NRC Generic Letter 1982-24

NRC Generic Letter 1982-024: Safety/Relief Valve Quencher Loads: Evaluation for BWR Mark II & III Containment

ML031080490
Person / Time
Issue date:	11/04/1982
From:	Eisenhut D Office of Nuclear Reactor Regulation
To:
References
NUREG-0802 GL-82-024, NUDOCS 8211080059
Download: ML031080490 (12)

UNITED STATES

NUCLEAR REGULATORY COMMISSION

J3 r VASHINCTON. 0 C. 205S5

~. / November 4, 1982 TO BWR APPLICANTS WITH MARK II OR 11I CONTAILMENT (EXCEPT WPPSSII)

SUBJECT: SAFETY/RELIEF VALVE QU'ENCHER LOADS:

EVALUATION FOR BWR

MARK II AND III CONTAINMENTS

(Generic Letter No. 62-24)

Enclosed is a copy of NUREG-0802, -Safety/Relief Valve Quencher Loads: Issued Evaluation for BWR Mark II and III Containments.* NUREG-0802 is being to provide acceptance criteria for hydrodynamic loads on piping, equipment, finds and containment structures resulting from SRV actuation. The NRC staff that use of these acceptance criteria satisfy the requirements of General Design Criteria 16 and 29 In Appendix A to 10 CFR Part 50. NUREG-0802, the however, is not a subsititute for the regulations, and compliance with NUREG is not a requirement. An approach or method different from the-accep- tance criteria contained herein will be accepted if the substitute approach or method provides a basis for determining that the regulations have been met.

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, July

1976). However, the staff, the Mark II Owners Group and GE recognized that these criteria were very conservative because they were established at the early stage of quencher development. Since then, extensive quencher test programs were performed resulting in a sufficient data base to justify re-evaluation the SSRV load criteria. In response to the request by theand Mark II Owners Group and GE, the staff has re-evaluated the SRV loads established the new acceptance criteria in NURE-0802. The staff also finds the earlier criteria acceptable. The acceptance criteria in NUREG-0487 supplement No. 1 (for Mark 1I plants) or the acceptance criteria in an attachment 2 (for Mark III plants) are conservative with.respect to the acceptance criteria proposed in Appendices A and B of NUREG-O802, respectively and they are acceptable.

The reporting andior recordkeeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L.96-511.

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.' XDarrell G. Elsenhut, Olrector Division of Licensing Office of Nuclear Reactor Rcgulation Enclosure:

NUREG-0802 Attachments 1 L-2

62oos

ATTACHMENT 2 .

ACCEPTAIICE CRITERIA JUL I d FOR QUENCHER LOADS FOR

THE MARK III CONTAINMENT

I . I

1. INTRODUCTION

'on September 2, 1975, the General Electric Company submitted topical reports NEDO-11314-08 (nonproprietary) and NEDE-11314-08 (proprietary)

entitled, "Informaticn Report Mark III Containment Dynamic Loading Conditions,' docketed as Appendix 3-B to the Amendment No. 37 for GESSAR, Docket No. STN-50-447. As part of this report, a device called a "quencher" would be used at the discharge end of safety/

relief valve (SRV) lines inside the suppression pool. Tests were performed ina foreign country to obtain quenchier load data that were used to establish the Mark II!data base. A statistical technique using the test data to predict quencher loads for Mark !IIcontarmnent was also presented. GE had submitted another topical report NEDE-21078 entitled. 'Test Results Employed by GE for BER Containment and Vertical Vent Loads," to substantiate their method to extrapolate the loads obtained from the tests to the Mark III design.

We reviewed the above topical reports and had identified several areas of concern. Meetings with GE were held to discuss these concerns. As a result, GE presented a modified method during the April 2, 1976, meeting held inBethesda, Maryland. Subsequent to the meeting, this modified method and prcposed load criteria were reported inAm.endment No. 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 PREDICTION

arrive at design quencher The statistical method proposed by GE to of a series of steps.

loads for the Mtark III containment consists analysis for the first Initially, a multiple linear regression base taken from three actuation event is performed wtth a data scale (70 points) and tests series: mint-scale (9 points), small large scale (37 points).

using quadratic Non-linearities are introduced where necessary~by The regression coeffi- variables and formed straight line segments.

data set. The resulting cients are estimated from the appropriate terms that take equation contains a constant term plus corrective into dccount the influence of all key parameters.

effect is determined by In the second step, the subsequent actuation the observed maximum postulating a direct proportionality between first actuation pres- subsequent actuation pressure and the predicted by considering the large- sure. The proportionality constant is found scale data.

the predicted future SRV

In the third step. the total variance of the total variance is subsequent actuation is found by noting that the uncartainty in the the sum of three terms: (1)a term due to

QI -.

3 first actuation prediction which ts calculated from standard (normal variate) 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 of predicted maxiium subsequent actuation pressure. The proportionality constant is found from the large scale subsequent actuation data (10

values).

In the fourth step, design values for Mark III are determined from the estimated (i.e., predicted) values of naxim'n subsequent actuation pressure and its standard deviation by enploying standard tables of so-called "tolerance factors." These tables are entered with three quantities: (1)n, the number of sample data points frosi which the estimate of the mean and standard deviations are obtained. GE has set n a 10, based on 10 maximum subsequent actuation points used in the third step, (2)the probability value, and (3)the confidence level.

The design value is then simply the predicted value plus the tolerance factor 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 which represents the most severe SRV operation condition; For the single actuation case, the calculational procedures are similar with the

I

0-4- 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 tolerance factor since these data points in the large-scale tests relate to single value actuation.

For negative pressure calculation, a correlation of peak positive and negative pressures is developed. The correlation is based on the principle of conservation of energy and verified by the small-scale and large-scale test results.

Based on the rethod outlined above, GE has calculated the SRV quencher loads for the Mark III and established the load criteria for six cases of SRV operation. 'The calculated load criteria based on 95-95% confi- dence level are given on Table 1 which is attached.

H'. EVALUATION SU-1MARY

As a result of our review, we have concluded that the statistical method proposed 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 is based essentially on the large-scale data with correction terms that take into account the influence of non-large-scale variables.

- Since the large-scale tests were performed in an actual reactor

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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 primary conservatisms are:

a. The calculation is based on the most severe parameters. For example, the maximum air volume initially stored in the line, the maximum initial pool temperature and the highest primary system pressure were selected to establish quencher load criteria.

b. For the cases of multiple valve actuation, the load criteria are based on the assumption that the maxizrmw pressures resulting frcm each valve will occur simultaneously. V.e believe that the assumcption is conservative since different lengLns of line and SRV 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 attached Table 1, are acceptable. The criteria were established by using

95-g5: confidence limit. Our consultant, the Brookhaven National Laboratory, has performed an analysis for the effect of confidence limit. The result of this analysis indicates that for 9S-950 confi- dence limit, approximately lI of the number of RSV actuations may result in -containment loads above the design value. WIe believe that

-6- this low probability is acceptable considering the conservatism of the method of prediction! i.e., the actual loads should not exceed the design value.

4. With regard to the subsequent actuation, the load criteria are based upon a single SRV actuation. G.E. has established this basis by regrouping the SRV's in each group of pressure set points.

As indicated in Amendment 43, there are three groups of pressure set 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 set at the loaest pressure set point. Based on this pressuwe set point arrangement for the 19 SRV's, GE has analyzed the mo;t severe primary pressure transient, i.e., a turbine trip withov: bypass.

4 Results of the aralysis shows that Initiation of reac .- isolation will activate all or a portion of the 19 SRV's which will release the stored energy in the primary system. Following the initial blowdown. the energy generated in the primary syste~m consists primarily of decay heat which will cause the lowest set SRV to reopen and reclose (subsequent actuation). The time duration between subsequent actuation was calculated to-be a min'mum of

62 seconds and increasing with each actuation. The time duration of each blowdovn decreases from 51 seconds for-the initial bl.w- down and decreases to 3 seconds at the end. of the period of subsequent actuatlons which is 30 minutes after initiation of

w7- reactor tsolation.

reasonable. There- The staff finds the result of the GE analysis set SRV operatir fore, the assumption of only the lowest acceptable.

subsequent actuation is justified and is based on the test The acceptance of the quencher load criteria that the tests lack data available to us. We realize. however, the quencher system for exact dynamic or geometric similarity with therefore, could not the Mark liI containment. The test results, lads for the Ilark III appear be applied directly. Though the quencher data, some degree of uncer- conservative in comparison with the test Is prirarily due to a sub- tainty is ack-nowledged. The uncertainty W:e therefore will require stantial degree of scatter of all test data.

in-plant testing.

!'. REGULATORY POS1T Ct INI containments using the It is our position that applicants for Mark specified below:

quencher device commit to the criteria should be designed to

1. The structures affected by the SRV operation Table 1. For the cases withstand the maximum ioads specified in loads from each line of multiple valve actuation, the quencher simultaneously and shall be assurmed to reach the peak pressure oscillate in phase.

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are for a parti-

2. The quencher loads as specified in Item I above tHEDO-

cular quencher configuration shown In the topical reports

11314-08 and NEDE-11314-08. Since the quencher loads are sensi- configura- tive to and dependent upon the parameters of quencher tion, the following requirements should be met:

be identical a. the sparger configuration and hole pattern should with that specified in Section A7.2.2.4 of NEDE-11314-08.

or less thao b. The value of key parwneters should be equal to that specified below:

Total air volume in eacbh SRY -nirc (ftt) 56.13 Distance from the center of quencher to the pcol surface at high water level 13.llu Maximu.m pccl te.vperature during 100

normal pla2nt operation (F¢

to or larcer c. The.value of those key para-eters should be ecual than that specified belcw:

2 295 Water surface area per quencher (ft)

SRY opening tir.e (sec) 0.020

be calculated

3. The spatial variation of the quercher loads should report NEDE-21078.

by the methods shown in Section ..4 of the topical specified in Figure

4. The load profile and associated time histories load frequency AS.11 of NEDO-113/4-C should be used with a quencher

8 of 5 to 11-Hz.

9 ,

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-9- for S. For the 40 year plant life, the nr.ber of fatigue cycles the destsn of the structures affected by the quencher loads should not be less than that specified in Section A9.O of NEDO-11314-08.

verify

6. In-plant testing of the quencher should be conducted to in- the quencher design loads and oscillatory frequency. The 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, and be strain of affected structures. A prototypical plant should selected for each type of containment structure. For example, not be the pressure responses from a concrete containm..ent should Tests used for a free-standing steel containment and vice versa.

and should be conducted as soon as operational conditions allow should be performed prior to full power operation.

performed

7. Based on tne in-plant test results, reanalyses should be the to ensure the safety margin for the structures, which include containment wall, basemat, drywell wall, submerged structures inside the suppression pool, quencher supports and components influenced b) S/R loads. If the analysis indicates that the the safety margin for the structures will be reduced because of

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modificatton or strengthening new loads tdentifted from the test, in order to maintain the safety of the structures should be made were originally designed. The margin for which the structures with quencEars for applicants for the Mark III containment topical report for approval.

S/R 'alves should submit a licensing pr~ogram and Identify the This report should present a test of the structures.

feasibility of modification or strengthening

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I hULt :

QUENCHER BUBBLE PRESSURE MARK 111, 238 STANDARD PLANT

95-95% CONFIDENCE LEVEL

Design Value Maxiumm Pressure (psid)

Case Description Po (4) P0 (H)

1. Single Valve First Actuation, at 100-F Pool Teaperature 13,5 -8,1

2. Single Valve Subse ent Actualton, at 1cbF Pool Tepe raLure 28.2 .12.0

3. Two Adjacent Valves First Actuation at 100lF Po1 Temperature 13.5 .8.1

4. 10 Valves (One Low Set and Nine Next Level Low Set)

First Actuation at loo 1f6 .

Pool Ye perature 16,7 r9,3

5. 19 Valves (All Valve Case)

First Actuation, at lG0F

Pool Teuperature 18.6. .9.9

6. 8 ADS Valves First Actuation at 120F Pool Temperature 17;4 *10.4 S.