ML20217H289

Regulatory Review of Passive Plant Design Certification Test Programs:Results & Lessons Learned, Presented at 970430 5th Intl Meeting on Nuclear Thermal Hydraulics, Operation & Safety in Beijing,China

ML20217H289
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Issue date:	04/30/1997
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(- 5th International Meeting on Nuclear Thermal Hydraulics, Operation & Safety Beijing, China, April 1997 REGULATORY REVIEW OF PASSIVE PLANT DESIGN CERTIFICATION TEST PROGRAMS:

RESULTS AND LESSONS LEARNED Alan E. levin U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Phone: (301)415-2890 Fax: (301)415-3577 E-mail: AE14NRC. GOV ABSTRACT References 1 and 2 discuss the testing reqmrements 1 cantainad in 10 CFR 52.47, describe the vendors' l As part of its review of advanced, passive light water testag programs, review progress in evaluating the reactor designs for certincahon, the U.S. Nuclear testing programs, and examane the NRC's regulatory R:;'", y Co==- (NRC) has alsnost completed perspective in those evaluations. This paper is written its review of test prograans developed by the reactor as the NRC's review of the testing programs nears an vendors to dema==trate passive safety system end. AP600 testing and Westinghouse's data analyses i J., + and to support d.r4 / and validation are complete, and all test reports have been submitted cf safety analysis e g : codes. Results of the test to the NRC. SBWR testing is complete and all test program reviews are discin==ad, focusang on the data reposts have been submitted to the NRC, as well.

and its analysis. 'Ihe NRC's ig'" g perspective and Although the SBWR certification application was important lessoas learned are also addressed. withdrawn, the NRC agreed to GE's request to complete and document a high-level review of the test INTRODUCTION program, to help establish the technology base for passive BWR designs. This paper su==arizes the Reactor vendors in the United States are currently NRC's exponence in reviewing the passive reactor test developing new reactor desages for future deployasset. programs, and discusses some of the lessons learned The U. S. Nuclear Regulmaary Comsmannon (NRC) is during more than 5 years of intensive review activities, i revieweg thers desagas for certificaanna as standardnad plants under Part 52 of Title 10 of the Code of Federal AP600 TEST PROGRAM R -;'% (10 CFR 52). 'Ihe advanced designs include " passive" light water reactors: Westinghouse's Westinghouse's AP600 test program is discussed in AP600 and General Electric's (GE's) Simplified Boiling several previous publications (see Refs. 3, 4); it is Water Reactor. In March 1996, GE withdrew its focused primarily on reactor and conaannmant passive

. " = for design certificahoe of the SBWR; the safety system thermal-hydraulic (T/H) performanca. j AP600 femannas under review for certificahon. Major separate-effects programs included an approximately 1/100-volume-scale Core Makeup Tank Since 1991, the NRC has been reviewing the test (CMT) test at Westinghouse's Waltz Mill site near programs developed by W"71 and GE to Pittsburgh; full-scale Automatic Depressurization support certificataca of the AP600 and SBWR. Part System (ADS) testing at the Central Research

$2.47 of 10 CFR (10 CFR 52.47) requires applicants Establishment, Casaccia, Italy; and a 3-tube for certincataan of reactors with passive safety systems representation of the Passive Residual Heat Removal to demonstrate that computer codes used for safety (PRHR) system at Westinghouse's Science and q aanlyses of plant designs are veladated over a range of Technology Center (STC), near Pittsburgh. Integral j condataoss -g" " to the specific plant. Testing is system testing was conducted in two test loops: the  ;

not expbcitly reqmrod to provide the data, but the lack 1/395-volume-scale, full-height, high-pressure SPES-2 j of data on the umque, passive safety features of the facility at SIET Laboratones in Piacenza, Italy; and the i AP600 and SRWR led the two vendors to establish test 1/192-volume-scale,1/4-beight, low-pressure (2.6 MPs l prograses to produce -ary data. Both vendors [400 psia] maximum) APEX facility at Oregon State 6" ; ' broad-scope experimental programs, with University (OSU). Scaled studies of con an====* heet s

both separate eflects and integral-eystems tests, to transfer related to Passive Contain= ant Cooling System support developeset and validahon of analytical codes (PCCS) performance were conducted at Westinghouse's used for safety smalyses of the plant designs. STC. j

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a Testing begna in the late 1980's on the PRHR test significant issues arose requiring resolution before the Progress, continuang until 1994, when the ADS, SPES-2 review could proceed GE's process for data submittal and OSU/ APEX tests were =--g' ^d. Extensive test to the NRC differed from Westinghouse's. Indaad of Program doc ====a=au= was submitted to the NRC for QLRs, GE provided brief Apparent Test Results (ATR) review prior to initistaca of testang, including facility reports. Integrated data and analysis reports were them designs, scalmg studies, anstr====ama==, and test submitted, constituting both data record and data mainces. Test data were initially subausted is " quick- analysis. All final test reports have been received; the look' reports (QLRs); w -

ve Saal data reports NRC's review will not be compr=hannive, but will (FDRs) and test analysis reports (TARS) were provided <v=aw of a high-level overview of the general aAer testang was e f -_' and the results analyzed. applicability of the data for code validation.

Completion of a high-level review of GE's major Westeghouse has also submitted a E -m ' - "ve SBWR systems code, TRACG, is a separate activity.

" closure" documsat exaamaang data in concert with Ph==<==== Idemaincasaoa and Ramiung Tables (PIRTs) la developmg the SBWR test program GE also developed by W ^ f- to help guide test progress generated PIRTa and performed scaling analyses of test Pl amaang, and the scahag M for the test facilities, but the withdrawal of the SBWR design programs. De objectives are to show that (a) the certification application, means GE will not produce a PIRTs are " validated" by test data, i.e., f ------- acaling/PIRT " closure" report similar to that developed idamsined and raakangs assagned are approxianstely by Westinghouse for the AP600.

renected by data, and (b) the scaled range of T/H parameters == , - ' by the data represents that NRC CONFIRMATORY TESTING espected in AP600, '- ^- ^': apphcability of the data in validatsag computer codes for AP600 safety The NRC's Office of Nuclear Regulatory Research analyses. Details of the NRC staff's test program (RES) is sponsoring confirmatory test programs for the review are discussed in a later section Review of passive designs, focused primarily on the AP600.

safety analysis codes is a separate activity, in progress. Testing covering design-basis accident (DBA) and beyond-DBA sceannos has been perfonned in the Japan SSWR TEST PROGRAM Atomic Energy Research lastitute's Rig of Safety A=======a (ROSA) large Scale Test Facility (LSTF),

GE's SBWR test progress is d==c===ad in previous madified to a., " AP600 passive safety systems.

papers (e.g., see Raf. 5); its guading philosophy was ne loop is about 1/30-volume-scale, full-height, and sender to W - f 's, f 4: T/H full-pressure. Despite configurstion differences

T of SBWR's passive anfety sysseses in both between AP600 and the modified LSTF, evaluation of integral and ; -__ ." s tests. Testing began in the data shows that integral T/H behavior of the test loop is late 1980's and concluded in 1995. Maior prograan similar to that expected in AP600. Further tests are i=cl= dad, low-pressure, full-heaght, approximately planned to study beyond-DBA scenanos with multiple 1/508-volumMicals testang of Gravity-Driven Cooling failures of passive safety systems. neee tests should Systems (GDCS)  ; -- = in the GIST facility at GE begis is early 1997 and continue for about 6 asonths in San Jose; low-pressure, full-height integral system testang of (a) late blowdown /early essergency core De NRC began in 1995 to conduct confirmatory coolant (ECC) isiectaan phases and (b) long-tenu testag in the APEX loop at OSU; tests are stillin

-cL- ' ^ wy conlang via the Passive c'n-an====a progress. Objectives are to obtam additional, Coolang System (PCCS) in the 1/400-volume-scale confirmatory data on selected DBA scenarios, and to GIRAFFE facility at Toshiba's Nuclear Engineenng investigate beyond-DBA events, similar to the 1 '-- - y in Kawasaka, Japan and in the 1/25-volume- ROSA /LSTF tests, to help evaluate the robustaess of scale, full-height, low-presses PANDA facility at the the passive safety systems with amitiple failures. A Paid Scherrer lastatute (PSI) in Wuerentingen, virtue of APEX is that it can be turned around rapidly Swataarland; and ; c^ ? ^- tests at full-scale and between tests: while ROSA is capable of a test about prototypic pressuses of imaame== <v=d===ar (IC) and every 3-4 weeks, APEX can perform about one test per PCCS heet exchnigers (HXs) in the PANTHERS werk (and occanaonally more), providing flexibility in focality at $1ET Laboratones in Piaceman, Italy. the confirmatory program. Confirmatory testing has contributed significantly to the NRC staff's design laisial review of the SBWR test program was certification review of the AP600.

similar to the AP600. evaluataca of test facility design, scalang, anstr====amana=, and test mainces; however Q4-2

'Ibe NRC anticipaaed similar confirassion test gas on contain=aat performance, with helium requirements for the SBWR, and contracted with simulatang hydrogen ("H" series); and (b)

• system Pwdue University for a 1/400-volume-scale,1/4-height, interactaan* effects dunag the late blowdown and early lowpenswe assegral test loop representang the SBWR ECC injection phases of SBWR DBAs (GIRAFFE / SIT).

reactor, contaanasset, and assety systems. The Purdue Univoresty Multi *h'lasegral Test Ama==hly As the staff reviewed results of completed testang, a (PUMA) was -==amad and performed its first concers was idsstified regarding quality assurance (QA) natsgral test in June 1996. N onginal test program, - - , '- - - ^% for design certification testing, la focused on DBA and beyond-DBA scenanos specific to contracts with the U.S. Departaset of Energy (DOE),

the SBWR. A revised program concentrates on integral the vendors comunitted to meet QA requirements of and , J-* T/H tests addressing esore general ANSI /ASME NQA-1 for design certification activities issues reissed to two-phase astural circidataan, on advanced reactors, the requirements applied, as well, consamummet response, and systeses imeeractaces- to contractors (including thana outande the U.S.). Initial evalustaan of QA i ;'--- ^% on selected test NRC REVIEW OF VENDORS

• TEST PROGRAMS programs ind. cased that key aspects of NQA-1 had not base included. "Ihese findings triggered QA h samous of the NRC staff's passive reactor design assessments on all major design certificahon T/H test certificateos scating review da=c===d in Refs. I and 2 programs for both AP600 and SBWR, including on-site (timough early 1995) is summariand below, with father == tare- of facilities and docu=aatation. Results of infor== ham os progress since Ref. 2 was prepared. the staff's QA F,- % and lessoas learned froni these activities are detailed in Ref. 7. Conclusions la 1991, the staff began ======amg in detail the test frons the QA review are doen== Wad in the AP600 psograses that had been ------ f =d is some cases, DraR Safety Evaluation Report (Ref. 8) and in the

- f ^- f b support design certificeham of passive staff's high-level review of GE's test program.

piames. laisial reviews focused on test facility designs, scalang analyses, instnumastaham, and test nashices. Post-test review of testing programs has For test programs under 6' , ^ the staff sande

, concentrated oo =======aat of data, ;'--- *gical rana ====dahn== regardaag facility configurahoes and insights, and application of data lo validation and

, '" - , and planned test mainces. verificahon of the vendors' safety analysis comiputer codes. Ha anna of the long delay in reachmg For AP600, the staff recommasaded that testang in a agreement with GE on an acceptable design certificahon  ;

high-presswe, full-height integral sysseems facility was test program, and GE's withdrawal of its application for aseded to provide data over a sufficiset range to certification, most of the staff's detailed post-test review valadste Westinghouss's accidset aaniysis cosaputer effort has been on the AP600.

codes. Discussions on this issue contimmed for about one year wish the vendor, and seemised in Both vendors prepared extensive doen===anham on Weseenshonnee's contractang wish SIET Laborasones to their safety analysis s-- a codes, to dama==trase (a) sendify amt perfons teshag in the SPES facility. applicability to the passive plants and (b) that the codes j have been validated baand os coseparisons to design For the SBWR, GE istaally 4 " SBWR certificahon test data. 7hese reports are also being T/H testang, i.e., programs in the GIST and GIRAFFE reviewed by the NRC separate from the test programs facaistaes and samall-scale university tests, as largely elia==3ves. Test program reviews, detailed below,

'na pa an;= piammed testang in PANTHERS and assess whether the data can be used with confidence to PANDA was desenbod as 'w.e.

-y,' i.e., not establish a database for code validahon and to moedad but useful. h sanff's passanon was that da-a==arese the p k, - a of the plaats' passive PANTHERS and PANDA tests wese required for safety sysseso=; code reviews are concerned with the dessem certificehos, and that fwther integral sysseems ability of safety analysis computer codes to represent tests represenhas early stages of DBAs might be 7 -- - - and T/H system behavior observed dunng requued. About two years of negatanhoes wish GE the testang. W code reviews are not discussed further.

were required to resolve these ia=== AAer a cosapiece reassessansat of the SBWR technology base and test An important participant in the test program programa, reported in GE's Test and Analysis Program reviews has been the Advisory Committee on Reactor Desenytaca (TAPD) report (Ref. 6), GE acceded to the Safeguards (ACRS), particularly the Subcomunittee ce staff's position and also agreed to run tests in Thermal-Hydraulic Phenoanena. W NRC staff and the GIRAFFE to study (a) effects of light non-enadammible vendors have met regularly with the ACRS since the Q4-3

test program reviews began, and the Comunitsee has behavior over an appropriate range of T/H parameters.

made a notable contribuhan to the review of the design h staff's review of tics program is now closed.

certificahon testing programs,

& ADS separate-effects test is still under review.

AP600 Test Programa 'lhe staff has asked queshoes about how test data will be used in selectag and testing ADS valves for the The first AP600 program to be reviewed was the AP600 plant. The staffis also continuing to review the PRHR - ;- 'fects test, wluck was the only rendor- data, in part hac==a of the difficulty of accurately systems related testing that had been complesed at the ma==urmg T/H parameters during the rapid blowdowns start of the reyww. Data were taken on three straight, that comprised much of the testing in this facility. W vertacal tubes, .., ^N the AP600 PRHR HX staff has not yet found any "show-stopper" issues, but design when the test progress was iv ':; d Westinghouse unast respond to the staff's questions and However, aAer perforumag the tests, W"f-- -- _

the data review anst be completed.

changed the HX design -a = muy, replacing the streght tubes with "C'-shaped tubes. & staff than Results frons the SPES-2 integral program indicated requued W- '-f -__ tojustify the use of the straight. that tbs facility appropnately i., :s AP600 passive tube data.is anodelang the new HX's J., = la safety system perfonnance over a wide range of e===nanar the dets more closely, the NRC observed misstated accidents and transients. Distortions were unexpected behavior of the heat transfer in the tubes, observed in the test faciiny's response, compared to which reannamed is single-phase liquid flow at all tinses; expected behavior of the AP600 plant, as a result of the outer surface of the uhas, e innamarsed is a pool non-prototypic configurahons of SPES-2 for some representing the AP600's in-contaansaset refueling water systems (e.g., one pump per loop, ia a-d of two) and storage tank (IRWST), in which the PRHR HX aits, the need to ec---;- "- for the facility's high heat could be in either two phase or single-phase natural losses early in the test and high weal heat input to the convectaan flow. When the reasons for the amosenlous coolant late in the test. Howevg .be effects of these data were investagnand further, it was determumed that an distortions could be characterized at lenat qualitatively, ermr in data conversion soAware had caused the and their impact on the data nadarstood 1he staff's observed prnhta==. W- 'f : has re-eanlyzed the review indicated that the data are appropnate for use in dass and revised the test repwt. m staff will complete validating accident analysis codes. Westinghouse has the review is early 1997. also responded satisfactorily to the staff's questions during the test program review, and the review of the The other four reactor / safety system tests-CMT, SPES-2 program has been closed.

ADS, and the SPES-2 and APEX integral programs-have been reviewed al-a=a =i==h==a==ty, due to their b most challenging integral test program to

= 1 ' ^'= = chad =1== h staff had mo:, by early review has been OSU/ APEX. b scaling approach for 1995, idsstified any 'insur-au=amhla problems" in these the APEX facility, especially its 1/4 height, has a programs (see Ref. 2). he is still anne, pendaag significant impact on the system's response, compared

- f ^ = of the soviews, includsag eval-anaa of to that expected in the AP600. This is to be expected Wesenghouse's Scaling and PIRT Clasme report is a passive plant, where, for inme==ce, elevation :l W ^ f :'s tassang .7

', for the esost part, differences play a major role in driving ECC injection.

tresmants and accidents up to the plant's design hania, A more subtle effect was the compression of scaled includeng en==darsham of single active failures as time during the tests. Test time is a factor of about 2 i requued by the NRC's ;g' ^%= ';--r.; a".as shorter than 'real* AP600 time; e.g., a 6-hour test in tests showed systems and/or ca-pa==a , Jo.--- + APEX represents about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of elapsed time in the

, is general, to pnHast c "" . plant. Another cocephcahon was the rad = cad pressure

^'

However, Ibase was -- ;-- -- behavior in the CMT of the APEX loop. Tests were started with the system test, soeultang in a change in the design of the CMT in water-filled, rather than trying to represent the abe plant. la sonne samadated accidents, stemas entered condition of AP600 at a reduced pressure early in an the CMT while it was full of cold weser, rendting in event, resulting in the need to " pressure scale" the data.

dynemac behavior caused by rapid cand===manan of large Also, the APEX program was designed to provide data steam bubbles. Westagbause placed a sparger at the relevant to the long-term cooling behavior of AP600.

CMT inlet, permitting enad=amana= to occur without While the containment was not completely represented, large condemention shocks. & staff's evalunhoe of the containment sump was modeled in APEX. A test dets indacated that the CMT test progrecs typical test, such as a small-break loss-of-coolant accomplished its obpective of providing data on CMT accident (SBLOCA) progressed relatively rapidly Q4-4

through blowdown, CMT injection at elevated & author is not reviewing program and cannot discuss

~

preamres, ADS actuation - r. m it in this paper in detail; however, the review is still in accuandator negoctaos, and blowdown through the ADS progrees, with several issues to be reso!ved.

to near- '= ,' -ic pressure. ECC ispecteam from the IRWST them began, ca=an==ng for several hours, with a SBWR Test Program j transition aAer as long as 20,000 somade to recirculatory cooling from the sump. With h course of events in the SBWR design approni==saly 800 instruments in APEX, the amount of certificataca testing review and the decision by GE to data from a single test is enorusous, and E m ve withdraw the design from the certification process have data review is extremely time-a==a==ng lianised in. depth review by the staff to only two test programs: GIST and GIRAFFE.

In addition to logistical aspects of the review I d=e====d above, APEX tests showed more unexpected GIST testang was performed in the late 1980's, l l behavior than any other AP600 program. Attentaca has prior to the NRC's fonmal review of the passive reactor I centered on a a==d==cy, especially dunas soone test programs h sta# thus had no opportunity to SBLOCAs, for syssess-wide flow and presswe review GE's facility design, scahag rationale, or test  !

oscillan- to occur dunag the long-tens cooling matrix before testing began. Review of the program portaos of the test (at times greater than about 10,000 raised significant questions about its claimed objectives.

l ==cand=). D;- with W"-f - -- and l

^ evahamanna by the NRC staK indicate that & GIST facility was based on an early SBWR j thsee types of oscillaanama are charactenstic of systemas ca=capan=3 design. It i.r. ' the SBWR's Gravity with large, interca===cead pools of water, as is inne for Driven Cooling Syssess (GDCS) and some aspects of APEX and AP600 during this phase of soone accidents. camaan====ap L,-- =, but did not include either the l N stan is studying this behavior with the help of the isolation raad==aar or the passive contanamant cooling NRC confinestory APEX test psogram, to detersune if systesa, and could not fully simulate the effect of the  ;

the oscillations could grow to create ~ =t' "- conamia-aat on the primary system. In addition, many syssess response. Other types of systeses interactions design details were substantially different from the final and now oscillaana== have been abnarved during SBWR design. GIST was originally represented as an diflerent penods of the tests, due in part to the integral systems test to investigate late blowdown and charactenstics of a two-phans matmal circulataos early ECC injection phases of SBWR DBAs.

systeen. Another ;' -- - = observed in APEX was Maxi === facility operstang presswe was about 0.7 CMT re6ll several thousand =arand= aAer the CMTs MPs (about 100 pain), and initial conditions for the tests had draened. A ========a of this behavior by were established hamed on calculated SBWR conditions W f _ ; and the staff indicates that this is likely a at that system presswe. W staff reviewed the test .

distortaos in system respones caused by the 1/4 program and concluded that it provided pasaaei=1ly elevation scale in APEX, and that driving presswo for useful data on the perfonmence of the GDCS early in CMT renti would not be pnasat in the full-height plant. the accident scenano, but that the configurational differences and inability to completely represent A Emal aspect of the APEX test review is the staff's potestaal systems interacteams readered it '- ' 7- as finding that some behavior in the test facility was a true integral systeams test program. largely on this caused. in part by samlM"- ca=pa===a= in the hanis, the staff recommandad that GE conduct the loop. His is da=e====d in a 7- ^ asetaos. additional 'syssess interaction" tests in GIRAFFE ansationed previously.

Because of the complex issues aneing frwa the APEX test prograni, the NRC's soview is not yet Early SBWR-related tests were niso conducted in compiste. Issues related to test facility scelsag must be GIRAFFE prior to the start of the NRC's fonaal test 1 resolved, as well as gn==ena== that the sta# has raised program review; thus, the staff did not review facility l concerning speci6c f----- and systeen response design, scaling, or test seatnces before testing began.

charactenstacs dunas the tests. He staff expects many Early separate <ffects tests investigated the perfonnance of these issues to be resolved by the Scalang and PIRT of a scaled (about 1/400) PCCS HX under vanous Clomsre Report, but some may requere further work, operating conditions, including several concentraisons of non-candansible gas. Integral systems tests were then A last AP600 test progress should be noted: testing perfonned, focusing re long-tenn cooling, with the  ;

of the passive contanansat coolang systeen, using a 1/8- primary system paa====g to the cona mia-a-t, and the linear-scale i.re- ^ %= of the AP600 contanament. PCCS condensing the naamm and returning it to the Q4-5

pressure vessel. Initial conditions for these tests were penod. Initial certification lasts 15 years and can be h===d on calculated SBWR conditions approximately renewed for 15 years; beyond that 30-year period, one hour aAer initinhos of a DBA. AP600's nonunal operahng life, for inasance, is 60 years. Certincahon is based, to a significant extent, on

• lhe staff's initial review of the early GIRAFFE the use of computer codes, validated with data from the tests led toe ach==8 d==c- with GE resenhag testang programs, to <i==a==e ste that the plant design dotads of the tests (e.g., choice of non-ca=d===ible gas meets regulatory limits for DBA and transient response.

concentrassons) and overall system T/H behavior. It is also expected that insights from the testang However, as QA '; A= of the design certificahos programs will have an impact on other aspects of the test programs i E--- ' '. the staff concluded that deman certification process, such as deternu==aion of Toshsba had not' / ' a fosusal QA~ program the i== pace , tests, analyses, and acceptance criteria ca==3=a==* with 10 CFR 50, Appendix B requaremsets (ITAAC) required under 10 CFR 52 to ensure that a or GE's casamstamsmes to DOE to seest ANSI /ASME plant, as built, conformas to the certified design.

NQA-1 (or equivalems) standards. Prior to -- f S of the NRC's i==paceaa= of this progress, GE Dunas the review, a number of " lessons learned" voluntarily withdrew the early GIRAFFE data from the have harame apparent, applicable either to the NRC formal design certificanos dan =h===, although the ophon staff or to the vendors and their contractors. Some of was ===*== a to use the data is a 'connrsantory* the more important items include:

fashaos. GE and Toshiba M ; ~'y ' M--9 a fonmal QA program for GIRAFFE, and the additional 1. Developanent of Phenon === identification and H series and GIRAFFE / SIT tests were fo. ' under Raalung Tables (PIRTs) is an essential early step in test

=, " QA oversaght. program plannang. PIRTs sumunarize the state of knowledge required for design and analysis of the plant.

The staffis i i ' ; high-level reviews of the They also provide a direct link to the capabilities of PANTHERS, GIRAFFE 'H", PANDA, and GIRAFFE / computer codes required for plant safety analyses, and SIT final test reports. The reviews focus on as overall andacate if testing is naartad, either to validate existing evat==*an= of the suatability of the data for use in code madals for new plant designs or to develop (and vahamanas, but do not inchade inglepth =========* of validate) new models for unique phenomena. PIRTs SBWR-related ;'----- -- N staff also reviewed are also a first step in perfornung scaling analyses for GE's TAPD report and the -i=*=d SBWR scahag the testing programs, and selecting test conditions for report (Ref. 9) and issued high-level --views of those either separate-effects or integral systems experiments.

dac====as, as well.

2. A comprehensive scaling analysis should be REGULATORY PERSPECrlVE AND LESSONS performed as part of the design activities for a test LEARNED facility. In several cases, the NRC staff found that the vendors had not performed formal scaling evaluations.

Regidatory review of passive plant design When questions were raised about the capability of the ,

cessismaanam test progresas and associated computer test facilities to achieve their objectives, the vendors codes was a new activity for the NRC whom istaaled in had to perfona such analyses aAer the fact, sometimies 1991. The review has involved at least 6 a=ch=ac=1 findaag significant scaling-related distorhons that review branches in the OfEco of Nuclear Reactor affected the usefulaees and applicability of the data.

Regulasson (NRR), NRR's propect daractorate and QA- Using a formaal scaling approach, such as the seisted orpaanne , entensive support froan RES and Hierarclucal Two-Tiered Scaling Methodology (Ref.

estesual contractors, and addataa==I resources of the 10), as part of the design process helps to guide the ACRS, its staff, and consukants. By the time it ends, design and, as stated in (1) above, helps also to the review will have requued shout 6 cal ==dar years integrate the PIRT process into the design, as well.

and aba====da of staff-hours of effort. 'lhe vendors also aa==itted substanhal resources, both financial and 3. Quality assurance (QA) is also an important part of persommel, to plasmang, r ' --';, and analyzing the design certificahon test programs. N NRC's QA design certificataos testang. " lie smaanive effort is .eview is the subject of Ref. 7 and is not discussed in justified, however, when the importance of the test data detail here. However QA in general provides a logical 6 considered is the content of design certific=*=aa process for performing and documenting those elements Whos a passive advanced light water reactor is of a test program that are essenhal to its success. Good certified, the NRC is, in effect, approving a desage that QA implementation does not guarantee that a test has not yet been buik, and is doing so for as extended program will provide good data, but poor QA is Q4-6

generally a key indicasar that a test program has serious 5 years of work by both the NRC and the reactor problems. QA aust be ig-- - - " at the initinhos of vendors, has helped provide the basis for the staffs a program; it is almost impossible to " retrofit

• good assessment of the performance of passive safety QA aAer the program is in progrees. systems. Valuable lessoas were learned during these reviews that abould be cM- .d in the establishment

4. Tests need to be perfonmed carefully, accordaag to of future testang prograses of this type.

a=aahheart, writsen pe=- " - , with chmages or unusual behovaor documnasted at the time of testang. 'Ihis is, to REFERENCES some extent, part of the " good QA' of(3), but encompasses amore thanjust quelaty assurance. 1. A. E. Levia, 'NRC Review of Passive Reactor C- , -

z-- observaama==--T/H behavior (where Design Certificehon Testing Programs: Overview and direct or video observation is poensbie), real-time data Regulatory Perspective," Proc. ARS '94. Int. Too.

trends, noises (such as water h===ar)-when Mts. on Adv====I n-*ars hf=*v, (1994).

m., " 'y logged in writing, can provide valuable post 4sst gudes to data evah=a a= and can help to focue 2. A. E. Levin, 'NRC Review of Passive Reactor on unusual or 7-' i behavior. Design Certificataca Testing Programs: Overview, Progress, and Regulatory Perspective," bgg,,

5. Sufficient time ea=LI be allowed between tests to NURETH-7. 7th Int. Mtn. on Nuclaar Rancear evaluate the data, especially to look for amosanties, Thannal-Hydrauhes, (1995).

unusual or umanp=cead behavior, and/or annifu=ca-=ing astruensees or facahty hardware. N vendors oAsa 3. L E. Hochreiter, et al., " Integral testing of the tended to try to run as sanny tests in as short a time as AP600 Passive Emergency Core Cooling Sy=sa==,

possible. In OSU/ APEX, NRC connrummaary tests Proc. ARS '94. Int. Too. Mts. on Advanced Reactors uncovered faulty hardware that aNected test results, as liafdr (1994).

well as unexpected f--- that, had they been idanaified dunng the test program, could have been 4. E. J. Piplica, et al., "The Westingbouse AP600 Test invastagated amore fully by anodifying plammed testang. Program - Moving Toward Commercialization of the Neat Generation of Advanced Reactors," Proc. ASME

6. Close ataananam unust be paid to data acqwsition and Joint Int. Power Conference (1991).

data converanos soAware. In the AP600 PRHR test programa, a subtle error in the data converanos program 5. A. S. Rao, et al., " Safety Research for the SBWR,'

Pac ===aaaaad consplete re-enelysis of the test data. The Proc. 20th Water n-ea, hr.,y M ei.=,'

best test facility is only as good as its abahty to NUREG/CP-0126,1,147-157 (1993),

accurately record the data froen its instrumsets, and analysis of the data depends ce proper .., * %= of 6. 'SBWR Test and Analysis Program Description,"

calabsetaos and converanos factors in transforsnnag the Rev. C, GE Nuclear Energy Report NEDC-32391 dass froen electncal signals to engineenng quantities. (1995).

1

7. Finally, abe value of c='- ' y teshag cammot be

7. A. E. Levin, " Quality Assurance on Desiga l overnessed. h senn hoheves that boek vendors' testang Certificahon Testing Programs," NUTHOS-5 paper QS prograses, as uh,e. manly execused, provided a weekh of (1997) ,,

data wish which to valadese analytical nadma, and gave valuehle insights imeo both 7 - and syssea 8. "DraA Safety Evalushon Report Related to the respones characeanstics that are relevant to the Certificataos of the AP600 Design,' USNRC Report, [I opcreemos of passive designs. Nevertheises, the NRC's DraA NUREG-1512 (1994); and Supplement 1 (1996).

=' ^ -y tests were able to focus on specific issues for amose in-depth data acqmaition and evalama-=, and 9. ' Scaling of the SBWR Related Tests,' Rev.1, GE to psovide insights that may not have been idamaified Nuclear Energy Report NE-32288 (1995).

froaa the vendors' test data.

10. 'An Integrated Structure and Scaling Methodology

SUMMARY

AND CONCLUSION for Severe Accident Techancal Issue Resolution, Appendix D: A Hierarchical, Two-Tiered Scaling The NRC is mannag the end of its review of design Analysis,' NUREG/CR-5809 (1991).

certification test programs for the advanced, passive light weser reactors. 'this effort, compnaang snore than Q4-7 ,