ML17244A699
| ML17244A699 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/21/1979 |
| From: | Jabbour K Office of Nuclear Reactor Regulation |
| To: | Desiree Davis Office of Nuclear Reactor Regulation |
| References | |
| TASK-02-04, TASK-03-01, TASK-03-06, TASK-RR NUDOCS 7907250055 | |
| Download: ML17244A699 (18) | |
Text
I
~I>LII>)REgot',i~O p JUNE g i-tl3Q Docket No. 50-244 t1Et10RANDUN FOR: D. K. Davis, Chief, Systematic Evaluation Ppogram
- Branch, DOR FROM:
K. N. Jabbour, Systematic Evaluation Program Branch, DOR
SUBJECT:
'UtltQRY OF fIEETItJG ON REVIEW OF SEISMIC t)UALIFICATION OF NECHAttICAL EgUIPttENT FOR R. E.
GINNA NUCLEAR POWER
,PLANT As a followup to the ttRC/SEP Seismic Review Team visit to R. E. Ginna site, I attended a meeting with Rochester Gas 6 Electric Corporation (RGSE) at Westinghouse Electric Corporation in Pittsburgh on June 12, 1979.
The purpose of the meeting was to review the progress to date on the seismic qualification of mechanical equipment for R. E. Ginna facility.
A list of
~,attendees is provided in Attachment l,and the meeting agenda is provided in Attachment 2.
The following paragraphs summarize the discussion of the seismic qualifica-tion of each mechanical component within Westinghouse scope of supply.
l.. Auxiliar E ui ment Review 7 907200055 A.
Boric Acid Storage Tank This tank is located in the auxiliary building (Elevation 271').
The supports of this tank were statically analyzed for 0;26 g
horizontal and nozzle loads were not considered.
The resulting stresses were compared to AISC allowables.
B.
Component Cooling Surge Tank This tark Has analyzed=for pressure load only and nozzle loads were not considered.
AStlE Code,Section VIII allowables were used.
Ho seismic analysis was performed for this tank and its supports.
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NRC PORN 518 (9-76) NRCM 0240
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- 0. K. Davis C., Component Cooling Water Heat Exchanger This component and its supports were statically analyzed for 0.62 g
in the horizontal and vertical di rections and the stresses were combined using the absolute sum 'method.
nozzle loads were not considered arid ASNE code, Section VI:I allawables were used.
2.
Pi in S stem Review 3 ~
A.
Residual Heat Removal System In response. to IE Bulletin 79-07, RGSE has reanalyzed the residual heat removal system.
The reanalysis is being reviewed by the Engineering
- Branch, DOR.
Primar E ui ment Review A.
Control Rod Drive Mechanism and Supports A static 'seismic analysis was performed in 1968 for 0.8 g in the horizontal direction..Although the stress margins were adequate, lateral supports were installed in the plant prior to operation.
.The pressure boundary was evaluated in accordance with AStlE Code,Section III, 1968.
B.
Reactor Coolant Pump and Supports The pump was statically analyzed for 1.0 g lateral load.
The pump.
support loads were statically calculated by Gilbert for 0.47 g (SSE) and 0.19 g (OBE) and were compared to pipe rupture loads.
The analysis showed that pipe rupture loads controlled the design.
C.
Pressurizer
, The pressurizer,.skirt and bolts were statically analyzed for 0.48 g (SSE) in the horizontal direction and 0.32 g (SSE) in the vertical direction applied simultaneously.
The loads were combined by the absolute sum method and nozzle loads from. all pipi'ng except the surge
.line were included in the analysis.
ASt<E Cod'e,Section III allowables were used.
OFFICR~
ODRNAIIKW DATR~
SEC PORbf 31S (976) NRCM 0240
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D. K. Davis 3
w R 1 1979 D.
Steam Generator.,and.Supports The steam gener'at'or is of.the 4400 series.
The internals were evaluated, for 0'.ll. g (SSE) in the Porizontal direction and 0.15 g
(SSE) in the vdrtical direction applied simultaneously.
ASl1E Code,
. Section'II allowables were used. 'he supports were statically
'nalyzed by Gilbert for 0.47 g (SSE) and 0.19 g (OBE)
\\
The scope of the NRC tgeneric activity A-2 was.briefjy discussed because the pipe rupture loads on the primary equipment supports may be more limiting than the seismic loads.'t
/he conclusion of the meeting, RG&E ag'reed to send a letter to NRC aHressing the seismic. qualification of each mechanical and electrical component, and fluid and electrical distribution systems as. stated in. the summary of the meeting held at the R. E. Ginna site on April 10-11, 1979.
In addition, the letter will outline the information'available and the submittal date to NRC. It is expected that the letter will be received by NRC about June 29, 1979.
Kahtan N. Jabbour.
Systematic Evaluation'.Program Branch Division of Operating Reactors t
Attachments:
As stated cc w/attachments:
See attached page h
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NRC FORM 518 (9-76) NRCM 0240 DQR
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DISTRIBUTION FOR MEETING SUMMARIES 244 NRC PDR Local PDR TERA SEPB Reading NRR Reading H. Denton E.
Case V. Stello D. Eisenhut R. Vollmer W. Russell B. Grimes T. Carter T. Ippolito R. Reid G. Knighton V. Noonan A. Schwencer D. Ziemann D. K. Davis G. Lainas J. Scinto, OELD OIEE (3)
ACRS (16)
Licensee'5 cc NRC Participants D. Knuth R. Schaffstall, KMC J.
Shea T. Wambach T. Cheng H. Levin P. Y. Chen, P-924 K. Jabbour C. Hofmayer L. Shao J.
- Covey, RG&E J. Hutton, RGEE G. Wrobel, RGEE J.
- Hung, RGhE G. Davis, RGEE C. Chen, Gilbert Associates F. Rehill, Gilbert Associates J. Stevenson, Woodward-Clyde F. Loceff,"Westinghouse S.
Hyde, Westinghouse C. Lin, Westinghouse W. La Pay; Westinghouse T. Campbell, Westinghouse kEISiiR7 00<K'f IBÃf
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Docket No. 50-244 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 June 21, 1979 K5tÃOR7 MI2<J HLk hW7 MEMORANDUM FOR:
D. K. Davis, Chief, Systematic Evaluation Program
- Branch, DOR FROM:
K. N. Jabbour, Systematic Evaluation Program Branch, DOR
SUBJECT:
SUMMARY
OF MEETING ON REVIEW OF SEISMIC QUALIFICATION OF MECHANICAL EQUIPMENT FOR R. E.
GINNA NUCLEAR POWER PLANT As a followup to the NRC/SEP Seismic Review Team visit to R. E. Ginna site, I attended a meeting with'Rochester Gas 8 Electric Corporation (RG8E) at Westinghouse Electric Corporation in Pittsburgh on June 12, 1979.
The purpose of the meeting was to review the progress to date on the seismic qualification of mechanical equipment for R. E. Ginna facility.
A list of attendees is provided in Attachment 1, and the meeting agenda is provided in Attachment 2.
The following paragraphs summarize the discussion of the seismic qualifica-tion of each mechanical component within Westinghouse scope of supply.
l.
Auxiliar E ui ment Review A.
Boric Acid Storage Tank This tank is located in the auxiliary building (Elevation 271').
The supports of this tank were statically analyzed for 0.26 g
horizontal and nozzle loads were not considered.
The resulting stresses were compared to AISC allowables.
B.
Component Cooling Surge Tank This tank was analyzed for pressure load only and nozzle loads were not considered.
ASME Code,Section VIII allowables were used.
No seismic analysis was performed for this tank and its supports.
,I
D. K. Davis June 21, 1979 C.
Component Cooling Water Heat Exchanger This component and its supports were statically analyzed for 0.52 g
in the horizontal and vertical directions and the stresses were combined using the absolute sum method.
Nozzle loads were not considered and ASME code,Section VIII allowables were used.
2.
Pi in S stem Review A.
Residual Heat Removal System In response to IE Bulletin 79-07, RGSE has reanalyzed the residual heat removal system.
The reanalysis is being reviewed by the Engineering
- Branch, DOR.
3.
Primar E ui ment Review A.
Control Rod Drive Mechanism and Supports A static seismic analysis was performed in 1968 for 0.8 'g in the horizontal direction.
Although the stress margins were adequate, lateral supports were installed in the plant prior to operation.
The pressure boundary was evaluated in accordance with ASME Code,Section III, 1968.
B.
Reactor Coolant Pump and Supports The pump was statically analyzed for 1.0 g lateral load.
The pump support loads were statically calculated by Gilbert for 0.47 g (SSE) and 0.19 g
(OBE) and were compared to pipe rupture loads.
The analysis showed that pipe rupture loads controlled the design.
C.
Pressurizer The pressurizer, sk and bolts were statically analyzed for 0.48 g
('SSE) in the horizontal direction and 0.32 g
(SSE) in the vertical direction applied simultaneously.
The loads were combined by the absolute sum method and nozzle loads from all piping except the surge line were included in the analysis.
ASME Code,Section III'llowables were used.
\\'I I
I f
D. K. Davi s 3
w June 21, 1979 D.
Steam Generator and Supports The steam generator is of the 4400 series.
The internals were evaluated for 0.19 g
(SSE) in the horizontal di rection and 0.15 g
(SSE) in the vertical direction applied simultaneously.
ASME Code,Section III allowables were used.
The supports were statically analyzed by Gilbert for 0.47 g (SSE) and 0.19 g
(OBE)
The scope of the NRC generic activity A-2 was briefly discussed because the pipe rupture loads on the primary equipment supports may be more limiting than the seismic loads.
At the conclusion of the meeting, RG&E agreed to send a letter to NRC addressing the seismic qualification of each mechanical and electrical component, and fluid and electrical distribution systems as stated in the summary of the meeting held at the R. E. Ginna site on April 10-11, 1979.
In addition, the letter will outline the information available and the submittal date to NRC. It is expected that the letter will be received by NRC about June 29, 1979.
Attachments:
As stated Kahtan N. Jabbour Systematic Evaluation Program Branch Division of Operating Reactors cc w/attachments:
See attached page
I 6
Attachment 1
MEETING AT WESTINGHOUSE ELECTRIC CORPORATION PITTSBURGH PA JUNE 12 1979 LIST OF ATTENDEES RGRE J.
Covey J. Hutton G. Wrobel J.
Hung G. Davis NRC K. Jabbour NRC Consultant/
~Wd dCi d
J. Stevenson Westin house F. Loceff S.
Hyde C. Lin R. Kelly W. LaPay (part time)
T. Campbell (part time)
Gilbert Associates C.
Chen F. Rehill
Ih II '
REVIEW OF SEISMIC QUALIFICATION OF EQUIPMENT FOR RGE June 12, 1979 Meetin Aaenda A.
Organization of Review Meeting B.
Auxiliary Equipment Review (1)
Boric Acid Storage Tank (2)
Component Cooling Mater Surge Tank (3)
Component Cooling Mater-Heat Exchanger C.
Piping System Review (1)
Residual Heat Removal System D.
Primary Equipment Review (1)
Control Rod Orive t".echanism and Support (2) 'eactor Coolant Pump and Support (3)
Pressurizer (4)
Steam Generator and Supports E.
Summary of Meeting
U 0
II
,I i
I' 7