have decreased significantly over the last several years and are scheduled to decreace further.
At this 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 research.
Discussion:
Thermal hydraulic research is intended to support the Staff in the following areas:
o Understanding reactor transient events and their broad implications for operational safety; o
Detection of previously unrecognized issues important to safety; o
Investigating and resolving specific
- issues, e.g.,
the effectiveness of decay heat removal via feed and bleed; o
Evaluating the effect of design and operations-related changes, including operating procedures and changes to technical specifications and setpoints; i
Analyzing the early phases 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 the plant performance program element, and includes three activities.
The status and schedule of each activity is summarized briefly in the following.
1.
Babcock and Wilcox Testing The basic objective of this work, endorsed by the ACRS, is to provide a data base for B&W designs that is comparable to that which exists for the other NSSS vendors. Test facilities such as LOFT, SEMISCALE and FIST provided integral facility data for Westinghouse, 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 codes 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 Secc 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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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 experiments on once-through steam generator (OTSG) performance.
Recently, agreement was obtained with the B&W Owners Group to conduct a cooperative experimental program on OTSG performance.
This experimental program, expected to be completed in FY 1992, will make the experimental thermal hydraulic data base for B&W plants comparable to that which exists for the 3 other vendors.
2.
Experiments and Analysis 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 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 be 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 needed to model the complex phenomena being studied.
Experimentation l
will be completed in the current fiscal year, l
analysis in FY 1990, and final reporting in FY 1991.
Following the TMI-2 accident, Japan decided to build a large scale (1:50) integral test facility to investigate small break LOCAs in PWRs.
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 l
1985, and the cooperation currently extends to 1992.
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Several past experimental programs such as LOFT, 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 1
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 modeling 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 assessed under the ICAP program through 1991.
No new versions of these codes are planned unless assessment results show significant deficiencies in the ability to predict nuclear plant performance.
In this code development activity starting with the effort to revice 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,
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data base, and range of applicability.
User guidelines for nodalization and applying the code were improved and documented.
An independent review was conducted of software quality assurance procedures in use at INEL and LANL to assure conformity with ANSI standards and prevailing industry practices.
Future Research Plans The governing objective for future thermal hydraulic research was developed as a joint, staff position.
RES and the principal user office, NRR, interact on a routine basis, both formally and informally (Reference 5).
AEOD also participates in this process.
The formal mechanisms are:
(1) the Reactor Systems Safety Senior Research Program Steering Group, headed by the Director, Division of Systems Research, RES and 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:
o The NRC staff will continue to need independent expertise and analytical capability for addressing transients in PRRs and BWRs; and o
The principal RES analysis codes (TRAC-PWR, TRAC-BWR, RELAP5, 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 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 issues typically require code calculations.
The codes are large, complex and require experienced users.
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Code development has proceeded in an interrelated fashion with experimental programs.
Most of this work is complete; however, certain projects remain.
The 2D/3D program will end in 1990.
The experimental programs associated with B&W designs will finish in 1992.
The Japanese ROSA-IV program 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-quent 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 chock.
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-lenging to attract and retain researchers and must be consistent with the NRC's mission.
We plan to:
1.
Complete the current projects within the three activities described previously, i
which should occur by 1992.
1 2.
Continue code maintenance for the RES FWR and BWR systems codes, and keep the j
software in pace with advances in j
computers.
3.
Perform research as appropriate, directed at improving the accuracy of the codes in areas determined to warrant improvement.
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4.
Interact with domestic and foreign
-research programs in the subject area.
5.
Continue and improve our utilization of university expertise.
We have made a determination of the minimum.
research support level'needed for the long 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 believes that, in the long term, the budget for j
contract research in this. program element 1
should be maintained at about $3M.
The current fiscal year budget for plant performance research is $8.lM and declines to
$5.8M in FY-1990.
For reference, Figure 1 shows the history of RES thermal hydraulic funding.
Planned RES professional staffing for the plant performance prog 2*m 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 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 experimental data and models in the future.
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 research be further consolidated?
Several years ago as part of its plan for k
technical integration of thermal hydraulics (Reference 6), the NRC established a policy of concentrating its remaining thermal hydraulic research at INEL.
This step was
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taken because research reductions had raised a concern that an effort dispersed among several national laboratories 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 group and has performed j
J well.
I Certain work continues to be performed at LANL and BNL.
At LANL, the research centers around the TRAC-PWR code, which was develope l1 at that laboratory.
The BNL effort is associated with the RAMONA code, used for i
analysis of reactivity transients in BWRs.
I To date and for the near future, this arrangement is the most efficient.
We would of course continue support for university and international test programs, as required.
The possible negative aspects of consolida-tion were considered, three of which can be identified.
First is the effect of com-petition, or lack thereof, of different laboratories submitting proposals on a given issue.
Second, is the question of assuring sufficient external input and review of the research.
Here, interactions with universities and domestic and international organizations can be utilized to overcome this difficulty.
In terms of project management, a benefit is obtained from consolidation in that the tasks of coordination and integration become less important ccmpared to planning and review of the research.
The third possible detriment of consolidation is the question of whether unique expertise becomes lost if laboratory programs are terminated 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 pursue cooptarative programs with industry and international organizations whenever subjects were identified which would be mutually beneficial to resolve
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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 CSNI; and temporary assignments of technical personnel.
The NRC has, in the past, played a leadership role in fostering international cooperation.
- Also, because of the extensive past research, the RES codes RELAP 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 programs underway in certain countries, notably France, Japan, and Federal Republic of Germany.
We propose to:
(1) continuo a multilateral program based on RELAP and TRAC following expiration of the current ICAP program in 1991; (2) continue to explore bilateral research agreements of benefit to the NRC; (3) continue to explore cooperative programs with domestic industry (including EPRI); (4) continue to participate in CSNI activities; (5) continue to accept temporary assignments of foreign technical personnel (although few of these are anticipated in the future); and (6) consider t;emporary assignments of U.S.
technical personnel to leading foreign research programs.
The multilateral program (Item 1) will be developed with the following objectives:
(1) share the effort and I
resources required for code maintenance; (2) retain access to foreign experimental facilities and expertise; 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 assist the Staff are envisaged?
A 11 A new program element, Reactor Applications, was initiated last year to apply the research results from the Plant Performance element.
Reactor Applications research consists of performing LWR systems studies for both operating reactors and those of advanced design.
Thus, Reactor Applications is concerned with resolving issues by application of information and computer codes developed through Plant Performance research.
This element includes three activities:
(1) Analysis of Operating Reactor Events; (2) Light Water Reactor (LWR) Systems Studies; and (3) Thermal Hydraulic Technical Support Center.
The first activity relates to issues highlighted by operating experience.
For example, a subject of recent interest, as a result of the LaSalle event of March 1988, is BWR stability.
The purpose of the research in progress is to determine the extent of safety concerns associated with BWR oscillations and whether any unacceptable conditions exist that might warrant regulatory action.
To accomplish this, we are:
(1) performing code validation for such applications; (2) performing additional ATWS analysis to determine whether oscillations would be expected and their consequences; and (3) determining the key parameters affecting onset and amplitude of oscillations and their mode (uniform or nonuniform).
The instability research is the most recent example of research needs identified as a result of operating experience, either directly or indirectly.
Issues continue to arise requiring analytical assessment in the area of plant response to off-noimal transients.
l The second activity (LWR Systems Studies) is 1
associated with advanced LWR designs.
The first task is to determine whether, and in what ways, the transient response of advanced LWRs differs from current designs.
From I
this, a determination will be made of whether the thermal hydraulic codes are validated for such applications.
The second task includes
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12 i-code validation and model development, as required, to establish or improve the applicability of current analytical tools to transient analysis of advanced LWRs.
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 experiments and 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 j
funded at $5.3M for FY 1990 and planned at about $3 to $4 million in later years.
This j
assumes that there will be no major thermal hydraulic test facility for advanced LWRs.
The RES staffing level is three FTEs.
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Coordination:
The ACRS has reviewed this approach for j
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
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13 arise in the future.
Also, they agree with the general leve) of funding projected for the next several years.
In addition, the ACRS provided several relatively-specific 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.
Recommendation:
That the Commission take note of the approach proposed herein for-the future goals and directions of thermal hydraulic-research.
Schedulina:
This paper is scheduled to be considered at an open meeting on August 3, 1989.
4 i tor S 1:
, Jr.
E cutive Dkrector 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, 1989.
Please refer to the appropriate Weekly Commission Schedule, when published, for a j
specific date and time.
DISTRIBUTION:
Commissioners OGC OIG LSS GPA REGIONAL OFFICES EDO ACRS ASLBP ASLAP SECY l
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4 References 1.
Compendium of ECCS Research for Realistic IDCA 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.
3.
Nuclear Power Plant Thermal-Hydraulic Performance Research Program Plan, NUREG-1252, July, 1988 4.
Thermal-Hydraulic Research Plan for Babcock and Wilcox 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.
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ATTACHMEiVT 1 I
APPENDIX 1 Historical Perspective of Thermal Hydraulic Research For many years following its emergence in the mid-1960s, 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 of the original ECCS rule in 1973 the Commission mandated that a research program be carried out to develop a better understanding of LOCA related phenomena.
A subsequent review of the research program by the American Physical Society highlighted phenomenological and modeling issues which needed to be addressed 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.
Fuel 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 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 accelerated 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 appropriate accounting for uncertainties.
The principal products of thermal hydraulic research were 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 heat transfer and fluid flow.
Code development and experimental programs proceeded according to a feedback process.
As different 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 experimental programs is such that one cannot exist without the other.
Events involving new phenomena were encountered periodically in operating reactors for which code applicability I
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1 had not been verified.
This necessitated a dual analytical and experimental approach to help resolve such issues.
The attached table provides examples of issues resolved using this approach.
Since the ability to determine the uncertainty of a large complex systems code was subject to sohe debate, RES undertook to develop a suitable methodology.
The approach is known as Code 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 parameters over their uncertainty ranges, as determined from l
separate effects experiments.
RES believes that the CSAU application produced final closure on the issue of large break LOCA.
It provided a best-estimate core peak temperature history l
for the event, along with a statement of uncertainty.
This was done in a scrutable, traceable, and auditable manner.
The process by which the CSAU methodology was developed was as essential as the product.
It was a cooperative effort among personnel from three national laboratories, Idaho National Engineering Laboratory (INEL), Brookhaven National Laboratory (BNL), and Los Alamos National Laboratory (LANL).
It also utilized 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.
While not relevant to the mission of the NRC, it is noteworthy that past RES thermal hydraulic research has produced other unintended benefits.
Currently, RES codes are being utilized for safety analysis of production and research reactors.
Research into two-phase flow has also been applied to petroleum and chemical industries and to space applications.
Role of International Cooperation 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 safety; obtain access to foreign experimental results and expertise; and save resources by carrying out large programs with other countries on a shared basis.
The principal means of international cooperation RES has undertaken are described in the i
following.
o.
3 It became apparent around 1975 that the cost to obtain large scale experimental data necessary to resolve the LOCA/ECCS issue l
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 Vill completed in 1990.
Three large facilities were constructed, two j
in Japan (Cylindrical Core Test Facility, Slab Core Test j
Facility) and one in Federal Republic of Germany (Upper Plenum j
Test Facility).
The RES contribution included advanced j
instrumentation for these facilities and the development of advanced analytical tools need to model the complex phenomena being studied in these facilities.
Following the TMI-2 accident, Japan decided to build a large scale (1:50) integral test facility to investigate small break LOCA's in PRRs.
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 thermal 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 Internatic5al 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 to:
o Develop a common understanding of the ability of the code to appropriately represent important physical phenomena; o
Share user experience on code assessment and present a well documented assessment data base; o
Share experience on code errors and inadequacies and to cooperate in removing the deficiencies to maintain a single, internationally recognized code version; and o
Establish and improve user guidelines for applying the code.
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The ICAP program is successfully accomplishing these goals.
Participating organizations 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.
Cooperation 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-FLECHT 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.
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l TABLE 1 EXAMPLES OF DUAL EXPERIMENTAL /
ANALYTICAL APPROACH TO RESOLVE ISSUES lasue Experiments Anatyses it Margin of Conservatism in LOFT, Semiscale, UPTF, CCTF Analyses of Test Appendix K; Revision to SCTF, TLTA, SSTF, FIST, LOBI, Facility Data and Appendix K (LOCA)
Marviken, PKL, Creare, BCL Full-Scale LWRa Creare, Purdue, Semiscale, TRAC-PWR and RELAPS Pressurized Thermal Shock UPTF, HDR, Finland, and Analysis of Oconee H. B. Robinson and Calvert Cliffs Small Break LOCA in W PWRs Semiscale, LOFT, ROSA-IV, LOBI TRAC-PWR and RELAPS Small-Break LOCA and Natural MIST, OTIS Circulation in B&W Reactors Analysis of Data
^~
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Small-Break LOCA in BWRs FIST, TLTA D a Feed-and-Bleed Procedures Semiscale S-SR-1, 2, and for Decay Heat Removal A A Analyses S-PL-3. LOFT LP-FW-1.
in PWRs Performance of Upper Head Semiscale S-UT series and Upper Plenum injection
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2D/3D, CCTF, SCTF in W Reactors Small-Break LOCA with less of ROSA-IV, LOFT, LP-SB-3, RELAP5, TRAC-PWR, High-Pressure injection Semiscale S-NH series SASA Analyses u
up in team Semiscate S-UT-6, S-UT-8, RELAP5, TRAC-PWR,
" "~
e[,e a 0' S-LH-1, S-LH-2, ROSA-IV NOTRUMP 0 s t
Steam Generator Tube llOSA-IV, Semiscale TRAC-PWR, RELAPS Rupture (SGTH)
Anticipated Transients LOFT L9-3, L9-4, RELAP5, RAMONA-3B, em scale SR-7, RST M AC-B M Rs nd BWRs lodine Behavior Following MB-2, ORNL, Northwestern CITADEL, TRAC-PWR, SGTR University RELAPS Y. MST, Natural Circulation TRAC-PWR and RELAPS g
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Vessel internals after LOCA BWR Containment Pressure MIT, GE, Livermore PELE, SOLA Suppression Pool Loads Marviken Stability Margins for BWRs DRESDEN, FRIGG NUFREO TMI-2 Accident
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75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 flSC/1 EAR
- Dollar Amounts not Adjusted for Inflation i
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ATTACHMENT'3 o Mao UNITED STATES o
!", q' g NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
/.. p W ASHINGTON, D. C. 20555
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June 15, 1989 7
The Honorable Lando W. ' Zech, Jr.
Chairman U.S. Nuclear Regulatory Comission Washington, D.C.
20555
Dear Chairman Zech:
SUBJECT:
NRC THERMAL-HYDRAULIC RESEARCH PROGRAM During the 350th meeting of the Advisory Committee on Reactor Safeguards, June 8-10, 1989, we reviewed the NRC's plan for continuing thermal-hydraulic.
research as related to the design and operation of nuclear power plants.
This matter was also considered by our Subcommittee on Thermal Hydraulic Phenomena at a ' meeting on May 23, 1989.
During these meetings, we 'had the benefit of presentations by representatives of the Office of _ Nuclear Regula.
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, 1988.
Thermal-hydraulic research has always been a central and major part of the NRC's research program. Much'of the work was inspired by the perceived need to better understand hypothetical large-break loss-of-coolant-accidents (LB-LOCAs) and the performance of emergency core cooling systems (ECCS).
Experiments and analytical models, such as the RELAP and TRAC codes, have confirmed compliance with the ECCS rule.
Continuing research on LB-LOCAs' culminated with a 1988 revision to the ECCS rule which permits licensees to use more accurate means of analysis and makes possible certain safety and operational 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.
In addition, the experimental information base and the codes have been found useful in assessing and predicting the consequences of transients and small-break loss-of-coolant-accidents (SB-LOCAs) which are now recognized to be much more risk significant than the LB-LOCAs. The codes are also being used L
to analyze the early stages of severe accident scenarios.
Proposed Research Prooram We understand the continuing NRC program in thermal-hydraulic research to have two principal purposes:
Bring development of the major computer codes to a successful comple-f
- tion,
e The Honorable Lando W. Zech, Jr. June 15, 1989 0 Maintain, within the NRC and its contractors, a capability for thermal-hydraulic analysis sufficient to deal with safety and regulatory con-cerns that might arise in the future.
This includes the continuing availability of a cadre of experts.
RES representatives indicated these general purposes would be re511 zed through achievement of several specific objectives:
0 The major codes will be maintained indefinitely and some further devel-opment will be carried out. The scope and depth of further development seems not to have been decided. Apparently, it will include appropriate reactions to new data from foreign experimental programs and assessments which are expected to continue for some time.
It may also include a review and redevelopment of the important constitutive equations in the codes.
The current experimental programs related to specifics of the Babcock and Wilcox (B&W) nuclear steam supply (NSS) system will be completed.
Beyond this, any further experimental programs will be carried out at universi ties, rather than by the creation or operation of any major facilities at national laboratories.
Relatively inexpensive " integral" facilities, of scope similar to the facility nuw operating at the University of Maryland, are being considered as contrasted with what have been called " separate effects" facilities.
These would be mockups of specific NSS systems and of an advanced LWR (600 MWe size) design.
O An expanded program of applications research is planned.
Apparently, much of this activity is expected to be in response to issues that arise from experiences with operating plants.
But, it will include prepara-tion of input data for several more plant types than are now aveilable to the NRC.
This will permit more rapid analysis than would otherwise be possible in response to future safety or regulatory issues.
This program may also include exploratory, in-depth studies of a range of possible transients for a variety of plants.
In addition, two other specific program elements were centioned:
0 A further demonstration of the " Code Scaling, Applicability, and Uncer-tainty" methodology will be carried out for an SB-LOCA with RELAP5/ MOD 2, 1
similar to that recently completed for an LB-LOCA.
0 Improvements will be made to the NSS system process models now incor-porated in training simulatcrs at the NRC Technical Training Center.
(
This will permit more accurate simulation of off-nonnal scer.arios for the study of emergency and accident management procedures.
I s
i l
V
- The Honorable Lando W. Zech, Jr. June 15, 1989 Before commenting on these research proposals, it is pertinent to consider l
two statements made by the NRC staff at the May 23, 1989 Thennal Hydraulic Phenomena Subcommittee meeting, because the ideas expressed have an influence on our recommendations:
A representative of the Office of Nuclear Reactor Regulation said, "NRR is not relying extensively on the codes to address current licensing issues."
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
[ plant] performance or [ accident] consequences."
ACRS Recommendations 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 there is too much emphasis on further development of the existing codes in the planned program.
Maintenance of the needed NRC capability is more a matter of ensuring the availability of a cadre of experienced and expert analysts and access to the general body of experimental data, than it is of improving or even ensuring the availability of large systems codes.
The Committee reiterates its comments in the report of June 7, 1988, that " marginal improvements that 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 cooe accuracy and validation."
To accomplish this general purpose, we recommend a program of four primary elements:
(1) Code Development Maintain the present large system codes, TRAC-Pf1/ MODI, RELAPS/ MOD 2, TRAC-BWR, and RAMONA-3B, for an indefinite period.
Limit improvements only to those required by:
(a) the discovery of important errors or (b) crucial new information from the fareign 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 or numerical algorithms in these codes.
We are not convinced by the arguments given for the need to develop TRAC-PF1/M002 and RELAP5/ MOD 3.
It is our view that the proposed modifications will not substantially improve the codes.
a+
+
The Honorable Lando W. Zech, Jr. June 15, 1989 Instead, consideration should be given to the development of a new type of systems code that will be more usefu' for analysis of extended plant transients involving interactions of piant systems. The Comittee also mace this recommendation in its June 7,1988 report.
TRAC and RELAP were originally designed to analyze the LB-LOCA, a rapid and severe reactor transient, in great detail. There is a need for a more empir-ical and efficient analytical 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-haven National Laboratory is an example of the type of code we are suggesting.
(2) Experimentation The staff proposal to develop relatively inexpensive " integral" test facilities at universities is sound. We see this as consistent with our recommendation for a new type of systems code. We agree that it would be inappropriate to build several such facilities at one time.
A gradual approach is warranted. The first such new facility might be one that would incorporate features of the advanced LWR designs.
Also, it will be better to completely assess the benefit that has been obtained from tests with the University of Maryland facility mentioned above.
In addition, a small program to deal with more fundamental research should be maintained. These are experiments of the sort that have been 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 to the investigation of:
fluid-elastic instability related to vibration of tubes in U-tube steam generators, departure from nucleate boiling with oscillating flow and power in
- BWRs, O dynamic instabilities and loads on valves.
(3) Data Analysis A major effort is needed to organize data from test programs into a useful form other than the large systems codes.
In particular, with the 2D/3D, ROSA-IV, and the B&W test programs all coming to closure, mea-sures are needed to ensure that these expensive and valuable bodies of data are preservec and used.
In addition, olcer data from, for example, 1
f
The Honorable Lando W. Zech, Jr. June 15, 1989 the FIST and FLECHT programs can be of greater value if they are effectively organized into more useful forms.
(4) Applications Research A program in this area should include three elements:
Analysis of transients indicated to be of interest as a result of-plant operating experience.
O Preparation of input data decks for several classes of plants so that turnaround time for analyses in response to experience is shortened.
O 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 for considera -
tion:
- feed and bleed scenarios
- secondary depressurization scenarios Finally, we suggest that RES broaden its perspective as to what other re-search in the themal 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, 4
l David A. Ward l
Acting Chairman
References:
1.
U.S. Nuclear Regulatory Commission, draft SECY Paper: " Status and Plans for Thermal Hydraulic Research Conducted by the Office of Nuclear P.egulatory Research," provided to the ACRS in May 1989 2.
U.S. Nuclear Regulatory Commission, NUREG-1252:
" Nuclear Power Plant Thermal-Hydraulic Performance Research Program Plan," Office of Nuclear Regulatory Research, July 1988
f(
k UNITED STAT ES y
9, NUCLEAR REGULATORY COMMISSION l
j WASHlNGTON, D. C. 20555 a
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JUL 2 41989 l'
Mr. David A. Ward, Acting Chairman Advisory Committee On Reactor Safeguards U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Ward:
SUBJECT:
NRC THERMAL-HYDRAULIC RESEARCH PROGRAM The Advisory Committee on Reactor Safeguards (ACRS) letter of June 15, 1989, provided comments on the subject program.
The ACRS agreed with the general objecti+2 to maintain expertise in thermal-hydraulics to meet future agency needs in this field.
The Committee also agreed with the general level of funding projected for the next several years.
Parallel to the staff's plans, a program comprised of four elements was recommended.
staff review provides the following responses:
(1)
Code Development The ACRS may have misinterpreted a staff statement to mean that older versions of computer codes were already nature.
This would have led the ACRS to recommend that no further work is needed on the final code versions, TRAC-PF1/ MOD 2 and RELAP5/ MOD 3.
The staff believes that the final versions of TRAC-PWR (TRAC-PF1/ MOD 2) and RELAP (RELAP5/ MOD 3), as completed in 1989, have an acceptable level of accuracy and that further development is not likely to produce major changes in our understanding of plant performance and accident consequences.
Previous versions of these two codes (TRAC-PF1/ MOD 1 and RELAP5/ MOD 2) were released in December 1984.
Since then, peer review and code applications studies identified a number of modeling 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 Program (ICAP) members for assessment to be completed in December 1991, when the ICAP program ends.
We believe that these final code versions, plus the other NRC codes already completed (RAMONA-3B, HIPA, TRAC-BWR, and COBRA / TRAC) provide NRC with sufficient analytical capability to meet its future needs,
N
.o
+
Mr. David A. Ward 2
including any needed analyses of extended plant transients involving interaction of plant systems.
This has been confirmed at several meetings among the RES staff and the user offices of NRR and AEOD that were convened for this purpose.
The staff thus concludes that no new systems code development is required at this time.
(2)
Experimentation We note that the ACRS endorses our plan to develop university experimental facilities.
We intend to clearly define what we expect from any new test facilities at universities before any proposals to perform this research are issued.
Concerning the suggestion to maintain a small program to deal with nore fundamental research, we intend to include this in a Broad Agency Announcement for proposals similar to the ones you have suggested.
In addition, we will continue our grants program which has supported such research efforts for several years.
(3)
Data Analysis You have raised a concern about how best to 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 publish research synthesis reports on special topics.
However, these existing programs may not fully address your concern.
The staff will review our current capabilities and identify any ways we can improve on putting completed research into the most useable form.
We would 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, and would add a fourth, code applicability studies.
Such studies better define the classes of reactor geometries and accident scenarios for which specific computer codes are presently applicable.
This would help to assure that code performance is well understood.
4
.w' Mr. David A. Ward 3
With regard to the suggestion of including other research in the thermal sciences, we propose to discuss this further at a subcommittee meeting in order to understand what the ACRS has in mind.
Sincerely,
/
1 i
or, el
, Jr.
ecutive rector for Operations cc:
Chairman Carr Commissioner Roberts Commissioner Rogers Commissioner Curtiss OGC SECY a
__.___________._________m_m
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UNITED STATES -
' y' NUCLEAR REGULATORY COMMISSION n
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- ,i l WASHINGTON, D.C. 20555 June 26, 1989 OFFICE oF THE
. COMMISSION E R MEMORANDUM FOR:
Victor.Stello, Jr.
Executive Director for Operations FROM:
Kenneth C. Rogers
SUBJECT:
ACRS LETTER OF JUNE 15, 1989 ON NRC THERMAL-HYDRAULIC RESEARCH PROGRAM-I consider the' subject ACRS letter excellent and lucid advice on the Office of Nuclear Regulatory Research (RES)' thermal-hydrau)ics research program.
I support the ACRS recommenda-tions in the letter'and the four primary. elements of this program.
Please inform me of any RES comments on the ACRS recommendations or their plans for implementation.
Kenneth C. Rogers Commissioner cc:
Chairman Zech Commissioner Roberts Commissioner C&rr Commissioner Curtiss OGC SECY A6 X ik Dete _
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