ML20202B653
| ML20202B653 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/06/1997 |
| From: | Miller H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Eliason L Public Service Enterprise Group |
| Shared Package | |
| ML20202B028 | List: |
| References | |
| FOIA-97-325 EA-97-204, NUDOCS 9712030138 | |
| Download: ML20202B653 (4) | |
Text
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FA 97 204 g- _q Mr. Leon R. Eliason d' '.' ,
Chief Nuclear Officer & President Nuclear Business Unit Public Service Electric and Gas Company Post Office Box 236 Hancocks Bridge, New Jersey 08038
SUBJECT:
01 REPORT NO. 1 95-013R (SALEM)
Dear Mr. Eliason:
This letter refers to the investigation conducted by the NRC Office of Investigations (01) to review whether Public Service Electric and Gas Company (PSE&G) intentionally operated outside its design basis at Salem and f ailed to make a timely notification to the NRC of an unanalyzed condition regarding the Pressurizer Overpressure Protection System (POPS). Based on its investigation, 01 concluded that PSE&G willfully operated outside the design basis and ,
failed to provide timely notice to the NRC, pursuant to 10 CFR 50.72, that it was operating in an unanalyzed condition. Additionally,01 noted that NRC inspection findings indicated that PSE&G made changes to the facility contrary to the requirements of 10 CFR 50.59. A copy of the synopsis of the Ol report is enclosed.
On March 15,1993, you were informed by Westinghouse of nonconservatisms in the setpoint methodology for the POPS for low temperature / overpressure transients. However, you took nine months to conclude that the corrected peak transient pressure would exceed the pressure / temperature (Pff) limits as described in each of the units' technical specifications.
Af ter completing the analysis in December 1993, you dispositioned the nonconservatisms in the POPS setpoint methodology by administratively limiting Reactor Coolant System (RGS) operation to two reactor coolant pumps whenever RCS was less than 200'F, and increasing each unit's P/T limit by 10E The latter corrective action was inadequate, because contrary to the requirements of 10 CFR 50.55a,it relied on the methodology of an ASME Code Case (N 514) which had not been reviewed and approved for generic application by the NRC.
Subsequently,in January 1994, after recognizing the unacceptability of using the unauthorized code case to disposition the POPS setpoint methodology, you elected to implement corrective action by taking credit for the relief capability provided by RHR system suction relief valve RH3 to augment the POPS relief capability. However, your FSAR, which describes the POPS to include two Power Operated Relief Valves (PORVs), did not describe the RH3 valve.
Therefore, this entrective action was also inadequate because, contrary to the requirements of 10 CFR 50.55, an evaluation was not performed to determine the acceptability of this change to the facility as described in the FSAR. In addition, you failed to identify that on receipt of a safety injection signal, a previously operating positive displacement charging pump's dischargo, combined with dischargo from the high head safety injection pump (which starts on receipt of a safety injectie signall, could have injected wate mass into the RCS at a rate that could have prevented POPS from performing its function.
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KEENAN97-325 PDR j7g
4 Public Service E!ectric and 2 Gas Company Two Severity Levelill violations of Criterion XVI of 10 CFR Part 50, Appendix B, related to the POPS issue, each assessed a $100,000 penalty, were included as part of a $600,000 total civil penalty package that was proposed on October 16,1995 (EAs95-062,95-C65 & 95-117). The two POPS related violations, items I.C.1 and I.C.2 of that Notice, involved the f ailure to identify and correct the condition adverse to cuality, as already described herein, related to nonconservatisms in the setpoint methodology of the POPS for low temperature overpressure transient conditions, which resulted in operating outside the design basis. You responded on November 15,1995, to tne two corrective action violations and admitted the violations and prov:ded corrective action. The additional violations of 10 CFR 50.72,10 CFR 50.59 and 10 CFR 50.55a noted herein were not cited at that time.
Notwithstanding these violations, I have decided, af ter consultation with the Director, Office of E1forcement to exercise enforcement discretion, pursuant to section Vll.B.6 of the enforcement policy, and not take any further action since (1) the 01 findings are the result of or closely related to the violations for which civil penalties have already been issued as part of the significant civil penalty issue to you in October 1995, (2) these 01 findings are matters which occurred in 1993 and 1994 prior to the extended shutdown which began in 1995, (3) there have been significant changes in the managemr.'it team and personnel at Salem, and significant positive changes in the approach to identificathn and correction of problems, (4) the Of conclusion does not involve a conclusion of deliberate wrongdoing.
Nonetheless, you should reemphasae to your staff the importance of operating the facility in accordance with the design basis, as well as promptly reporting conditions that are contrary to that design basis.
No response to this letter is required. In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice," a copy of this letter,its enclosure, and your response will be placed in the NRC Public Document Room (POR).
Sincerely,
/' f h H ert J. Miller l Regional Administrator Docket Nos. 50-272 50-311 License Nos. DPR-70; DPR-75
Enclosure:
Synopsis of 01 Report I
l l
l
d Putnic Service Electric anu 3 Gas Company cc w/ encl:
L. Storz, Senior Vice President - Nuclear Operations E. Simpson, Senior Vice President - Nuclear Engineering E. Salowitz, Director - Nuclear Business Support A. Kirby, Ill, External Operations - Nuclear, Delmarva Power & Light Co.
D. Garchow, General Manager - Salem Operations !
- J. Benjamin, Director - Quality Assurance & Nuclear Safety Review D. Powell, Manager, Licensing and Regulation
-R. Kankus, Joint Owner Affairs A. Tapert, Program Administrator J. Keenan, Esquire M. Wetterhahn, Esquire J. Isabella, Manager, Joint Generation Atlantic Electric Consumer Advocate, Office of Consumer Advocate W. Conklin, Public Safety Consultant, Lower Alloways Creek Township Public Service Commission of Maryland State of New Jersey State of Delaware e
. ENCLOSURE SYNOPSIS 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
?rotection system'(P0PS). NRC Region I inspection findings indicate that
>SE&G changed the POPS design basis transient for mass addition without evaluating the change pursuant to the requirements of 10 CFR 50.59.
Based on the evidence developed during this investigation. it is concluded that PSE&G willfully oxrated outside its desian basis and failed to provide a timely notice, to the ARC, xirsuant to 10 CFR $0.72 that it was operating in an unantlyzed condition. T1e results of this investigation were presented to Region I staff and were considered, along with the staff's insxction findings, in the issuance of a notice of violation against PSE&G on October 16, 1995, and subsequent escalated enforcement action.
. NOT FOR PlBLIC' DISCLOSURE WITHOLIT APPROVAL OF =
FIEth 0FFICE DIRECTOR, OFFICE OF INVESTIGATIONS, REGION I o
Case No. 1 05 013 1
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EXHIBIT 3 l
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Case No. 1 95 013. Exhibit 3
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+ b ef UNITED STATES E' ,w ; NUCLEAR REGULATORY COMMISSION 5 a REGION I o
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[ 475 ALLENDALE ROAD KING oF PRUSSIA PENNSYLVANIA 19406 1415 March 30, 1995 Mr. Leon R. Eliason Chief Nuclear Officer & President Nuclear Business Unit Public Service Electric and Gas Company P. O. Box 236 Hancocks Bridge, New Jersey 08038
SUBJECT:
SALEM ENGINEERING FOLLOWUP INSPECTION 94-32
Dear Mr. Eliason:
This letter refers to the safety inspection conducted by Mr. B. McDermott of this office on December 5-19, 1994, and March 14-15, 1995, at the Salen Generating Stations and on February 13-24, 1995, at the NRC Region I office.
The inspection was focused on issues important to public health and safety, specifically, your actions taken in response to industry and NRC information regarding nonconservatisms in the setpoint of the Pressurizer Overpressure Protection System (POPS) for brittle fracture protection of the reactor vessel. The preliminary results were discussed with Mr. J. Sumners, and others of your staff, on December 19, 1994. A second exit meeting was conducted by telephone with Mr. Summers on' March 23, 1995.
Four apparent violations were identified during our review of your actions to resolve the POPS setpoint nonconservatisms, specifically: (1) your reliance on an exemption from the requirements of 10 CFR 50.60 without NRC approval; (2) failing to report a condition outside the plant's design basis (10 CFR 50.72 and 50.73); (3) revising the plr'it's design basis without the safety evaluation required by 10 CFR 50.59; and (4) failing to take appropriate currective actions for a significant condition adverse to quality (10 CFR 50, Appendix B, Criterion XVI). The details of these apparent violations are discussed in the attached inspection report.
We acknowledge your position that the engineering significance of the problem was considered to be low by engineering personnel and that the plant was adequately protected. However, the fact that the issue was not entered into an appropriate system for resolution of engineering discrepancies for over a year indicates a significant weakness in your ability to understand and effect prompt resolution of problems. Even after the. issue was entered into an appropriate system, seven months passed before you detemined that the plant was outside it's design basis. Further, failure to perform a safety evaluation for changes to the POPS design bases, as required by 10 CFR 50.59, reflects a continuing weakness in your process for developing and implementing effective corrective action, 1-95-013 ~ EXHlBIT 0 CASE fD.
((/
PAGE / OF/S PAGE(S) e YLWEN _ OlY
Mr. Leon R. Eliason 2 Merch 30, 1995 The apparent violations discussed above are being considered for appropriate enforcement action and, accordingly, no Notice of Violation is presently being issued for these findings. You will be informed by separate correspondence of any subsequent action required in response to the apparent violations.
Your cooperation in this matter was appreciated, and we would be pleased to discuss the conclusions described in this report with you.
Sincerely, 3
James T. Wiggins, Director Division of Reactor Safety Docket Nos. 50-272; 50-311
Enclosure:
Salem Inspection Report 94-32 cc w/ encl:
J. J. Hagan, Vice President-Operations S. LaBruna, Vice President - Engineering and Plant Betterment C. Schaefer, External Operations - Nuclear, Delmarva Power & Light Co.
R. Burricelli, General Manager - Informations systems & External Affairs J. Summers, General Manager - Salem Operations J. Benjamin, Director - Quality Assurance & Safety Review F. Thomson, Manager,' Licensing and Regulation R. Kankus, Joint Owner Affairs A. lapert, Program Administrator R. Fryling, Jr., Esquire M. Wetterhahn, Esquire P. J. Curham, Manager, Joint Generation Department, Atlantic Electric Company Consumer Advocate, Office of Consumer Advocate William Conklin, Public Safety Consultant, Lower Alloways Creek Township Public Service Comission of Maryland D. Screnci, PA0 (2)
PUBLIC Nuclear Safety Infonnation Center (NSIC)
NRC Resident Inspector State of New Jersey State of Delaware EX BlT PAGE opp p4gg(g)
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U.'S. NUCLEAR REGULATORY-Com !SSION l REGION I -
DOCKET / REPORT NOS: 50-272/94-32
-i 50-311/94-32 LICENSEE: Public Service Electric & Gas Company FACILITY: Sales Generating Stations Hancocks Bridge, New Jersey ,
^
1 DATES: December 5-19, 1994 February 13-24, 1995 March 14-15, 1995 INSPECTOR: -F^- .
i
. Brian L_NcDeralgtt, Reactor Engineer Dafe f Systems Section-Division of Reactor. Safety APPROVED BY:
EugdnT M J. elly, Chief / Dat'e Systems Wetion Division of Reactor Safety 5
$UMMARY: PSE&G worked to resolve nonconservatises.in the Pressurizer Overpressure Protect ton System (POPS) setpoint calculations for approximately two years (March 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. During the inspection, engineering personnel. stated that from the time the issae was identified in March 1993 they considered its safety significar.co 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 P0PS was changed and the evaluation required by r
10 CFR.50.59 for-identification of a possible unreviewed safety question, was not performed. Several apparent violations of NRC requirements identified
-during this inspection are being considered for enforcement action. The licensee's calculations for-the revised POPS design-basis transient have been referred to the NRC 0ffice Of Nuclear Reactor Regulation for review and per. ding their evaluation, this aspect of the issue will' be unresolved.
3 i
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. 1 DETAILS ;
1.0 INSPECTION SCOPE ,
This inspection evaluated.the response of Public Service Electric and Gas (PSE&G) to information regarding the nonconservatisms identified in the-technical specification setpoints for the pressurizer overpressure protection system (POPS) at both Salem units. The nonconservatisms in the original Westinghouse setpoint methodology were communicated to PSE&G in a letter from the vendor dated March 15, 1993. NRC Information Notice 93-58, "Nonconservatism in Low-Temperature Overpressure Protection for Pressurized-Water Reactors," dated July 26, 1993, reiterated the problems identified by Westinghouse.
2.0 FIDSINGS
Background
The POPS uses two pressurizer power-operated relief valves (PORVs) to mitigate low tem prature (<312*F) overpressure transients, keeping the peak pressure below tie 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 (780 gpe) and its injection into a water solid reactor coolant system (RCS). POPS was 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.
An NRC safety evaluation report, dated February 21, 1980, associated with Amendment No.' 24 to the Unit 1 TS, approved the Salem POPS setpoint of 375 pounds per square inch gage (psig), based 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.
The P/T limits for all reactor vessels decrease with successive operating cycles due to irradiation effects on the vessel materials. Therefore, margin between the peak transient pressure and the P/T limit will change as
- subsequent revisions of P/T curves are reviewed and approved by the 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 lov 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 i effective full power years- of operation.
- Setnoint Nonconservatism On March 15,.1993, Westinghouse-issued a Nuclear Safety Advisory Letter (NSAL-g3-0058)--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 s
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2 to 'the reactor vessel midplane, were found not to have been considered in the original setpoint methodology. The static head error for Sales is1 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 increase the differehce between pressure at the reactor vessel midplane and
<that sensed by the P0PS instrumentation hy approximately 25 psi, i
i Consequently, for a four-loop plant such as Salem, the sensed pressure (with all four RCPs running) could be as much as 100 psi less than_the actual pressure at the reactor vessel midplane (the area of concern for P/T curves).
1hese errors simply can be added to the original peak transient pressure since their effect is to offset (nonconservatively) the pressure at which POPS will actuate. NRC Infomation Notice (IN) 93-58, 'Nonconservatism in Low-l Temperature Overpressure Protection for Pressurized-Water Reactors,' 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.
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 Itmiting 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), administratively limiting the maximum number of RCPs in scryice to two when RCS temperature was below 200'F (limiting the dynamic error in the most restrictive area of the P/T curve), and increasing each unit's P/T limit by 10% using an unapproved American Society of.
Mechanical Engineers (ASNE) Code Case N-514. The inspector noted that, at
- temperatures above 200*F and up to 312*F, the Appendix G P/T curves allow for much higher pressure limits.
The inspector considered that - at the point PGE&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 Appendix G limits could be potentially exceeded) - both Salem Units could be potentially operated in an unanalyzed condition (whenever below 312'F) which would be outside the plants' design bases. Therefore, the condition was reportable, and the licensee's failure to make such a report is.an apparent violation of the reporting requirements-of 10 CFR 50.72 and 73. (EEI 50-272;311/94-32-01) Further, the inspector noted that an exemption request for use of ASME Code Case N-51a had not been submitted by PSE&G until late December 1994. Use of the ASME code case would require 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 ASNE Code Case N-514 for over one year without the-required exemption 'is an-apparent violation of 10 CFR 50.60.
(EEI.50-272;311/94-32-02)
Less than one month after the issued had been dispositioned in Memorandum MEC-p- 93-917, the licensee recognized.that the-ASME code case could not be used L 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 EXHIBIT
! PAGEdOFM PAGE(S) l l_-_______.________
3 anslyzed POPS relief casacity. The spring-operated relief valve (RH3 has the same setpoint as POPS, sut has a greater effective flow area and will) actuate- '
faster than a PORV once its setpoint is reached. A subsequent analysis by PSE&G confirmed the licensee's initial ju('gement that, with RH3 available, the i peak pressure would remain below the Appeadix G limit. The issue of crediting RH3 as part of PDPS (without either a 50.59 safety evaluation or prior NRC-approval by changing the POPS TS) was under consideration from ald-January .
through mid-April 1994. On April 19, 1994, a Discrepancy Evaluation Form !
(DEF 94-0060) was written to document the fact that relief valve RH3 was not credited the original POPS analysis for Sales 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 fallere to initiate corrective actions for this significant condition adverse to quality, is an apparent violation of 10 CFR 50, Appendix B, Criterion XVI, " Corrective Action." (EE! 50-272;311/94-32-03)
Corrective Action Process initiated l The April 1994 DEF addressed the imediate safety concern by assuring the availability of relief valve RH3 and considering the safety margins discussed l
in the ASME code case. At this time the licer.see also initiated a procedure r 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
- l. 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 Sales. The inspector noted j that, even after the issue was entered into the PSE&G DEF process, the i condition outside the design basis was still not reported to the NRC.
The inspector. independently assessed the availability and capability of relief valve RH3 for supplementing the POPS. Valve RH3 is available for RCS pressure relief when the RHR 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 p reviewed by the NRC under Westinghouse Topical Report WCAP-ll736, " 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 EXHIBIT PAGE h 0F/M PAGE(S) 1
l 4
4 requirements and would substantially reduce the peak transient pressure based on its design. ' However, crediting valve RH3 as part of POPS would,- in the inspector's estimation. require a change.to the Sales Technical Specifications.
The issue was once again closed (by memorandum, dated 5/26/94), 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 necessary (relatively a small effect); therefore, the original analysis was concluded by PSE&G to be
~
still valid. The inssector noted that the procedural requirement to have a i pressurizer " bubble' Izefore starting any RCPs was in place, and had been previously reviewed in the 1980 NRC safety evaluation report for POPS.
- Although the DEF was closed by PSE&G via this memorandum, further 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 detemined that the affects of a running RCP on the POPS analysis should also be considered. However, no femal 50.59 safety evaluation had as yet been performed.
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 cnalysis would not provide acceptable results after the effects of running RCPs were considered, engineering personnel established what they considered a more
' realistic" transient as the design basis event for POPS.
L The original transient was simply the start of a safety injection pump (the intermediate head pump delivers 780 gpe) and its injection into a water-solid l RCS. The licensee's revised transient is mechanistic and relies upon procedural controls for limiting possible injection sources. The revised transient begins with the reactor in Mode 5 (<200*F), the positive displacement (PD) charging pump in service and one RCP running, whereafter an L inadvertent safety injection (SI) signal would cause the centrifugal-charging
! 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 I the GOTHIC computer code resulted in a predicted peak pressure of 438 psig, L below the P/T limits of- each unit. Therefore, by limiting the magnitude of L the' mass addition, the licensee was.able to reduce the predicted peak transient pressure and justify the existing TS setpoint for POPS.
L By the end of September 1994,. the licensee believed they had 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 by revising the limiting transient upon which the setpoint is based. The relevant Technical Specifications for the Salem Unit 1 POPS are 3.4.9.3 and Bases 3/4.4.9.3; For
- Sales Unit 2 they are TS 3.4.10.3, and Bases 3/4.4.10.3. Further, the new n
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transient, which changed the design basis for POPS, also invalidated the NRC's -
SER upon which Amendment No. 24 - and both units' POPS TS - was based. The description of the limiting transient and the design bases for POPS in Salem '
- FSAR Section 7.6.3.3 were, therefore, no longer correct and current.
The inspector noted that, since March 1993, several PSE&G corrective action programs were used but none was effective in resolving the Sales 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 perfomed to detemine if the change in the POPS design-basis transient had created an "unreviewed safety i question.' 10 CFR 50.59 requires licensees to evaluate changes to the plant or its procedures (including methods and modes of operation), prior to those ,
changes being effected, to assure no unreviewed safety question exists. The licensee's failure to perfom this safety evaluation is, therefore, an
- apparent violation of 10 CFR 50.59. (EEI 50-272;311/94-32-04)
"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 POPS mass addition transient. The configuration ,
document had incorrectly assumed that the positive displacement (PD) charging l
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). After 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.
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 nc longer mitigate the transient. PSE&G reported this to the NRC under 10 CFR 50.72 ar. an unanalyzed condition for Sales Unit 1. IR 94-419 provided justification for the continued operation of Unit I based on RHR relief valve RH3 being availabli to augment POPS. With the three valves (two PORVs and RH3) available below 312*F, sufficient relief capacity was reasoned (by the licensee) to be provided and the single failure criterion could be met, i 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.
.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.
2 DAGE JEXHiT__OFdO .
6 RCS Vent Path-TS for both Salem units require cold overpressure protection be provide:1 by
- either the redundant PORVs (the POPS system) or a reactor coolant system vent ,
of greater than or equal to 3.14 square inches (in'). Venting the RCS is an alternative to having the POPS operable and would be accomplished after depressurizing the RCS.- The TS action statement for POPS requires that, in _
the event a PORV fails and cannot be re=*ared within seven days, the reactor ;
must be depressurized and vented througn one 3.14 in' vent within the next l eight hours.
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 adequate based on: (1) the flow from an unrestricted opening of 3.14 in' would 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. 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-ll640, March 1988) of a full-open PORV would be adjusted for hydraulic resistance, and factors affecting this correlation were provided in a December 8,1992, memorandum to PSE&G from the 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 stroke. The Salem PORVs are 2-inch diameter Model D-100 " plug-in-cage" valves, with flow coefficients on the order of 50. This flow coefficient can be used, along with previously compiled EPRI test data for these type valves, to calculate a so-called equivalent area-corresponding to a smoothly convergent nonflashing (sonic flow) nozzle - that licensee themal hydraulic engineers estimated to be 1.21 in'.
The RCS vent area of 3.14 in', by itself, has no hydraulic meaning unlecs a geometry can be assumed so that resistance and loss factors can be calculated.
Nonetheless, a single PORV gagged-open is clearly enveloped in the RCS vent configuration by a simple two-inch diameter flanged opening 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. :
Code' Case'Anoroval The-inspector reviewed the licensee's documentation and interviewed personnel
- - involved with the POPS issue during the 20 months between the NSAL issuance in March 1993-and PSE&G's 50.72 notification in November 1994. The ins metor-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 ASNE Code Case N-514.. Based on the inspector's discussions (prior to February 1995) with representatives from the NRC Office of Nuclear Reactor EX IT PAGE- OFMPAGE(S)
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Regulation (NRR) ASME Code Case N-514 represented a- technically acceptable position, although a plant specific exemption would be required. ,
On December 16, 1994, a conference call between PSE&G and NRC representatives was held to discuss the licensee's more inmediate 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 injecting into the RCS. These commitments i were formally submitted in a letter to the NRC from PSE&G issued later that same day.
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 for POPS. Without the code case, PSE&G credited valve RH3 on Unit'I to meet the design basis single failure criterion for POPS. However, for Unit 2, the licensee did not credit valve RH3 because the peak pressure, based on a single PORV, was predicted to be 1.0 psi below the unit's P/T limit. By letter dated February 13, 1995, the NRC issued an exemption from the requirements of 10 CFR 50.60 for Salen Units 1 and 2. This 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 I and 2 limits became 495 and 522 psig, respectively.
3.0 CONCLUSION
S The inspector considered several aspects of PSE&G's actions to resolve the POPS issue over the past 20 months as inadequate or inappropriate:
e The POPS issue was initially dispositioned to she b,9 the requirements of 10 CFR 50.60 were met, invoking an ASME code case that had not received prior NRC approval.
- When inclusion of the setpoint nonconservatism put the Sales Units outside the POPS design basis, reports to the NRC were not made pursuant to 10 CFR 50.72 and 73.
e No safety evaluation pursuant to 10 CFR 50.59 was perforised prior to revising the POPS design-basis transient (described in the Salem FSAR) in September 1994. As of the December 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. Theilowest
. priority possible was assigned to this issue within the Operational ~
Experience Feedback program in March 1993, and the issue was not entered into:an appropriate PSE&G quality program for resolving engineering .
discrepancies for over a year.
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The adequacy of the new design basis for POPS is currently under review by the NRC's Office of Nuclear Reactor Regulation. Fending NRC assessment of the licensee's proposed limiting design-basis transient for POPS, this issue is unresolved. (URI 50-272;311/94-32-05)
The licensee's: (1) reliance on ASME Code Case N-514 without NRC approval, (2) failure to report a condition outside the Sales design basis, and (3) failure to perform an adequate. safety evaluation of the revised POPS design-basis transient are all apparent violations of NRC requirements.
Further, the process used to address the issue (memoraadum superseding memorandum) was considered to be fragmented and not appropriate for potentially safety-significant issues. The inspector further concluded that the corrective action processes that were engaged (a year late) did not appropriately resolve a condition outside the plant's design basis.
4.0 MANAGEMENT REETINGS Licensee representatives were informed of the scope and purpose of this inspection at an entrance meeting conducted on December 5, 1994. Findings were periodically discussed with the licensee throughout the course of this inspection. A telephone conference call was conducted between NRC and PSE&G representatives on December 16, 1994, to discuss the licensee's plans to resolve several aspects of the POPS issue. During this call, PSE&G committed to taking several actions and subsequently documented these commitments in a letter to the NRC issued later that day.
The inspector met with the principals listed below to summarize preliminary findings on December 19, 1994. The licensee acknowledged the preliminary findings and conclusions, with no exceptions taken. Further, the bases for the preliminary conclusions did not involve proprietary information, nor was any such information expected to be included as part of the written inspection report.
Public Service Electric & Gas Comoiny.
l l L. Catalfomo Operations Manager J. Morrison Salem Technical Department Manager M. Morroni Controls Maintenance Manager J. Ranalli Nuclear Mechanical Engineering Manager l 0. Smith Principal Engineer, Nuclear Licensing l J. Summers General Manager, Salem Operations Manager E
C. Marschall Senior Resident Inspector, Sales D. Moy Reactor Engineer, DRS J. White Chief, RPS2A, DRP i
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A final sumary was provided to PSE&G representatives by telephone on -
March 23, 1995, discussing the results of the NRC inspection and evaluation that took place after the December 1994 site visit. Another telephone conversation was held on March 29, 1995, to discuss the issue of PORY flow characteristic's (refer to page 6, "RCS Vent Path') between E. Kelly of the NRC and the following PSE&G representatives:
C. Lambert Manager, Nuclear Engineering Design Vijay Chandra Technical Consultant, Thermal Hydraulics Gita Narasimhan Mechanical Engineer ,
Mahesh Danak Mcchanical Engineer i
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