ML20071M621

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Provides Comments on Zuber Rept to Catton on ROSA-IV/AP600 Meeting on 920603-04 & 23
ML20071M621
Person / Time
Site: 05200003
Issue date: 07/21/1992
From: Modro S
EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To: Shotkin L
NRC
Shared Package
ML20024G666 List: ... further results
References
SMM-38-92, NUDOCS 9408040306
Download: ML20071M621 (7)
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/NEA idaho National Engineering laboratory Q 00 July 21,1992 Dr. L. M. Shotkin U. S. Nuclear Regulatory Commission 5640 Nicholson Lane MS NL/N-353 Washington, DC 20555 COMMENTS ON DR. ZUBER REPORT TO DR. CATTON ON ROSA-IV/AP600 MEETING, JUNE 3-4, FROM JUNE 23, 1992 - SMM-38-92

Dear Dr. Shotkin:

I have read Dr. Zuber's report to Dr. Catton from June 23, 1992. Upon l reading his comments, I have the impression he discusses application of the ROSA-IV test facility for demonstration type tests or AP600 simulation rather than for generation of data for code assessment. His argumentation .

of significance of distortions focuses on the capabilities of ROSA-IV to I duplicate the AP600 behavior. Irregardless of ROSA-IV's shortcomings, duplication of reference system behavior in scaled facilities under two-phase transient flow conditions is a priori impossible. On page 13 of his memorandum, Dr. Zuber observed critically that ROSA-IV will generate conservative as well as non-conservative results. I believe test results must be understood, but the aspect of conservatism is irrelevant for gathering test data for code assessment.

The work we have performed so far is aimed at evaluation of ROSA-IV as a facility that should provide data to assess capabilities of computer codes to model the overall system behavior during postulated transients. Our analyses showed that ROSA will exhibit most of the AP600 processes. The sequence and relative magnitude of events will be reproduced by ROSA-IV for most transients. However, ROSA will not be a AP600 simulator or a demonstration facility. Based upon our analyses, we believe that ROSA-IV can provide very useful data for assessment of ccm.nuter codes to simulate AP600 integral system behavior during high pressure and depressurization phases.

On page 11 of his memorandum, Dr. Zuber discusses the issue of asymmetrical behavior of AP600 and respective code capabilities:

...RELAP5, being a one-dimensional code, cannot properly model the effects of flow asymmetries." Because the ROSA shortcomings in simulation of the asymmetries was considered as the most important, I would like to discuss this aspect in more detail.

9408040306 940629 PDR COMMS NRCC CORRESPONDENCE PDR jfEGcG,..Inc. P.O. Box 1625 Idaho Falls, ID 83415

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Dr. L. M. Shotkin  !

July 21, 1992  ;

SMM-38-92 i Page 2 Potential asymmetry in AP600 response is associated with flows in such components as the cold legs, pressure balancing lines, direct vessel l injection (DVI) lines and downcomer. The asymmetry is a result of i interactions between these components for a set of transients. Most of ,

4 the DBA transient will be of a symmetrical nature in AP600 with respect to l safety systems response. Only transients with breaks in DVI lines or pressure balancing lines will exhibit asymmetrical behavior not typical for current generation reactors.

Flows in all the components of interest, except in the downcomer, can be treated one-dimensionally as it is practiced for present generation reactors (all current system codes such as TRAC or RELAP treat piping flow as one-dimensional). Because of potential multi-dimensional effects in the downcomer we have nodalized it applying interconnected mesh of cells.

However, to increase the fidelity of simulation, we suggested a rigorous two-dimensional modeling of the downcomer. Furthermore, we are currently analyzing the downcomer behavior of both systems with a 3-D CFD code I (FIDAP), to obtain an independent evaluation of ROSA /AP600 comparison.

1 Current analyses indicate that ROSA-IV can exhibit the local phenomena '

that control AP600 response. With the currently proposed configuration .

. ROSA-IV will be able to provide data on symmetrical system response and on l asymetrical response. The asymmetrical transients will not duplicate AP600 system behavior but will provide, in an integral system environment, interactions and phenomena that govern the asymmetric AP600 behavior.

Codes validated using these data should be able to simulate AP600 symmetric and asymmetric system response.

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i S. M. Modro l

NRC Thermal Hydraulic Analysis Programs dap cc: D. Bessette, US/NRC W. H. Rettig, DOE Field Office, Idaho, MS 1134 J. C. Okeson, EG&G Idaho, MS 3600

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APPENDIX II I

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, )e( : UNITED STATES  !

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"** miwf( f Nuclear Safety Research Review Committee

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.' : 1cly 1992

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i Mr. Eric S. Beckjord Director Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission l Washington,DC 20555

Dear Mr. Beckjord:

Enclosed please find a copy of a letter report of NSRRCs ALWR Subcommittee on AP600 thermal hydraulic t: sting. This letter report has been received and reviewed l

l by the NSRRC and is accepted as a statement of the Committee's cunent position on AP600 thermal hydraulic testing.

l If you have any questions on this NSRRC report, please contact Dr. Neil Todreas i

or me.

l Sincerely,

. / n.

David L. Morrison l

Chairman Nuclear Safety Research Review Committee l

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e Nuclear Safety Research Review Committee S, # Vvashington. O.C. 2o555 July 20,1992 Dr. David Morrison

! The MITRE Corporanon 7525 Cc! shire Drive. MC W766 M: Lean, VA 12102 I

Dear Dr. Momson.

The NSRRC Advancec Reactors Subcommittee (Messrs T Boulette. S. Burstem.

H. Ishin and N. Todreas (Chairman) in attendance) met on July 1 and 2.1992, and reviewed Office of Nuclear Regulatory Research (RES) programs pertammg prmespally to the AP 600 l program. Among these programs the RES prt>posaj to conduct integral systems tests at the l ROSA facility of the Japanese Atomic Energy Research Insutute (JA2RI) was examined m i detail. Because of the timeliness of this NSRRC review regartiing the forthcoming  ;

Commission decision whether to prcceed with this program, this letter has been prepared to ,

I set forth the relevant conclusions of our review. A supplementary report of the full scope of the July 1 and 2 meettng will follow which will contam detailed obsenanons and suggesuons '

relevant to the RES programs examined.

THE RES PROPOSAL The RES proposal examined was to conduct USNRC sponsored confirmatory integral systems tests on AP-600 using a full. pressure, full. height factilty. The purpose of these tests is to develop a sufficient data base with which to enhance the assessment of an analytical tool

! that could then be used with confidence to assess full size plant responses to initiating accident sequences. The selection of a facility is constrained by the: (a) desire to obtain test results prior to the currently scheduled preparation (Summer,1994) of the Draft Safety Evalus: ion Report (DSER) and the issuance (November,1994) of the Final Design Assessment (FDA),

and (b) need to obtain these results within the currently anticipated budget for this work of approximately $10 million. The selected facility is ROSA modified as pro psed by RES and agreed to by JAERI to a configurat;on (ROSA V) representing a 1/30 by vo: ume scaled model of AP 600 with the mm.ior rnodel deviation being the use of a single versus the actual two cold legs per loop.

l The NSRRC Subcommittee examined this proposal by posing and resolving a series of -

questions, starting with the need for this testing and culminating in the examination of the efficacy and adequacy of the proposed solution. These questions, restated specifically for AP 600 integral systems testing, will be sequentially reviewed next by summamias the NRC position and then stating the Subcommittee's conclusion.

"What are the NRC's needs for confirmatory systems research r eP.600"?

Integral systems tests in a full pressurs; full height facility areerewi beeense the response of the iystems to mitiating' events canno', be. analytically .with ggf 94 conddenos:by the ass of existing anatydcaLapons (computerW His is to bedrtbs-q of interacticos.betwoon systems.and cc= rend tbs low driving.beeds-

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( nec to quality an anatvit:al tool to assess plant resconse Tnis approsca is taken because r:o

aied fac:hty can se:Ye as a cemonstration of full s:ze piant response to imtiating events.

I system behavior uncer three acc; cent sequences is of pamcular interest because the passive l Mety systems are called upon to operate at hign pressure.

, Small break loss-of-coolant c:ident. l Steam generator tube rupture, and

  • Steam hne break.

i Independent NRC testing at low pressure is not considered essential since the planned l vendor test program is deemed to y1cid sufficient data. However,it is anticipated that the tugh i

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pressure ROSA fac1hty to be used by the NRC as discussed later can be run to yield  !

t supplementary lower pressure cata.

I 45RRC Subcommittee concurs with the NRC's need for indepencent confi. . story l sys  ; search to insure that its analytic tools are qualified to assess full plant response.

! The availability of integral systems low pressure cata to insure performance of the gnvity I dram / core cooling system behavior is recognized as equally important as high pTessure test results.

"What integral systems testing program has the vendor proposed for A P-600" ?

l high pressure test program will be conducted in the full pressure, full height,1S95 by vt- e, scaled SPES 2 facility in Italy. A low pressure (400 psi maximuro) test program will be conducted in the 1/200 by volume scaled Oregon State University (OSU) facility. The extension of SPES tests below 400 psi so as to initialize OSU tests is being explored.

De NSRRC Subcommittee took note of this planned vendor test program.

l "Why should the NRC conduct confirmatory integral systems tests on AP 600 using a full pressure, full-height facility when the vendor will conduct a similar test program"?

! The NRC stated that they had a need to extend the expected vendor test matrix beyond the design ba. sis to develop confidence that the design basis is a satisfactory limit. His wuld be l achieved by experiments at or slightly beyond design basis conditions to ensure that no l unanticipated phenomena or major effects occurred in this operating band, and thereby l confirm the adequacy of the design basis limit.

The NSRRC Subcommittee concurs with the NRC need to develop confidence in the

! design basis in this manner. However, it is emphasized that we do believe that vendor l demonstration of satisfactory plant performance within the design basis should clearly remam

! the required standard for design approval.

l l "Why did the NRC select ROSA as the test facility rather than use the Italian

! SPES. faelifty in which the vendor will conduct tests or construct a newv dotnestic fae,llty"7 ,

The NRC couki have chosee so cocmetseparately with tlw SPES cv.ie= fecondoct.

of warnoan !@t w7g.-. 4 imo n==e r-gy prescribed-NRC test matrix; thereby-av

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;n caon m T-wc_ spa Esc 1e e :=ce =.02 Morrison,7/20/92 V Page 3 Vendor access to the facility will take precedence over NRC access. Delay in conducting the vendor program or extension of the vendor program is possible and would severely upset the NRC schedule for acquisition of NRC independendy-produced test data, ne value of the test results will be muimiwd if they can be used in the assessment of codes requued for NRC safety analyses.

The vendor has not presented analysis to the NRC to firmly establish that the data from SPES is valid by itself to qualify an analytic tool for use on a full scale plant.

Scale effects probably need to be assessed and confirmed, as they have in past NRC l

thermal / hydraulic test programs, by tests at different scales.

The NSRRC Subcommittee concurs that plans to conduct NRC tests in SPES would not be prudent because of the cited schedule and test scale concems.

"What is the NRC doing to ensure that the ROSA facility will be configured correctly and will simulate the performance of the AP 600 passive safety features with acceptable fidelity"?

The NRC has performed an extensive cmuj-Etive assessment, using the RELAP 5.

MOD 2.5 analytic tool, of the behavior of the ROSA facility and the AP-600 plant to the same set of initiating events. From thsee analyses, desired L.y....- sis in the ability of ROSA to simulate the phenamena appearingin the plant weeidentified. Costs for these Levi w specifically changes in the facdhy config-iaa. were estimated and suh===tly negotiated with the ROSA owner. The final negotiated configuradon has been analyzed and is --;- Ed to satisfactorily spresent all full plant phenomena including many, but not all, aspects of asyns esl loop behavior. The cost for ROSA modifications and the schedule for their

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implementation and the conduct of the test program meet NRC criteria. Further, the NRC

' stated to us that no domestic facility could come close to n==nns the NRC cost and schedule criteria in that a cost of $40 50 million and a tiene of appr =ie='dy three (3) years would be required to construct a domestic facility meeting or i..y.s.hg on the ROSA V facility criteria.

ne NSRRC Subcommines reviewed the technical basis for the proposed modifications to ROSA and its consequent suitability as the NRCs selected high presses test facility. The Subcommittee concludes that the following key facers need to be balanced in reaching a decision:

. The i. w of obtaining '=4, - %t NRC data to confirm the adequacy of the ,

design limit. l i

. The of obtaining these data in a timely meaner to allow their use in i assassing usedin safety analysgs. .

. Ihs need to avoid tha possibility of'=+=aMa r 204iy into the assessmaat < zooess from experimental data taken on a test facil.ty which may not represent ful plant phenmennin allaspects.

After reviewing the data pressoned1 and weighing these facers, the Subcomsmitess concurs with the RES swamraandanan to M with the ROSA V program for integral systems testing of the ANiOO plant design. Tids e.24ri needs no be part of a wellinesgrated 1

1 Dming the of this repost, the SatKh reesived and soviewed the commeno e Acas consehenes sensaming the SPES sad 105A hassel test $scilkka-

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rogram involving careful coce enhancement ans assessment. anc possloly well-selec:cd separate-effects tests for phenomena that cannot be tully explorce in t .ese Intepal facdi:ll J Such a program is r.eeced since the purpose of the integral testing is not a demonst acon:

AP-600 performance, but rather it is to gatner data for code assessment. Rese aspects v ciscussed more fully in our supplemental report 1

Sincerely,f l'

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dlsi Neil E. Todreas  :

l Chairman NSRRC Advancec Reacters Subcommir.ee .

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6 APPENDIX 111 1

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Appendix III

Subject:

Reply to Memorandum by V. K. Dhir on the June 3-4 Meeting in Idaho Falls The following responds point-by-point to comments provided by Dr. Dhir on the subject meeting. It is organized according to Dr. Dhir's memo.

Meeting Summary No comment. l l

Observations j

1. The ROSA tests are planned to cover the full pressure range from full system pressure to IRWST injection. They will, thus, overlap both the SPES high pressure and the OSU low pressure testing. We have stated we are interested in both high pressure and low pressure confirmatory testing. We explained to the ACRS that at the time when the staff was proposing to do separate low pressure l confirmatory testing, the facility we envisaged was identical to l OSU. At the time we were precluded from interacting with OSU due to conflict of interest considerations. Once these were resolved, we could no longer justify pursuing a duplicate facility.

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( In contrast, based on our interest in high pressure systems interactions phenomena and processes, by June,1991 we identified ROSA as the best candidate for performing high pressure confirmatory testing. We have been working since that time towards formulating ,

a technically sound program to modify the facility and conduct  !

testing.

l i In terms of experimental programs in scaled facilities intended to l model full scale power reactors, it is a well-established principal l that facilities of difftirent scales and scaling approaches should be used to ensure that the effects of scale are well-understood. By definition, scaling introduces distortions in all scaled facilities.

Testing programs must be - formulated accordingly. The integral system test program carried out to study small break LOCAs in Babcock and Wilcox reactors was an example of such a program.

Research Information Letter 164, describing the results of this program, is attached. We would like to refer Dr. Dhir to Commissioner Rogers' memorandum (attached) on thir, program. In addition, to quote from MIT Professor Peter Griffith's review of the ROSA program:

"This is a well structured program which clearly benefitted from our experience with the LOCA work done on LWRs during the 70's and 80's. Because the experimental program consists of l three integral tests being run on three quite different rigs l i l

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2 APPENDIX III each scaled according to three different rationales, I can't imagine that there will be many questions outstanding about the system performance when these experiments are completed.

We have, in addition, got an operating, documented computer

! code, RELAP-5, which can be used for the prediction and analysis of the ROSA-IV experiments. The pieces of the program will come together in a timely way so that the results of this program can be used to design out-any problems which might arise in the course of this research."

2. The costs and schedule would be effectively prohibitive. We have already given serious consideration to a domestic facility. We have stated that, ideally, this is oui preference, however, such a l facility could not be built within the FDA schedule, and would be considerably more expensive than the ROSA program.

In its letter of March 10, 1992, the ACRS stated " Inasmuch as FHFP integral system testing will require at least three to four years to complete, there is a risk that the present certification schedule will be affected unless the test program is begun now. We believe the likelihood of such an impact is great. If the present l certification schedule is to be adhered to, we recommend that a FHFP j testing program be initiated now." RES agrees with the ACRS' l

assessment.

3. The use of RELAP to perform comparative calculations of ROSA and AP600 is not circular and certainly is logical. We completed a i RELAP code applicability review in early 1991. RELAP models and correlations were reviewed from the perspective of new AP600

, phenomena and features that could be important to reactor safety.

l The purpose was to identify those areas in which new mathematical f

models of physical phenomena would be required to be added to RELAP5. In most cases, the AP600 design and its systems and the planned and off-normal operations were found to be similar enough to current PWRs that RELAP safety analysis applicability was unchanged.

There were basically no new phenomena involved in the AP600, however the physical parameter ranges and applications of the phenomena may be different than those in present generation reactors. Therefore, a validation program was laid out accordingly.

Review of ROSA with respect to our separate scaling study for a FHFP facility showed that ROSA meets all requirements. If Dr. Dhir is aware of a better method to demonstrate facility similitude and scaling adequacy, we should like to know of it.

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l 3 APPENDIX III

4. Distortions will be present in ROSA, as they will be in each and every scaled facility. The true question is whether the major phenomena and processes are preserved. Our analyses have shown that they are.

In addition, JAERI has agreed that the facility will remain available beyond the initial set of 10 experiments.

5. The experiments must meet the objectives and specifications.

l determined prior to the experiments. Otherwise, they will be repeated.

l 6. RES has assessed the cost-benefit ratio for the ROSA program, along l with schedule requirements of the FDA, and found that this facility meets all requirements identified by NRC staff and contractors for confirmatory integral system testing.

7. We never planned to stop the experiments at the actuation of 3rd L stage of ADS. Rather, we plan to run the ROSA experiments to full i

depressurization, including the initiation of IRWST injection.

j 8. Aside from the problem of schedule, we could not achieve the same

! data base for the same cost with a new U.S. facility.

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9. In an ideal world, we wodid also prefer a dedicated U.S. facility, j however, this is not possible within the constraints of schedule and budget. Pragmatic alternatives must be sought that will successfully meet the same needs within the given constraints.

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  • UMTED STATES t i NUCLE AR REGULATORY COMMISSION
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"E*0RANDUM FOR: Thomas E. Murley. Director  !

! Office of Nuclear Reactor Regulation i

CM: Eric S. Beckjord. Director Office of Nuclear Regulatory Research

SUBJECT:

RESEARCH INFORMATION LETTER 164, " THERMAL HYDRAULIC DATA BASE RELEVANT TO PLANTS OF THTBABC0CK AND WILCOX LOWERED-LOOP DESIGN" eferences: 1. ALAB-708, 16 NRC1770 December 29. 1982.

2.

Clarification of TM1 Action Plan Requirements. NUREG-  !

0737, November 1980.

3. Letter from H. R. Denton to R. B. Minogue. " Request for the Conceptual Design of a facility for the Study of B&W and CE Integral System Characteristics," December 30, 1981.

4 J. Gloudemans and D. P. Birmingham, " MIST Program:

Summary of Key Results," NUREG/CP-0097, Vol. 4, March 1989.

! 5. K. Almenas, et al., " Scaling of Integral Facilities l

at Reduced Pressures " MDNE/061589, June 15, 1989.

6. K. Almenas, et al., " Evaluation of four MIST

' Atypicalities," MDNE/041089, April 10, 1989.

7. Letter from H. R. Denton to R. B. Minogue, " Request for Follow-on Program in the B&W Integral System Test l Facility (MIST)," October 31, 1984 This memorandum transmits results from research conducted in the Integral System Test (IST) program. IST includes the Multi-loop Integral System Test (MIST) facility at the Alliance Research Center in Alliance Ohio, and the University of haryland at College Park (UMCP) 2x4 Loop facility. This l research provided thermal-hydraulic experimental data relevant to plants of i

the Babcock and Wilcox (B&W) lowered-loop design. MIST was jointly sponsored by the U.S. Nuclear Regulatory Commission, the Electric Power Research Institute (EPRI), B&W Owners Group (B&WOG) and B&W. The UMCP 2x4 Loop, a reduced-pressure and small-scale facility, was designed to address scaling atypicalities of the MIST facility and to provide data for code assessment.

I Contacts: ,

R. Y. Lee, RES/DSR, 49-23560 H. Scott, RES/DSR, 49-23563

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=culatorv !ssue:

r ailowing the Three Mile islano Unit 2 (TMI-2) accident, a numoer of reculatory issues concerning the design of the B&W reactors -ere raised. The effectiveness of feed ana bleed ano the boiler condenser mooe (SCM) of natural

rculation was cnallengea during the THI-l Restart Hearing conoucted by the
mic Safety ana Li:ensing Appeal Board [1]. In BCM, heat is removed from u.e or1 mary system througn vapor concensation in the steam generator and the accompanying primary-to-secondary heat transfer. Clarification of THI Action Plan Reouirements (NUREG-0737) Item II.K.3.30 (2) required that small-break i loss-of-coolant accident (LOCA) calculational models be compared to applicable data.

In response to NRR's request for integral system characteristics for B&W reactors (3), the NRC and industry formed a Test Advisory Group to make recommendations regarding the type of data base required to validate small-treak LOCA models. The IST program was formed in 1983 to acoutre the desired

ata. The primary experimental facility in the IST program is the MIST facility.

lthough MIST is designed as a full-height and full-pressure integral experiment facility, it is still a scaled model of a B&W plant. Thus, it

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entails various design compromises such as an atypical downtomer. These design compromises are a potential source of distortion of some of the physical phenomena (e.g., variation of flow regimes in the hot legs) which in turn could lead to atypical transient behavior (e.g., premature interruption of natural circulation). The UMCP 2x4 Loop, a reduced-height and reduced-pressure, integral experiment facility employed an alternate design approach (e.g., a more typical downcomer) to assess the impact, of some of the MIST design compromises on transient behavior.

Conclusion:

IST produced an integral experiment data base for natural circulation, small-break LOCA, feed and bleed, steam generator tube rupture, effects of non-condensible gases, and pump operations on small-break LOCA behavior [4, 5].

Key observations (1 to 6 for MIST, I and 7 for UMCP) are summarized below.

, (1) Natural circulation was studied under varying degrees of loss of primary l inventory. A key question about natural circulation was whether the BCM l would remove decay heat effectively and depressurize the reactor coolant j system. This mode of heat transfer was consistently observed.

(2) During small-break LOCA, heat removal from the primary system was

! further augmented by the staan yenting from the upper plenum to the i downconer through the reactor vessel vent valves (RVVV). Adequate heat j

removal was. observed in MIST tests for a wide range of pripry boundary conditions (i.e., variation of break sizes from 5 to 50 cm , variation of break locations, and both full- and half-capacity HPI flows). In all cases tested, the MIST system depressurized and attained primary circuit mass equilibrium without uncovering the core.

(3) MIST results show that the feed and bleed technique can be utilized to cool the core and 4tpressurize the primary system.

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4) for multiple simulated steam generator tube ruptures. the primary system  ;

cepressurized rapidly due to tube rupture discharge. SCM like activity was observed between primary system break flow to the seconaary sloe of the ruptured steam generator. For a smaller numoer of tube ruptures.

  • he primary system depressur12ed by single-loop cooldown of the intact  ;

000. )

5) The presence of non-condensible gases reduced BCM cooling, but did not )

prevent primary system cooldown and depressurization.

! (6) Reactor coolant pump operation is a m tageous during small-break LOCA. ]

With forced flow, primary-to-secondar. neat transfer is maintained '

longer and more energy is removed from the break. Therefore, the primary system pressure decreases more rapidly and the primary system refills more quickly.

) The comparisons of the experimental result:; from the two facilities l indicated that the UMCP 2x4 Loop is able to simulate the thermal- l l

hyoraulic behavior observed in MIST. First. it reproduced the l cualitative aspects of the flow modes. That is. it exnibited similar local flow regimes, flow regime transitions, the presence of both steady state and boiler-condenser natural circulation flows, and loop

. asymmetries [5). Second, it reproduced the sequence at which these flow modes occur during an inventory depletion transient. Inventory scaling was used to estimate the quantitative aspects of the flow modes (e.g.,

duration, magnitude of pressure changes). A precise parameter to parameter mapping between the UMCP and MIST data is not implied and is, in fact, precluded by the stochastic nature of some flow mode

transitions. However, key phenotaena of inventory transients can be simulated and the effect of pressure on the characteristics of these phenomena is understood. Despite the differences between the design of l the two facilities, similar thermal-hydraulic characteristics were l observed. The MIST design atypicalities do not affect the expected thermal-hydraulic behavior during a small-break LOCA [6).

Reaulatory Imolications:

i The MIST and the'UMCP 2x4 Loop experimental data provide a sufficient small-l break LOCA data base to satisfy the requirements of NUREG-0737. The integral system data is self-consistent, comprehensive, and suitable for benchmarking computer codes used to calculate B&W plant transients. Such benchmark calculations for MIST were performed and are in good agreement with experimental data. The data, as well as code calculations, show that various methods, such as feed and bleed and BCM, are effective modes of decay heat removal in a B&W plant during a small break LOCA.

Castricttons on OD olications: 0

'he scaling evaluation has shown that key thermal-hydraulic behavior (e.g.,

3CM) coserved in these facilities can be expected to occur in full scale B&W

lants of the lowered-loop design. However, these test facilities are scaled coels of a B&W lowered-loop nuclear steam supply system. As such, various

', ;caling atypicalities in simulating a plant were required. Hence, these cata snould not be applied directly to a full scale plant. Rather, validated computer coces (TRAC-PWR, RELAP5) should be used to calculate plant small-break LOCA. The requisite code validation was performed as part of the IST program.

Further work:

At the request of NRR [7], additional testing was performed in the MIST facility to obtain data for: small-break LOCA without high pressure injection: station blackout; and examining scaling questions. Analysis of these tests is expected to be completed by the end of 1989. Additional experiments are being performed at UMCP to further test the scaling concepts ancer more complicated boundary conditions, i.e., with HP! flow. Any new

ignificant results will be reported in a future RIL.

Eric S. Beckjord($0irector Office of Nuclear Regulatory Research

Enclosures:

(1) MIST Program: Summary of

  • Key Results (2) Evaluation of Four MIST Atypicalities e

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