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rpr ? 1 1977 DOCKET NOS.: 50-275 and 50-323 APPLICANT:
PACIFIC GAS AND ELECTRIC COMPANY (PG&E)
FACILITY:
DIABLO CANYON NUCLEAR POWER STATION, UNITS 3 AND 2
SUBJECT:
Su!9%RY OF MEETING HELD ON MOVEMBER 31,1977 TO DISCUSS DIABLO CANYON SEISMIC DESIGN - COMBINED LOAD ANALYSIS OF REACTOR COOLANT SYSTEM We met with the applicant on November 11, 1977 in Bethesda, Maiyland to discuss the analysis of the reactor coolant system with earthquake loads and loss-of-coolant accident loads combined.
A list of attendees is provided in Enclosure No.1.
PG&E was reanalyzing the plant's seismic design capabilities to detemine what modifications might be necessary to withstand an earthquake with a reference horizontal ground acceleration of 0.75g as compared to the original value of 0.4g.
PG&E was also performing a dynamic, time-history, non-linear analysis of the reactor coolant system to resolve mcent generic questions about the effect. of a pipe break at the reactor vessel nozzle safe end (asymetric loads).
PG&E believed that an earthquake at the design levels was so unlikely to cause a loss-of-coolant accident (LOCA) that earthquake loads and LOCA loads <need not be combined in the reevaluation of the reactor coolant system. Nevertheless, we had requested that they be combined and Westinghouse was perfoming the combined load analysis for EG&E.
The analysis had progressed to a detemination concerning the two critical points in the reactor coolant system. Although the detailed i
evaluation of the remainder of the system was yet to be completed, Westinghouse believed that such detailed analyses would indicate that existing reactor coolant system design was acceptable.
The two critical points in the system appeared to be the reactor pressure vessel supports and the reactor fuel spacer grids. The critical loads ocoured when earthquake loads were combined with asymmetric loads from a pipe break at the reactor vessel nozzle safe end. The msults for these two points and the status of the entire l
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reactor coolant system analysis are summarized in a status report provided by the applicant at the meeting (Enclosure 2).
For the reactor vessel supports, absolute combination of earthquake and LOCA loads gave a load about 1% over allowable. Since such a combination is conservative and since the allowable is 80% of the maximum load measured in the test of a support, the applicant believed this result to be acceptable.
For the fuel spacer grids. Westinghouse had perfonned tests and determined a lower bound value (95/95) at which pennanent deformation of the grids would occur. For the Diablo Canyon analysis, peak earthquake loads were calculated to be 86% of the test strength and peak L96A loads were calculated to be 40% of the test strength.
Absolute combination of peak values would give a load in excess of the test strength. Separating the loads or using a square root of the sum of the squares combination of peak values would give a load
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less than the test strength. However, we had been requiring that calculated loads be less than 50% of the test strength to allow for uncertainties in analysis and so the calcu. lated loads were large enough to indicate that permanent deformation might occur, as measured against our acceptance criterion, for any method of handling the con 6tnation of loads.
In any event, the possibility of permanent deformation wascoidy indicated for fuel assemblies at the edge of the core since the calculated l'oads are less for assemblies closer to the center of the core.
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Westinghousehadperformedemergencycorecoolingsystem(ECCS) 98!**!""!
performance calculations assuming the hot fuel assembly had partially deformed spacer grids. They are summarized in slides peesented at the meeting (Enclosure 3). Two factors mitigate against a loss of coolable geometry, aside frem conservatism in the analyses or load combinations, even if permanent deformation of the spacer grids should occur. The first factor is that the deformation would be i
minor in extent. For example, a photograph shown by Westinghouse at the meeting indicated a minor amount of deformation for a spacer i
grid that was subjected 23 times to a test impact load great enough to cause defonnation. The second factor is that restraints have been installed that would severely limit the size of the pipe break for a nozzle safe end break. Among other thhngs, this ensures that the particular LOCA under consideration here is not limiting in the ECCS performance calculations.
The applicant intended to prepare formal submittals on these subjects along the lines indicated in Enclosure 2 and 3.
We raised the following questions and comments to be addressed inppreparing the submitt&ls.
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We would need a complete description of the analysis used to calculate reactor cavity pressure including sensitivity studies on the node arrangements.
2.
We were interested in whether or not asymmetric loads due to subcompartment pressure developed by LOCA's in other areas could have a significant effect on other areas of the reactor coolant system design, such as the steam generator support design.
The applicant did not believe such effects would be significant.
We agreed to discuss this further in the near future and/or review j
the applicant's rationale in the forthcoming submittal.
3.
Our Division of Operating Reactors (DOR) had recently completed a review of a similar LOCA analysis for another 4-loop Westinghouse plant (Indian Point 3), including independent calculations.
The independent calculations had shown the Westinghouse results to be conservative. OurDivisionofSystemsSafety(DSS)would review this naterial with DOR to detemine how much of this effort could be utilized dn the Diablo Canyon review. We agreed to inform the applicant of the results of this check. We might then need another meeting to discuss similarities and differences I
between the Indian Point 3 analysis and the Diablo Canyon analysis.
4.
Our review could proceed in parallel. For example, the Mechanical Engineering Branch (EB) could proceed with a review of the stresses in the reactor vessel supports at the same time Containment Systems Branch (CSB) was reviewing the reactor cavity pressure analysis used aa,an input to the stress analysis.
5.
We would need a full description of the tests establishing allowable loads for the reactor vessel supports.
i 6.
We would need justifications for the size of pipe breaks assumed and a description of the restraints.
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7.
We woul'd need a full discussionoof the adequacy of Oe reactor coolant system.
i 8.
With regard to the MJLTIFLEX code used to calculate hydraulic blowdown loads:
a.
We would need a comparison of results fvte using 10 mass l
. points and using 5 mass points.
j b.
We were interested in the effects of out of plane motion of the core barrel. We would check this item with the Indian Point 3 confimatory analysis.
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We would need details concerning the results including a l
time history of the largest displacement.
d.
We would need a statement that the calculations were performed in accordance with the conditions in the Safety Evaluation Report (SER) for the MULTIFLEX code.
9.
A response spectrum method of analysis had been used to calculate seismic loads and a time history method had been used to calculate blowdown loads. We were interested in comparing the methods and had previously issued a question on this subject. We could discuss this subject at our forthcoming seismic design audit but it would be better if the applicant cotid provide a submittal prior to the audit.
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- 10. The analyses discussed above would be completed prior to the tentative Mechanical Engineering Branch audit date (early December) but the reports would not have been submitted by then.
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Dennis Allison, Project Manager i
Light Water Reactoss Branch No. 1 l
Division of Project Management
Enclosures:
- 1. Attendance List i
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- 2. Status Report M
- 3. Slides i
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cc: Pacific Gas and Electric Company Mr. James 0. Schuyler, Nuclear Projects ATTN: Mr. John-C. Morrissey Engineer Vice President & General Counsel Pacific Gas and Electric Company 77 Beale Street 77 Beale Street San Francisco, Califon.ia 94106 San Francisco, California 94106 Philip A. Crane, Jr., Esq.
Mr. W. C. Gangloff Pacific Gas and Electric Company Westinghouse Electric Corporation 77 Beale Street P.O. Box 355 San Francisco, California 94106 Rittsburgh;ePe6nsyNania. ?1523Gm Janice E. Kerr, Esq.
Brent Rushforth, Esq.
California Public Utilities Connission Center for Law in the Public Interest 350 McAllister Street 10203 Santa Monica Boulevard San Francisco, California 94102 Los Angeles, California 90067 Mr. Federick Eissler, President Arthur C. Gehr. Esq.
Scenic Shoreline Preservation Snell & Wilmer Conference, Inc.
3100 Valley Center 4623 More Mesa Drive Phoenix, Arizona 85073 Santa Barbara, California 93105 I
Bruce Norton, Esq.
-j Ms. Elizabeth E. Apfelberg 3216 North 3rd Street 1415 Cazadero Suite 202 San Luis Obispo, California 93401 Phoenix, Arizona 85012 l
l Ms. Sandra A. Silver Michael R. Klein, Esq.
425 Luneta Drive Wilmer, Cutler & Pickering San Luis Obispo, California 93401 1666 K Street, N.W.
g Washington, D.C. 20006 i
Mr. Gordon A. Silver 425 Luneta Drive David F. Fleischaker, Esq.
l San Luis Obispo, California 93401 1025 15th Street, N.W.
Sth Floor j
Paul C. Valentine, Esq.
Washington, D.C.
20005 l
400 Channing Avenue Palo Alto, California 94301 i
Yale I. Jones, Esq.
100 Van Ness Avenu~e 19th Floor San Francisco, California 94102 Ms. Raye Fleming 1746 Chorro Street San Luis Obispo, California 93401 l
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ENCLOSURE 1 LIST OF ATTENDEES PACIFIC GAS & ELECTRIC COMPANY DIABLO CANYON MEETING NOVEMBER 11, 1977 NRC STAFF PG&E WESTINGHOUSE D. Allison H. Gormly T. Esselman I. Sihweil M. Williamson M. Torasso R. Bosnak J. Hoch W. Gangloff P. Chen J. Petron L. Gesinski F. Litton J. Bandstra V, Noonan W'. Milstead R. Kelly H. Levin V. Esposito P. Kuo R. Kemper G. Bagchi R. Mattu a
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ENCLOSURE 2 flovember 11, 1977 DIABLO CANYOff REACTOR PRESSURE VES?EL LOCA ANALYSIS STATUS REPORT 1.0 Introduction The purpose of this report is to describe the reactor pressure vessel (RPV) loss-of-coolant-accident analysis performed on Diablo Canyon and to summarize the results that are presently available. Also considered will be the effect of various methods of combination of the blowdown loads and the seismic loads.
T.hese analyses are being performed for breaks at the vessel nozzles and at the pump outlet nozzle.
The postulated pipe break at the reactor vessel inlet nozzle is, on the basis of similar p.lant analyses, generally the most severe break.
2.0 Reactor Vessel Analysis The analysis performed for the Diablo Canyon RPV loss-of-coolant-accident (LOCA) loading condition consists of a dynamic, time-history, nonlinear analysis.
The applied loads during a postulated LOCA at the RPV nozzle safe-end locations come from three basic types of loading:
internals forces due to pressure transients, external forces due to asymmetric cavity pressure, and mechanical loads on the nozzles
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due to reactor coolant loop response. The vessel internals loads result from a depressurization wave travelling into and around the vessel barrel and core.
These loads are calculated by considering the fluid-structure interaction of the water in the downcomer annulus between the vessel and the barrel and the beam vibra-tions of the barrel. The external asymmetric cavity pressure loads result from the mass and endrgy release from a postulated break at the nozzle safe-ends.
The pressure builds up higher on the nozzle and vessel shell in the vicinity of the treak than is experienced opposite the break, resulting in a net force on the vessel.
The reactor coolant loops also respond to the effects of the LOCA and this response is felt by the reactor vessel, especially the break release forces. These forces are summed as time-histories and are applied to the mathematical model of the RPV and internals.
The RPV model contains the vessel she.ll, barrel, fuel assemblies, thermal shield, and upper and lower internals. The horizontal response is modeled by use of beam elements, non-linear gapped springs, dampers and lumped raasses.
The vertical response is modeled using non-linear, gapped springs, dampers and lumped masses.
The horizontal and vertical models are coupled at the vessel.
The applied loads have been generated for a break at the RPV inlet nozzle ;afe end with a break opening area of 115 square inches and a break opening time of 1 millisecond.
This limited break area has been assured by addition of a piping restraint into the primary shield wall. This restraint limits the developed area of the postulated break and.is effective in reducing the severity of the event.
The RPV inlet nozzle location was analyzed first since it is normally the location that for A So-W A - too t
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gives the highest applied loads and the highest vessel response. The break opening area was calculated based on vessel and loop displacement estimates supplied by Westinghouse and has been verified to be conservative by the'results of the dynamic analysis.
The internals hydraulic forces have been calculated using a 5 mass model of the barrel for the fluid-structure interaction calculation in the MULTIFLEX computer code. A ten mass MULTIFLEX study has"been completed and the changes in the hydraulic force are not significant. The external cavity pressure loading was calculated using the TMD computer code.
The time history plot of the horizontal cavity pressure force is shown in Figure 1.
The RPV dynamic analysis was performed for a postulated break,at the reactor vessel inlet nozzle by applying the summed time history loads to the RPV model en the DARI-WOSTAS computer code in order to determine vessel and internals displacements and vessel support loads. The horizontal vessel displacement at the nozzle centerline displacement resulting from this analysis is shown in Figure 2.
The horizontal vessel motion was approximately 0.18 inches.
3.0 Loop Analysis and System Evaluation The results from the reactor vessel analysis for the vessel inlet nozzle break were used in evaluating the Reactor Coolant System. The vessel horizontal and vertical displace-g ments and rotation were applied in a time history manner in conjunction with the hydraulic loads in the loop piping to perform the Reactor Coolant Loop LOCA analysis.
This analysis provides primary equipment nozzle loads, primary equipment support loads, and piping stresses ~.
On the basis of previous similar plant analyses, it was determined that the reactor vessel supports and the fuel will be the most limiting components of the primary system.
These components were studied in detail. The loads used in the evaluation of the system were the seismic loads resulting from the DDE or the Hosgri earthquake and the results of the vessel inlet break analysis.
Based upon a fracture mechanics evaluation of the reactor coolant loop piping (to he issued shortly), combination of these loads is an unrealistically conservative event combination.
However, in response to an NRC request, the consequences of an absolute sunmtion of the seismic load and the LOCA loads will be discussed.
The vessel suppo'rt loads are used in the evaluation of the supports for the faulted condition.
The highest vertical support load time history plot is shown in Figure 3 i
and the i.arizontal support load per support is given in Figure 4 The maximum l
RPV support loads for seismic and LOCA are summarized in Table 1.
The horizonta!
allowable value was determined by scale model testing and analysis and the vertical i
load capacity was determined by PG&E.
As can be seen from the table, the seismic and LOCA loads are acceptable when considered independently and even an absolute combination of the DDE and LOCA loads are only approximately IS above the allowable.
The allowable does not represent the load at which failure will occur, but is a value based upon 80% of the maximum load measured in a test of the vessel support.
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The fuel grid loads were determined by applying the upper and lower core plate motions to a detailed fuel model consisting of 15 fuel assemblies representing one complete row of fuel.
This time history analysis provides displacements
'of the fuel and fuel grid loads.
The results will be presented as a percentage of a lower bound load (statistical 95/95 load) which represents the load at which permanent deformation of a grid will occur.
The peak fuel grid loads are at 86%
for seismic and at 40% for the vessel inlet break.
In performing a conservative absolute summation combination of the. seismic load and the LOCA load indicates that grid deformation may occur:
This deformation will occur only in an isolated area of the core (peripheral assemblies near the baffle plates) and this de-formation is acceptable from a standpoint of coolable geometry.
Other components not expected to be limiting have also been considered.
A preliminary evaluation was made of the nozzle loads applied to the primary components. Based upon a conservative absolute summation of seismic and LOCA loads, the nozzles appeared to be adequate when compared to faulted condition allowables.
Further and more detailed evaluation is expected to confirm this statement.
The stresses in the reactor coolant piping are acceptable.
The stresses in the reactor coolant pump and steam generator supports have not yet been calculated for the vessel nozzle breaks but they are expected to be adequate with evaluation.
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The CRDM's are evaluated by applying the vessel head notion to a mathematical model of the CRDM's.
The time history evaluation, although not yet completed for Diablo Canyon, is expected to demonstrate acceptability since the evaluation has been perfonned and shown to be ccceptable for other similar plants which had larger vessel motion than Diablo Canyon.
4.0 Summary The analyses completed to date include evaluation of the nest severe break for the asymmetric loads issue.
Evaluation of the components of the prinary system which, based.upon previous analyses, are the most highly loaded has also been completed.
These analyses indicate that the primary system of the Diablo Canyon plant is adequate.
It is expected that completion of the analyses will demonstrate that this is the case.
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ALLOWABLE (KIPS)
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' Vertical 3800 2526 3400 8000 5
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. ENCLOSURE 3 EFORED GRID IMPACT ON ECCS ITEMS CONSIDEED:
- 1. PERIPERAL ASSEMBLY (F O XY
- 2. BEAK LOCATION AND SIZE (IltET N0ZZLE/SMALL BEAL 0
- 3. APPENDIX K EVAUJATION MODEL (DNSERVATIVE APPLICATION l
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m l# T EVAllRTION WIE 17 X 17 RJEL PLANT
- 1. EDJIVALENT LENGTH INCEASED TO REFLECT FLOW ESISTANCE INC ASE (SATAN) IN Alttutu ASSEMBLY,
- 2. ASSLE HDT ASSEMBLY HAS DEFOR1ED GRIDS
- 3. CASES CONSIDERED: DEFOR% TION MAGNITUDES (D 1 R0W WITH 10% FAR + 0.6% ASSEMBLY FAR (ID 2 R0WS WITH 22% FAR + 2.6% ASSEMBLY FAR f4. HDT ASSEMBLY HAS 2 GRIDS DEFORED AS AB0VE.
- 5. STEAM COOLING DURING EFLOOD BEF0E PCT OCCURS,
- 6. AT EACH DEFORTD GRID LOCATION, MASS VELOCITY BEIRVIOR IS b
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it APPEM)IX K APPLICATION TO EFORED GRIIE OCTTER 1975 VERSION OF THE WESTIER)USE EVALIRTION MODEI.,17 X 17 FUEL, 3 LOOP STEAM COOLIE PLM
- FEE, PCT,*F APCT, *F BASE 2181 1
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Ek PF0WlISTICBREAK LAEEST IPPACT LDADS AE CALCl1ATED FOR A BEAK AT INLET t0ZZL BREAK AEA.
2 ASSlNIE 144 IN BEAK AEA:
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3 LOOP 17 X F PlM, OCr0BER 1975 F0 DEL LIMITING ECLG - 2067*F
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2 1 FT BEAK - 1670*F WCAP-8566-A 4 LOOP 17 X 17 PLANT, MARCH 197510 DEL LIMITING ECLG - 2178*F 2
1 FT BREAK - 2050*F l
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RFl QNSIDEPATIONS CDRE BARREL PDdlBIT RESULTIE Fim LOCA/ SEISMIC EVENT DEFORMS 2 GRIDS OF PERIlhERAL FUEL ASSEMBLY TO TE MAXIltM EXTENT CALC 11ATED BASED ON GRID TESTIE (2 RDWS, 22% FAR),
FOR 17 X F PLANTS, t0 CLEAR DESIGN SAYS Fxy < 0.95
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F Q g < 0,95* Fg TE BENEFIT IN PCT IS ESTIMATED TO BE, PER WCAP-8986, (5,2*/.01)
(,115) = 60*F FOR AN httLitu PLANT AS-BUILT PEl_LET AVERME TBfERATUE IS 40*F LESS THAN 17 X F GENERIC FUEL VALUE PCT BENEFIT = 10-15*F SWNING UP PCT EFFECTS IN PERIPERAL ASSEMBLY, BLOCKAGE FROM DEF0ff% TION
+20*F FxyREDUCTION
-60*F RLLET TEMPEPATURE
-10*F OEPAIL PERIPHERAL ASSEMBLY PCT < 2200*F C&BINING E0iANISTIC BPEAK AND FUEL C04SIDERATI0tlS GIVES AN EVEN L0hER PCT
- VALUE, FbTk Ro-39/
A im-