ML20147F650
| ML20147F650 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 10/06/1978 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1546, NUDOCS 7810200095 | |
| Download: ML20147F650 (56) | |
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P d."U Ld. L U L L s.'Q td r4ItRTIES OF THE DIABLO CANYON i
SUBCOMMI7 TEE f4CETI!;G NASHING' ION, D. C.
${(jQj - jf Qh JUNE 14-15, 1978 fO A lof6(77 A rrcoting of the ACRS Subcommittee on the Diablo Canyon Nuclear k)wer Station was held in Washington, D. C. at 1717 11 Street, N. W. on June 14, and June 15, 1978. She purpose of the meeting was to review the status of the seismic design criteria and status of the seismic re-evaluation.
Notice of the meeting appeared in the Federal Register Vol. 43, No.104, May 30, 1978. 7he schedule for discussion and a list of attendees at the meeting are attached. Written statements were received from members of the public shown on the attached list.
No request., were received from members of the public to make oral statements.
The Subcommittee did not issue, approve, or receive any written reports during the meeting.
A copy of vu-graphs shown during the meeting is attached to the office copy of these minutes.
EXECUTIVE SCSSION (8:15 a.m. - 8:30 a.m.- JUNE 14, 1978)
The chairman summarized the status of the seismic re-evaluation and said that the meeting had two purposes: 7b review the status of the scismic design criteria and the status of the seismic re-evaluation.
He said that this meeting would not review new unresolved items relating to gen-cric issues or non-seismic issues.
He said also that the Subcommittee would again review the seismic design criteria, formulated two years ago and looked at several tines since. There are no outstanding issues relat-ing to the seismic design criteria; the outstanding issues relate to:
1.
Seismic re-evalnaHon and 2.
non-seismic matters.
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2-The chairnan noted that the Applicant has responded to the ACRS consul-tants' reports but that the IEC Staff has no formal mechanism for respond-ing to ACRS consultants' comments, thus making response difficult.
The ACRS has not yet reached a conclusion on the seismic design criteria, but hopes to do no in the near future.
MEUTIIG UITil THE APPLICANT AND Tile NRC STAFF (8:30 a.m. - 6:07 p.m.)
ItfrRODUC'10RY STATEMENT BY MR M. H. FURBUS!!, GENERAL COUNSEL, PG&E Mr. Furbush said that PG&E believes the project is ready for a decision.
There are still some outstanding issues remaining, but they can be address-ed in a manner similar to those in other ACRS reports. The review of seismic matters parformed by IG&E and the NRC Staff has been unprecedented and complex.
Judgement used in arriving at criteria and solutions has been based on vast experience and reputation.
11e noted that there will always be differences in judgement among various people.
lie closed his statement saying that PG&E would do all they could to place the ACRS in a position to make a favorable recommendation for the project.
OVERVIEW OP NRC STAFF REVIIM, D. B. ALLISON, NRC STAFF The Applicant does not fully agree with the NRC Staff design criteria, but has acquiesced to them and has re-evaluated the plant against the Staff design criter:
In situations where the original design criteria were more limiting, the Appli-cant has chosen not to take credit for the lower llosgri earthquake loads.
The basic criteria are a 7.5M earthquake on the llosgri fault with an effective acceleration of 0.75g for the input to design analysis.
Some reduction in ground response spectra due to building size effects was allowed (called the i
S
, '/
effect).
Ibgulatory Guide 1.61 damping values were used.
The one exception to Reg. Guide 1.61 criteria is the reactor coolant system where Westinghouse has justified, by test, 4% damping instead of 3%.
NRC Staff review has been completed through Amendment 57 to the FSAR (dated 1-23-78) and other information submitted through 3-27-78.
The NRC Staff be-lieves that the outstanding issues can be resolved to their satisfaction.
In addition to normal equipment and structures reviewed for seismic design, the Staff also reviewed and is requiring the systems, components, and instru-mentation needed for cold shutdown assuming loss of offsite power be quali-fied for the seismic criteria.
Diablo Canyon structures are founded on rock, and fixed base analytical mod-cls were used in the seismic re-evaluation.
This is the normal procedure for rock sites although a soil-structure analysis was performed originally.
Containment analysis was performed using the sum of absolute values of normal, seismic, and LOCA loads.
Seismic loads were combined by the square root of the sum of the squares.
The Staff has accepted the use of average test results for some structural materials rather than code specified minimum values.
Ductility ration up to 1.3 were approved.
She only two places in the plant where the Applicant could claim' credit for ductility are in the piers in front of the intake structures and on the ends of the turbine building.
l For mechanical systems, seismic loads were combined by adding one horizontal e
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4 component'and the vertical component.
A sensitivity study is being pei formed of this method vs the square root of the sum of the squares of thr(
components.
Electrical equipment that is being requalified is being quali-1 f
fied to IEEE-344, 1971.
The Applicant will install a seismic scram.
Because the seismic scram is tied into the reactor trip system, the NRC Staff will review it for l
comformance with class IE circuit requirements.
l i
l NRC STAFF REVIEW OF SEISMIC DESIGN CRITERIA l
DAMPING, P. T. KUO, NRC STAFF 1
1 l
The Applicant used 5% damping for reinfort d concrete structures in the original l
analysis for the DDE and 7% damping for the re-evaluation as permitted by Reg. Guide 1.61.
The NRC Staff reviewed the data and concluded that 7%
is a reasonable estiinate.
Dr. Kuo illustrated the conclusion by showing re-sults of several tests and analyses of building response measurements l
l
( Figs.1 and 2 ).
All testing included effects of soil structure interaction, l
l material damping, and system damping, and not strictly structural response.
Ibmping values as illustrated in Dr. Kuo's presentation varied from 1.5%
to 16% and showed an increase with amplitude.
The data were from ordinary structures as well as from in situ testing of containment structures at low-level respenses.
In response to a question from the Chairman, Dr. Kuo said that he believes the tests illustrated also represent a range of stress levels.
He also said structural damping cannot be separated from other damping.
The 7% damping may be high for struc-tural damping, but it is adequate for the analysis because it takes into account other damping not included in the analysis.
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Dr.'Newmark commented that if the reinforced concrete structure remains uncracked and stresses are low there is very little damping in it.
With any stress of the steel approaching working stresses damping goes up to the order of St and when stresses approach the design allowable stresses, damping becomes in excess of 7%.
1 I
Dr. Kuo showed a slide illustrating computed stresses under-various load combinations at several points in the containment (Fig. 3 ).
In answer to Mr. Bender's question, he said that the results illustrated are representa-tive of the bulk of the structure. They showed calculated seismic reserve capacity above the yield stress varying from 123% to 161.
Seismic reserve capacity is defined as (allowable stress minus non-seismic stirss) divided by seismic stItss.
Dr. 'Kuo also showed a slide illustrating stresses in the auxiliary building
( Pig. 4 ).
Stress levels for the combined loads varied from 43% to 92% of the yield stress.
In the cases of the intake structure, turbine building and diesel generator building,. analysis showed them to ba overstressed. The Applicant is making modifications to these structures and they are still under review by the NRC Staff.
Dr. Kuo showed results of a sensitivity study performed by Dr. Newmark for changes in damping factors from 5% to 7%.
The results indicate that by changing structure damping from 7% to 5% and keeping equipment damping at 31, structural response would change by 11.3% for frequencies greater than 1.5 hz.
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PR3SENTATION ON DAMPING BY DR. J. BLUME Dr. Blume presented a summary of damping value tests in structures and structure models. !!e said that the models (made by the Portlant Cement Association) represent shear walls and shear wall action that predominate in the structures at Diablo Canyon. The results for the average damping vs. stain level showed damping values increasing from about 2% at first cracking to about 9% at or near yield.
Ile indicated that the 7% used for Diablo Canyon falls far below the yield value.
1 Dr. Blume also described a parametric study performed by URS/Blume to investigate the response of equipment high in the structure to a small earthquake. For the examples illustrated, no combination exceeded the basic criteria of.75. 7% damping.
Considering that Dr. Newnerk's study 9
was based on an average value plus one standard deviation and Dr. Blume's work was for simple models and based on peak values, th two studies seem to compare favorably.
ADEOUACY OF CRITERIA FDR SAFE OPEIMTION OF TilC PIANT, N. NEM1 ARK, NRC STAFF CONSULTANT Dr. Newmark pointed out that engineering design is an art, and an earthquake design engineer must take fato account various kinds of information from various cources.
Ile said for many structures that have to be designed to prevent loss of life (hospitals, apartment buildings, etc), current criteria are 8 to 20 times less conservative than for a nuclear power plant.
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, lie. continued by saying that experience with blast, shock and earthquakes must be taken into account along with theoretical calculations in design-ing earthquake-resistant structures.
Ile cited some examples of structures that have withstood earthquakes larger than designed for.
He said that not enough is known about source and propagation mechanisms to accurately predict what htppens in an earthquake.
Ele also cited some examples of inaccur-ate response of instruments, especially those for vertical motion.
Dr. Newmark feels that maximum acceleration is practically independent of magnitude close to the earthquake source.
At a distance from the source, magnitude has an important effect because it defines how much energy is focused tc the structure. Velocities that are generated by motions at the source are limited by fracture strain to 5-6 fps.
In his opinion, close to the source, velocity is a better measure of what happens to the response spectrum than peak acceleration.
An effective peak accelera-tion, which is consistent with velocity, can be inferred from the energies available.
lie continued by saying that the use of an effective acceleration different than the peak can be determined on a qualitative basis, but should only be used close to the cource.
Although uncertainties are great, observa-tions of damage are consistent with a lower effective acceleration.
Dr. New-mark concluded by saying that, based on his experience and judgement, 0.75g is amply conservative.
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- Dr. Trifunac questioned the philosophy of using a magnitude that may be unreasonably high and then being forced to justify so many things.
He suggested using proper judgement and a more formal approach on which every-one agrees.
Dr. Newmark responded that he doesn't believe magnitude is particularly important when the earthquake occurs very close to the plant. Ile also said that he does not use a formalistic method to get from magnitude to accaleration, but uses what he thinks is a reasonable l
basis for acceleration level and duration to obtain an appropriate effec-
'tive acceleration.
TAU AND 70RSION EFFECTS, N. NEM MRK, NRC STAFF CONSULTANT Dr. Newmark described the tau effect as a variatio.n in acceleration over an area because of nonhemogeneity in the ground and wave propagation effects.
The tau offect was developed by Yamahara in 1970 to try to explain rela-tive damage among structures in a Japanese earthquake.
The tau offect is one way of trying to account for relative damage, and its validity depends on comparison with observations.
Dr. Newmark said he knew of only two examples of the use of the tau effect.
Shere will always be a torsional effect connected with earthquakes and it is generally agreed that some allowance must be made for it.
Damage from torsional effects has generally been confined to structures where there is come eccentricity or where all of the torsional resistance is concentrated in a central core.
In reactor structures, the effect is not as important
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as'it is for a structure in which all of the torsional resistance is concen-trated around a central core.
Dr. Newmark believes calculations made for torsional effects should be taken into account qualitatively ' but not quantita-tively.
In studien made by Dr. Newmark and his associates, using time histories from several earthquakes, good agreement was found among various combinations of translational and torsional effects.
In general, the torsional frequency is different from the translational frequency, and the maximum effect in time agrees well with the square root of the sum of the squares of the combinations.
In response to a question from Dr. Trifunac, Dr. Newmark said that the only other effect present is tipping and this is not significant for Diablo Canyon.
In subsequent discussion with Dr. Trifunac, Dr. Newmark said that the tau effect is already included in earthquake records made in building foundations.
For this reason and because there are very few free-field records in the USA, Dr. Newmark would not use the tau effect except very close to the source.
ACRS CONSULTANT COelENTS ON THE DIABID CANYON DESIGN BASES Dr. Thompson: The evidence is fairly strong that the degree of motion on the l
Hosgri Fault has nearly died out in the last few million years.
Because of the segmented discontinuous appearance of the Hosgri Fault, it seems to Dr.
Tiompson that the breaks on the Hosgri Fault have been relatively short.
In his opinion, a magnitude 7.5 carthquake is very conservative.
Dr. Trifunac: On the question of spectra he cannot say whether it is safe or not; in his judgement, the 7.5 magnitude earthquake is unduly conserva-tive, using the standard approach of 10CFR 100, Appendix A.
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Dr. Trifunac does not understand what is meant by effective peak accelera-tion and has difficulty accepting the term.
lie feels that the tau effect has not been accepted in the scientific community and has been used in an over-simplified manner, lie would prefer to do a 3-dimensional soil-structure interaction analysis, which he feels is a simple and more direct approach.
Dr. Trifunac is still not convinced that 7% damping represents good judgement.
lie bases this on the lack of good experimental evidence on damping Associated with structures.
Dr. Maxwell: % e assumption of magnitude 7.5 earthquake is extremely con-servative. !!e feels that extending the llosgri Fault for 400 kilometers and putting the 1927 ear thquake on it is artificial.
Dr. Page:
Dr. Page is saticfied with the 7.5 magnitude earthquake and he thinks the llosgri Pault is 400 km long and has had 80-115 km of displace-ment.
Ilowever, most of the displacement has occurred between 25 X 10 6 6
and 5 X 10 years ago.
Ile thinks that the normal stress across the llosgri Pault is not an' great as it was once and stress build up now would not be as great as it might have been. Therefore, Dr. Page thinks it unlikely that a magnitude 7.5 earthquake will occur during the life of the plant.
Although the implications of an earthquake of that size on structures is be-yond his field of knowledge, Dr. Page thinks there is validity in the con-cept of effective acceleration based on what he has read and experienced about damage after an earthquake.
. DN.LUCO:
Dr. Luco does not feel that he is qualified to decide if the 7.5 magnitude is adequate.
Ilowever, in his cpinion the design spectra adopted are 20 to 30% too low for a magnitude 7.5.
Dr. Luco feels the spectra reflect a magnitude 6.5.
For the effective acceleration, Dr. Luco feels that the real inelastic response of the structure is being forced into a linear elastic mode, lie thinks it is not rational to reduce input; instead he would rather change the method of analysis if the level of damage is not as high as predicted.
Dr. Luco believes that the reduced effective acceleration should be eliminated and that the reduction of 20 - 30% for the tau effect is too high, even pre-suming the waves are horizontal.
Also,'the torsional effects are too small; the eccentricity should be closer to 25%, not the 5% used.
Dr. Luco recommended that a complete 3-dimensional soil-structure analysis be undertaken.
He does not feel that the fixed inse model with the tau effect is adequate.
Ile also indicated that he feels the 7% damping used for concrete structures should be reduced.
Dr. Pickel: '1he concept of effective acceleration should be considered in terms of a few single peak accelerations that might be spaced out.
Therefore, the response spectrum would be less than with smaller peaks.
The same results would be expected for atx>ut 1.16 ductility ratios and 5% damping, as for no yle] ding and 7% damping.
In his opinion, there is no question that the structure will survive given the spectrum and 7% damping.
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t DR. WillTE:
11e said that he has no concerns about the structural behavior.
Ilis earlier concern about the effect on equipment for smaller earthquakes and reduced damping has been satisfied.
REFORT ON RESEARCII AT IAWREt!CE LIVERMORE LABOlWlORIES ON SEISMIC MARGINS IN OPERATIt!G PIA!TrS.
P. SMITil, LLL Dr. Smith sumenarized the results of work performed at LLL to quantify the seismic conservatisms of operating reactors.
IIe showed results in three areas of the study:
1.
Broad band spectra (RG 1.60) 2.
Enveloping artificial time histories 3.
Floor spectra in 3-dimensional structures.
For the broad band spectra, the study calculated what Dr. Smith called the factor of conservatism for the Reg. Guide 1.60 spectrum.
'Ihe factor of con-servatism is defined as the RG 1.60 spectra divided by the spectra from 45 real earthquakes.
In some cases, factors of conservatism esem to be excessive (up to 1.43).
After much discussion, it became apparent that the data used by LIL was corrected data from that used for Reg. Gui6a 1.60.
Some conservatisms are in the Reg. Guide spectra because of the way they were drwn and were normalized.
In some plants, Reg. Guide 1.60' spectra are not used directly.
To try to show that, on the average, the methodology is conservative, LIL obtained 20 artificial time histories and compared them with the Reg. Guide 1.60 spec-tra. 'Ihe artificial time histories matched the Reg. Guide spectra fairly well
on average. She Chairman pointed out that there is no average plant so the use of an average in this situation is not too useful.
LLL also did a study on a 3-dimensional structure.
In this study, the square root of the sum of the squares for the Reg. Guide 1.60 artificial time histories were combined,and the average of 23 sets of real time histories were combined.
The factbr of conservatism was computed by taking the ratio of the two results at different frequencies at various points in the structure.
Dr. Smith said that the results show how conservatisms compound for floor spectra since the factor of conservatism for the 3-dimensional case analyzed ranged up to 2.1 as comparbd to 1.1 fdr the envelope of the artificial time histories.
NRC REVIEW OF STRUCTURAL RE-DVALUATION, P. T. KUO, NRC STAFF The NRC Staff reviewed, in detail, the following major structures against the seismic criteria:
1.
Containment 2.
Auxiliary bul.1 ding 3.
Intake structure 4.
Turbine building 5.
Outdoor water storage tanks and diesel oil fuel tanks.
6.
Buried piping Dr. Kuo summarized the NRC Staff review of each of the six structures:
De-tails are as follows:
Containment:
Dr. Kuo showed slides of the vertical and horizontal response spectra used for the re-analysis. (Figs. 5 and 6 )
The Applicant used a com-bination of the Blume and Newmark spectra depending on which is the most crit-ical. She horizontal response spectrum was based on tau = 0.04 and anchored at
, O.69
%ere was no tau effect included in the vertical spectra.
Dr. Kuo next showed a slide (Fig. 7) summarizing some of the results obtained for the containment exterior for both horizontal and vertical response.
Ratios of Ilosgri accelerations to original DDE analysis for various elevations ranged from 0.71 to 1.28 for the horizontal, and 2.04-5.93 for the vertical.
The reason for the large increases in the vertical response is because a dynamic a-nalysis was used in the Ilosgri analysis.
%e resulting stresses in the contain-ment are, in all cases, less than the allowable stress.
In response to a questicr.
from Mr. Bender, Mr. Sibweil said that, if the tau effect had not been used, there would still be some margin in stress (except at one point), but the margin would be reduced.
In answer to a question from Mr. Etherington, Dr. Kuo said that the Applicant assumed that the maximum II)CA force and scismic force were coincident.
%c Chairman noted that realistically, the seismic and II)CA loads would not occur simultaneously and combining them is a form of conservatism on which everyone agrees.
There are still nine areas outstanding (Fig. 8) in two categories involving methodology and requests for additional work.
W e NRC Staff expects these itmos wil.1 be resolved.
AUXILIARY BUILDING For this case, tau =.025 was used and the spectrum was anchored at 0.56.
9 Results obtained for the auxiliary building showed the same trends as the containment (Fig. 9).
Were are three outstanding areas (Fig. 10) remain-ing in the review of the auxiliary building; the NRC Staff expects to re-e
solve these shortly also.
TURBINC BUILDING Because the turbine building houses seismic category I equipment, it was re-evaluated for the Ilosgri earthquake, and modifications were made to the 4 kV switch gear room, diesel-generator engine room, and component cooling water heat exchanger room.
She response spectrum used for the turbine building re-analysis was ahchored at 0.5g. Several modifications are being made to the turbine building as a result of the reanalysis.. Six areas in the NRC Re-view remain outstanding (Fig.ll), and the NRC Staff expects to resolve these items. In answer to a question from Dr. Luco, Dr. Kuo said that a damping value of 7% was used throughout all of the buildings.
Mr. Seawell added that in the case of welded steel structures, a value of 5% damping was allowed.
INTARC STRUCTURE e
The intake structure was also re-evaluated because it houses seismic category I equipnent.
In the response spectra used for the re-analysis, a tau value of 0.04 was used, and the spectrum was anchored at 0.69 Near the seaward end, the stresses exceeded the capacity of the piers, and the NRC Staff requested further analysis of the seaward piers.
She NRC Staff suggested there are two ways Ulat the Applicant might demonstrate that the piers are satisfactory:
one is to take credit for structural ductility, and the other is to show that the circulating water system would work even if the pier failed.
Mr. Ghio replied that the Applicant is submitting both approaches.
Shere are four outstanding areas remaining in review of the intake structure (Fig.12).
'1ho NRC Staff expects to be able to resolve these four items.
% N
OUrDOOR S'IOTAGE TANKS The five water storage tanks were modified because.the NRC Staff was not con-vinced that deflections occurring on the base slab would be within allowable limits. 'Ihe re-analysis was made with no tau effect. and with the spectrum anchored at 0.75g with 7% damping.
The tanks were modified by anchoring the bases to the rock, and by placing an 8-inch concrete shell around the tanks.
The NRC Staff is still reviewing the Applicant's re-evaluation of the outdoor storage tanks.
The two fuel oil storage tanks, which are about 5 feet below the ground surface, were re-analyzed (FLUSH analysis), and a de-convolution procedure was used.
The Applicant also used procedures proposed by Dr. Newmark for the burst pipe analysis, that has been accepted by the NRC Staff.
The results from the FLUS I analysis were more conservative than the burst pipe analysis.
This item is stil under review by the NRC Staff.
NRC REFORT ON THE STATE OF CALIFORNIA GEOLOGICAL INVESTIGATION, R. MC MULIEN, NRC STAFF The California Division of Mines and Geology has been investigating in the Diablo Canyon site area. They had noted an early map of the geology of Cali-fornia showing a fault coincident with the San Miguelito Pault zone ex-tending across the peninsula on which the site is located, from the San Luis Obispo area to Sierra Bay.
A map published by Dr. Hall in 1973 showed the San Miguelito Pault zone t'erminating at a point about 2 miles east of the Diablo Canyon site. The California Division of Mines and Geology took it upon themselves to investigate this conflict, and their mapping confirmed that the fault, based on all the evidence, seems to end 2 miles east of the Diablo Canyon r*
. During their investigation, they found anomalies suggesting faulting in two locations on the seacliff. One is about 14,500 feet west-northwest of the site (#1), and the other 500 southwest of the first (#2). NRC Staff members along with members of the USGS and the California Division of Mines investigated the area and concluded that the # 2 area does not pose a threat to the site because it is discontinuous and does not affect the overlying terrace surface or the marine terrace deposits overlying it.
Mr. Page commented that he believes examination of power plant site should be an ongoing project both before and after granting of a license.
He added that the San Miguelito Fault should be. watched and studied.
Dr. Thompson said that it is impossible to resolve these potential problems on the basis of what Mr. McMullen presented.
Dr. Maxwell indicated that he feels there is nothing in this exposure to worry about.
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CRAND ANALYSIS, D. ALLISON, NRC STAFF The Applicant has committed to perform seismic analyses of the cranes and to submit a combination of operating restrictions and analyses to show that there wi]l be no safety problems from the cranes.
NRC Staff review of these analyses is not complete.
Mr. Bender asked if the NRC Staff would consider a
. time restriction on the use of cranes with large loads as a possible alter-native.
Mr. Allison replied that except in the case of the turbine missile study, the Staff is reluctant to utilize probabilistic analyses.
In the case of spent fuel pool crano, the Applicant has proposed an operating restriction limiting the crane to a one-ton load, and has sulnitted a crane analysis for the one ton load.
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Mr'. Hoch commented that the cranes have been qualified without loads for the Hongri event. Ile said this is essentially standard procedure for plants and the NRC Staff has now backfitted a p rtion of the Reg. Guide on cranes. The Applicant has agreed to analyze the cranes for the Hosgri carth-quake under load, has completed the analysis for two of the cranes, and is pro-coeding with the analysis for the remaining cranes.
The Applicant has asked the NRC Staff to reconsider its requirement, but if the NRC Staff does not re-1ent, the Applicant will proceed with the analysis of the remaining cranes.
Mr. J. Pray, PG&E, described the probabilistic analysis for the turbine building crane. Mr. Fray showed a slide (Fig.13) illustrating the turbine building crane event tree under a large earthquaka load, with the crane falling and. damaging safety related equipment. Combining the probabilities for each of the events in the sequence leads to a probability of 1 X 10-13 of producing e
doses greater than allowed by 10CFR 100.
NRC STAFF REVIEW OF SYSTEMS NEEDED TO MAINTAIN COLD SHUTDDWN, S. ISRAEL, l
NRC STAFF Mr. Israel summarized the NRC Staff review of the systems that are needed to f
maintain cold shutdown in the event of an earthquake, and assuming loss of offsite power.
Previously, the NRC Staff has required only that hot shut-down be maintained with safety grade equipment, however, recently the R C Com:nittee has approved a new NRC Staff position requiring Applicants to have the capability to achieve cold shutdown using only safety grade equipe nt.
The R C has approved partial implementation.of this position for plants
]
i currently undergoing licensing and will require full implementation for new
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. L CP applications. The criteria under which the Staf f reviewed Diablo Canyon, are as follows: Credit can be given only for equipment qualified for the llosgri carthgeake, all offsite power is lost, the systems must meet the single failure criterion, and failure of non-qualified equipment must not affect the ability to achieve and maintain cold shutdown. The only system which did not nect the single failure criterion was the suction line from the hot-leg to the two R11R pumps. 3his line has two normally closed isolation valves. The only single failure that could present a problem would be the valve disc separating from the valve stem. The Applicant subnitted an analysis of the probability of this failure conincident with an earthquake, and the NRC Staff concurred -
that the probability of this single failure was lov enough to be acceptable.
The NRC Staff has found the provisions for maintaining and achieving cold shut-down to be acceptable except for one remaining issue dealing with the raw water reservoir. She condensate storage tank provides a supply of auxiliary feedwater for the steam generator for about 9 to 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> of operation.
The NRC Staff-required an extended cource of water supply.
The NRC Staff has accepted fire tanks and condensate transfer and storage tanks, which will provide a total capacity of about 450,000 gallons as the extended source of raw water supply.
Shose tanks, combined with condensate storage tanks, will provide in excess of thirty-two hours of operation of the steam generators.
There are still a few l
details to be worked out concerning the fire tanks and transfer and storage tanks, but the NRC Staff feels they can be easily resolved.
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NRC STAFF REVIEW OF Tile MECilANICAL SYSTEMS, P. CilEN, NRC STAFF Dr. Chen reviewed the status of the review at the time of the last ACRS meeting, and the history of the review procedure since that time. 'Ihe systems that the NRC Staff reviewed are: the safe shutdown system, piping and components, the reactor coolant system including components and supports, and some other systems affected by loss-of-coolant-accidents (such as the auxiliary feedwater system,-portions of the main feedwater system, residual heat removal _ system, etc.)
The NRC Staff also reviewed proposals for test-ing of equipment and reviewed the electrical instrumentation and control systems.. In order to conduct an efficient review,_ the Applicant was asked to complete Seismic Qualification Summary Forms.
These forms identify the compon-ent and indicate how the equipment was qualified
- that is,whether by test, anal-ysis, or a combination.
Dr. Chen showed examples of the seismic Qualifi-cation Suarnary Forms and their results.
The systems that the NRC Staff reviewed in this manner are the emergency diesel, auxiliary feedwater, steam generator atmaspheric relief valves, main steam isolation valves, portions of the component cooling water system and auxiliary salt water systems, residual heat removal system, portions of the chemical and volume control system and the pressurizer relief and spray system.
Dr. Chen concluded by showing a slide illustrating the portion of each of the components studied that has the least margin (Fig. 14).
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NRC REVIEM OF PIPING SYSTEMS -T. SULLIVAN, NRC STAFF Mr. Sullivan descrit>ed the review of the auxiliary feedwater piping from the condensate storage tank to the steam generators. 7he Applicant re-viewed the entire piping system and 64 supports to compare the stresses resulting from the llosgri spectra to the allowable. There are two high stress points in the piping of about 20,500 psi, which is about 76% of the yield strength.
Most of the piping stresses are about 9,000 to 10,000 psi, which is about 40% of the yield stress.
Mr. Sullivan showed several slides
- illustrating the 64 auxiliary feedwater train hangers that were analyzed, and the ratio of actual to allowable stress. The margins range from 0.02 for member bending to 0.99'for some snubbers.
Because of the low margins on some of the snubbers, PG5E had tests performed to determine.the difference between the rated load and actual failure load.
She test.s were performed on one hy-draulic and two mechanical snubbers.
For the hydraulic snubber, the tested failure load was 4.7 times the one-time load rating, and for the..iechanical snubbers, the failure load was about 21/2 times the faulted load rating.
Member bending was limiting in 31 of the 64 hangers analyzed, and snubbers were limiting in 10 of the 64 hangers.
Mr. Sullivan showed a slide illustrating the peak stress points in other piping systems (Fig. 15). She peak stresses, as a fraction of allowable stress, range from 0.211 for the containment h'ydrogen purge system to 0.999 for the contain-ment-spray system.
Mr. Sullivan said that only about 3% of the piping lines in the plant have stresses which exceed 70% of the allowable stress.
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.l In, answer to a question from Mr. Bender, Mr. Iloch said that in Unit 1, 902 out of approximately 2,615 large hangers and about 500 out of approximately 1500 small' hangers are being rrodified.
In response to a previous request from the subcommittee, the Applicant made a failure study of the upper steam generator support snubters and a study of other selected piping systems in the DOP.
'Ihe snubber study was performed assuming that one of the four snubbers on each steam generator was totally in-effective in resisting seismic motion. Results of the analysis showed that all stresses in the system were within the allowable limits.
Ioads on other snubbers in the main steam line did exceed catalog-rated loads by a small fraction, but they were within the load for which each of the snubtr.s had been tested.
The study of the BOP piping, assumed that. a single snubber failed to work during the earthquake.
Conclusions of this study were that the proba-bility was 10-7 for a pipe break occurring from.a single snubber failure during d
the Hosgri earthquake,and 10 for a pipe break occurring from a locked-up snubber assumi_ng 30 cycles of heat-tp per year.
Mr. Sullivan said that the NRC Staff feels that thesc' values are too optimistic and they areaskingthe$pplicantformoredetails.
Dr. Siess commented that it is his opinion, along with some ACRS consultants, that the question of snubber re-liability is a generic matter.
He said that there are reservations about the assumptions made regarding future snubber failures as compared to past snubber failures, because of the. lack of data.
Dr. Pickel commented'that he
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agreed with Dr. Siess and he feels the snubber study is a good first step in an area where there has not been a lot of work performed. He recommended that the NRC Staff continue snubber studies on a generic basis.
He feels that it is not clear whether there is much conservatism in the failure a-nalysis.
NRC STAFF REVIEW OF Tile COMBINED IDAD ANALYSIS OF THE REAC'IOR COOINTf SYSTEM-P. CHEN, NRC STAFF j
IDCA loads and seismic loads were combined for the three most severe cases in the spectrum of pipe breaks.
The three pipe breaks chosen control the loads in the reactor vessel internals and the control rod drive mechanisms.
In the unbroken loops, IDCA loads are combined with either the DDE or the Hos-q gri spectra, depending on which is the most limiting.
Four different load combin-ations are evaluated. They are: 1) normal plus seismic; 2) normal plus ID'A/*
3) normal plus the square root of the sum of the squares of the IDCA and seismic; and 4) the absolute sum of the normal, seismic, and LOCA.
Mr. Chen re-viewed the results of the load combinations (Fig.16).
'In a few instances, the absolute sum of the normal, seismic, anci IDCA loads exceeds the allowable.
Ilow-ever, the NRC Staff has determined that this is acceptable based on the allow-abic stress being less than the actual strength,which is greater than the code allowable.
VIBRATION TESTING PROGRAMS - P. CilEN, NRC STAFF
.The Applicant performed in-place vibration tests on eleven pieces of equi >-
t ment, and five valves. Te valves were tested on a shaker table.
Dr. Chen gave a brief summary of the results of the tes. ting, including the frequency,
experimental damping versus analytical damping, the acceleration Jevel l
- v. s.
at the resonant frequency, an] the Hosgri acceleration value.
Dr. Chen also discussed the Westinghouse vibration tests on the Indian Point No. 2 reactor coolant system. W'estinghouse submitted a report on these tests that was reviewed and approved by the IEC Staff in 1974.
Dr.
Chen showed a slide plotting damping versus deflection for various pieces of primary coolant loop equipment at several reactors (Fig.17).
It shows a The trend of increasing damping as the response of the equignent increases.
J response of the Diablo Canyon steam generator is expected to be about 2,000 mills, and that of the reactor coolant pumps about 800 mills.
Dr. Chen sug-gested that the plot, which includes what he called a linear-fit curve, justi-fies as reasonable thc 4% damping for the Diablo Canyon steam generators.
Dr. Chen also discussed the Applicant's vibration testing of electrical equip-i 1
ment.
bbot of the natural frequencies observed were above 33 hertz, and are considered to be rigid.
In instances where structures are not rigid, it was i
decided to requalify the equipment according to current criteria.
The Appli-cant requalified 22 items divided into 7 groups (Fig.18).
As a result of the requalification and re-evaluation, three pieces of electrical equipnent re-quired modifications for the Hongri earthquake (Fig.19).
For mechanical i
component qualifications,14 of the 61 components required for shutdown were modified, and eight of 1700 valves required modifications.
~~
i FUEL ASSEMBLY DEFORMATION FOR COMBINED EARTHOUAKE AND LOCA-R. PEYER, NRC STAFF There are two essential elem 7ts to the method of analyzing fuel assemblies for.
their ability to withstand the combined earthquake and UDCA loads.
One ele-ment is a computer analysis to calculate the forces in the fuel assembly, ard the other is the strength and deformatio'n properties of the fuel element grids, as measured in laboratory tests.
From the computer analysis, the NRC Staff has determined that the most critical component in the fuel assembly is the spacer grid.
Westinghouse submitted models nearly four years ago, for the response of fuel assemblies to loads from both earthquakes and loss-of-coolant-accidents.
At first, the.NRC Staff believed the vertical components were dominant; therefore, verification was performed by comparing predictions with tests made of vertical impacts on fuel assemblies. Khen the asymmetric load problem in the North Anna project developed, it became apparent 'that the horizontal component was im-portant.
Independent verifications from audit calculations performed by the Idaho National Engineering Laboratory have led the NRC Staff to conclude that the' Westinghouse model for calculating fuel assembly response is adequate.
Ilowever', because of uncertainties associated with the calculations, the NRC Staff has applied a safety factor on the order of 1.5 to 2 to the calculated load.
Mr. Meyer compared the analysis for North Anna with the analysis for Diablo Canyon.
For North Anna, the loads on the fuel assembly were calculated by the Westinghouse code and multipled by a safety factor of 1.74.
'Ihey were then i
compared to the 95-95 Jower confidence limit of the measured grid strength, i
4
Calculated loads were found to be less than the measured grid strength loads, and it was concluded that the deformations will be in the elastic range.
In the case of Diablo Canyon, the earthquake forces are larger and the calcu-lated loads on the fuel assembly without a safety factor are less than the measured grid strength, flowever, when safety factors of 1.5 to 2 are applied to the 95-95 lower canfidence limit strength, and the seismic and LO'A loads are combined, the measured loads on the outer two rows of fuel assemblies are larger than the grid strengths.
The major impact to the fuel assemblies occurs at the interface between the fuel assembly and the core inrrel.
The NRC Staff concludes that, for all fuel assemblies except those in the two outer rows, there is no permanent deformation. In response to a question from Dr.'Siess, Dr. Eccelman of Westinghouse said that the cold-leg-break gives the higher loads, and it is only the fuel assemblies adjacent to the blown nozzle that experience permanent deformation.
Uhere are about 23 fuel assemblies in-volved.
The NRC S,taff has not been able to agree with Westinghouse that the tests conduc-ted represent the Diablo Canyon loads.
7herefore, Westinghouse will use the max-imum theoretical deformation (22% flow area reduction) in the safety analysis for Diablo Canyon.
EFFF,CT OF SPACER GRID DEPOMATION ON PEAR CIAD TEMPERATURE-W. IlODGES, NRC STAFP l
Mr. Ilodges explained that a model is used to calculate the effects of reduc-tion in flow area, because there is no FLECilT-type heat transfer data for de-
't
formed bundles.
FV. Hodges reviewed the nine assumptions made for the cal-culations and presented the results. A deformed bundle at maximum power re-sults in an increase in peak clad temperature of 17 degrees F.
When the effect for a double-ended cold-leg guillotine break is added, the increase in peak clad temperature goes to 25 degrees.
However, there are several credits that can be used to offset the effect of the deformation, since only the outside fuel assemblies are affected.
Power in the deformed bundles dur-ing reload is lower than the hot bundle, so credit.of 15 degrees can be taken, and a 35 degree credit can be taken because there is a lower clad temperature at the beginning of reflood in the hot rod. The NRC Staff concludes that the peak cladding temperature for a deformed bundle vill be less than the peak cladding temperature for the bundle which has the hot rod.
3he bundle which has the hot r.>d will have no deformation, so the ECCS analysis made for an undeformed bundle, with the hot rod still represents the limiting peak clad temperature.
OPERATING BASIS EART110UAKE-D. ALLISON, NRC STAFF The OBE for this plant was originally 0.2.
The NRC Staff 9
considers the Hosgri carthquake to be the SSE and asked the Applicant to justi-fy the OBE in light of 10 CFR 100, Appendix A.
Based on a probability study, the Applicant conservatively estimated that the probability of exceeding a 0.29 earthquake would be 14% in'50 years, or a 275-year return period.
This meets the NRC Staff's criterion for OBS's which is a return period of 110 years or a probability of exceedance of 30% in 40 years.
SEISMIC INSTRUMENTATION - O. STEINHAP&P, PG&E The Diablo Canyon seismic instrumentation system consists of a basic system, which was installed to satisfy the minimum requirements of the NRC Staff, and a much more extensive supplemental system, which was installed at the suggestion of the ACRS. The purpose of the basic system is to determine if the seismic design bases have been exceeded; the purpose of the supple-mental system is to enable the Applicant to verify the dynamic structural analysis procedure and to evaluate soil-structure and structure-to-structure interaction.
The Unit 1 containment, the auxiliary building, and the turbine building are each quite heavily instrumented.
Instruments are also placed at sufficient distances from the structures to record the free field motion.
Triggering to start the system is accomplished at three locations in the Unit 1 containment structure. The trigger initiates recording of the carthquake motion and keeps the recording system operating until all significant response has ended. A central recording and playback system is installed in the control room for the basic system, and another central recording and playback system is provided for the supplementary system.
Power supply is through batteries backed by a charger connected to the AC power supply, and a common timing signal is provided on all records. hhen the systems trigger, annunciation is provided in the control room, and the maximum accelerations in each of the three orthogonal sensor directions are indicated on an earthquake force monitor in the control room.
In answer to a question from Dr. Trifunac, Mr. Steinhardt l
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said that the seismic instrumentation system is not a safety related system, and reaundancy is not provided.
However, the system i's exercised and tested Once per month. A certain amount of redundancy is provided because the basic and supplementary systems are separate, tied together only by the common tim-ing signal. All of the censors are on fixed structures; none are placed on equipment.
Mr. Iloch described the installation of the seismic scram.
%e instrumentation systemconsistsofthreetriaxialsensorsarrangedi[12out3 logic,andset to trip the reactor when the acceleration is in the range of 0.4 to 0.5g.
The sensors are equally spaced on the ledge around the base of the Unit 1 contain-ment.
The 2 out of 3 logic will trip the reactor if any two censors sense an acceleration within a six second time interval.
We NRC Staff is reviewing the seismic trip sensor circuits to make sure they meet NRC Staff requirements and will not affect proper functioning of the reactor protection sy.cem.
NCTTE: A complete transcript of the meeting (except Executive Session) is on file at the NRC Public Document Room at 1717 !! Street, N. W., Washing-ton, D. C..or can be obtained from ACE Federal Reporters, Inc., 415 Second Street, N. E. Washington, D. C.
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INFORIMTION REGARD"4G EQUIVALBU STIFRESS OF INERIOR STPUCTURE 4
2..
EVALUATION OF JUSTIFICATIO" FOR DECOUPLING 3.
EVALUATION OF NMLYSIS FOR INTERNAL STRLETURES (DISTRIB'JTION OF SHEAR NO BENDING IGENT)
II.
EVALUATION OF C0!TTAllFENT STABILIW AGAll4ST OVERTURNING
- 5. N4ALYSIS OF FOU' OAT 10tl t%T I..
- 6. CRITERI A NO METlDDS FOR SCP.EENING CRITICAL ELEMENTS AND A COMPLETE LIRT OF TtESE ELEMENTS NO THEIR PROPLEM AREAS 7.
EVALUATION OF CONTAINMENT PENETRATirN DESIGN
- 8. NiALYSIS OF STRESSES IN DIAGONAL REINF0DCING f
9.
EVALUATION OF N4NULUS AREA b
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HQRIZONTAL ACCELERAT!ONS HOSGRI s
Analysis 6 ado DDE HOSGRl"
- DDE HOSGRI
^" y)
' ~ @cyation
-(g (g)
(ft.)
apnlysis Analysis Rati DDE (g)
(g)
NORTH. SOUTH INPUT ~~
~~
NORTH.S_OUTH INPUT 0.56 (44% D2 crease) 1 O.01 0.018
~
183.0 1.57 1.08 '
1.13 (13% increase) 0.13 0.18 0.72 (2S% Decrease) 0.84 (1G% Decrease) 0.09 0.1S 0.47 (33% Decrease) 163.0 1.65 1.96 1.06 ( 6% increase) 0 55 (47% Dgcrease) 140.0 1.23 1.16 0.03 0.11 1.09 ( 9% increase) 0.05 l
0.08 0.63 (37% Opctease) 115.0 0.92 0.84 103.0 0.74 0.62 1.19 (19% Increase)
EAST-WEST INPUT EAST WEST INPUT 0.01 1.31 (31% increase)
For DDE accidental 188.0 1.45 1.10 0.12 0.63 (31% Decrease)
O.07 cceentricity of 5%
163.0 1.65 2.40 0.71 (29% Decrease) was used.
140.0 1.15 1.60 0.05 0.77 (23% Decrease) 115.0 0.84 1.08 j
i 0.03 i
100.0 1
0.69 0.74 1 0.93 ( 7% Decrease)
Auxiiipry Building MAXIMUM ABSOLUTE VERTICAL hCCELERATIONS E
HOSGRi DDE*
HOSGRI Elevation Analysis Analysis Rati DDE (ft.)
(g)
(g) 168.0 f
0.80 0.27 2.96 163.0 0.66 0.27 2.44 2.13 140.0 0.59 0.27 2.07 l
115.0 0.56 0.27 2.00 100.0 0.54 G.27
" 140.0 i
1.45 0.27 5.37 g
4) tion was used in design -
i
- No vertical dynamic analysis was made for the DDE; % the peak ground accc ora
% x 0.40g = 0.27g "Contiol Room Stab
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it.!c P.r., c l0 I-e SLMMRV 0F DJTSTel0ING AREAS 1, DEIU4STRAT10N OF THE CCNSERVATISti 0F BfE MET 1GD USED TO ACCOUNT FOR TORSIONAL EFFECTS IN THE ROOF 0F FUEL CJX.ING
- 2. EVALUATION OF N4ALYSIS OF SHEAR KEYS IN FOUNDATION
[
L.
3 FURTHER ANALYS1S OF Soll BEARING PRESSURE G
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SLM%RY OF OUTSTN0lm AEAS
- 1. VERIFICATION no JUSTIFICATION OF CGM INS.
2.
lWESTIGATION OF VHETHER OR NOT CORRECTIVE ACTION l
l F
l IS NECESSARY AS A RESULT OF TORSION SROIES.
l i
l
- 3. FuRmER INFORI% TION REGARDING FED bel 6S. '
)
- 4. I'bRE DETAILED ~ DESCRIPTIONS OF TRUBINE BUILDING VODIFICATIONS.
l
(.
- 5. IhSCRIPTION OF ENLARGEtElR OF THE GAP BETWEEN lJ TURBINE PEDESTAL NO ANACEf6 STRLCTURAL ELDGtES.
1
- 6. REVIEW 0F SEVERAL 119ts FOR WHICH INFORi%TECN 4%S f BEEN 1%DE AVALIABLE.
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- n. -
- 2. FURTHER ANALYSIS FOR PIPERS e
3 EVALUATION OF FOUNDATION iMT ANALYSES 4
STN31LITY ANALYSES - SLIDING i
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- U DJICT --------- L CATALOG NO.151000 4
,fA ADE IN U.S.A.
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IA A.RGIU S FOR C Lass I.
FQO1PMDJ" REGOIREb AFTER HoS6 tr C.o tApo N E NT*
PNahme tcR kjtTH L E A sT-MARGtra
). AUX. FEEDWATCR P) UMP swr otstticTion is Tne cair ics.
I r unBiNe-drive d Thit AmtTE A Aub \\s son es ALLowAELs
- 2. AOX. FEEDWATER PUMP
'THROLT T1EhRit@ LOADED To G5 %
TUKb t 4 E.
OF R ATE D Lotsb
- 3. D tE.L EL-Ci EWER AToRE M Ax. ST R C L5 E b ME vitER t s F ix.E b EtJG INE
~
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IT 15 G5 % cF ALLOW 6BLE.
- 4. DiESE L G EW GRATOR.S AtXBOR EctTS STRE SSE D To LM */o LTim.T itXr hiR bcE WERS O F A).L oWABt.E E. L".ES E L-G ENERATOR S McMoR. BoL75 STRESSEb 7o 890/6 oF A.t E L otL FILTER ALL oWABL E.
}
(. DIE SE L-G EHERAToR S SE tSMtc BRACES STR E.SsE D To FVEL est PRIMit<r TAIM 13 % oF /wLOWA O L E.
i
- 9. D lE 5 G L - G E tJ E'WTTORS
!>\\ $ 0 M ARCr C PIP iNG-ST RE SSED TO P.i E L o tt TRhuSFEit Pump 51 %
oF allow'A BL E
- 9. C ro rR iroG At PUMP HE AD FL tdG E AMb D t5C tf ARG E-C " A% i tKr PUMPS t@tt.i.E STREssE b ro <. 9 5 % ALL c WABL E M. Pr R tc Acio ThtKS hold DowtJ BOLTS STRESSED To CitM'o OF % Low'r\\BL C l
l l. B ORic. ACib TR6MSFCR SU "T told tsjo t E LE.
STREGSE b To 9'l %
P MP5 ALLOWABLE
~
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SHELL/ SAODLE WELD STRESSED TO 30 %
i f
M AT inc.)l A tG ER OF ALLOWABL E
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- 13. Stat WATe R B hse 9L AT E sS tness ED To 80 % oF TNJEcTt04 FiLT t R S ALLOWAB L E.
I. 3cRic Ac iD FiLTE R AWCHoR EotTs STREsWD To 31 /o OF ALLOW ASL E 15 RE stou At HEAT P \\) M P MOLD-DoW N B OLT S STRESSED R7moVAL Puto?S TO 60 % OF ALLOWAILL E.
l(. E Estount. HEhT RO/cVAL ANClioR EcLTs, ShsEPLMES i SUMORT' HE AT tiXCB ANGERS GU SSETS STREtSu b To < 65 '/o allow'ABLE I
M. Ce t$bNryT Cootihtr WE ARlW& RtNCr C LEhR ANC C.
WATER, PuMFS I E C o Mi': tS tJT c o o t N G-b)KHOR BOLT STRESS LimiTtt3G-PARAuETER.
At WATT R, MEhi-Exctitst4FR NAmic AMAyys15 lN FROGRE 55
- d. h04\\L tARY Shs.TWATER TnRUST iBEARiOG LoADGb 7o 95 %
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ETafMyllhjection
_ 2705 (line t1900)
J.5,8 u.9 n e Cs, )
m. e i n 6..aamt s
o 1.
CromponentiCocling Water L871 (line 2342)
R'oeT&al pbat!Romoval
!.749 (!ine 224) i; CPlomi'calland!Volumo Control i.680 (line 243) e
,1.
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'P.ioccur.i2er R.elief and Scray r...
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Maheup,Vlh2eri
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'i di! pid Rhdvasie I
Contaibmont H2 Perge[.
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.211 (!!no 4030)
's Auxiilary.Sall;>dter.
i.617;(30!1.Sirescos)
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STRESS TABt.E FOR PRIMARY SYSTEM
{
{
(Stresses Presented As % of Total Allowable)
% Norm +-
g I
4:
Co:conent 1
% Norm +
% flor:1 Soiskic+
"j Critical Location Normal Seismic LOCA Seismic
+LCCA
(.0CA
% SRSS
$we ff RE;fTOR PRESSURE VESSEL 33 12 51 45 84 96 85 l.10
[;
UiLET IDZZLE SAFE EliD g,4 CROM 5
42 18 47 23 65 51 liEAD ADAPTOR INTER MLS 0
16 92 16 92 108 93 1.15 LOWER RADIAL SUPPORT FUEL ASSEliBLIES 0
86 77 86 77 151 108
~
~ GRIO IfiLET N0ZZLE SAFE END
- 29 MTa) 15 49 44 78 93 P4 l.og STEAM GENERATOR OUTLET N0ZZLE SAFE END i32")40 92 74 56 96 81
! 078 REACTOR COOLANT PUMP 34 3
{URGEN0ZZLE RESSURIZER 35 5
48 40 83
-88 83 g
g 33 41 44 74 55
(,((d.
PRIIGRY PIPIllG 11 g )30 RPV SUPPORTS 0
44 64 44 04 108 78 Il0RIZONTAL LOAD (38%)
~
S G SUPPORTS (2) 0 50 51 50 51 80.
71
% D FRA!;E HEMBERS (22%)
g 3
RCP SUPPORTS (2)
FRAME !! EMBER
- (55%)60 56 67 63 103 E9 7
N CRD:t SUPPORT /3)
D
.- ' liEAD LIFTII:G LEGS 5
81 12 86 17 98 ga i
'f 7
tiOTES:
(1) Fuel results presented as % of 95/95 strength, tiie LOCA & Seismic events had maxi-
)
rr.un loads in different grids.
e b ' Who$
=
~
~~
- w_-
- ~ --
~----
., n. n-1-r
._y
'T1
~
7
- ~,.
O sar. Onofre steam Generator - S. Fernando Earthquake C San Onof re Stean Generator - Vibration Cencrators O Indian Point No. 2 Steam Generator C Gak Ridge ECCR Steam Generator San Onofre Pressurizer - Vibration Generators X
A Sen Onofre Puep - Vibration Generators V Indian Point No. 2 Coolent Pump e-o P
E
~
u 9
T Si 6({
we e
a E-5!
/
6
/
~
Linear Fit Curve s
/
4
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/
o O
/
I 3
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O O. -
2 X b.,y-
-- q
~ ~~ ~'~ ~ g 4
y V
U g
]
IMI N
f L
8 9
1_thl A I - - E I
9 f
i 30*
1 2
3 4
5 10 20 30 40 50 100 200 y
5YO O
wn>
DE'cT.ECTION, HII S
- Zr'
-dO Figure '.. - Sycten Dn= pica Values Associated with Ericary n
- O L
z Coolant Loop Equip ent
> 1, O y
y.
g 9
I[AIAA l 0
(
i t
- t f
ii
- ?
TABLE II IDE!GITICATION OF TEST STECI)C)lS a
/.
CROUP I VITAL REIAY ED 4160 Sit 1TCEC2A1 I
CROUP 11 DIESIL ChTRAT02 C0;rl?.0L car,112T DOOR DIESEL C21.T,r.AT03 EXCITATIO!! CUDICl2 DIESEL Cl2tATC3 C0!rr?.0L CASIL3T SUS-PA!GL
)
i
\\
CROUP III VE!CII.ATIO!! SYSTZit TILAY PAIEL VEh*IIIAT10:1 SYSTE!41DCIC ?A}ZL j
HAlli A!"ilTNCIAIC2 1
GROUP IV LCCAL STAnt0R LFC37 FIPI MtP C01CT.0LLZR 12 5-2 50VDC M.C.C.
DC SUITCir'.Aa MTTERY CECC22 i
I MTTZ3Y C".JLLS (2) i I
CRot'P V FISCICR CO!Tr20LLtn IDCAL START 2a LIC66 VITAL La\\D CE!!TI: ( 40,0V)
TW AUZIL7/3Y TILAY CASICTS CROUP V1 IDCAL STA.".TER LPY36 LIMIT SUITCEIS CR01'? VII CC?CT.0L E0ARD EQUIn22rr SIX SktITCICS AND CTB Al?2TE2
.y
.__.,a..
p y
b 91 'RJ ECT 3E0 CATAL.O w crin
.i 6
? -9 9 II e,
.... L.J J l
i ELECTRICAL EQUIPMENT MODIFICATIONS
_REQUIPED FOR 1105CR1 SEISMIC EVEllT 1.
4t10 VOLT VITAL ADDED CLIPS TO RESTRAIN LOAD CEMTER DRAW-OUT UNITS 2.
BATTERY RACKS ADDED BRACING 3.
4KV SWITCllGEAR STRUCTURAL MCDIFICATIOMS BEING DES,1CHED i
3@.cATat.ooNo.
SUBJECT
^
3M CENTCit,3 r sl(3 MADf IN tl 1
~-
1 DIABIO CNWON SUBCDbMITTEE lET'ING WAS!!INGTON, D. C.
JLNE 14-15,1978 WRITTEN STATEMENTS l
Written statements were sthmi.tted by,the following nenbors of the l
ptblic:
i I
J. R. Bakalian, Executive Director, Nuclear hbratorium Task Force dated June 8,1978 i
(
D. S. Fleischaker, Attomey for Intervenors dated Jtne 9,1978 4
4 e
e a
' s 4
4 5
4 k 1
pm
. h4.IJ7
\\
l Tan'ATIVE SilCEDULE DIABLO CN4 YON SUDCO @11TTEE JUNE 14-15,1978 WASilING' ION, DC 100M 1046-11 i
I
- J_ U11B 14 i
8:00 A.M.
Executive Session (OPEN) 9:00 A.M.
Introduction and Status Report by NRC Staff (OPEN) 9:20 A.M.
Introduction by Pacific Gas and Electric Company (IG&E)
(OPEN)
J t
9:30 A.M.
Overview of NRC Review of the Seismic Reevaluation (OPEN) i 10:30 A.M.
Conments by PG&E (OPEN) l 11:00 A.M.
Review of the NRC Seismic Design Bases (OPEN) 1.
Basic inputs (magnitude, acceleration, spectra) 2.
Tau effect 3.
Danping 4.
Torsion 12:00. NOON Coments by PG&E (OPd) 12:30 P.M.
LUNCil 1:30 P.M.
Public comments (OPEN) 2:00 P.M.
Subcommittee and Consultants Discussion of Seismic Design
', Bases (OPEN) 1 3:30 P.M.
Break 3:45 P.M.
Report on research Results at Lawrence Liverriere Laboratory (OPM) 4:00 P.M.
Continued Discussion of the Seismic Design Bases (OPEN) 5:00 P.M.
Recess 3
a s
i 4
t w
-a e= *e. 4 -
J; a 4Js,.
E
_a.
J
.c d
4 e
--... - + -
e J.we o
a m
4 2-t JUNE 15 8:00 A.14.
NRC Determin'ation of Systems and Components Required to Achieve Cold Shutdown (OPEN) 8:30 A.M.
Comnents try PG&E (OPEN)
~9:00 A.M.
NRC Description of Structural Review (OPEN) 10:15 A.M.
Comments by PG&E (OPEN) 10:45 A.11.
Dreak I
NRC Description of Mechanical Review (OPEN) 11:00 A.M.
1 12:30 P.M.
LUNCil 1:30 P.M.
Comnents by PG&E (OPEN) 2:00 P.M.
NRC Description of Reactor Fuel Response (OPEN) l 2:20 P.M.
Comments by PG&E (OPEN) 2:30 P.M.
NRC Report on Recent State of California Geological Investigations (OPEN) i 2:45 P.M.
Conments by PG&E (OPEN) 3:00 P.M.
Break 3:15 P.M.
Discussion of Seismic Scram Provisions by PG&E (OPEN) 3:30 P.M.
Discussion of Other Aspects of Seismic Design and Roanalysis (OPEN) 4:00 P.M.
Discussion of Nonseismic Matters (OPEN) 1.
Fire Protection 2.
Circulating water system Sumnation 4:45 P.M.
5:00 P.M.
Adjourn C
5 g
e 4
9 l
.5
.y.,,,.
+ t n
ATTf!l0EES - WFONESDAY, JUNE 14,197B ACRS_
NRC C. P. Siess, Cliairman N.11. Newmark, Consul tant
- 11. Etherinqton J. P. Knight S. Lawroshi D. B. Vassallo
- j 11, Bender D. P. Allison E. Luco, Consul tant J. F. Stolz H. P. Ilhite, Censultant J. Tourte11otte T. Pickel, Consultant _
E. J. Sullivan, Jr.
- 8. !4. Page, Consul tant J. R. Morgan, Consultant l
J. C. Maxwell, Consultant W. J. Hall, Consul tant H. D. Trifunac, Consultant R. J. Bosnak J. C. McKinley, Sta f f D. C. Jeng i
R. L. Wright, Staff Pei-Ying Chen H. E. Pol k S. P. Chan f
H. A. Dameron A. llafsz P. T. Kuo R. E. Lipinski I. Sihweil
(
G. Bagehi C. P. Ta n C. H. Hofmayer J. H. Pittman J. A. Murphy R. B. Ha fmann J. T. Greeves C. Stepp llestinahouse Fleischaker R. J. Sero J. M. Madden A. Ball, Jr.
D. Popp D. II. Alexander itHB S. J..Jarocki e
R. W. Brandoa R. Hubbard i
G.-J. Bahm W. C. Ganglo f f G. Fidler Associated Press i
H. Torcaso i
T. C. Esselman R. L. Cloud
- H. Cole
. I.
k
+
ij 1
J.
4
[,
~
~~'
4 1
ATTEUGEES - WEDilESDAY, JUllE 14, 1978 - CONTPlVED PUBl.1C - C0tlTitlVE0
.Snell & Wilmer URS/Bl ume A. C. Gehr L. E. Malik D. P. Jhaveri D. A. Lang Lawrence Livermore Lab.
J. A. Bl ume F. J. Tokarz
}larding-Lawson Assoc.
Telodyne Eng. Serv.
T. Idaler A. L. Buchignani D. F. Landers R. Lawson Gil bert/ Commonweal th LLL G. E. Clyde P. D. Smith EES, Inc.
PG&E CO.
R. E. Bacher T. Kohli N. Chanhan J. O. Schuyler R. E. Hamati P. G. Antiochos S. Refai M. V. Williamson R. A. Young M. H. Furbush GE A. Cornell, Censultant G. Frazier, Consultant
~
N. C. Shirley S. W. Smith J. A. McLaughlin
- 0. Steinhardt EG&G Idaho.
R. M. Laverty D. Nicisen B. F. Sa ffell, Jr.
W.
'B.
Ka e fe r S. Hanusiak P. Crane USGS R. R. Fray H. J. Gormly J. F. Devine R. Bettinger B. Norton J. B. Hoch F. F. Mautz G. A. Blanc D. H. Hamil ton J
V. J. Ghio R. L. Kelmenson 3,
h
1
.w s
OTFICIAL USE ONLY Diablo Canyon 6
ATTENDEES - THURSDAY, JUNE 15, 1978 ACRS NRC C. P. Siess, Chairman J. F. Stol z M. Bender _
D. Allison H. Etherington D. B. Vassallo S. Lawroski J. P. Knight M. P. White, Consultant E. J. Sullivan, Jr.
T. W. Pickel, Consultant I. Sihwell B. H. Page, Consultant R. J. Dosnak J. C. Maxwell, Consultant D. C. Jeng G. A. Thompson, Consultant J. L. Kno x J. C. McKinley, Staff P. Y. Chen R. L. Wright, !taff
- n. Fischer E. Luco, Consultant R. E. Lipinski H. D. Trifunac, Consultant C. P. Tan C. H. Hofmayer P. T. Kuo PACIFIC GAS AND ELECTRIC H. E. Pol k S. P. Chan
~M. V. Williamson A. Itafz P. Antiochos S. Israil R. Kolmenson H. A. Pambron F. F. Mautz U. Hodges D. Nielsen J. R. Morgan, Consultant W. Kacfor W. J. Hall, Consultant
- 0. Steinhardt D. H. Hamilton R. R. Fray J. B. Hoch J. Gornly R. Bacher 1
P. Lbanez J. F.*Bonner B. Norton e
H. H. Furbush A. C. Gehr G. Blanc S. Hanosiak 1
o' R. H. Laverty J. A, McLaughlin i
J. 0. Schuyler l
P. Crano -
_'i R. A. Young V. J. Ghio J
i
.g OFFICIAL USE ONLY l
~
n-.
- - - -e
....2
. A Li
..._...,-y OFFICIAL USE OfiLY
- l Diablo Canyon i
l ATTEllDEES - THURSDAY, JUllE 15,1978 (C0ffT'D)
WESTIllGHOUSE-EES, If1C.
I W. C. Ganglo ff S. Refai R. L. Cloud i
R. E. Hamati S. J. Jarecki fl. Chanhan M. A. Torcaso I. D. Kohli G. S. Sidier R. 11. Brandon i
D. Popps URS/BLUME R. J. Sero A. Ball, Jr.
R. Yokoyam D. Alexander L. E. Malik D. A. Lang R. R..Villatuya EG&G Idaho D. P. Jhaveri B. F. Saffell, Jr.
R. L. Grubb GEllERAL ELECTRIC N. Shirley liarding Lawson Associate _
i T. Udaka GILBERT /C0f'.M0t! WEALTH A. L. Buchignam G. E. Clyde R. A. Lang
)
4
+:
i l
e-8 l
1 i
l
(
)
A I
i 1
I
.i t
I
.n i
d
.J OFFICIAL USE Ofil.Y
.u j I
REFERENCES 1.
3 Vu-graphs use,, NRC Staff to illtstrate ovarview of NRC Staff review.
2.
ACRS consultants reports by T. W. Pickc1, J. E. Luco.
3.-
12 Vu graphs used by NRC Staff to illustrate danping values used for reinforced concreto structures.
4.
14 Vu-graphs used by Dr. B1tre illustrating sensitivity studies of danping in structures and equiprent.
5.
17 Vu-graphs used by LLL to illustrate study of scismic conserva-tisms of operati1g plcats.
6.
44 Vu-graphs used by KRC Staff illustrating NRC myiew of structural re-evaltution.
7.
3 Vu-graphs used by NRC Staff to illustrato California Division of Mines and Geology investigation of.
8.
3 Vu-graphs used by NRC Staff to illustrate systens and conpanents mquired to achieve cold shutdown.
9.
46 Vu-graphs used by NRC Staff to illustrate nochanical and electrical equipnent review.
30.
7-35 mm slidos illustrating variots ncchanical conpanents myiewed.
- 11. Vidcotape presentation tsed by PGSE to illustrate vibration testing of nochanical and c1cctrical conponents.
12.
14 Vu-graphs used ly NRC Staff to illustrate piping systems review.
13.
7 Vu graphs used by NRC Staff to illustrate fuel msponse to asymnetric loads.
3 14.
1 Vu-graph used by PGGE to illustrate ttnbine building event tmc.
f 15, 12 Vu-graphs used by PG GE to illustrate scismic instrumentation.
'16.
12 Vu-graphs used by NRC Staff to illustrate ftel assenbly defonnation
?
and effect on PQ'.
- %.