NRC Generic Letter 1982-24
| ML031080490 | |
| Person / Time | |
|---|---|
| Issue date: | 11/04/1982 |
| From: | Eisenhut D G Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0802 GL-82-024, NUDOCS 8211080059 | |
| Download: ML031080490 (12) | |
UNITED STATESNUCLEAR REGULATORY COMMISSIONr J3 VASHINCTON. 0 C. 205S5~. / November 4, 1982TO BWR APPLICANTS WITH MARK II OR 11I CONTAILMENT (EXCEPT WPPSSII)
SUBJECT: SAFETY/RELIEF VALVE QU'ENCHER LOADS:EVALUATION FOR BWRMARK II AND III CONTAINMENTS(Generic Letter No. 62-24)Enclosed is a copy of NUREG-0802, -Safety/Relief Valve Quencher Loads:Evaluation for BWR Mark II and III Containments.* NUREG-0802 is being Issuedto provide acceptance criteria for hydrodynamic loads on piping, equipment,and containment structures resulting from SRV actuation. The NRC staff findsthat use of these acceptance criteria satisfy the requirements of GeneralDesign Criteria 16 and 29 In Appendix A to 10 CFR Part 50. NUREG-0802,however, is not a subsititute for the regulations, and compliance with theNUREG is not a requirement. An approach or method different from the-accep-tance criteria contained herein will be accepted if the substitute approachor method provides a basis for determining that the regulations have beenmet.The NRC had issued SRV load acceptance criteria for both Mark II (NUREG-0487, Supplement No. 1, September 1980) and Mark III (SER for GESSAR, July1976). However, the staff, the Mark II Owners Group and GE recognized thatthese criteria were very conservative because they were established at theearly stage of quencher development. Since then, extensive quencher testprograms were performed resulting in a sufficient data base to justifyre-evaluation the SSRV load criteria. In response to the request by theMark II Owners Group and GE, the staff has re-evaluated the SRV loads andestablished the new acceptance criteria in NURE-0802. The staff also findsthe earlier criteria acceptable. The acceptance criteria in NUREG-0487supplement No. 1 (for Mark 1I plants) or the acceptance criteria in anattachment 2 (for Mark III plants) are conservative with.respect to theacceptance criteria proposed in Appendices A and B of NUREG-O802, respectivelyand they are acceptable.The reporting andior recordkeeping requirements contained in this letter affectfewer than ten respondents; therefore, OMB clearance is not required underP.L.96-511.V 2 -.)_,. .,S, i, i_. ..-E* **.' XDarrell G. Elsenhut, OlrectorDivision of LicensingOffice of Nuclear Reactor Rcgulation
Enclosure:
NUREG-0802Attachments 1 L-262 oos ATTACHMENT 2 .ACCEPTAIICE CRITERIA JUL I dFOR QUENCHER LOADS FORTHE MARK III CONTAINMENTI .I1. INTRODUCTION'on September 2, 1975, the General Electric Company submitted topicalreports NEDO-11314-08 (nonproprietary) and NEDE-11314-08 (proprietary)entitled, "Informaticn Report Mark III Containment Dynamic LoadingConditions,' docketed as Appendix 3-B to the Amendment No. 37 forGESSAR, Docket No. STN-50-447. As part of this report, a devicecalled a "quencher" would be used at the discharge end of safety/relief valve (SRV) lines inside the suppression pool. Tests wereperformed in a foreign country to obtain quenchier load data that wereused to establish the Mark II! data base. A statistical techniqueusing the test data to predict quencher loads for Mark !II contarmnentwas also presented. GE had submitted another topical report NEDE-21078entitled. 'Test Results Employed by GE for BER Containment and VerticalVent Loads," to substantiate their method to extrapolate the loadsobtained from the tests to the Mark III design.We reviewed the above topical reports and had identified several areasof concern. Meetings with GE were held to discuss these concerns. Asa result, GE presented a modified method during the April 2, 1976,meeting held in Bethesda, Maryland. Subsequent to the meeting, thismodified method and prcposed load criteria were reported in Am.endmentNo. 43, which was received on June 22, 1976. Our evaluation, therefore,--is baied on the modified method and the load criteria calculated by I'this Method.11. SUmMARY OF THE METHOD OF QUENCHER LOAD PREDICTIONThe statistical method proposed by GE to arrive at design quencherloads for the Mtark III containment consists of a series of steps.Initially, a multiple linear regression analysis for the firstactuation event is performed wtth a data base taken from threetests series: mint-scale (9 points), small scale (70 points) andlarge scale (37 points).Non-linearities are introduced where necessary~by using quadraticvariables and formed straight line segments. The regression coeffi-cients are estimated from the appropriate data set. The resultingequation contains a constant term plus corrective terms that takeinto dccount the influence of all key parameters.In the second step, the subsequent actuation effect is determined bypostulating a direct proportionality between the observed maximumsubsequent actuation pressure and the predicted first actuation pres-sure. The proportionality constant is found by considering the large-scale data.In the third step. the total variance of the predicted future SRVsubsequent actuation is found by noting that the total variance isthe sum of three terms: (1) a term due to the uncartainty in the QI-.3first actuation prediction which ts calculated from standard (normalvariate) formulas. (2) a term due to the uncertainty in the propor-tionality factor as was calculated in the second step above, and (3)a term due to the variance of the residual maximum subsequent pressure.It is now assumed that this variance is proportional to the square ofpredicted maxiium subsequent actuation pressure. The proportionalityconstant is found from the large scale subsequent actuation data (10values).In the fourth step, design values for Mark III are determined fromthe estimated (i.e., predicted) values of naxim'n subsequent actuationpressure and its standard deviation by enploying standard tables ofso-called "tolerance factors." These tables are entered with threequantities: (1) n, the number of sample data points frosi which theestimate of the mean and standard deviations are obtained. GE hasset n a 10, based on 10 maximum subsequent actuation points used inthe third step, (2) the probability value, and (3) the confidence level.The design value is then simply the predicted value plus the tolerancefactor times the estimated standard deviation.T.e approach as outlined above is used to calculate the positive -pressures for a single SRV considering multiple actuations whichrepresents the most severe SRV operation condition; For the singleactuation case, the calculational procedures are similar with the I0-4-method mentioned above with the following exceptions:1. The calculation whtch involves subsequent actuations is eliminated;and.2. Thirty-seven data points were selected for establishing the tolerancefactor since these data points in the large-scale tests relate tosingle value actuation.For negative pressure calculation, a correlation of peak positive andnegative pressures is developed. The correlation is based on theprinciple of conservation of energy and verified by the small-scale andlarge-scale test results.Based on the rethod outlined above, GE has calculated the SRV quencherloads for the Mark III and established the load criteria for six casesof SRV operation. 'The calculated load criteria based on 95-95% confi-dence level are given on Table 1 which is attached.H'. EVALUATION SU-1MARYAs a result of our review, we have concluded that the statistical methodproposed by GE and the load criteria shown on Table 1 are acceptable.This conclusion is based on the following:1. The method has properly treated all available test data and isbased essentially on the large-scale data with correction termsthat take into account the influence of non-large-scale variables.--Since the large-scale tests were performed in an actual reactor
.-5.with a suppression containment conceptually similar with GE contain-ment, extrapolation from the large-scale by statistical technique,therefore, is appropriate and acceptable.2. The method has been conducted in a conservative manner. The primaryconservatisms are:a. The calculation is based on the most severe parameters. Forexample, the maximum air volume initially stored in the line,the maximum initial pool temperature and the highest primarysystem pressure were selected to establish quencher loadcriteria.b. For the cases of multiple valve actuation, the load criteriaare based on the assumption that the maxizrmw pressures resultingfrcm each valve will occur simultaneously. V.e believe that theassumcption is conservative since different lengLns of line andSRV pressure set points will result in the occurrence of maxi-mum pressures at different tines and consequently lower loads.3. The proposed load criteria, whic!h are provided on the attachedTable 1, are acceptable. The criteria were established by using95-g5: confidence limit. Our consultant, the Brookhaven NationalLaboratory, has performed an analysis for the effect of confidencelimit. The result of this analysis indicates that for 9S-950 confi-dence limit, approximately lI of the number of RSV actuations mayresult in -containment loads above the design value. WIe believe that
-6-this low probability is acceptable considering the conservatismof the method of prediction! i.e., the actual loads should notexceed the design value.4. With regard to the subsequent actuation, the load criteria arebased upon a single SRV actuation. G.E. has established thisbasis by regrouping the SRV's in each group of pressure set points.As indicated in Amendment 43, there are three groups of pressureset points for the 19 SRY's for the 238-732 standard plant,-namely,one SRV at a pressure set point of .1103 psig, 9 SRYVI a: 1113 psig,and the remaining 9 M's at 1123 psig. Vwiy one SFV is now setat the loaest pressure set point. Based on this pressuwe set pointarrangement for the 19 SRV's, GE has analyzed the mo;t severeprimary pressure transient, i.e., a turbine trip withov: bypass.Results of the aralysis shows that Initiation of reac4.- isolationwill activate all or a portion of the 19 SRV's which will releasethe stored energy in the primary system. Following the initialblowdown. the energy generated in the primary syste~m consistsprimarily of decay heat which will cause the lowest set SRV toreopen and reclose (subsequent actuation). The time durationbetween subsequent actuation was calculated to-be a min'mum of62 seconds and increasing with each actuation. The time durationof each blowdovn decreases from 51 seconds for-the initial bl.w-down and decreases to 3 seconds at the end. of the period ofsubsequent actuatlons which is 30 minutes after initiation of w7-reactor tsolation.The staff finds the result of the GE analysis reasonable. There-fore, the assumption of only the lowest set SRV operatirsubsequent actuation is justified and acceptable.The acceptance of the quencher load criteria is based on the testdata available to us. We realize. however, that the tests lackexact dynamic or geometric similarity with the quencher system forthe Mark liI containment. The test results, therefore, could notbe applied directly. Though the quencher lads for the Ilark III appearconservative in comparison with the test data, some degree of uncer-tainty is ack-nowledged. The uncertainty Is prirarily due to a sub-stantial degree of scatter of all test data. W:e therefore will requirein-plant testing.!'. REGULATORY POS1T CtIt is our position that applicants for Mark INI containments using thequencher device commit to the criteria specified below:1. The structures affected by the SRV operation should be designed towithstand the maximum ioads specified in Table 1. For the casesof multiple valve actuation, the quencher loads from each lineshall be assurmed to reach the peak pressure simultaneously andoscillate in phas .2. The quencher loads as specified in Item I above are for a parti-cular quencher configuration shown In the topical reports tHEDO-11314-08 and NEDE-11314-08. Since the quencher loads are sensi-tive to and dependent upon the parameters of quencher configura-tion, the following requirements should be met:a. the sparger configuration and hole pattern should be identicalwith that specified in Section A7.2.2.4 of NEDE-11314-08.b. The value of key parwneters should be equal to or less thaothat specified below:Total air volume in eacbh SRY -nirc (ftt) 56.13Distance from the center of quencherto the pcol surface at high waterlevel 13.lluMaximu.m pccl te.vperature duringnormal pla2nt operation (F¢ 100c. The.value of those key para-eters should be ecual to or larcerthan that specified belcw:Water surface area per quencher (ft2) 295SRY opening tir.e (sec) 0.0203. The spatial variation of the quercher loads should be calculatedby the methods shown in Section ..4 of the topical report NEDE-21078.4. The load profile and associated time histories specified in FigureAS.11 of NEDO-113/4-C8 should be used with a quencher load frequencyof 5 to 11-Hz._ _ , 9
-9-S. For the 40 year plant life, the nr.ber of fatigue cycles forthe destsn of the structures affected by the quencher loadsshould not be less than that specified in Section A9.O ofNEDO-11314-08.6. In-plant testing of the quencher should be conducted to verifythe quencher design loads and oscillatory frequency. The in-plant tests should include the following:a. single valve actuation;b. consecutive actuation of the same valve; and,.c. actuation of multiple valves.Included should be measurements of pressure load, stress, andstrain of affected structures. A prototypical plant should beselected for each type of containment structure. For example,the pressure responses from a concrete containm..ent should not beused for a free-standing steel containment and vice versa. Testsshould be conducted as soon as operational conditions allow andshould be performed prior to full power operation.7. Based on tne in-plant test results, reanalyses should be performedto ensure the safety margin for the structures, which include thecontainment wall, basemat, drywell wall, submerged structuresinside the suppression pool, quencher supports and componentsinfluenced b) S/R loads. If the analysis indicates that thesafety margin for the structures will be reduced because of the_ _
-10.new loads tdentifted from the test, modificatton or strengtheningof the structures should be made in order to maintain the safetymargin for which the structures were originally designed. Theapplicants for the Mark III containment with quencEars forS/R 'alves should submit a licensing topical report for approval.This report should present a test pr~ogram and Identify thefeasibility of modification or strengthening of the structure II hULt :QUENCHER BUBBLE PRESSURE MARK 111, 238 STANDARD PLANT95-95% CONFIDENCE LEVELDesign ValueMaxiumm Pressure (psid)Case Description Po (4) P0 (H)1. Single Valve First Actuation,at 100-F Pool Teaperature 13,5 -8,12. Single Valve Subse entActualton, at 1c bF PoolTepe raLure 28.2 .12.03. Two Adjacent Valves FirstActuation at 100l F Po1Temperature 13.5 .8.14. 10 Valves (One Low Set andNine Next Level Low Set)First Actuation at loo 1f6 .Pool Ye perature 16,7 r9,35. 19 Valves (All Valve Case)First Actuation, at lG0FPool Teuperature 18.6. .9.96. 8 ADS Valves First Actuationat 120F Pool Temperature 17;4 *10.4S.