ML20069A648
| ML20069A648 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Oyster Creek |
| Issue date: | 03/20/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Finfrock I Public Service Enterprise Group |
| Shared Package | |
| ML20069A188 | List: |
| References | |
| FOIA-82-399, TASK-03-06, TASK-03-11, TASK-3-11, TASK-3-6, TASK-RR LSO5-81-03-048, LSO5-81-3-48, NUDOCS 8105110512 | |
| Download: ML20069A648 (3) | |
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f March 20, 1961 Docket No. 50-219 L505-81-03-048 hbOLj@j,, y a -, a i l
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1 Mr. I van R. Fi nf rock, J r.
Oyster Creek Nuclear Generating Station P. O. Box 388 Forked River. New Jersey 08731
Dear Mr. Finfrock:
SU8 JECT: SEP TOPICS III-6. SEISMIC DESIGN CONSIDERATlal AMD 111-11, CO*FONENT INTEGRITY - OYSTER CREEK ltlCLEAR POIER STATION As you are aware, the staff and its cengultants have cogleted the seismic
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review of Oyster Creek nuclear pouer plant. Enclosed (Enclosure 1) are copies of draft of IREG Consultant Report, "Setssic Review of the Oyster Creek i
Nuclear Power Plant as part of the Systeestic Evolvattee-Program,* and EG4G piping raport. EGG-EA-5211. " Summary of the Oyster Creek Unit 1 Piping.
Calculations Performed for the Systematic Evaluarten Preram." These reports will serve as the principal input for staff's fini assagement for Systematic
- Evaluation PreFan Topics 111-81 Seismic Desip Considerations and III-11 Cogonent Interity. Please' Inform us if your as-tuilt facility differs
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from the licensing basis assumed in.our~ assessment.
According to our review, some spen items have been identified (Enclosure 2) related to these topics. The detailed evaluation of these open items can be found in the attached reports.
You are required to respond to item 6 of, Enclosure 2 within 30 days from the date of receipt of this letter. For the remaining items of the same enclosure, we require additional information from you. You should submit this information within 45 days from the date of receipt of'this letter. dn the event that analysis is necessary for you to coglete our evaluation, you should submit a scheele for coe letion of each open item.
Proposed modifications identified in our ivport are repres*entative of the types of corrective ' actions which should be considered to upgrade seismic safety margins. Pursuant to 10 CFR 50.59 you s5ould independently evaluate the necessity of any modifications to your facility.
Sincerely.
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2 f.li.i0 71c81> '8 Dennis M. crutchfielo,c Chief 6.
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Operating Reactors Branch No. ;
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Division of Licensing mo
Enclosure:
M As stated
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ng-fr 9 OPEN ; N is OYSI! R CMii.1! !* D lii e lhe folinwing list documer.ts issues that developed as a result of o g se e n ;11t re-f ew of Oyster Creek faci'ity.
These issues have been highlighted
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a <artety of reasons.
In sore cases, sufficient documentation was not ave'l 4fe to the staf f to rake an evaluation.
Designation here does not necessary 16 a safety deficiency.
However, the NRC staff has determined that further ir'-
mation rer;arding the scis-ic resistance capacity of these (ters, includim further evaluation, if necessary,is warranted.
JCPLCo should develop an tcticn plan for esolation of these open issues including a schedule for completi n.
1.
Energency Service Water Pump - Functional integrity was not evaluated cae to lack of design detail.
A determination of the material used for the pump housing should be made.
If the material is cast iron. a justificatim should be provided for its adequacy.
7.
Erergency isolation Condenser - The audit analysis indicated that the anchor bolts at the center saddle were found to be overstressed in seismic shear from the postulated loading.
Provide detailed analyses to demonstrate design adequacy of these anchor bolts.
3.
Containment Spray Heat Exchanger - The anchor bolts were found to be over-stressed from the postulated seismic loads.
Provide detailed analyses to demonstrate design adequacy of these anchor bolts.
4 Recirculation Pump Support - No information was provided for evaluation.
5.
Motor Operated Valves - The seismic accelerations used in your evaluatio'1 (refs. 77 and 75 of Enclosure 1) are unrealistically low.
Detailed reevalua-tion with proper seismic accelerations should be provided to demonstrate its design adequacy.
No infomation was provided to evaluate functional adequacy.
6.
CRD Hydruulic Control Units - Support system of the free standing (cantilever type was found to be overstressed from the postulated seismic loads.
Provi&
detailed reanalyses to demonstrate design adequacy.
7.
Reactor Vessel. Support and Internal - No detailed information was available to evaluate design adequacy.
B.
Motor Control Centers No information was available to evaluate either structural integrity or functionability of these components.
9.
Transforrers - No information was available to evaluate dm ian adequacy of these components.
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, n _~ w n w1 t chyar Panels - Prss ue i n f-mt1r to t t.on that the pam
,1si tively ant!.ored to res is t se isr ic.wer torqing rgcents dr.d t i j d > '
wr:or Diesel Generator - No informtion vm available to evaluat.
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' maty of either the anchorage sys t em or the functionabilit f the ic ol accerator.
i' Control Room Preh - No infor mtion was available for evaluatim.
Ea ttery Room Di3 >, it ution Panels - No inforra t ion was a va ilable f 0-evaluation.
It isolation Phase Ductwork Supports - The duct support was found to be m.erstressed from the postulated seismic forces.
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Cordens ite Storage Tank - The anchor bolts wert-found to be overstressed f m the po,tulated seisnic loading conditions.
A similar conclusion wn dran from the results of analyses reported in your FDSAR (ref. 40 o' Er closure 1), but no corrective action we taken.
Provide _Juiti'ica t mn to show the design adequacy of the tank ar.d the outlet pipin1 M
iorus - Insuf ficient information was provided to evaluate the design a >quacy of the supporting columns and its connections to the lateral t-rvings li.
Peattor Building - Provide detailed analyses to demonstrate that the cable',
e re slack enough to accomodate differential displacenents between the butidings l'
!iping - Provide a detailed reanalysis to demonstrate desion adequacv t
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.u following piping systems:
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A Main Steam Line - Several snubbers were found to be overstreswd t' t v potential seismic loads.
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F he results of the audit analyses showed loading conditions at uveral
.atations to exceed the ASME Code allowable limits.
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4 11 E l ec t r ica l Cable Raceways - No information was available Nr evaluatio1.
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NUCLE AR REGULATORY COMMISSION
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March 20,1981 Dodtet No. 50-219 L505-81-03-048 MOWyfg 9 >-. a
.,a. u w. t Mr. Ivan R. Finfrock, Jr.
.0yster Creek Nuclear Generating Station P. O. Box ~388 Forted River, New Jersh 08/31 I
Dear Mr. Finfrock:
L
~SU5 JECT: SEP TOPICS !!!-6, SEISMIC DESIGN CONSIDERATI91 AM) 111-11. C0ff0NENT
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' : INTEGRITY - OYSTER CREEK WCLEAR POWER STATICM
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As you are awar;e, the staff and 'QR;ltants have completed the seismic 3
its consu k. 'vP ' review"of Ofst 1
. Creek" nucleaf.^ power'. plast. Enclosed (Enclosure 1) are copies
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of' draft 'of
~ Consultant Report..' Seismic Review of the Oyster Creek
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r_. Pmerflant asjrt of the' Systematic Evaluation Program." and EGAG
~.4 pip report. EGG-EAMi1 L"Suunary;of the Oyster Creek Unit 1 Piping
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- 4 al ations Performed 4o,r,thei systematic Evaluation Program." These reports
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Q: 1. (
C
- will. serve as the principal input for staff's final assessment for Systematic i,
Z, Evaluation Program Topics !!lj-615eismic Design Considerations and III-11, 4
('4 Cosponent Integrity.. Pleas g'ermR1f your as-built facility differs lpfrom thi' licensing basis as(sumed.in,our, assessment.
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- ,'According to our, review, Mfig.Of '
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some,apen items have been identified (Enclosure 2) l 1
3related to these, topics. The. detailed evaluation of these open items can f l<
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he found in the attached reports.
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You are required to respond to. item 6 of Enclosure 2 within 30 days from the 1
date of receipt of this letter. For the remaining items of. the same enclosure,
' we require ~ additional information from you. You should 'subait thts infornation within 45 days from the date of receipt of this letter.
In the event that r
analysis is necessary for you to complete our evaluation, you should submit
~
a sche &Ie for cospletion of each. open item. Proposed modifications identified in our report are repres*entative 'of the types of corrective' actions which should tie considered to upgrade seismic safety margins. Pursuant to 10 CFR 50.59 you" should independently ' evaluate the necessity of any modifications to your facility.
b Sincerely,
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M/J 0 71981m. 8 Dennis M. crutchfielo., chief Q
Operating Reactors Branch No.' 5
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Enclosure:
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As stated
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See next page i
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E tiCl ' !R E ?.
OPEN ISSUES OYSTER CREEL SEISM:C iii.V I[ a The following list documents issues that developed as a result of our wisric audit review of Oyster Creek facility.
These issues have been highlighted fer a variety of reasons.
In some cases, sufficient documentation was not available to the staf f to make an evaluation.
Designation here does not necessary ii-ply a safety deficiency.
H0 wever, the NRC staff h;s determined that further inf or-mation regarding the seismic resistance capacity of there itees, including further evaluation, if necessary, is warranted. JCPLCo should develop an action plan for resolution of these open issues including a schedule for completion.
1.
Emergency Service Water Pump - Functional int.1rity was not evaluated due to lack of design detail. A determination ' ' the material used for the pump housing should be made.
If the mate
.I is cast iron, a justification l
should be provided for its adequacy.
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2.
Emergency Isolation Condenser - The audit analysis indicated that the anchor bolts at the center saddle were found to be overstressed in seismic shear from the postulated loading.
Provide detailed analyses to demonstrate design adequacy of these anchor bolts.
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3.
Containment Spray Heat Exchanger - The anchor bolts were found to be over-A i
stressed from the postulated seismic loads.
Provide detailed analyses.to j
demonstrate design adequacy ~of these anchor bolts.
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t 4.
Recirculation Pump Support - No information was provided for evaluation.
5.
Motor Operated Valves - The seismic accelerations used in your evaluation c
(refs. 77 and 78 of Enclosure 1) are unrealistically low.
Detailed reevalua-tion with proper seismic accelerations should be provided to demonstrate its design adequacy. No information was provided to evaluate functional adequacy.
6.
CRD Hydraulic Control Units - Support system of the free standing (cantilever) type was found to be overstressed from the postulated seismic loads. Provide detailed reanalyses to demonstrate design adequacy.
7.
Reactor Vessel, Support and Internal - No detailed information was available to evaluate design adequacy.
8.
Motor Control Centers - No information was available to evaluate either structuri.1 integrity or functionability of these components.
9.
Transformers - No information was available to evaluate design adequacy of these components.
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Swi tchgear Panels - Pro vide infoma t icr. to show that the panels or, positively anchored to resist seismic overturning moments ar.d slid ><,,
forces.
11.
E..e* ;ency Diesel Generator - No infomation was available to evaluat.
design adequacy of either the anchorage system or the functionability c3 the 'iesel generator.
12.
Control Room Panels - No information was available for evaluaticn.
13.
i Battery Room Distribution Panels - No infornation was available fo-evaluation.
o 14.
Isolation Phase Ductwork Supports - The duct support was found to be l
overstressed from the postulated seismic forces.
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15.
Condensate Storage Tank - The anchor bolts were found to be overstressed l.
from the postulated seismic loading conditions.
A similar conclusion was I
l drawn from the results of analyses reported in your FDSAR (ref. 40 o' ), but no corrective action was taken.
Provide justification to show the design adequacy of the tank and the outlet piping.
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16.
Torus - Insufficient information was provided to evaluate the design j
adequacy of the supporting columns and its connections to the lateral i
bracings.
1 17.
Reactor Building - Provide detailed analyses to demonstrate that the cables were slack enough to accomodate differential displacements between the butidings.
18.
4 Piping - Provide a detailed reanalysis to demonstrate design adequacy of the following piping systems:
Main Steam Line - Several snubbers were found to be overstressed tv A.
the potential seismic loads.
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The results of the audit analyses showed loading conditions at several B.
locations to exceed the ASME Code allowable limits.
19.
Electrical Cable Raceways - No infomation was available for evaluation.
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