United States Nuclear Regulatory Commission - Protecting People and the Environment

Safety/Relief Valve Quencher Loads: Evaluation for BWR Mark II and III Containments (Generic Letter No. 82-24)



                                UNITED STATES
                        NUCLEAR REGULATORY COMMISSION
                           WASHINGTON, D.C. 20555

                              November 4, 1982

TO BWR APPLICANTS WITH MARK II OR III CONTAINMENT (EXCEPT WPPSS II) 

SUBJECT:  SAFETY/RELIEF VALVE QUENCHER LOADS: 
          EVALUATION FOR BWR
          MARK II AND III CONTAINMENTS
          (Generic Letter No. 82-24)

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 substitute for the regulations, and compliance with the 
NUREG is not a requirement. An approach or method different from the 
acceptance 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 SRV 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 NUREG-0802. The staff also finds 
the earlier criteria acceptable. The acceptance criteria in NUREG-0487 
supplement No. 1 (for Mark II 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-0802, 
respectively and they are acceptable. 

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


                                   Darrell G. Eisenhut, Director 
                                   Division of Licensing 
                                   Office of Nuclear Reactor Regulation 

Enclosure:
NUREG-0802
Attachments 1 & 2



8211080059 
.

                                                              ATTACHMENT 2 

                            ACCEPTANCE CRITERIA
                          FOR QUENCHER LOADS FOR
                         THE MARK III CONTAINMENT

I.   INTRODUCTION 

     On September 2, 1975, the General Electric Company submitted topical 
     reports NEDO-11314-08 (nonproprietary) and NEDE-11314-08 (proprietary) 
     entitled, "Information 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 quencher load data that were used to 
     establish the Mark III data base. A statistical technique using the 
     test data to predict quencher loads for Mark III containment was also 
     presented. GE had submitted another topical report NEDE-21078 entitled, 
     "Test Results Employed by GE for BWR 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 proposed load criteria were reported in Amendment 
     No. 43, which was received on June 22, 1976. Our evaluation, therefore, 
     is based on the modified method and the load criteria calculated by 
.

                                     -2-

     this method. 

II.  SUMMARY OF THE METHOD OF QUENCHER LOAD PREDICTION 

     The statistical method proposed by GE to arrive at design quencher 
     loads for the Mark III containment consists of a series of steps. 
     Initially, a multiple linear regression analysis for the first 
     actuation event is performed with a data base taken from three tests 
     series: mini-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 
     coefficients are estimated from the appropriate data set. The resulting 
     equation contains a constant term plus corrective terms that take into 
     account 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 
     pressure. 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 uncertainty in the 
.

                                     -3-

     first actuation prediction which is calculated from standard (normal 
     variate) formulas, (2) a term due to the uncertainty in the 
     proportionality 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 maximum-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 maximum subsequent actuation 
     pressure and its standard deviation by employing standard tables of 
     so-called "tolerance factors." These tables are entered with three 
     quantities: (1) n, the number of sample data points from which the 
     estimate of the mean and standard deviations are obtained. GE has set 
     n x 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. 

     The 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 
.

                                     -4-

     method mentioned above with the following exceptions: 

     1.   The calculation which 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 method 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% 
     confidence level are given on Table 1 which is attached. 

III. EVALUATION SUMMARY 

     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 
          containments 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 maximum pressures 
               resulting from each valve will occur simultaneously. We 
               believe that the assumption is conservative since different 
               lengths of line and SRV pressure set points will result in 
               the occurrence of maximum pressures at different times and 
               consequently lower loads. 

     3.   The proposed load criteria, which are provided on the attached 
          Table 1, are acceptable. The criteria were established by using 
          95-95% 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 95-95% 
          confidence limit, approximately 1% of the number of RSV actuations
          may result in containment loads above the design value. We 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 SRV's for the 238-732 standard plant, namely, 
          one SRV at a pressure set point of 1103 psig, 9 SRV's at 1113 
          psig, and the remaining 9 SRV's at 1123 psig. Only one SRV is now 
          set at the lowest pressure set point. Based on this pressure set 
          point arrangement for the 19 SRV's, GE has analyzed the most 
          severe primary pressure transient, i.e., a turbine trip without 
          bypass. Results of the analysis shows that initiation of reactor 
          isolation will activate all or a portion of the 19 SRV's which 
          will release put the stored energy in the primary system. 
          Following the initial blowdown, the energy generated in the 
          primary system 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 minimum of 62 seconds and increasing with each actuation. The 
          time duration of each blowdown decreases from 51 seconds for the 
          initial blowdown and decreases to 3 seconds at the end of the 
          period of subsequent actuations which is 30 minutes after 
          initiation of 
.

                                     -7-

          reactor isolation. 

          The staff finds the result of the GE analysis reasonable. 
          Therefore, the assumption of only the lowest set SRV operating in 
          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 III 
     containment. The test results, therefore, could not be applied 
     directly. Though the quencher loads for the Mark III appear 
     conservative in comparison with the test data, some degree of 
     uncertainty is acknowledged. The uncertainty is primarily due to a 
     substantial degree of scatter of all test data. We therefore will 
     require in-plant testing. 

IV.  REGULATORY POSITION 

     It is our position that applicants for Mark III containments using the 
     criteria specified below: 

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

                                     -8-

     2.   The quencher loads as specified in Item 1 above are for a 
          particular quencher configuration shown in the topical reports 
          NEDO-11314-08 and NEDE-11314-08. Since the quencher loads are 
          sensitive to and dependent upon the parameters of quencher 
          configuration, 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 parameters should be equal to or less than 
               that specified below: 

               Total air volume in each SRV line (ft3)      56.13 

               Distance from the center of quencher 
                 to the pool surface at high water 
                 level                                      13'-11" 

               Maximum pool temperature during 
                 normal plant operation (F)            100 

          c.   The value of those key parameters should be equal to or 
               larger than that specified below: 

               Water surface area per quencher (ft2)        295 

               SRV opening time (sec)                       0.020 

     3.   The spatial variation of the quencher loads should be calculated 
          by the methods shown in Section 2.4 of the topical report 
          NEDE-21078. 

     4.   The load profile and associated time histories specified in Figure 
          A5.11 of NEDO-113/4-08 should be used with a quencher load 
          frequency of 5 to 11 Hz. 
.

                                     -9-

     5.   For the 40 year plant life, the number of fatigue cycles for  the 
          design of the structures affected by the quencher loads should not
          be less than that specified in Section A9.0 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 containment 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 the in-plant test results, reanalyses should be performed
          to ensure the safety margin for the structures, which include the 
          containment wall, basemat, drywell walls, submerged structures 
          inside the suppression pool, quencher supports and components 
          influenced by S/R loads. If the analysis indicates that the safety
          margin for the structures will be reduced because of the 
.

                                    -10-

          new loads identified from the test, modification 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 quenchers for S/R 
          valves should submit a licensing topical report for approval. This
          report should present a test program and identify the feasibility 
          of modification or strengthening of the structures. 
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