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 STATES NUCLEAR REGULATORY
COMMISSION
r J3 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: Evaluation for BWR Mark II and III Containments.*
NUREG-0802 is being Issued to provide acceptance criteria for hydrodynamic loads on piping, equipment, and containment structures resulting from SRV actuation.
The NRC staff finds 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, however, is not a subsititute for the regulations, and compliance with the 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 the Mark II Owners Group and GE, the staff has re-evaluated the SRV loads and 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.V 2 -.)_,. .,S, i, i_. ..-E* **.' XDarrell G. Elsenhut, Olrector Division of Licensing Office of Nuclear Reactor Rcgulation Enclosure:
NUREG-0802 Attachments
1 L-2 62 oos 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 in a 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 !II contarmnent 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 in Bethesda, Maryland.
Subsequent to the meeting, this modified method and prcposed load criteria were reported in Am.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
The statistical method proposed by GE to arrive at design quencher loads for the Mtark III containment consists of a series of steps.Initially, a multiple linear regression analysis for the first actuation event is performed wtth a data base taken from three tests series: mint-scale
(9 points), small scale (70 points) and large scale (37 points).Non-linearities are introduced where necessary~by using quadratic variables and formed straight line segments.
The regression coeffi-cients are estimated from the appropriate data set. The resulting equation contains a constant term plus corrective terms that take into dccount the influence of all key parameters.
In the second step, the subsequent actuation effect is determined by postulating a direct proportionality between the observed maximum subsequent 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 SRV subsequent actuation is found by noting that the total variance is the sum of three terms: (1) a term due to the uncartainty in the 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
.-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 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.Results of the aralysis shows that Initiation of reac 4.- 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.
The staff finds the result of the GE analysis reasonable.
There-fore, the assumption of only the lowest set SRV operatir subsequent actuation is justified and acceptable.
The acceptance of the quencher load criteria is based on the test data available to us. We realize. however, that the tests lack exact dynamic or geometric similarity with the quencher system for the Mark liI containment.
The test results, therefore, could not be applied directly.
Though the quencher lads for the Ilark III appear conservative 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 require in-plant testing.!'. REGULATORY
POS1T Ct It is our position that applicants for Mark INI containments using the quencher device commit to the criteria specified below: 1. The structures affected by the SRV operation should be designed to withstand the maximum ioads specified in Table 1. For the cases of multiple valve actuation, the quencher loads from each line shall be assurmed to reach the peak pressure simultaneously and oscillate in phase.
-8-.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 identical with that specified in Section A7.2.2.4 of NEDE-11314-08.
b. The value of key parwneters should be equal to or less thao 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 normal pla2nt operation (F¢ 100 c. The.value of those key para-eters should be ecual to or larcer than that specified belcw: Water surface area per quencher (ft 2) 295 SRY opening tir.e (sec) 0.020 3. The spatial variation of the quercher loads should be calculated by the methods shown in Section ..4 of the topical report NEDE-21078.
4. The load profile and associated time histories specified in Figure AS.11 of NEDO-113/4-C
8 should be used with a quencher load frequency of 5 to 11-Hz._ _ , 9
-9-S. For the 40 year plant life, the nr.ber of fatigue cycles for 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.
6. In-plant testing of the quencher should be conducted to verify the 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, and strain of affected structures.
A prototypical plant should be selected for each type of containment structure.
For example, the pressure responses from a concrete containm..ent should not be used for a free-standing steel containment and vice versa. Tests should be conducted as soon as operational conditions allow and should be performed prior to full power operation.
7. Based on tne in-plant test results, reanalyses should be performed to ensure the safety margin for the structures, which include the 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 safety margin for the structures will be reduced because of the_ _
-10.new loads tdentifted from the test, modificatton or strengthening of the structures should be made in order to maintain the safety margin for which the structures were originally designed.
The applicants for the Mark III containment with quencEars for S/R 'alves should submit a licensing topical report for approval.This report should present a test pr~ogram and Identify the feasibility of modification or strengthening of the structures.
I I hULt : QUENCHER BUBBLE PRESSURE MARK 111, 238 STANDARD PLANT 95-95% CONFIDENCE
LEVEL Design Value Maxiumm Pressure (psid)Case Description Po (4) P 0 (H)1. Single Valve First Actuation, at 100-F Pool Teaperature
13,5 -8,1 2. Single Valve Subse ent Actualton, at 1c bF Pool Tepe raLure 28.2 .12.0 3. Two Adjacent Valves First Actuation at 100l F 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.