ML20199L503
| ML20199L503 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 08/23/1995 |
| From: | Logan K NRC OFFICE OF INVESTIGATIONS (OI) |
| To: | |
| Shared Package | |
| ML20199L462 | List: |
| References | |
| FOIA-97-325 NUDOCS 9712020040 | |
| Download: ML20199L503 (10) | |
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EXHIBIT 12 l '
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f 9712O20040 971119 PDR FOIA KEENAN97-325 PDR i Case No. 1 95 013 Exhibit 12 t.
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INTERVIEW REPORT OF GITA NARASIMHAM On August 16, 1995. Investigator Keith Logan, U.S. Nuclear RegulatMy Commission (NRC), Office of Investigations, Region I, King of Prussia, PA 19406, interviewed Gita NARASIMHAM, Senior Staff Engineer NSSS (Nuclear Steam Supply System), Salem Generating Station (Salem), Public Service Electric &
Gas Co. (PSE&G) Hancocks Bridge, N.J., regarding her role in the POPS (Pressurizer Overpressure Protection System) Issues. During this interview she was represented by attorneys Mark J. WETTERHAHN and Marcia GELMAN, who also represent PSE&G and other individuals interviewed during the course of this investigation. Mr. WETTERHAHN and Ms. Gelman are with the law firm of Winston & Strawn, 1400 L Street, NW, Washington, D.C. 20005, and also represented NARASIMHAM during her first 01 interview on March 14, 1995.
NA1ASIMHAM reviewed the transcript from her first interview and made some corrections before attesting to its accuracy and signing it. NARASIMHAM stated essentially as follows:
In 1993, she was researching the POPS issue raised by Westinghouse in its March 1993 notice and she was also res>onsible for preparing a draft of a letter on that same issue for Howard BERRICK's signature. She does not recall whether she discussed with BERRICK that code case approval was needed at Salem. She contacted Tom ROBERTS, QA Department and discussed with him a pending American Society of Mechanical Engineers (ASME) code case and how she could get copy of it. She is not sure if they discussed the need for NRC a> proval before it could be used at Salem; but, she did get the draft copy of t1e ASME code case from him. She knew from the draft she read that it was not plant specific.
l'er initial focus was on safety significance. Westinghouse had indicated that to her that exceeding the limits by 100 psi or less was not safety significant. Westinghouse had also indicated that the delta P due to pumps running had not been taken into consideration. She remembers asking Westinghouse, Mr. (fnu) DiTOMASS0, in March 1993, why Westinghouse had not made this a Part 21 reportable. In May 1993, she spoke with Roger WATERS, at Westinghouse, on POPS. (She made some notes on her POPS related activities on December 13, 1994, after BERRICK asked her to prepare a chronology of her actions during that time frame. This was also about the same time the NRC was conducting a POPS related inspection at Salem.
Westinghouse told her what other plants were doing, but does not know if the code case had been approved by that time, because she had initially read only a draft. She learned that Flcrida Power and Light (FPL) was going to use the code case; but she did not discuss this matter it with anyone at FPL. l However, she did discuss it with a Zion representative.
After reviewing the initial Westinghouse letter and speaking with a Westinghouse representative, she requested plant specific information for Salem. She recalls that there was a recuisition document that she arepared for the plant specific work she wanted hestinghouse to do on the P0)S system, but does not recall whether Mahesh DANAK helped her prepare it.
CASE NO.
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She does not know why the draft she prepared for BERRICK did not reflect that NRC code case a> proval should be sought. This was the first time she dealt s/
with an issue wwre an ASME code case was involved.
She is now aware that Salem was operating outside its design basis limits because of the POPS issue, but she does noc recall at what point she realized that.
She indicated that she may have first discussed the POPS issue with DANAK back in April 1993 because, within NSSS, DANAK was the cognizant engineer for that system and had responsibility LTOR. At the time she was only a contractor working within the group and she did not think that she needed the technical information that Charles LASHKARI would have provided. She did not see the need to talk with him: she believed she could get everything she needed from DANAK.
She knew LASHKARI from having worked with him on another issue when she first went to work at SALEM. But their discussion were limited to telephone contact. At the time she worked on the POPS issue, she did not have unescorted access within the plant. She worked in a cubicle, the next isle over from DANAK and would have seen him every day. But, she does not recall how often she discussed issues related to POPS with him. She had also spoken with other members of the group on other issues.
In the December 1993, through June 1994 time frame, she worked part time as a contractor employee for Sargeant & Lundy, a Salem contractor. She does not recall discussing, during this period the POPS issue and the work she was doing was not related to POPS.
She did not recall telling anyone at the time she was leaving Salem at the end of October 1993, that NRC approval was needed for the ASME code case (N 514).
She did not know when she left that DANAK would be responsible for finishing her work on the memorandum for BERRICK. She simply turned over all of her work on this matter to BERRICK. Although she did initiate a change in procedures before she left, she may not have done anything further on code case approval because her knowledge of using code cases was limited.
This report was prepared on August 23, 1995.
Reported by:
w K G. L an Special Agent Office of I stigations .
Field Office, Region I Case No. 1 95 013 2 EXHIIT b l PAGE OF2.t_ PAGE(S)
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i EXHIBIT 13 9
Case No. 1-95 013 Exhibit 13
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{ ' COMMITMENT NUMBER " ' ^ ~' ~ ~~~'~~
', " INCIDENT REFORT .No.
i USE CONnrTUATION EMErrS TT NECESSARY -
SECTION I (Initiator) r REPORT
SUBJECT:
UNIT , (S1, S2,53,'HC) : ~61 DATE OF INdIDINT:
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SUMMARY
OF IVENT (IF ESF ICTUATION, ' INCLUDE S05 PRINTUUT): ' ~ '
REvIEU OF EU8 LETTEA MEC-930]7DA7FD DEC,39,/fB
/ND/C#lES THRT PORS sETi'olMT* tJAS 3Et E&rPb Aio/J-ColsmvMWEb lA)EsTINGHOUSE .DIDN0Y CoMCGER MllE MATIC Ewvarlod MFF lhEnse su nauswrreR TAP LOCAT/cnJ .4@ 2EACTOR/CcRE mtDKAdt :
AN'b 'WMAvnc fFFEcTs ,7%A= To RCP OPERAT/onJ. t'WesriaGlazc HAs Mah GALCl)L A72% '
SYMAm Ir En=7=cT OF 3/ PJI* tbR 'e8 pcP DPE5'ATion.idHEA) THis 3tPSE Is AbbED 7D 44GPLrA/Wic/f/S l'E' AK lPR.E350RE 20E= 7D Pof5 SerPCIMTnr27SP5Z~nt& A TRA/JS/EMT CAos d G RCS PPES.50RE 7tJcR GA2 . 77 i.s asses Anumm THRT S6LEM .hs Es Alb7 (EE MoRE 7 HAM OATE 6?CP dh;ZdQ 2<raf.
SY/LL US TscHJPEC MfGr0P/tcal.,2%btd CdAVES/UMIf1 ARE V/Olesc.p.
- PORTED BY: 4./h DEPT: GCN ~
. PHONE EXT: _ Z 7M ,
SECTION II (SNSS/ BPS MG)
RX PWR AT TIME OF EVENT: % UKIT LOAD: MWe op Con /Mede REPORT MADE PER ECG7 (Y/N): (IF YES, ATTACH ECG COPY)
LCO #: A/S #: DATE IN:
_ TIME IN: W.R.#: ,
INITIAL OTHER: CAUSE DETERMINATION: EQUIP ~ DESIGN PERSONNEL PROCEDURAL REPORTABLE: YES/NO, RFASON SNSS/NSS. SIGNATURE:
DATE: .
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COMMENTS:
l OPERATIONS MANAGER REVIEW :
DATE: / /_
+ sp Nuclear Cosuten Page 1 of-4 Rev. S E 'EF '
EXHIBIT /
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l EXHIBIT 14 l
Case No. 1 95 013 Exh1 bit 14
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, Tot The Technical Department Manager FRON: C P L6?hkari, system Engineer SUBJECT POPS setpoint Impact on PTS Concern DATE: January 30, 1994 Vice President stanley LaBruna required salen Generating Station to investigate possible violation of Tech spec on RCS Heatup and Cooldown at Diablo Canyon 1/2 as reported in Industry Briefs Volume II, Number 29 ( copy attached ).
An EWR Uma written to E&PB to evaluate this concern.
Additionally, Westinghouse identified in PSE-93-204 a Potential Issue (PI) regarding non-conservatism in the POPS setpoint deve)opment. The pressure difference from the wide range pressure transmitter to the reactor vessel midplane (where the Tech. Spec. heatup und cooldown pressure /
temperature limits are defined) was not considered.
In OE 5691, Comanche Peak SES ( 4-loop Westinghouse PWR) also reported that indicated wide range pressure could be lower than actual pressure at reactor vesself core midplane e3svation by approximately 50 psi.
Based on above concerns, the NRC issued INFO NOTICE IN-93-53 which shows similar concern has been reported to the NRC by Byron, Elon, Diablo Canyon, Kawanee,sequoyah and Point Beach salem analysis provided by EEPB letter MEC-93-917 dated Decesbar 30,1993 confirmed that calculated pressure at core mifplane would violate the Tech Spec Heatup & Cooldown curves. Cooldown Curve at 20 F/HR was considered in determining the POPS setpoint. E&PD has taken credit for ASME Code case N514 which provider. 10% margin over Appendix 0 of ASME Code. Appendix G is also the bases of Tech Spec heatup and cooldown curves. The Code case is not approved by the NRC and its use to defend a Tech Spec violation is ust allowed. Licensing Department needs to submit an emergency revision to the Tech. Spec. on heatup and cooldown curves.
If limitation on number of reactor coolant pumps ir CASENO.
1-95-013 $ b'0 [ d
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ij t operation during POPS applicability temperature range is implemented as a result of this concern evaluation, a 10CFR21 risport may still have to be made to the NRC based on calculation error discovered by Westinghouse ,
and confirmed by the Licenses /PSE&G.
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EXHIBIT 15 l
4 il Case No. 1 95 013 Exhibit 15
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\- A T *rt9CMM EhlT* &-
! TO:
John Morrir,on, Manager !
SalemTechnicalDepartment !
C P Lashkan, Salem SystemEngineer j FROM:
Salem Technical Department 4
SUBJECT:
Non Conseivation m POPS Set Point
REFERENCES:
Westinghouse letter PSE 93-204 & EAPB letter MEC-93 917 ,
DEF No. 94-0060 f DATE: April 22,1994 l Above referenced DEF was faxed to me on April 19,1994. This DEF and Wl inster and E&PB letter relate to potential de6ciency reported by Westingho:
roepect to a non- coservatism in the development ofP ,
and the'effect cf oynamir. prnsure head which would edst withlor 2 RCP in
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during Mode 5 were not cor.idered. Mode 5 temp !
mass addition cases as required in calculating the POPS setpoint. This issue under study at PSE&G for appmximately a year. ;
In December 93 E&PB informed Salem Station that calculated peak pre the Tech. Spec heatup & cooldown curves in Mode 5 . However, ASMEl '
or 93 edition of ASME Code Section XI allows peak pressure to exceed by
) Appendix G curves which are br.sically the curves in Salem Tech S ,
l acceptable to Salem Techn* cal Deps.i..= since prior permission of t '
l to use a Code Case or later edition of ASME Code. TomRoberts of QA Dept conc l
with this position. Per instructions of my Supervisor, this whole issue was sent to licensing Dept, to determine if an Incident Report should be delivered it l
Room or OPS Staff. AAer sweral meetings on details about how use ofRH-3 ca bedp to show that there is no Tech Spec violation. It was also agreed .
that it can be appropriately tracked, It was concluded by Ucensing Depar ,
issue is not covertd by the existing DESIGN BASIS for POPS as reDecte' and Tech Spec Basis.
Therefore, Licensing decided that an IR should be delivered to the Con t
a draft ofIRwas faxed to the System Engineer which required minor commei System Engineer and E&PB Responsible Eegineer. At about 12.3 Licensing Manager and E & PB Nuclear / Mechanical Manager will meet and I should remain near my telephone so that they can patch me in ifit w' ,
Next day I was informed that Licensing Dept has reveued itself and IR f delivered on a DEF which exists and is being worked expeditiousneu by E&
calculation which will form basis of change to UFSAR and possibly Tecl t
several weeks away. This does not meet either the NRC or PSE&G 7 EXHET/[ l' CASE NO. - 1-95.-0l'3
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Nudeu Department reportibility requirements for safety signi6 cant issues.
4 System engineer duties under NC.DE.-AP.ZZ 0018 requires timely notific Operations Department if any it may affect operability of any system or compon which may have any safety concern . I was instructed by H.Berrick/ J Ranalli t this de6dency to control room. Such delays or not reporting at all creates co misleading impressia-s of the management expectations retarding complian PS&EG procedures and regulatory requirements. Getting used to this typ ,
can create a mind set where a system engineer may not report a minor deficienc t On another issue I was reprimanded under Positive Disdpline program for no erred on conservative side even though I had violated no existing procedures. It!
paidul physically and mentally . It also affected m -
position with respect to existing Design Bases has already issued an INFO NOTICE. PSE&G undue delay will not make u leader but can also cause unfortunate operating incidents and possibly loose of our regulator in being trustworthy, i
- Ihe issues which must be reviewed are related to 10CFR50.5, and 10CFR50.72 I must request your urgent attention since this issue must be reviewed pri change for Salem 1. ,
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EXHIBIT 17 l
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Case No. 1 95 013 Exhibit 17
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WeSilnin0Use : *.: : ir . a n: ,
Electric Corporation ~ 2 r ::
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3 September 29.1993 PSE 93 707 RFQ88 9314355 GO# NA-41746 PSE&G PO # Pl 430450 Howard Berrick. Principal Engineer Public Service Electne and Gas Company P. O. Box 236. Mail Code N50 Hancock's Bridge. NJ 08038 3
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Subject:
Salem Core Delta P Calculanon i
Dear Mr. Berrick:
An issue was raised regarding a nonconservatism in the Cold Overpressure Mitigation System (COMS) serpoint de.elopment. The pressare difference from the reactor vessel midplane region (where the Appendix G Limit is defined) to the wide rang pressure transmitter tap was not considered in the original Westinghouse analyses.
This pressure difference effectively results in the pressure'in the reactor vessel being- -
greater than that seen by the wide range pressure transmitter (s) which are used to actuate the PORV's when in the cold overpressure mode. The reactor vesselintegrity limits for ,
which COMS provides protection are defined by the Appendix G curves generated for
, each plant A potential exists for Appendix G Limit viola, tion when considering the aforementioned conditions. -
To address the concem regarding the nonconservatisms in the COMS analysis provided by Westinghouse, Westinghouse calculated the dDelta P associated with the difference in the location of the wide range pressure transmitter and the limiting material (weld)in the core midplane region of the reactor vessel. The delta P values were calculated based upon a set of conservative assumptions in order to bound possible plant operating conditions.
The key assumptions used in the generic analysis were as follows:
- 1. RCS Temperature of 70 degrees F
- 2. No pressurizer steam bubble
- 3. All RCP's operating
- 4. Best Estimate Flow + 2Cr l
~ EXHIBIT /f CASE No 1-95-013 g /g yy PAGE / OFd__ PAGE(S)
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i Further, it had been assumed that the pressure transmitter used for cold overpressure l protection is zerced out to at least the hot jeg elevation if the actual elevation of the transmitter is above the hot leg. .
Westinghouse calculated a generie delta P for four loop plants (with 12 foot core) of 74 psi. This additional pressure is equivalent to the increase in pressure at the entical location for reactor vessel integrity of the core midplane as opposed to the location of the wide range pressure transmitter in the RCS hot leg which controls operation of the relief valves during a COMS event. ,
)
Per the cited purchase order, Westinghouse has calculated a Salem Specific delta P value considering the operation of one, two and four Reactor Coolant Pumps during cold conditions. Also, outlined below are some other key assumptions used in the analysis:
RCS Temp. (DF) 70 70 70 RCS Pressure (psia) 500 500 ' 500 Power level (%) 0.0 0.0 0.0 SGTP Level (%) 3.0 .3.0 3.0 Loop Flow (Active GPM) 101.613 99,359 90,375 Limiting Weld Elevation (ft) 75.58 75.58 75.58 Hot leg Centerline Elevation (ft) 97.0 97.0 97.0 Wide Range P* 403 Elevation (TAP ft) 90.75 90.75 90.75 Wide Range 17T-405 Elevation (TAP ft) 92.4 92.4 92.4 Distance from RX Outlet Nozzle .
, to RHR Suction TAP (ft) 10.35 10.35 10.35 Based on the above input assumptions, Westinghouse has calculated the delta-Ps of interest to be as follows:
- of Operating RCPs Delta P (psi) 1 29 2 37 4 71 Note that these pressure, drops include losses due to the effect of the difference in velocity head between the hot leg and the downcomer region in the reactor vessel. Also, these pressure drop values assume that the pressure transmitter (s) used for cold overpressure protection (I'T-403 PT-405) are zeroed out to at least the pressure tap elevation of the RHR suction line.
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I Also. hydraulic losses from the RHR suction line tap off of the hot leg to the pressure !
transmitter were not considered. Past analyses have shown this pressure drop to be very ;
small when compared to the overall pressure drop of concem. Furthermore, when s.nsidering tne inherent conservatisms used in the calculation of the overall pressure drop.
this term is not requned to provide a bounding worst case pressure drop value.
These pressure drop values are considered to be bounding for both Salem Units due to '
similarities which exist between RCS loop layouts and also due to the inherent -
conservatisms applied in the analysis. i.e., SGTP level and increased BEF-(4 RCP case only). ,'
If you have any questions. or require any additional information, please contact Robert Tiskus at (412) 374 5946 or me at (609) 339 5343.
Very truly yours, I t
'\ s 3 v f, h , " ,
JeffI uckabee '
hec,ill Sales Representative .
Power System Field Sales ec: Gita Narasimhan,N50 O
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b Case No. 1 95 013