ML20202D097
| ML20202D097 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/31/1997 |
| From: | Letts B, Logan K NRC OFFICE OF INVESTIGATIONS (OI) |
| To: | |
| Shared Package | |
| ML20199L462 | List: |
| References | |
| FOIA-97-325 1-95-013, 1-95-13, NUDOCS 9712040143 | |
| Download: ML20202D097 (40) | |
Text
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Title:
SALEM GENERATING STATION UNIT 1:
INTENTIONALLY OPERATED OllTSIDE ITS DESIGN BASIS AND FAILED TO MAKE TIMELY NOTIFICATION TO THE NRC 0F AN UNANALYZED CONDITION REGARDING SALEH'S PRESSURE OVER PRESSURE PROTECTION SYSTEM Licensee:
Case No.:
1 95 013 Public Service Electric & Gas Company Report Dato: March 31, 1997 P.O. Box 236 Hancocks Bridge, New Jersey 0803B Control Office: 01:RI Docket No.:
50 272 Status: CLOSED Reported by:
R'eviewed and Approved by:
e Keith/G. Logan, cial Agent Barry R.[Letts. Director Office of Invest 1Aons Office of Investigations Field Office. Region I Field Office, Region I Participating Per.ionnel:
i Brian J. McDermott. Resident Inspector Reactor Projects Branch 4 Division of Reactor Projects WA$NING DO
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5 SYNOPSIS On February 21, 1996, the NRC Office of Investigations initiated this investigation to determine whether Public Service Electric & Gas Co. (PSE&G) intentionally opercted outside its design basis and failed to make a timely notification to the NRC, pursuant to 10 CFR 50.72, of an unanalyzed condition regarding Salem Generating Station's (Salem's) pressure over.pressum protection system (POPS). NRC Region I inspection findings indicate that PSE&G changed the POPS design basis transient for mass addition without-evaluating the change pursuant to the requirements of 10 CFR 50.59.
Besed on the evidence developed during this investigation, it is concluded that PSE&G willfully o xrated outside its design basis and failed to provide a timely notice, to the ARC, >ursuant to 10 CFR 50.72, that it was operating in an unanalyzed condition. T1e results of this investigation were presented to Region I staff and were considered, along with the staff's inspection 6'ndings, in the issuance of a notice of violation against PSE4G on October 26, 1995, and subsequent escalated enforcement action.
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TABLE OF CONTENTS Eine SYNOPSIS 1
LIST OF ACRONYMS S
LIST OF INTERVIEWEES 7
OETAILS OF INVESTIGATION....'
9-Applicable Regulations 9
Purmse of Investigation 9
Bac(ground 9
Interview of A11eger.......................
10 Coordination with Regional Staff.................
18 Review of Licensee's Investigation................
18 Coordination with Regional Counsel 19 Allegation (Intentionally Operated Outside Its Design Basis and Tailed to Make Timely Notification to the NRC of an Unanalyzed Condition Regarding Salem's Pressure Over Pressure Protection System (POPS))
19 Sumar 19 Agent'y/ Evidence s Analysis
......................33 Conclusion 34 LIST OF EXHIBITS............................
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LIST OF ACRONYMS ASME American Society of Mechanical Engineers ATS Action Tracking System r9 Code of Federal Regulations DEF Discrepancy Evaluation Form DR Deficiency Report FSAR Final Safety Analysis Report LCO Limited Condition for Operation LER Licensee Event Report LTOP Low Temperature Overpressure Protection NSAL Nuclear Safety Advisory Letter NSSS Nuclear Steam Supply Systems IN Information Notice IR Incident Report liSAL Nuclear Safety Advisory Letter P E&G Public Service Electric & Gas P0i S Pressure Overpressure Protection System
. OR\\'
Power 0perated Relief Valves P/T Pressure / Temperature RCF Reactor Coolant Pump RCS Reactor Coolant System TS Technical Specifications NOT PUB _I DISN.0ShEWI APPROV 0F /
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LIST OF INTERVIEWEES EXHIBIT 1
BERRICK, Howard. Supervisor, Mechanical Group. NSSS.
Salem, Public Service Electric & Gas Company (PSE&G) 5 & 27 CHANDRA,Vijay,ThermodynamicsAnalyst,TechnicalConsultant, PSE&G...............................
29 DANAK, Hahesh, Senior Staff Engineer, Engineering and Plant Betterment Department, Salem, PSE&G.............
18 LASHKARI, Chandra " Charles", former Senior Engineer, Salem PSE&G..... 2 H0RRISON, John Technical Manager, Salem, PSE&G,............. 4 NARASIMHAN, Gita, Senior Staff Engineer, Hechanical (a former contractor with Sargent & Lundy) NSSS Group, Salem, PSE&G.. -....................
12, 23, & 28 0'GARA, Ken. Principal Engineer. Ratheon Engineers and Constructors (formerly Ebasco Services), a PSE&G contractor at Salem....
10 & 25 SMITH, David. Principle Engineer, Nuclear Licensing, Salem, PSE&G.............................
6 & 24 THOMSON. Francis former Manager. Licensing and Regulation (Salem & Hope Creek), PSE&G..................
9 & 26 WIEDEMANN. John. former Supervisor, Salem, PSE&G 7&8 N'TORF ICDI]SCLhl JT AL OF blJ.
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DETAILS OF INVESTIGATION ADolicable Reoulations 10 CFR 50.5:
Deliberate misconduct.
10 CFR 50.59:
Changes, tests, and experiments.
[ Paragraph (a)(2) requires licensees to evaluate proposed changes, under certain conditions, to determine whether an unreviewed safety' issue exists.]
10 CFR 50.72:
Immediate notification requirements for operating nuclear power reactors.
10 CFR 50.73:
Licensee event report system.
Purpose of Investiaation On February 21, 1995, the NRC Office of Investigations initiated this investigation to determine whether Public Service Electric Gas Co. (PSE&G) intentionally operated outside its design basis and failed to make a timely notification to the NRC, pursuant to 10 CFR 50.72, of an unanalyzed condition regarding Salem Generating Station's (Salem's) pressure over pressure protection system (POPS). Region I inspection findings indicate that PSE&G changed tF? design basis transient for mass addition without evaluating the change pursuant to the requirements of 10 CFR 50.59.
Backaround On August 8,1995, the NRC received an allegation from Chandra " Charles" LASHKARI 1dentifying 23 separate concerns with Salem's operations, engineering and management. The alleger was first interviewed by Region I staff and, subsequently, by 01. On September 30, 1995, 01 initiated an investigation into potential discrimination against LASHKARI (see 01 Case No. 1 94 043) and LASHKARI was interviewed by 01 as aart of that investigation (Exhibit 2). One of the safety issues raised by LASiKARI with the Region 1 staff concerned the plant's pressure overpressure protection system (POPS).
It appeared from the preliminary results of an inspection by the Region I staff that on or about April 19, 1994, PSE&G changed the POPS design basis transient for mass addition, without evaluating the change pursuant to 10 CFR 50.59.
On February 16, 1995, a Region 1 panel convened to discuss the alleger's concerns, specifically whether the licensee failed to notify the NRC for 11 months & continued to operate the plant outside its design basis for POPS.
In the 01 interview LASHXARI stated that he brought, to the attention of his supervisor, John WIEDEMANN, information regarding the POPS issue in the form of an Incident Report (IR) dated January 31, 1993.
LASHKARI stated that when he wanted to bring the IR to the control room, he was told: *No.
You do not file this.
It's not your job
..." (Exhibit 2, p. 32).
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Case Noi.1 013 9
The POPS issue was eventually re>orted to the NRC in Licensee Event Report (LER) No. 94 017 00. dated Decem 2r 14, 1994 (Exhibit 30).
Interview of the Allecer On December 15. 1994 LASHKARI was interview by 01. With regard to his POPS allegation. LASHKARI stated that on Januaiy 31, 1994, at 4:00 p.m. he prepared an Incident Report (IR) on the POPS issue (Exhibit 13) and wanted to report the problem to the control room.
LASHKARI recalled that WIEDEMANN said to him: "No. You do not file this.
It's not your job. You turn over all the paperworktotheEngineeringDepartmentandletthemmakealongterm assessment of the situation (Exhibit 2, p. 32).
LASHKARI said that WIEDEMANN told him to " Forget it. This can shut down the present running units.
Let's have the whole issue go out to the Engineering and Plant Betterment Department... and let the licensing people make a determination are we in violation, are we... having problem with the plant not complying with the regulation.* According to LASHKARI. the whole issue, including his incident report, was turned over to the Licensing Department (Exhibit 2
- p. 34).
LASHKARI stated that he turned over all of his paperwork on POPS.to the Engineering De>artment and he attended several meetings on POPS with Ken O'GARA, a.icensing Engineer, who agreed with him (Exhibit 2
- p. 32).
LASHKARI stated that Frank THOMSON (Licensing Manager) originally agreed to report the matter, but the next day THOMSON did not file the IR (Exhibit 2.
- p. 33)
LASHKARI stated that on about A)ril 21, 1994 O'GARA sent LASHKARI a co)y of a proposed IR: he reviewed it, macing some minor changes, and sent it bac( to O'GARA, believing that they would be taking it to the Control Room (Exhibit 2.
- p. 31).
LASHKARI stated that Engineering determined that Salem was in violation of its own design basis and was required to go to the NRC for relief (Exhibit 2.
- p. 34).
LASHKARI stated that when Licensing reviewed the BERRICK memorandum (Exhibit 11), they basically agreed with him that an IR was due (Exhibit 2.
pp. 34 36).
Insoection Report POPS Summary The NRC completed an inspection of Salem and the results are contained in the Ins 3ection Report dated Aarch 24, 1995 (Exhibit 3). The report noted that PSE4G worked to retnlye nonconservatisms in the POPS setpoint calculations for ap)roximately two years (itarch 1993 through February 1995).
In the process.
PSE&G relied on an exemption from the requirements of 10 CFR 50.60, without NRC approval, failed to report a condition outside their plants' design bases, and revised the POPS design basis transient (described in the FSAR and Technical Specification Bases) without performing a safety evaluation pursuant to 10 CFR 50.59 (Exhibit 3).
IJ RE OF I
Case No 1 95 013 10 1
During the inspection, engineering personnel stated that from the time the issue was identified in March 1993 they considered its safety significance to be low and that the plant was adequately protected.
However, the POPS issue was not entered into an appropriate system for evaluating operability, safety significance, or reportability for over a year while options to assuage the problem were explored. After the issue was entered in an appropriate system, the design basis for POPS was changed and the evaluation required by 10 CFR 50.59, for identification of a possible unreviewed safety question, was not performed (Exhibit 3).
Technical Backaround 4
The POPS uses two pressurizer power operated relief valves (PORVs) to mitigate low tem >erature (<312'F) overpressure transients, keeping the peak pressure below tw limits of.10 CFR 50, Appendix G, " Fracture Toughness Requirements,"
for brittle fracture protection. The Appendix G limits are incorporated in technical specifications (TS) as pressure temperature (P/T) curves specific to each unit's reactor vessel. The original design basis mass addition transient for the POPS was based on the start of a safety injection pump (POPS was780 gpm) and its injection into a water solid reactor coolant system (RCS).
designed to meet the single failure criterion, with either PORV having sufficient relief capacity to limit the peak pressure to less than the P/T curve limit (Exhibit 3, p. 1).
An NRC safety evaluation report, dated February 21. 1980, associated with Amendment No. 24 to the Unit 1 TS, oproved the Salem POPS setpoint of 375 pounds per square inch gage (psig), )ased on the calculated peak transient pressure of 446 psig and a 14 psi margin (at that time) below the Unit 1 Appendix G limit of 460 psig. Requirements for the Unit 2 POPS were incorporated into the unit s TS prior to initial startup and were approved based on the Unit 1 POPS safety evaluation (Exhibit 3 p.1).
The P/T limits for all reactor vessels decrease with successive operating cycles due to irradiation effects on the vessel materials. TP 'efore, margin between the peak transient pressure and the P/T limit will cha-as subsequent revisions of P/T curves are reviewed and approved by sne NRC. The Salem Unit 1 P/T curves were revised in February 1990 in TS Amendment No.108, which established a more restrictive limit of 450 psig at low temperatures.
The Unit 2 P/T curves were approved (at the same time) in TS Amendment No. 86, which established a limit of 475 psig. These curves are valid for up to 15 effective full power years of operation (Exhibit 3
- p. 1).
Setooint Nonconservatism On March 15, 1993, Westinghouse issued a Nuclear Safety Advisory Letter (NSAL 93 005B) informing PSE&G about the nonconservatisms in the setpoint methodology for POPS. The dynamic head, resulting from running reactor coolant pumps (RCPs) and the static head, due to elevation of sensors relative to the reactor vessel midplane, were found not to have been considered in the original setpoint methodology. The static head error for Salem is relatively small, resulting in a 4.7 psi increase in the peak transient pressure.
However, the dynamic head error is more significant.
Each operating RCP will PW C D b IT M AP M. OF/\\ f F-C CT, OFF N
p REG (ON I Case No. 1 95 013 11 l
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increase the difference between pressure at the reactor vessel midplane and that sented by the POPS instrumentation by approximately 25 psi.
Consequently, for a four loop plant such as Salem, the ser. sed )ressure (with all four RCPs running) could be as much as 100 psi less than tle actual pressure at the reactor vessel midplane (the area of concern for P/T cuFves).
These errors simply can be added to the original peak transient pressure si.&
their effect is to offset (nonconservatively) the pressure at which POPS will actuate, NRC Information Notice (IN) 93 58, "Nonconservatism in Low-Temperature Overpressure Protection for Pressurized Water Resctors," was issued on July 26, 1994.
The IN noted that administrative restrictions,'
recommended by the Westinghouse NSAL, were intended to provide interim actions until either setpoints were verified to be accurate, or appropriately revised in TS (Exhibit 3 pp. I and 2).
In December 1993, after reevaluating (over a nine month period) the original POPS analysis to address the NSAL concerns, PSE&G determined that the corrected peak transient pressure would exceed the P/T limits of both units.
Even with limiting the number of running RCPs to two, the corrected peak pressure would be 485 psig (applicable-for either unit since the analyzed transient is the same). On December 30, 1993, the licensee dispositioned the issue by memorandum (MEC 93 917) (Exhibit 31), administratively limiting the maximum number of RCPs in service to two when RCS temperature was below 200*F (limiting the dynamic error in the most restrictivo area of the P/T curve),
and increasing each unit's P/T limit by 10% using an unapproved American Society of Mechanical Engineers (ASME) Code Case N 514. The inspector noted that, at temperatures above 200'F and u) to 312*F, the Appendix G P/T cu ves allow for much higher pressure limits (Exhibit 3, p. 2).
The inspector considered that at the point PSE&G became aware that the margins to TS P/T limits for Appendix G brittle fracture considerations were not only reduced but, in fact, lost (and the Ap>endix G limits could be potentially exceeded) both Salem Units could >e potentially opu ated in an unanalyzeo condition (whenever below 312*F) which would be outside the plents' design bases. Therefore, the condition was reportable, and thy licensee's failure to make such a report is an a requirements of 10 CFR 50.72 and 73. p)arent violation of the reporting urther, the inspector noted that an exemption request for use of ASME Code Case N 514 had not been submitted by PSE&G until late December 1994.
Use of the ASME code case would reouire preapproval by the NRC, either generically via regulatory guide or specifically for Salem by exemption from 10 CFR 50.60. The licensee's reliance on the then unapproved ASME Code Case N 514 for over one year without the required exemption is an apparent violation of 10 CFR 50.60 (Exhibit 3.
- p. 2).
Less than one month after the issue had been dis)ositioned in Memorandum HEC 93 917, the licensee recognized that the ASME code case could not be used without prior NRC approval. The licensee then sought to credit the capacity of the residual heat removal (RHR) suction relief valve RH3 to augment the analyzed POPS relief ca>acity. The spring operated relief valve (RH3) has the same set >oint as POPS, )ut has a greate effective flow area and will actuate faster t1an a PORV once its setpoint is reached. A subsequent analysis by PSE&G confirmed the licensee's initial judgement that, with RH3 evailable, the 1n3T F PL5 b DISCL I
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, RE 13 I Case No. 1 95 013 12
3eak pressure would remain below the Appendix G limit. The issue of crediting U13 as part of PDPS (without either a 50.59 safety evaluation or prior NRC a> proval by changing the POPS TS) was under consideration from mid January t1 rough mid April 1994. On April 19, 1994, a Discrepancy Evaluatica Form (DEF 94 0060) was written to document the fact that relief valve RH3 was not credited the original POPS analysis for Salem or in the existing licensing and design basis (the NRC safety evaluation) for the system. The inspector noted that at this point, PSE&G had attempted to resolve the issue for over a year without entering the fundamental engineering question (the adequacy of the POPS setpoint) in either of the two existing PSE&G quality systems for resolution of Salem engineering discrepancies (the Incident Report System or DEF process). The inspector concluded that the licensee's failure to initiate corrective actions for this significant condition adverse to quality, is an apparent violation of 10 CFR 50, Appendix 8. Criterion XVI, " Corrective Action" (Exhibit 3, pp. 2 and 3).
Corrective Action Process Initiatefj The licensee's April 1994 DEF addressed the immediate safety concern by assuring the availability of relief valve RH3 and considering the safety margins discussed in the ASME code case.
At this time, the licensee also initiated a procedure revision to limit the number of running RCPs in Mode 5 (below 200*F) to one pump, thus further minimizing the dynamic head error in the most restrictive region of the P/T curves. Since the licensee concluded that there was no immediate operability concern, they sought to find other reasons why the Westinghouse nonconservatism did not apply to Salem.
The inspector noted that, even after the issue was entered into the PSE&G DEF process, the condition outside the design basis was still not reported to the NRC (Exhibit 3, p. 3).
The ins)ector independently assessed the availability and capability of relief valve R-13 for sunlementing the POPS.
Valve RH3 is available for RCS pressure relief when the RiR system is aligned for shutdown cooling. Review of Salem integrated operating procedures 10P 2, " Cold Shutdown To Hot Standby," and 10P 6, " Hot Standby To Cold Shutdown" showed that RHR shutdown cooling will be in service when POPS is required to be operable (<312'F). One reason valve RH3 was not credited by the NRC in the original 1980 POPS analysis was that an automatic closure interlock would shut the RHR suction valve on high RCS pressure, isolating RH3 from the RCS. However, this interlock was removed from both Salem units in the late 1980's. This change was generically reviewed by the NRC under Westinghouse Topical Report WCAP 11736. " Residual Heat Removal System Autoclosure Interlock Removal Report," and was subsequently approved by the NRC in a safety evaluation, dated August 8,1989.
The inspector also reviewed the valve's relief capacity, actuation response time, and calibration schedule. The inspector concluded that valve RH3 would be available to supplement POPS based on the procedural requirements and would substantially reduce the >eak transient 3ressure based on its desigri.
However, crediting valve (H3 as part of 20PS would. in the inspector's estimation, require a change to the Salem Technical Specifications (Exhibit 3, pp.-3 and 4).
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Case N 1 95 013 3
The issue was once again closed (by memorandum, dated 5/26/94) (Exhibit 33),
l based on a procedural requirement to achieve a pressurizer " bubble" (saturated conditions with a steam space) before starting a RCP. Because of this requirement, it was reasoned that only a correction for static head was l
necessary (relatively a small effect): therefore, the original analysis was concluded by PSE&G to be still valid (Exhibit 3, p. 4).
The inspector noted that the procedural requirement to have a pressurizer
" bubble before starting any RCPs was in place, and had been previously reviewed in the 1980 NRC safety evaluation remrt for POPS. Although the DEF was closed by PSE&G via this memorandum, furt1er analyses to support license changes for (crediting valve RH3 and using the ASME code case) were continued in anticipation of future, more restrictive revisions to the P/T curves.
During this analysis, the licensee determined that the effects of a running RCP on the POPS analysis should also be considered. However, no formal 50.59 safety evaluation had as yet been performed (Exhibit 3. p. 4).
Revised Desian Basis Transient On September 27, 1994, Problem Report (PR) No. 940927126 was initiated after the licensee determined that they could not rely on the establishment of a pressurizer
- bubble" to resolve the problem.
Since the original POPS analysis would not provide acceptable results after the effects of running RCPs were considered. engineering personnel establi:.hed what they considered a more
" realistic" transient as the design basis event for POPS (Exhit,:t 3, p. 4).
The origit al transient was simply the start of a safety injection pump (the intermediate head pump delivers 780 gpm) and its in ection into a water solid RCS. The licensee s revised transient is mechanist c and relies upon procedural controls for limiting possible injection sources. The revised transient begins with the reactor in Mode 5 (<200'F). the positive disp'acement (PD) charging pump in service and one RCP running, whereafter an inac.ertent safety injection (SI) sigr,al would cause the centrifugal charging pump (high head SI at 560 gpm) to start, the PD charging pump to trip, and the isolation of letdown to the chemical and volume control system.
Evaluation of this transient (mitigated by a single PORV having a 375 psig setpoint) using the GOTHIC computer code resulted in a predicted peak pressure of 438 psig, below the P/T limits of each unit. Therefore, by limiting the magnitude of the mass addition, tha licensee was able to reduce the predicted peak transient pressure and justify the existing TS setpoint for POPS (Exhibit 3,
- p. 4).
By the end of September 1994, the licensee again reached a final resolution and closed the issue (for the third time in nine months) because the revised transient could be mitigated by the original POPS hardware with the existing 375 psig TS setpoint. Although the licensee had not changed the TS setpoint, they had changed its technical justification bv revising the limiting transient upon which the setpoint is based. TherelevantTechnical Specifications for the Salem Unit 1 POPS are 3.4.9.3 and Bases 3/4.4.9.3: and for Salem Unit 2 they are TS 3.4.10.3, and Bases 3/4.4.10.3.
Further, the new transient, which changed the design basis for POPS. also invalidated the NRC's SER upon which Amendment No. p and th units' POPS T was based.
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description of the limiting transient and the design bases for POPS in Salem FSAR Section */.6.3.3 were, therefore, no longer correct and current (Exhibit 3, pp. 4 and 5).
The inspector noted that, since March 1993, several PSE&G corrective action programs were used but none was effective in resolving the Salem POPS issue.
Another (more recent) opportunity to evaluate all the relevant considerations of this issue was missed: as of the conclusion of the exit meeting on December
- 19. 1994, no safety evaluation was performed to determine if the change in the POPS design basis transient had created an "unreviewed safety question." 10 CFR 50.59 requires licensees to evaluate changes to the plant or its 3rocedures (including methods and modes of operation), prior to those changes
>eing effected, to assure no unreviewed safety question exists.
(The licensee's failure to >erform this safety evaluation was an apparent violation of 10 CFR 50.59) (Exhi ait 3, p. 5).
- New" Transient Amended In November 1994, the licensee recognized that an error recently identified in their configuration baseline document would adversely effect their assumptions for the revised PODS mass addition transient. The configuration document had incorrectly assumed that the positive displacement (PD) charging pump trips off on a SI signal: however, if off site power is available when the SI signal occurs, the pump continues to run and trip signals are blocked (until the SI signal is reset). Atter discovering this error, analysis for the limiting POPS transient was revised to include the mass addition of the PD charging pump and resulted in a calculated peak pressure of 474 psig (Exhibit 3, p. 5).
PSE&G Incident Report (IR) 94 419, dated November 17, 1994, documented this latest discovery and concluded that the Unit 1 POPS no longer met its design basis single failure criterion because a single PORV could no longer mitigate the transient.
PSE&G reported this to the NRC under 10 CFR 50.72 as an unanalyzed condition for Salem Unit 1.
IR 94 419 provided justification for the continued operation of Unit 1 based on RHR relief valve RH3 being available to augment POPS. With the three valves (two PORVs and RH3) available below 312*T, sufficient relief capacity was reasoned (by the licensee) to be provided and the single failure criterion could be met.
However, the licensee considered Unit 2 to be "not reportable" because with a single PORV the peak transient pressure was still 1.0 psi below its P/T curve limit (Exhibit 3. p. 5).
The inspector concluded that, since the margins to safety for overpressure protection (viz. peak pressure versus Appendix G P/T limits) had either been significantly reduced or lost altogether (depending upon which transient and assumptions are adopted as limiting), the new ' limiting" transient represented a potential unreviewed safety question (Exhibit 3, p. 5).
RCS Vent Path TS for both Salem units require cold overpressure protection be provided by either the redundant PORVs (the POPS system) or reactor c lant system vent APb !
T Pl DI L OF F CE DIRECT F CE ST k I I
Case 460.
95 013
of greater than or equal to 3.14 square inches (in'),
Venting the RCS is an alternative to having the POPS op' rable and would be accomplished after e
depressurizing the RCS. The TS action statement for POPS requires that, in the event a PORV fails and cannot be restored within seven days, the reactor must be depressurized and vented through the 3,14 in' vent within the next eight hours (ixhibit 3, p. 6).
The inspector could find no specific justification for the TS required vent area of 3.14 in'.
However, the inspector concluded that the vent area required in TS should be adequato based on:
(1) the flow from an unrestricted openir,g of 3.14 in' nuld encounter less resistance than that through a single PORV; (2) a single PORY must be shown to provide sufficient relief capability even with its delay for actuation; and (3) the vent area is passive protection and, therefore, does not need to be redundant (Exhibit 3, p. 6).
The inspector noted that while no formal analyses were available to support the 3.14 in' area or compare it to actual PORV capacity, the full open port area of a single PORV is approximately 2.2 in'.
The " equivalent throat area" (a term used by Westinghouse in WCAP 11640, March 1988) of a full ow n PORV would be adjusted for hydraulic resistance, and factors affecting &G from the t11s correlation were provided in a December 8,1992, memorandum to PSE valve vendor, Copes Vulcan.
The vendor's memorandum depicts the estimated flow coefficient as a function of valve lift or opening during its 1.5 second strott. The Salem PORVs are 2 inch diameter Model D 100
- plug in cage" valves with flow coefficients on the order of 50. This flow coefficie" can be used, along with previously compiled EPRI test data for these type valves, to calculate a so called eouivalent area corresponding to a smoothly convergent nonflashing (sonic flow) nozzle that licensee thermal hydraulic engineers estimated to be 1.21 in' (Exhibit 3, p. 6).
The RCS vent area of 3.14 in', by itself, has no hydraulic meaning unless a geometry can be assumed so that resistance and loss factors can be calculattu.
Nonetheless, a single PORV gagged open is clearly enveloped in the RCS vent configuration by a simple two inch diameter flanged vening w'. hen corrected for flow losses: this effective area is on the order of 2 2.5 in The licensee, in fact, utilizes several options to establish the RCS vent, including removal of a steam generator primary side manway, removal of one or more code safety valves, or gagging open the PORV's as an alternative to POPS valves set to automatically relieve at 375 psig (Exhibit 3
- p. 6).
Cofje Case Acoroval The inspector reviewed the licensee's documentation and interviewed personnel involved with the POPS issue during the 20 months between the NSAL issuance ia March 1993 and PSE&G's 50,72 notification in November 1994. The ins wetor concluded that there was an adequate assurance of safety, based on tie additional relief capacity of valve RH3 and.the margin that can be gained with-use of ASME Code Case N 514.
Based on the ins metor's discussions (prior to Februa?y 1995) with representatives from the NRC Office of Nuclear Reactor Regulation (NRR) ASME Code Case N 514 represented a technically acceptable position, although a plant specific exemption would be required (Exhibit 3, pp. 5 and 7).
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I On December 16,19')4, a conference call between PSE&G and NRC representatives was held to discuss the licensee's more immediate actions to resolve certain aspects of the POPS issue. During this call, the licensee committed to limit the number of RCPs in service (per existing procedures) when RCS temperature is below 200'F. and to maintain procedural controls preventing an intermediate head safety injection pump from injectinc into the RCS. These commitments were formally submitted in a letter to the NRC from PSE&3 issued later that same day (Exhibit 3, pp. 6 and 7).
On December 22, 1994, PSE&G submitted an application for NRC approval of ASME Code Case N 514.
Included in the submittal were the calculations supporting the new design. basis transient fon POPS. Without the code case, PSE&G credited valve RH3 on Unit 1 to meet the design basis single failure criterion for POPS. However, for Unit 2, the licensee did not credit valve RH3 because the unit' pressure, based on a single PORV, was predicted to be 1.0 psi below the peak r P/T limit (Exhibit 3, p. 7).
By letter dated February 13, 1995, the NRC issued an exemption from the requirements of 10 CFR 50.60 for Salem Units 1 and 2.
Tnis exemption permits using the safety margins recommended in ASME Code Case N 514 in lieu of the safety margins required by Appendix G to 10 CFR 50. Therefore, each unit's P/T curve limits for POPS were increased by 10%: the Unit 1 and 2 limits became 495 and 022 psig, respectively (Exhibit 3, p. 7).
e Insoection Conclusj_Oni The inspector considered several aspects of PSE&G's actions to resolve the POPS issue over the past 20 months as inadequate or inappropriate:
o The POPS issue was initially dispositioned to show that the requirements of 10 CFR 50.60 were met, invoking an A ME code case that had not received prior NRC approval.
e When inclusion of the setpoint nonconservatism put the Salem Units outside the POPS design basis, reports to the NRC were not made pursuant to 10 CFR 50.72 and 73.
No safety evaluation pursuant to 10 CFR 50.59 was performed prior tc e
revising the POPS design basis transient (described in the Salem FSAR) in September 1994. As of the Dccember 19, 1994, exit meeting, a 50.59 evaluation had not been completed. Several times during the licensee's attempts to resolve the POPS questions, the margins to safety for POPS (defined in the February 1980 NRC safety evaluation) were found to be reduced, but not appropriately evaluated.
e It took almost two years (March 1993 to February 1995) for PSE&G to take appropriate actions to address the NSAL nonconservatism. The lowest priority possible was assigned to this issue withia the Operational Experience Feedback )rogram in March 1993, an.1 the issue was not entered into an appropriate )SE&G quality program for resolving engineering discrepancies for over a year (Exhibit 3, p. 7).
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The re rt indicated that the licensce's:
(1) reliance on ASME Code Case N 514 w thout NRC approval. (2) failure to report a condition outside the Salem design basis, and (3) failure to perforn an adequate safety evaluaticr.
of the revised POPS design basis transient are all ap)arent violations of NRC requirements. Further, the process used to address tie issue (memorandum superseding memorandum) was considered to be fragmented and not appropriate for potentially safety significant issues. The report further concluded t5at the corrective action processes that were engaged (a year late) did not appropriately resolve a condition outside the plant's design basis (Exhibit 3,
- p. 8).
Coordination with Reaional Staff -
Several allegation panel meetings were held with the regional staff, including one attended by the Regional Administrator.
Brian McDERM0TT, a regional inspector (now a resident inspector at Susquehanna Steam Electric Station),
participated in the O! interviews and provided technical assistance during this investigation.
The results of this investigation we'
- easented to Region I staff and were considered, along with the staff's inspection findings, in the issuance of a notice of violation and proposed imposition of civil penalties against PSE&G on October 16, 1995.
Review of Licensee's Investiaation On April.i2, 1995, Synergy Consulting Servicts Corp. (Synergy), under contract with PSE&G. issued its technical investigation of several issues at Salem which had been identified by LASHKARI.
Synergy sustained two of U.SHKARI's concerns:
(1) " assuming credit for an ASME Code Case that had not been approved by the NRC, and... (2) the potential violation of the Technical Specifications /Apwndix G pressure / temperature limit by 0.7 psig." Synergy also noted that t1e licensee had "not yet submitted to the NRC a license amendment request" that LASHKARI r.sd initiated some manths prior to his departure (Exhibit 16, p. 22).
Synergy indicated that their investigation of LASHKARI's " concerns led to the identification of directly related potential violations. These involve (the) failure to recognize and report operation of the plant outside the licensing / design basis wherein analysis predictN that certain pressure transients could exceed the Appendix G/ Technical Specifications limit." Other already existing potential violations were identified, which involve (1) a failure to recognize the need to perform 50.59 safety evaluations related to adjusting the flow rate assumed for the safety injection pump in the analysis:
(2) a failure to recognize when the plant was operating outside its licensing / design basis when changing the analytical code used for the design basis calculations for the POPS without NRC a) proval; and (3) attempting to change the design basis events for which the 9PS is intended to provide protection without NRC approval. Synergy reported that because of the failure l
to recognize these situations, the required reporting was not made and corrective actions were not taken (Exhibit 16, p. 22).
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Ken O'GARA, a Principal Engineer assigned to Nuclear Licensing, indicated in an internal draft memorandum to Dennis SMITH that the SYNERGY findings noted Engineering's failure to
- initiate an Incident Report documenting that the POPS was outside its design (licensing; basis when Engineering received a letter frcn Westinghouse (9/93) presenting the plant specific results for Salem" and that this should nave been reported "to the NRC pursuant to 10 CFR 50,72/73 for plants being outside the Design Basis" (Exhibit 19, p. 7).
Coordination with Recional Counsel The investigative findings were discussed with Regional Counsel in conjunction with the proposed PSE&G enforcement action.
Alleoation: Intentionally Operated Outside Its Design Basis and Failed to Hake a Timely Notification to the NRC of an Unanalyzed Condition Regarding Salem's Pressure Over Pressure Protection System (POPS)
Summarv/ Evidence The following individuals were interviewed regarding LASHKARI's allegation that PSE&G failed to notify the NRC that Salem was operating outside its design basis:
HLms Position Date of Interviews Howard BERRICK Supervisor Hechanical Group, March 14 &
Group, NSSS, Salem, PSE&G July 13, 1995 Hahesh DANAK Senior Staff Engineer, March 14 & 18.
Engineering and Plant 1995 Betterment Departmer,t, Salem, PSE&G VijayCHANDRA Thermo dynamics Analyst, July 13, 1995 Technical Consultant, PSE&G Chandra LASHKARI former Senior Engineer, December 15, 1994 Salem, PSE&G John H0RRISON Technical Manager, February 15, 1995 Salem, PSE&G Gita NARASIMHAN Senior Staff Engineer, Mechanical July 13 &
(a former contractor with Saraent August 22, 1995
& Lundy). NSSS Group Salem, #SE&G Ken O'GARA Principal Engineer, Ratheon March 14 &
Engineers and Constructors August 22, 1995 (formerly Ebasco Services),
a PSE&G contractor at Salem
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David SMITH Principle Engineer, Nuclear March 14 &
Licensing, Salem, PSE&G August 22, 1995 Francis THOMSON former Manager, Licensing March 14 &
and Regulation (Salem & Hope July 13, 1995 Creek), PSE&G John WIEDEMANN former Supervisor, Salem, PSE&G February 15 &
March 15, 1995 Interview of WIEDEMANN (Exhibits 7 and 8)
WIEDEMANN recalled that he received a Westinghouse notice from Nuclear Engineering around June / July 1993 (Exhibit 8, p. 5). WIEDEMANN stated that
'the Westinghouse bul'.etin indicated that your vessel was susceptible to exceeding the maximum allowable stress because this pressure spike would develop." But if you could show that the pressure spike could not develop, then Salem was not outside of the heat u) and cool down curves and was not outside of a POPS non conservative (Exhiait 8, p. 30).
WIEDEMANN stated that the spike (seak pressure) described in the Westinghouse letter would not occur because of 1H3. But as far as what was currently written in the FSAR, Salem did not take credit for the RH3 (Exhibit 8, pp. 36 and 37).
WIEDEMANN stated that if Salem was in violation of its design basis, the a)propriate process for an engineer to get that information out to the rest of tie organization would be an IR (Exhibit 8, p. 8). WIEDEMANN recalled that LASHKARI presented him with a draft IR "in response to the concern Westinghouse had raised, that under a specific scenario you could achieve a high pressure spike that would lead to a failure of the reactor yessel.... This particular issue as described by Westinghouse differed slightly in how the Salem plant was configured. And that the means of mitigating the pressure spike at Salem was provided by not only the POPS system, but also the RH3 valve on the RHR system" (Exhibit 8, pp.10 and 24).
WIEDEMANN stated that when he discussed the POPS issue with LASHKARI, they "had an understanding" that Salem was not challenged by this particular sequence of events, because the RH3 would lift well in advance before Salem would achieve a pressure spike as identified in the Westinghouse bulletin (Exhibit 8, pp. 9.10, and 24).
WIEDEMANN understood that the concern was whether Salem could take credit for the RH3 valve, because it was not specifically defined as a POPS valve (Exhibit 8, p. 10). And, it was not aart of the POPS system, as defined in tech spec and the tech spec basis (Ex11 bit 8, p.11).
He recalled reviewing with LASHKARI the P0FS issue and "it was understood that that failure mechanism could not occur at Salem, the failure mechanism of achieving 479 psi pressure that would cause a failure of the reactor vessel.
If the POPS were challenged, then we had an equally sized RH3 valve that could also relieve.
So we had at least three valves that would relieve at a lower pressure well below the anticipated maximum spike" (Exhibit 7, p.12). And, he stated that LASHKARI agreed (Exhibit 7, p.13).
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,l WIEDEMANN stated he directed LASHKARI to take the POPS issue to the Licensing i
and Regulation Group to determine if the RH3 credit could be taken, and let them determine what would be the next step (Exhibit 8, p. 10). WIEDEMANN stated that he asked LASHKARI to contact licensing to determine whether Salem (PSE&G) would take credit for the RH3 or the code case (Exhibit 8, p. 43).
WIEDEMANN also stated that he turned the POPS issue over to the Licensing group because the question was whether you could take credit in Licensing l
space for the RH3. It was his understanding that if an incident report was needed, it would be submitted by the Licensing group or Nuclear Engineering after they had done their research on it.
It *may have been premature to submit an incident report when we didn't have all our homework done...."
(Exhibit 7, pp. 24 and 25). WIEDEMANN stated that the Nuclear Engineering Mechanical Section's analysis indicated that the RH3 was a viable option.
And, this issue was also being looked at by the Licensing Department (Exhibit 7, pp. 13 and 27).
WIEDEMANN felt that in the January time frame Salem was not complete in its evaluation and still needed to look at the RH3 and whether they could take credit for it. He told LASHKARI not to issue m incident report at that time and let Licensing look at the issue and follow up on it (Exhibit 8, pp.10, 11, and 27).
WIEDEMANN recalled that LASHKARI had been working on the Westinghouse PDPS issue for about six months before LASHKARI brought the IR to him. He knew LASHKARI had assistance from Nuclear Engineering, from the thermal hydraulic swcialist all looking at what would be the magnitude of the spike (Exhibit 8, p. 18).
WIEDEMANN stated that the Salem would not physically experience a pressure spike because of the RH3 valve. And even if the RH3 wasn't there, it was his understanding that it would not be a safety issue because of the additional margin afforded by the ASME Code case.
Salem could apply for a code case that I
would extend the allowable stress. So that depending on which avenue Salem wanted to pursue. Salem was covered on both casec as far as any safety challenge to the vessel. The approach that Salem would take was to be determined by Licensing and Regulation (Exhibit 8 pp. 17 and 18).
WIEDEMANN acknowledged that the RH3 valve was not written as the basis for POPS, but it appeared to him to be the way that the RCS would relieve its pressure, should that sequence of events occur (Exhibit 8, p. 37). WIEDEMANN stated that the situation that was postulated in BERRICK's memo would not occur at Salem because the RH3 would relieve at 375 pounds (Exhibit 8, p. 39).
WIEDEMANN stated that he believed that Salem had to apply for the acceptance of the code case that would allow for the extension of the allowable stress.
And. if the code case was acceptable. Salem could take credit for it even without POPS without the RH3 valve (Exhibit 8.-pp. 40 and 41). WIEDEMANN stated that the code case, as identified in the Westinghouse bulletin, allowed you the additional margin that showed that the pressure spike did not channel into the vessel. But regardless of whether the code case was applicable or not. Salem did not challenge the vessel integrity because of RH3 (Exhibit 8,
- p. 41).
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WIEDEMANN stated'that when LASHKARI first raised the issue of submitting an IR, Salem did not have a safeb concern as far as vessel integrity and that this was more of a Licensing space issue to determine what direction Salem wanted to take.
"Did we want to apply for a code case? Did we want to take credit for the RH3" (Exhibit 8, pp. 42 and 43),
Another point that WIEDEMANN made was that when LASHKARI first presented the IR to him it was not at the stage where it should be submitted. And as it turned out. Salem had not violated the heat up and cool down curves that challenge the vessel integrity (Exhibit 8. p). 44 and 45). The way the'IR was written it appeared to WIEDEMANN that LASHKARI's concern was that Salem had violated the Appendix G cool down' curves. WIEDEMANN did not feel that was Salem's position in January, that the pressure spike, should the scenario occur, would have been mitigated and Salem would not have violated the :urves (Exhibit 8, p. 45).
The way WIEDEMANN read the IR (Exhibit 13) it tells the Operations De)artment
_ j that Salem violated the heat up and cool down curves and he does not t11nk that is the case. He did not want the IR to go forward because there was t
additional work to be performed to show that Salem did not violate the curves (Exhibit 8, pp. 46 47).
WIEDEMANN denied having seen the two LASHKARI memoranda, dated January 30, 1994 (Exhibit 14), and April 22, 1994 (Exhibit 15), within several months of their issue dates (Exhibit 8, pp. 54 and 55).
WIEDEMANN stated that an operability determination would be made on an IR during the same shift that received it (Exhibit 8, p. 59).
WIEDEMANN stated that if he believed that Salem was outside its design basis that an IR would have been an cpproariate vehicle to raise that issue to Operations (Exhibit 8, p. 64). WIE]EMAb'N stated that the RH3 or the code case insures that Salem is well within its POPS requirements (Exhibit 8, p. 72).
AGENT'S NOTE: WIEDEMANN' statement still belies the fact that the code case had not been approved; his main thrust seems not to be addressing the units design basis (See Exhibit 33).
WIEDEMANN acknowledged that Salem needed NRC appro,'al for the RH3, and he discussed with LASHKARI that Salem had to apply to the NRC for the code case (Exhibit 8, p. 72).
WIEDEMANN stated that, as a result of the training he received on operability determinations in 1994, as a conservative gesture, Salem could have written an IR when the Westinghouse bulletin came in.
But, he would have handled hit decision en the submission of LASHKARI's IR the same way. He still would have told LASHKARI that they needed to do a little more homework to determine if the heat up and cool down curves were violated (Exhibit 8, pp. 62 and 63).
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Interviews of THOMSON 6xhibits 9 and 26)
THOMSON stated that he was told in April 1994 that the issue raised in the Westinghouse letter "was dispositiond initially at the end of the 1993 year taking credit for an ASME code case with the NRC approval" (Exhibit 9 p. 9).
As he understands it, the letter indicated that there was not a problem with the Salem limits beciuse of the code case (Exhibit 9 pp. 9 and 10).
THOMSON stated that O'(MRA communicated to Engineering in the end of January / February 1994 that Salem could not take credit for a code case that was not approved by the NRC (Exhibit 9 p.10).
It was THOMSON's understanding, at the time, that Engineering communicated to O'GARA that there was adequate margin to meet the design basis even without the code case and O'GARA was concerned that they should be writing an IR because Salem was cut of its design basis (Exhibit 9, pp.1113).
THOMSON had a meeting near the end of April 1994 with Jerry RANALLI. O'GARA.
Rich VILLAR (Station License Engineer). Vijay CHANDRA (Engineering), and Dave SMITH (0'GARA's supervisor) regarding the POPS issue (Exhibit 9, p. 12).
THOMSON stated that there was never a question in his mind that there was a safety issue.
nowing that the NRC had allowed other people to take credit for the code case which gives you 10 percent more margin under cales and knowing that they had the RH3 valve available, which is an extra relief in the system (Exhibit 9. p. 13). THOMSON represented that after the meeting no one present believed that this was a safety issue (Exhibit 25. p. 5).
THOMSON stated that it was his understanding that there was never a question in his " people's minds or in Design Engineering that there was a safety issue.
The reason for that... [and he agreed) was knowing that the N.R.C. had allowed other people to take credit for the code case which gives you 10 percent more margin under cales and knowing that... [ Salem) had the RH3 valve available which is an extra relief valve in the system, recognizing that... [they) had those two things available that... [they) had not credited yet in... [their] calculations, there was a very high level of confidence that there was not a real problem.
It was more a compliance problem with tech specs" (Exhibit 9. p. 13).
THOMSON stated that when they left the April meeting. "it was the consensus that... (they) had c. very high level of assurance a reasonable assurance that
.. [they) were within our design basis from a compliance point of view. That was based on a couple of things. One was recegnizing that Design Engineering was going to do more cales and could take credit for the RH3 valve. At that point in time... [they) thought that was legitimate.
Recognizing that Engineering had told Ken had told... [him) they had done a very simple calculation a more detailed calculation that could be done on... the Gothic Code.
[T] hey had a very high level of assurance that they would be able to show that their peak pressures would be below the limits required by the tech specs. THOMSON felt very comfortable that... [they]
had reasonable assurance that... [they) were within... their] design basis and there was no 1 iden report required at this point in time."
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Th0MSON noted that he also felt comfortable with the fact that there was a DEF, THOMSON stated that everyone left the April 1994 meeting with a high.
level of reasonable assurance that they were within the design basis from a compliance point of view (Exhibit 9, pp.14 and 15).
AGENT'S NOTE: The key point is that THOKSON knew that there was a coda case (N bl4) which would give Salem the margin it needed and that, because of this, he did not believe there was a safety question. He then sought to find an alternative method of coming within current specifications without having to go to the NRC for a code case acceatance. While it may have been the judgement of THOMSON and others at tie A)ril meeting that the code case would be approved by the NRC, it had not >een done. The report that Salem believed that it was operating outside of its design basis and in en unanalyzed condition should have been 'nade to the NRC. THOMSON made a decision to explore other possibilities, without any immediate concern for time. Salem had already relied on the code case, since BERRICK signed the memorandum on December 30, 1993. This writer believes that he had one hour to resolve the problem, when Salem learned the code case could not be used, or call the NRC: THOMSON did not notify the NRC, but re examined the issue.
THOMSON stated that a DEF " kicks off a process where you evaluate potential design deficiencies and throughout that process it requires you at any time you have enough information to warrant an incident report....
it is something where the engineering deficiency could be safety significant if it were true then you are required to analyze it in a fairly quick time frame" (Exhibit 9, p. 15).
THOMSON stated that because of the code case, the RH3. and the more detailed cales that were being prepared, he had a high level of assurance that Salem did not have a problem with its design basis (Exhibit 9, p.17). THOMSON stated that in April 1994 he " thought it was legitimate to also credit RE3 (sic) based on generic approval of Westinghouse W cap" (Exhibit 9, p). 38 and 42).
But frora a compliance point of view he knew Salem could not tace credit for the code ca.ce (Exhibit 9, p. 19).
In May 1994 the results of the calculations indict ted 450.7 peak pressure with the transient and a time when the limit was 450 (Exhibit 9, pp. 21 and 22).
THOMSON stated that he agreed that 10 CFR 50.59 requires a safety evaluation for proposed changes to the plant as described in the FSAR and that type of evaluation was necessary to add RH3 to the POPS system (Exhibit 9, pp. 20 and 21).
AGENT'S NOTE: Mc0ERM0TT stated that what Salem was doing was taking credit for a valve that was not part of the system as originally analyzed (Exhibit 8, p. 37).
Although THOMSON stated that in April 1994 he thought Salem could take credit for RH3 as long as the autoclosure lock was removed, he made it very clear that "you absolutely can't take credit for the code case.... Based on the code case, credit for RH3 and the fact that this Gothic Code would Af F
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l give... more margin... everyone agreed that there was not a safety issue here" (Exhibit 9, p. 39).
In April. THOMSON stated that he did not consider the plant to be outside its design basis because he was given reasonable assurance that even without taking credit for the code case more detailed cales would show they were inside design basis (Exhibit 9, p. 40). THOMSON recalled that RANALLI, CHANDRA, DANAK, and VILLAR all gave him reasonable assurance of this (Exhibit 9, p. 41: see also Exhibits 18 and 29).
THOMSON stated that, until recently, he was not aware of the draft IR prepared by LASHKARI on the POPS system (Exhibit 9, pp. 26 and 27). At the April 1994 meeting, THOMSON did not remember discussing a draft IR by O'GARA (Exhibit 9,
- p. 43). But, he subsequently learned that O'GARA had drafted an IR, prior to the meeting, stating that Salem was outside its design basis (Exhibit 9,
- p. 44). THOMSON acknowledged that in November 1994, PSE&G made a 50.72 notification to the NRC on the POPS system being outside of its design basis (Exhibit 9, p. 47: see also Exhibit 30). THOMSON statN that it was his understanding that a safety evaluation was done for the change in the mass addition assumptions (Exhibit 9
- p. 54).
AGENT'S NOTE: During his interviews, O'GARA seemed very confident that he had prepared the irs (Exhibits 20 and 32) and one of them was discussed. McDERM0iT stated that during his inspection of Salem, he was told that the safety evaluation THOMSON referred to was not done (Exhibit 9. p. 54).
THOMSON stated that he recalled discussing at the April meeting that " people felt good that they were able to resolve the problem and demonstrate to a high level of assurance that we were within our design basis without going having an incident report and cause an unnecessary shutdown." He did not recall discussing whether an IR should be filed at a time when an NRC AIT team was present, or if an AIT was present (Exhibit 9, p. 61).
THOMSON denied that there had ever been any discussion as to the cost of shutting the plant down versus the fine the licensee would receive from the NRC (Exhibit 9, p. 62).
Interviews of O'GARA (Exhibits 10 and 25) 0'GARA stated that on January 25, 1994, LASHKARI told him that in BERRICK's December 30, 1993, memorandum (Exhibit 11), Mechanical Engineering had taken credit for an ASME code case (N 514) which LASHKARI did not believe was approved by the NRC in general terms for the nuclear industry (Exhibit 10, pp. 6 8). After reviewing the code case. O'GARA agreed with LASHKARI and tried to have Mahesh DANAK in Mechanical Engineerir.g write a DEF to document the concern (Exhibit 10, p. 9).
AGENT'S NOTE: DANAK prepared DEF #94 0060, dated April 19, 1994.
regarding LTOP design and removal of non conservatism by including the pressure limits (Exhibit 21).
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According to O'GARA. the April 20, 1993, meeting in THOMSON's-office started with a discussion of whether to write a DEF or an IR: it was Hechanical's position that Salem could take credit for the RH3 valve, then the concerns they had which formed the basis for the DEF would not be valic (Exhibit 10, pp. 10, 11, and 38). D'GARA stated that DANAK's group (Mechanical) felt that by taking credit for RH3 it would make the problem go away and a DEF would not have to be written (Exhibit 10, p. 12).
However, they realized sometime in May, that they could not take credit for the RH3 (Exhibit 10, p.18). SMITH, his supervisor, agreed with O'GARA (Exhibit 10, p. 45).
Prior to the three hour Am 11 meeting, O'GARA felt that an IR was necessary and prepared a draft (sec'ond draft) which he copied to several people and passed it out at the April meeting: those THOMSON, D.J. CHANDRA, Rick VILLAR, SMITH,preent at the meeting were DANAK, BFootCK, and RANALLI. He recal %d the purpose of the meeting was to diser or not the draft IR reached a
valid conclusions and should be issued
- 1 pp. 14 and 36). They alst, discussed whether or not this 1ssue wa; c'
bh o the NRC: the final reportsbility determination is made by t-
%t.
M shift was not advised and they did not send the IR forward (E). 1 10, 42). O'GARA stated that it was the consensus of those present at cht ct ting that the P0FS issue that formed the basis for the DEF and IR was J. t aperability concern
(Exhibit 10,
). 16). As a result, Engineering.
do some final evaluations (Exhibit 10, p. 43).
O'GARA stated that he would not go forward with the IR without his supervisors and managers agreeing with the decision he made (Exhibit 10, p. 17).
O'GARA recalled that Salem changed the design basis mass addition transient that wou?
be used to evaluate the system. The original basis included the intermed ne head pump and the r.evised basis included the high head. They went back and changed the tech specs basis to clarify which pumps ware going to be in operation (Exhibit 10, p. 28). But they did not recognize unt'l the December 1994 time frame that it was actually a design basis change that required a tech sxc change in order to take credit for the centrifugal charging pump (Ex11 bit 10, pp. 29 and 30).
O' GAP.A also stated that when the decision was made to change the flow rate they did not consider that to be a des 9n basis change (Exhibit 10, p. 31). A 50.59 review was not done in the Sep: mber time frame. All they were doing was taking credit fcr the way they currently operated the plant (Exhibit 10, p. 32).
O'GARA indicated that he had problems with the fact that between December 1993 and May 26, 1994 he was being told that Salem could take credit for the RH3, yet he was not told if it had been appropriately approved for use at Salem (Exhibit 25, p. 57).
O'GARA stated that he did not recall anyone discussing the comparative costs of shutting down Salem versus being fined by the NRC (Exhibit 10, p. 52).
O'GARA recalled giving a draft a copy of an IR he prepared to LASHKARI and telling LASHKARI that it was his intention that it be issued the day of the April 1994 meeting (Exhibit 10, p. 55)
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AGENT'S NOTE:
Four versions of 0'GARA's draft irs are attached (Exhibits 20 and 32) and show similarities with the final IR prepared by O'GARA, with the LER attached (Exhibit 35).
-In an April 6, 1995, memorandum to his supervisor, SMITH, and, in respcnse to an NRC inspection report (94 32), O'GARA indicated that, with the information he had on that date, he believed Salem was operating outside its design basis as of December 30, 1993 (Exhibit 25, pp. 46 and 47). At this point, O'GARA had not met with, nor had he discussed this issue with NARASIMHAN (Exhibit 25,
- p. 58),
Interview of H0RRISON (Exhibit 4).
H0RRISON recalled a discussion with the outage manager and plant manager during a walk around the plant: about the fact that the present sizing of the POPS valves and the tech specs would require Salem, in some cases, during outages, to remove pressurizer safety valves in order to have adequate over pressure protection, because of the sizing of the valves. There was a perception that, with further analysis, they could show that the POPS valves themselves provide adequate relieving capability and then they would not be burdened with physically removing other valves to provide that.
It was in the context of, during these outages, "here's a step we have to do.
Is this something we can pursue through engineering and through tech spec change to provide for more efficient outages." He recalled that it was a conceptual type of discussion (Exhibit 4, p. 13).
H0RRISON stated that he had not :,een the two memoranda dated January 30, 1994,
.and April 22, 1994, from LEHKARI and did not recall getting any memoranda i
directly from LASHKARI (Exhibit 4, pp. 11 and 12)
Interviews of BERRICK (Exhibits 5 and 27)
BERRICK stated the first time he became aware of the POPS issue was as the NSSS Group supervisor in 1990 when the issue was raised by Westinghouse (Exhibit 5, ). 6).
NARASIMHAN prepared a res)onse memorandum (Exhibits 11 and
- 31) to SCHNAIR, for BERRICK's signature (Exhiait 5, p. 7).
BERRICK noted that a code case had been develo>ed within the industry, N 514, and allowed a 10 percent increase above the P0)S 3ressure; so based on that 10 percent allowance. it was felt that Salem was witlin that window, and therefore, the op6rability of the systems was not impaired (Exhibit 6, p. 9).
BERRICK recalled that he became aware of the need for NRC code case approval from Licensing (Exhibit 5, p. 9).
BERRICK did not recall seeing a draft IR prepared by LASHKARI (Exhibit 5, p. 11).
However, upon reviewing the draft IR, BERRICK stated that LASHKARI's IR did not present any information that his "3eople" were not aware of, in that same time frame (Exhibit 5, p.12).
BiRRICK stated that if "one could not resolve the question in a reasonable manner, then it would be raised in an IR" (Exhibit 5, p.12).
BERRICK was not sure if the POPS system was operable: they had not reviewed the calculations to find out whether th was s p conserv tisms that they IF IC DI L WI AP A 0F i
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Case No. 1 95 013 27 l
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had taken, some methodology that had been previously used, that was conservative and would have rendered Salem being ckay rather than not okay (Exhibit 5, p. lA).
By relying on a bubble in the pressurizer that was required for the operation of the reactor coolant pumps to minimize the effects of the Westinghouse information, this was a change in the design basis condition assumed for the POPS system.
(The technical specifications basis for the POPS system indicates that the POPS system should negate pressure caused by the injection of a single safety injection pump into a cold solid reactor system.) (Exhibit 5, p. 15).
BERRICK stated that, typically, his engineers would review any background documents they could find in order to assess whether or not there was margin that they could gair or conservatisms they could take out, It was also the responsibility of his engineers to identify when changes were being made to the licensing basis or the design basis of the plant (Exhibit 5, pp. 20 and 21).
BERRICK stated that he would have expected that his engineers would have known the original flow rate and would have expected 7em to identify a decrease in that flow rate (as a change in the design bat b V the system) (Exhibit 5,
- p. 22). BERRICK did not recall any discussion et tne meeting about the need to do a "50.59" (Exhibit 5, p. 23).
BERRICK admhted that he considered the fact there was that change, that shift in what was the assumed mass addition; and if that was known by the engineer, that should have been identified as something that would require a 50.59 review. And, at that time, a 50.59 review was not done (Exhibit 5, p. 24).
BERRICK noted that the engineers that work for him would typically be responsible for searching out design basis and more specifically licensing basis information. However, in the 1993 and 1994 time frame, when this was done, it was more the expectation that the Licensing Engineers would provide that licensing basis, since they had a more extensive data base.
Since then, they have been advised to work more closely with them and to participate more actively with them.
But, at that )oint in time, his group would have relied on the Licensing Department, and t1e Licensing Engineers (Exhibit 27, pp. 5 and 6).
BERRICK indicated that at the April meeting there was a discussion long the lines of whether the plant was operable. There was no discussion about reportable.
It was, "is this something that would prevent the unit from starting up?" BERRICK did not recall any discussion about the cost of failing to start, or the cost of being fined at a later date (Exhibit 27, pp.11 and 12).
With regard to the December 30, 1993, memorandum to SCHNARR, BERRICK indicated that NARASIMHAN and DANAK both worked on its areaaration. When NARASIMHAN left Salem, the matter was reassigned to DANA(:
1e thinks that NARASIMHAN's initials were simply left on the memorandum. Although NARASIMHAN was aware that NRC approval was needed, BERRICK denied knowing that a code case approval was required before Salem could take credit for it in its analysis. The first time he acknowledged that there was a problem was in 1994, when LASHKARI raised the issue.
BERRICK stated that had he been aware of t e code case i
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requirement, he would have not said, this was closed. He would have indicated that this is pending the approval of that code case, and it would have gone off to Licensing with a request that they initiate the paperwork (Exhibit 27, pp. 12 15).
BERRICK indicated that the response memorandum went to SCHNARR because SCHNARR was the originator of the ATS item. He relied on SCHNARR to take any other necessary action (Exhibit 27, p 16).
BERRICK stated that had he known that in December 1993, that he could not rely on the code case exception, that it would have been a reportable event. He stated, "We would have, at that point without the code case, it's outside design parameters." When asked if he considered it reportable, he noted that:
"Well, in order to... have a discussion, should we use the code case or not, you have to look to see, at that point, if you do a calculation and the number is below your design number, then there wouldn't have been even a -
there would have been a long term consideration for maybe we ought to use the code case for long term to give us more margin, but we were looking at the fact that we had exceeded some number and we needed the cede case in order to give us that margin" (Exhibit 27, pp 20 and 21).
BERRICK acknowledged that, "because you could rely on the code case, you were within design basis, and therefore it was not a reportable event. But had you not been able to use that, then you would have had a reportable event."
BERRICK stated that he " failed to see that the code case needed... to have prior NRC approval " so they exceeded the limit (Exhibit 27, pp. 21 and 22).
BERRICK indicated that he did not prepare an IR "[b]ecause there were other -
we were looking at other calculations to see if, in fact, we would not have to even rely on the code case. Were we being too conservative and restrictive to start with in our thinking?
Was the the initial work that we did brought us above this limit, actually too conservative, too restrictive. Did we take
- s) we started to look at things apin, and rather than it was let's go back and do our homework one more time.
Let's sharpen our ancils and make sure that we understand completely what the whole story is wre" (Exhibit 27, pp. 17 22).
BERRICK stated that the purpose of the April meeting was to discuss the POPS issue O'GARA's IR and see if it would prevent the start up of the unit. But he did not recall if a draft IR was passed out at the meeting (Exhibit 5, pp. 24 26).
AGEKT'S NOTE: BERRICK's recollection seems to be much closer than THOMSON's, as indicated above, as to the purpose of the meeting.
BERRICK recalled that it was said that an IR was not required. but as they continued to do the work, if it was determined that an IR was required, it would be issued at that point in time (Exhibit 5. p. 26).
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Interview of SMITH (Exhibit 6)
SMITH stated that without the use of the code case, unless there were some other margins available. "that would mean that we would have to report that situation because it would render the system inoperable." But he was not concerned that they would not have the margin (Exhibit 6, pp. 6 and 7). SMITH indicated that he was not concerned because his expectation was that they would have the margin, because the indication he had was that the easy thing to do was to take the code case, but there was actually a lot of other margin that could be applied to the LTOP (Exhibit 6, p. 7).
SMITH recalled that O'GARA came to him with information that neither the IR process nor the DEF process were tracking the POPS issue. This issue was placed in the DER process. He believes that it was noted in the DER that if there was a problem with taking credit for the code case, they would have to report it. SMITH stated that in April 1994 the information he was getting was that there were other avenues to pursue to be able to say that Salem did not need the code case (Exhibit 6, pp. 8 and 9).
In March 1994 he became aware of the POPS issue which he referred to as LTOP.
There was a concern with the Engineering group taking credit for a code case and had " inappropriately taken credit for a code case" and O'GARA was investigating it and whether or not Salem should apply for that code case (Exhibit 6, p). 5 and 6). SMITH stated that they were looking for things that they could tace credit for (Exhibit 6, pp.14 and 15).
SMITH recalled that when he discussed the LTOP issue with O'GARA, O'GARA was concerned that this issue was not being tracked through an IR or DEF. But, he believes that one was initiated in April 1994 (Exhibit 6, pp. 7 9).
SMITH indicated that based on what he was getting for information was that there were other avenues to pursue to be able to say that Salem did not need the code case (Exhibit 6, p. 9). SMITH noted that if Salem had to take credit for the RH3, it was his understanding at the time that an FSAR change was needed, if at all possible to take credit for it (Exhibit 6, p.10).
SMITH noted that in September the POPS issue was still in the evaluation stage and if it turned out to be inoperable, we'd have to make the repcrt then.
SMITH knew that there was a difference between what he was told in May and what he was being told in September.
It appeared that they were back in the evaluation process (Exhibit 6, p. 25). SMITH felt that they had an acceptable analysis to su) maintain opera) port a code case application and an acceptable analysis to ility, albeit it was a very tight situation. That's why it was imperative that they get the code case application out quickly (Exhibit 6, pp. 26 28).
SMITH stated that he knew 0'GARA wanted to write an IR, which was written for a conle reasons. One was they " felt as though Engineering was not getting off t1eir duff" and they " wanted to show that this thing is very serious." He noted that "we could be outside our design basis. This is how serious it is.
We're looking at reportability with this issue.
You have a DEF now on your NQT LI DI OSURE "P
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hands. We've got to drive this to closure. So, yes, I was aware that there was an IR drafted" (Exhibit 6, pp. 35 37).
SMITH did not recall anyone at the April 20 meeting indicating that if the IR went forward, that it might shut the plant down (Exhibit 6, p. 40). SMITH did not recall anyone at that April 20 meeting mentioning that if the unit did have to shut down, it would result in a loss of revenue of a million dollars a day, but if it kept on producing, and if the NRC came after you later, the fine would only be a couple hundred thousand (Exhibit 6, p. 41).
SMITH stated that he never saw the aroblem report dated September 27, 1994, although he knew of its existence,(Exhibit 6, pp.17 and 18: see also Exhibit 34).
Interviews of NARASIMHAN (Exhibits 12. 23. and 28)
NARASIMHAN stated in 1993 she was a contractor and she was researching the POPS issue raised by Westinghouse in its March 1993 letter.
She was also responsible for preparing a draft of a letter on that same issue for BERRICK's signature (memorandum from BERRICK to SCHNARR, dated December 30, 1993)
(Exhibit 11, pp. 11 15: and Exhibit 23, p. 5). She does not recall whether she discussed with BERRICK that code case approval was needed (Exhibit 12 and Exhibit 28, p. 9).
BERRICK assigned her to investigate the Westinghouse NSAL issue in approximately April 1993. At that time, she was not aware of any letters that had been written from L'estingbouse to.PSE&G.
She left BERRICK in October 1993.
Prior to her departure, she gave him her draft and working papers. She recalled stating in her draft (Exhibit 31) that if Salem could rely on a code Case exception, that it "would have to invoke the code case." At the time she left in October, she was not sure whether the code case had been issued by ASME. But, she recalls having had discussions with Westinghouse and with other people on the applicability of the code case to Salem's particular situation.
She recalled that they said it was applicable to Salem's situation and that Salem would have to have approval prior to invoking the code case.
But, she does not recall if she specifically pointed that out to BERRICK.
She did noce that the draft letter said they were invoking that code case (Exhibit 28, pp. 6 9).
She also recalled having discussions with DANAK in about the March April 1993 time frame and later with Fred SERWAN in Operations.
She was hired as a direct PSE&G employee in September 1994 and worked on the POPS issue again in December 1994. At that time, she was doing a peer review on calculations performed by Vijay CHANDRA (Exhibit 28, pp. 9 12: see also Exhibits 18 and 29).
She contacted Tom ROBERTS, QA Department, and discussed with him a pending ASME code case and how she could get copy of it. She is not sure if they discussed the need for NRC approval before it could be used at Salem: but, she did get the draft copy of the ASME code case from ROBERTS. She knew from the draft she read that it was not plant specific (Exhibit 12).
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Case No. 1 95 013 31
She stated that her initial focus was on safety significance.
In a conversation Westinghouse indicated 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 remembersaskingWestinghouse(DiT0 MASS 0)inMarch1993,whyWestinghousehad not made this a Part 21 report, In May 1993, she s)oke wita Roger WATERS, at Westinghouse, on POPS. She made some notes on her )0PS 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 PDPS inspection at Salem (Exhibit 12).
Later, Westinghouse told her what other plants were doing, but she does not know if the code case had been ap med by that time, because she had initially read only a draft.
She learned that Florida Power and Light (FPL) was going to use the code case; but she did not discuss this matter it with anyone at FPL. However, she did discuss it with a Zion representative. She requested plant specific information for Salem. She recalls that there was a requisition document that she prepared for the plant specific work she wanted Westinghouse to do on the POPS system, but does not recall whether Mahesh DANAK helped her prepare it (Exhibit 12).
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 with an issue w1ere 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 not recall at wnat point she realized that (Exhibit 12).
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. At the time, she was only a contractor working within the groua and she did not think that she needed the technical information that LASHKAll would have provided. She did not see the need to talk with him: she believed she could get everything she needed from DANAK (Exhibit 12).
At the time she was working 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 otten she discussed issues related to POPS with him. She had also spoken with other members of the group on other issues (Exhibit 12).
She did not recall telling anyone that she was leaving Salem at the end of October 1993 or 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 BERRICK memorandum. 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 (Exim.it 12).
In the December 1993, through June 1994 time frame, she worked part time as a contractor employee for Sargeant & Lundy, a Salem contractor. During this period, she does not recall discussing the POPS issue; the work she was doing was not related to POPS (Exhibi 12).
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Aoent's Analysis Mana9ement had numerous opp tunities to notify the NRC that it was operating outside its design basis. To first notification to the NRC should have been made when the matter, as idern ified by Westinghouse was being reviewed by NARASIMHAN.
In her resolution of the POPS issue, she mistakenly concluded that the matter was resolved by taking credit for a code case that had not been approved at Salem. This was memorialized in the BERRICK memorandum (Exhibit 11). Salem realized it had a problem, investigated it, and arrived at the wrcng conclusion. BERRICK indicated that had he known in December 1993 that NRC code case approval was required, it would have been a reportable event.
The next o>portunity for a notification, and the basis for the willful call, was when tie licensee learned that the code case could not be used at Salem.
2 The licensee dissuaded several engineers, including LASHKARI, in January 1994, and OURA, that an IR should not have been prepared. The IR would have raised the issue to the control room for a possible immediate resolution.
WIEDEMANN's reasoning, that the IR which LASHKARI )repared was not necessary, was that Salem would not have a safety problem witi POPS because it could rely on the RH3 valve or take credit for the code case. He knew that neither one had been approved by the NRC, and PSE&G would have to go forward to request approval. Since neither was approved, the licensee was o>erating outside of its design basis. WIEDEMANN's reasoning focused on what le saw as the safety requirements, not the regulatory requirements.
Even THOMSON, the Licensing Manager, indicated that in the January / February 2
time frame he became aware that Salem could not accept the code case without NRC approval. SMITH admits knowing in March 1994 that Salem could not use the code case, even though O'GARA knew earlier than that. Between January and about April 20, 1994 THOMSON and others were aware that the code case had not been approved but was relied upon. THOMSON chose not to make an NRC report, while looking for another solution to the problem. However, during this entire period, 10 CFR 50.72 required the licensee to make a one hour NRC notification.
It should be kept in mind that Salem relied on the unapproved code case for a year and operated outside its design basis.
In his interviews, THOMSON stressed the fact that the plant was safe: and like WIEDEMANN, his focus seemed to lose perspective with what he knew were the regulatory requirements.
O'GARA discussed with several people his belief that an IR was necessary to address this issue. He believed that this information should be presented to the shift, so that the shift supervisor could make a decision on whether or not to notify the NRC. THOMSON ran the April 1994 meeting, and by the end of it, it was concluded that an IR would not be prepared and sent to shift, thus, taking the decision away from shift management.
Failure to accept / process LASHKARI's or 0*GARA's draft irs enabled Salem to forestall making an operability determination based on the on the POPS system.
Months were spent trying to find a way to avoid making this issue known to the
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DWEST GAT II Case No. 1 9 013 33
The efforts by WIEDEMANN and THOMSON were to prevent the IR from going to shift. The Shift Supervisor would have had to make an operability determination and, during that shift, determine whether the plant was operating within its design basis, This action would have forced the issue into resolution. By not submitting the IR, Salem was able to postpone having to make an immediate decision, and attempt to find some way to prove that Salem was operating within its design basis and NRC action was not required.
This was not successful.
The reporting agent believes that this was a willful decision by Salem management to avoid reporting this to the NRC. THOMSON stated that there was never a question in his mind that there was a safety issue, " knowing that the NRC had allowed other people to take credit for the code case which gives you 10 percent more margin.
The re)orting agent believes that it was THOMSON's belief that safety was not a pro)lem, which drove him, although unsuccessfully, to look for a way not to have to report this to the M.
His focus was not on his regulatory responsibilities.
From the beginning, the licensee recognized that the code case provided sufficient margin, if approved by the NRC, and the plant was safe. Salem management chose to investigate other avenues to determine why the plant may not be outside its design basis, in spite of the fact that there.had been a detrimental reliance on an unapproved code case in 1993. The facts show that actions were directed at avoiding an NRC notification. The longer it took to actually call the ERC, the longer the licensee would have to indicate that it operated outside its design basis.
In spite of all that is known, the licensee still reported in their LER that the date of the event was November 17, 1994, and not a date in 1993.
The reporting agent believes that this entire issue took years to resolve.
For several months, management had engineers do additional mathematical computations, in the hopes that one might conclude that they were not in fact outside their desigr. basis. During this period, Salem continued to operate outside its design basis.
The preponderance of the evidence before management showed that Salem was outside its design basis, The evidence also showed that management, after having relied to its detriment on an unaparoved code case, still chose to seek other remedies which they knew would not )e found within the one hour reporting period. Having lost several opportunities to call the NRC, the reporting agent determined that managt wnt made a willful decision not to call the NRC and operate outside its design basir..
Conclusion Based on the evidence developed during this investigation, it is concluded that PSE&G willfully operated outside its design basis and failed to provide a timely notification to the NRC, pursuant to 10 CFR 50.72, that it was operating in an unanalyzed condition.
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-i LIST OF EXHIGITS Exhibit No.
DfuLc.tuf. t.m 1
Investigation Status Record, dated February 21, 1995.
2 Interview of LASHKARI, dated December 15, 1994.
3 Letter from WIGGINS to ELIASON dated March 30, 1995 w/NRC Inspection Report, dated March 24, 1995 (50 272/94 32 and 50 311/94 32).
4 Transcribed Interview of MORRISON, dated March 15, 1995.
5 Transcribed Interview of BERRICK, dated March 14, 1995.
6 Transcribed Interview of SMITH, dated March 14, 1995.
7 Transcribed Interview of WIEDEMANN, dated February 15, 1995.
8 Transcribed Interview of WIEDEMANN, dated March 15, 1995.
9 Transcribed Interview of THOMSON, dated March 14, 1995, 10 Transcribed Interview of O'GARA, dated March 14,199!..
11 Letter from WETTERHAHN to Logan, dated August 24, 1995, with attachment.; from NARASIMHAN and O'GARA.
12 Interview Report of NARASIMHAN, dated August 16, 1995.
13 Draft Incident Report by LASHKARI, dated January 31, 1994.
14 Memorandum from LASHKARI to the Technical Department Manager, dated January 30, 1994.
15 Memorandum from LASHKARI to MORRISON, dated April 22, 1994.
16 Independent Technical Investigation by Synergy Consulting Services Corp., dated April 12, 1995.
17 Memorandum from HUCKABEE to BERRICK, dated September 29, 1993.
18 Transcribed Interview of DANAK, dated March 14, 1995.
19_
Draft Memorandum from O'GARA to SMITH, dated May 16, 1995, 20 Draft IR pre)ared by O'GARA with three different drafts of the
" Summary of Event" section of the IR.
lidR PUllLIChlSCk.
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e 21 Discrepancy Evaluation Farm #94 0060, dated April 14, 1994, and prepared by H. DANAK.
22 Operating Experience Feedback Meeting, dated April 7, 1993: ATS assignment priority "4" (POPS) with related documents attached.
23 Transcribed Interview of NARASIMHAN, dated August 22, 1995.
24 Transcribed Interview of SMITH, dated August 22, 1995.
25 Transcribed Interview of O'GARA, dated August 22, 1995.
26 Transcribed Interview of THOMSON, dated July 13, 1995.
27 Transcribed Interview of BERRICK, dated July 13, 1995.
28 Transcribed Interview of NARASIMHAN, dated July 13, 1995.
29 Transcribed Interview of CHANDRA, dated July 13, 1995.
30 Licensee Event Report No. 94 017 00, with Transmittal Letter from HAGAN to the NRC, dated December 14, 1994.
31 Draft memorandum from BERRICK to SCHNARR, dated October 29, 1993.
32 Early Draft of O'GARA's Incident Report.
33 Memorandum from BERRICK to WIEDEMANN, dated May 26, 1994.
34 Memorandum from SMITH to RINALLI, dated September 28, 1994.
35 Incident Report #94 419, dated November 17, 1994.
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i Case No. 1 95 013 Exhibit'I
= -.
4 k
Q TR/byJ T4 kJ ltE INVESTIGATION STATUS RECORD
[
Case No.:. 1-95-013 Facility: SALEM UNIT 1 3
Allegation No.:
RI-94-A-0159 Case Agent:
LOGAN i
Docket No.:
50-272 Date Opened:
02/21/95 Source of Allegation: ALLEGER (A)
Notified by:
B. McDERMOTT (DRS)
Priority: HIGH (Coordinated with T.
Martin,RA) v Category:
AE Case Code: RP (Power Reactor)
Subject / Allegation: POTENTIAL INTENTIONAL OPERATION OF PLANT OUTSIDE ITS DESIGN BASIS AND FAILURE TO EVALUATE AN UNRESOLVED SAFETY i
QUESTION Renat ks:
Monthly Status Renort:
91/21/95: On August 8,1994, the NRC received an allegation identifying 23 separate concerns with Sales operations, engineering, and management. The alleger was initially interviewed by Region I staff. On September 30, 1994, 01 initiated an investigation into the alleger's claims of harassment, intimidation, and discrimination (1-94-043). The alleger was interviewed by 01 on December 15, 1994.
One of the safety issues reportedly raised by the alleger concerned the plant's pressurizer overpressure protection system (POPS).
It appears from the preliminary resultt of-an inspection by the regional staff that on or about April 19, 1994, PSE&G changed the 4
POPS design basis transient for mass addition without evaluating the change pursuant to 10 CFR-50.59.
PSE&G did not make a 10 CFR 50.72 report to the NRC on this issue (i.e., unanalyzed condition) until November 1994.
On February 16, 1995, an enforcement panel convened to discuss the alleger's' concerns, specifically whether the licensee " failed to
. notify the NRC for 11 months & continued to operate (the) plant outside (its) design basis for POPS'(50.72)."--In the 01 interview, the alleger stated that he brought to the attention of his o
supervisor-informaticn regarding the POPS issue in the form of an Incident Report (IR) dated January 31, 1994. He stated that when he-4
- wanted to bring the-IR to the control room, he was told, 'No.
You
- do not file this. It's not. your job..." Because-this matter is
.. viewed as a high priority by the Regional Administrator, and involves a distinct issue of. alleged intentional wronodoing by the licensee, 01 is opening a separate investigation to a6 dress this matter, and will coordinate with continuing regional inspection efforts. Status:
FWP ECD: 09/95.
4131T kk4 T FOR LIC'D OkREdl PP L
j j
CASEND.
1-95-013
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i EXHI3IT 5 Case No. 1 95 013
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Exhibit 5