IR 05000272/1994027
| ML18101A449 | |
| Person / Time | |
|---|---|
| Site: | Salem, Hope Creek  |
| Issue date: | 12/21/1994 |
| From: | Calvert C, Ruland W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18101A448 | List: |
| References | |
| 50-272-94-27, 50-311-94-27, 50-354-94-25, NUDOCS 9501090057 | |
| Download: ML18101A449 (7) | |
Text
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DOCKET/REPORT NOS:
LICENSEE:
FACILITY:
INSPECTION AT:
DATES:
INSPECTORS:
INSPECTOR:
APPROVED BY:
u~.s. NUCLEAR REGULATORY COMMISSION
REGION I
50-272/94-27 50-311/94-27 50-354/94-25 Public Service Electric and Gas Company*
Salem 1 & 2 and Hope Creek Generating Stations Hancocks Bridge, New Jersey 08038 Hancocks Bridge, New Jersey September 26 - October 7, 1994, and telephone exit on December 12, 1994 J. A. Cal.~e~ctor Engineer, DRS D. T. Moy~. tor Engineer, DRS
~~-L->~~*~.~ee~:-~-~rz:~1q~
n ctor Safety
~Rular&t.%Y Electrical Section Division of Reactor Safety
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Areas Inspected: This was an announced inspection of the Salem and Hope Creek station engineering program by regional personnel to determine the effectiveness of the licensee's engineering staff in providing technical support to the safe operation of the Salem and Hope Creek Nuclear Generating station No significant concerns were identifie ~DR ADOCK 05000272 PDR
- DETAILS PURPOSE The objective of this safety inspection was to determine the effectiveness of the Public Service Electric & Gas Company's (PSEG) engineering organizations in providing design changes, modifications, and technical support to the safe operation of the Salem and Hope Creek Nuclear Generating Station This objective was accomplished by performing a review of selected plant major and minor modification packages.. The inspectors reviewed the packages for the proper preparation, installation, and closeout and verified the following:
Design changes and modifications were controlled by approved plant procedures;
Modification packages were reviewed and approved by onsite and offsite review organizations;
Installation procedures were adequately reviewed;
Changes to the design as described in the FSAR and plant operating procedures were adequately controlle.0 DESIGN CHANGE PROCESS The inspectors reviewed the licensee's design modification process, which is described in procedure "Control of Design and Configuration Changes, Tests and Experiments," NC.NA-AP.ZZ-0008(Q), Revision 6, approved on March 30, 199 This procedure established a uniform method or road map for design and configuration changes to determine the appropriate change package workboo This procedure directs the user to the proper workbook from a series of six (6) books that discuss the following types of changes: standard design changes; engineering change authorization; equivalent and obsolete piece part replacement; document change only; as-built documentation. This procedure was previously reviewed and found to be adequate by NRC inspectors in Combined Inspection Report 50-272/94-07, 50-311/94-07, and 50-354/94-0.0 DESIGN CHANGES AND MODIFICATIONS The licensee implements design changes and modifications using the "Standard Design Changes Workbook One",.Procedure NC.DE-W~.ZZ-OOOl(Q), Revision 4, which was last reviewed by the NRC staff in Inspection Report No. 94-0 Major modifications are prepared, documented and implemented using the above procedur The inspectors reviewed the modification packages to verify the effectiveness of the licensee's design change process described belo.1 DCR IEC-3316, Throttling Service Water Flow to Number 11 Component Cooling Heat Exchangers (CCHX)
This design change was intended to limit the maximum potential flow for service water pump in order to mitigate net positive suction head (NPSH)
concerns. These concerns were originally identified in an engineering *
- discrepancy (DEF DES-90-01434).. The actual determination that available NPSH may be insufficient under certain design conditions was confirmed upon performing various hydraulic evaluations using the service water system hydraulic flow model (Calculation No. S-C-SW-MDC-131710).
The licensee used this flow model which indicated that available service water pump NPSH with a single active component failure of CCHX control valves failing to the full open position, may not be maintained within the pump manufacturer's requirements. Therefore, this DCP introduced an additional flow restriction into the service water flow path for the No. 11 CCH The additional flow restriction was primarily accomplished by the throttling. The number 11 CCHX service water inlet isolation valve 11SW121 from the existing full open locked position to a partially open locked positio *
The inspectors reviewed the modification package documents and interviewed design engineering, Station QA and operations staff to determine the quality of the engineering process for this modification. During the course of the audit, the inspectors had the following observations: The purpose of the design verification procedure (NC.DE-AP.ZZ-OOlO(Q))
is to establish a controlled method and standard guideline for performing an independent design verification. As required by this procedure, the design verification should include a summary statement sufficiently clear to identify the method, extent and depth of verifications. Furthermore, any engineering judgement used during the verification process should also be included in the summary statement.
The inspectors noted that the summary verification statement for this DCP was not detailed in compliance with the above procedur The inspectors brought the concern to the licensee and they acknowledged the inspectors' observation * The "Certification for Design Verification" section of NC.DE-AP.ZZ-OOlO(Q) procedure stated that the design verification signature indicates that the questions on the generic checklist have been reviewed for applicability. However, when the inspectors reviewed the checklist; they found the following statements were incorrectly checked "not applicable," but would apply to this modification:
(a}
Are assumptions necessary to perform the design activity adequately described and reasonable; and (b}
Is the output reasonable compared to inputs?
The inspectors brought the above two observations to the licensee and PSE&G acknowledged the observation.2 Small Des;gn Change Packages The inspectors reviewed the small change or equivalent replacement design process by document examination and personal interviews. This workbook (DE-WB.ZZ-0003(Q) is* used to replace present plant components when the component is obsolete, unavailable, or unreliable. This procedure applies to
"one-for-one" component replacement only.. The replacement component must be
evaluated to be equal to the original component or to the design application requirement If this criteria is clearly documented as meeting the equivalent replacement 10 CFR 50.59 review and safety evaluation, then no unreviewed safety question is involved or revision to the safety analysis report (SAR) is require The inspectors selected the following replacement modification packages for review:
- Change Package IE0-2353, package number 001, which replaced 12MS131 Mark No. FA109 with Equivalent Valve Mark No. F-9. This valve is the root valve for PT 526 and serves as part of the containment isolation pressure boundar.
Change Package IE0-2309, package number 001, which replaced 11SJ17 with an equivalent valve of a non-seal welded canopy style, W.0-93070917, Change Package 2E0-2334, packag~ number 001, which replaced Valve Internal (stem/plug/cage) 2PR1 and 2PR2 17-4 pH material with material more suitable to the actual applicatio.
Change Package IE0-2362, package number 001, which was concerned with an air operator that did not cycle solenoi The replacement was obsolete and the new model required is K831654 The inspectors reviewed the equivalent replacement evaluation form for each of the above package Some of the system parameters of the present design/specification and replacement component were different within each modificatio For example, component overall size and material, flow characteristic and the flow coefficient Cv were differen Each modification detailed the justification for those changes and a safety analysis was not required to be completed by Workbook ZZ-000 The inspectors did not identify any technical concerns with the small charge process as applied to the above change package.0 STATION QUALITY ASSURANCE (SQA) AUDITS.
The inspectors reviewed station quality assurance procedure, "QA Interface for E&PE Projects," ND.QA-AP.ZZ-0024(Q); Revision 1. This nuclear quality assurance procedure provides direction for review and reports to configuration change or design change package However, SQA is not responsible for complete review of the design package, only spot check the modification package and change The inspectors reviewed several SQA station QA audit and surveillance reports relating to plant operation.1 Station Quality Assurance Audit on PORV Internal Part Replacement - DCP-2EC-3190 During an unusual event Salem Unit 1 on April 7, 1994, the pressurizer power-operated relief valves (PORV) had cycled over 200 times in*35 minute The licensee decided to remove and inspect these valves to determine the material condition of the valve after thfs even The licensee determined that the
- stem exhibited cracking around the roll pin area where it penetrates the pin collars. The purpose of this audit was to review the development and implementation of DCP-2EC-3190, which permitted installation of parts different than that orig.inally intended for the Unit 2 PORVs (see Salem Inspection Reports 94-80, 94-11, and 94-13 for additional information). The licensee identified following observations relating to this event:
(a)
(b)
(c)
(d).
Shared responsibility between E&PB and Salem maintenance departme~t for installation of DCP was not well-defined nor understood by participants; When receiving material from storeroom for field installation, there was no existing process or procedure in place to verify that the "MMIS Issue Ticket" material is correct or accounted for; The modification instructions in the DCP'did not provide suffici~nt detail to identify the upgrade material for the new valve internal to be installed; nor did these installations identify the proper procedure to be used to perform these activities. Because of the shared responsibility to implement this DCP and verifying its proper installation was deficient; Station maintenance personnel for DCP 2EC-3190 had not attended nor have they been scheduled for DCP training; and (e)
At the time of 100% DCP submittal to SQA, code job package (CJP) for valve internal parts replacement for 2PR1 and 2PR2 were not identified in Section 13~0 of DCP 2EC-319 The inspectors reviewed the audit report and concluded that the licensee's findings were appropriat.0 INTEGRATION OF REGULATORY DOCUMENTS IN THE OPERATIONAL PROCESS Regulatory documents (generic letters) are processed in accordance with PSE&G Nuclear Procedure, "Nuclear Licensing and Reporting," NC:NA-AP~ZZ-0035(Q),
Revision The licensee's commitment control fovolves a three-step process:
- (1) the identification of items requiring formal responses; (2) the preparation and approval of those responses; and (3) the tracking a~d closeout of commitment associated with the response The identification of items requiring formal responses is achieved through various administration program For generic letters, the licensing representative reviews the documents and notifies the applicable department managers/designer of the required action Also, the representative enters the commitment into a database called "Action Tracking Task System," which is part of the MMI The inspectors requested status information with regards to NRC Generic Letter 92-04, "Resolution of the Issues Related to Reactor Vessel Water Level Instrumentation in BWR pursuant to 10 CFR 50.54(F}." The information provided by the database indicated that the licensee had responded to GL 92-04 and a subsequent request for additional informatio Because of the safety
- signi.ficance of the issue, additional information was required by the licensee, and this generic letter was re-issued as NRC Bulletin 93-0 The inspectors verified the licensee's response (NLR-N93126) to this bulletin and the acceptance by NRC staff on February 23, 199 The inspectors also reviewed the data for Generic Letter 94-02, "Long-Term Solutions and Upgrade of Interim Operating Reconunendation for Thermal-
Hydraulic Instabilities in Boiling Water Reactors." The inspector verified Generic Letter 94-02 and responded to the staff on September 9, 1994.. The.
inspector did not identify any discrepancy between the licensee's response and their implementation, and no unsafe conditions were note.0 USE OF PROBABILISTIC RISK ASSESSMENT (PRA)
The 1 icense'e has started a program for the integration of PRA into the engineering organization. The inspections verified that there is a PR department engineering specialist familiar with the concepts of PR The licensee had submitted both Salem and Hope Creek's individual plant examination (IPE) to the NRC staff for approval, and the external event analyses (IPEEE) will be submitted for Salem and Hope Creek by May 1995 and February 1996, respectivel The licensee stated that present training consists of the following:
(a) PRA overview course, which is held quarterly; (b) PRA applications course, which was a I-day course for nuclear engineering and reliability/assessment engineering for both Salem and Hope Creek Stations; (c) IPE descriptions and findings 2-hour seminar; and (d) discussion of IPE result~ in the licensed operator requalification training. The inspectors consider~d the training as adequate at this stage of the IPE submitta.0 EXIT MEETING Following the conclusion of the inspection on October 7, 1994~ the inspectors had a telephone conference with licensee representatives denoted in Attachment 1 on December 12, 199 The inspectors summarized the scope and
- results of the inspection at that tim The licensee acknowledged the inspection finding The inspectors neither received or reviewed any proprietary material during the inspectio ATTACHMENT 1 PERSONS CONTACTED Public Service. Electric and Gas Company C. Atkinson J. Bailey R. Chranowski J. Clancy A. Culliton R. Donges
- G. Englert E. Karpe *
- c. Lambert A. Pasricha D. Patel
- J. Ranalli
- R. Ritzman
- W. Scholl D. Smith F. Thompson Instrumentation & Controls Supervisor, Hope Creek Nuclear Engineering Science Manager Electrical Technical Engineer, Salem Technical Manager, Hope Creek Standard and Assurance Supervisor Senior Staff Engineer, Standards Nuclear Engineering Standards Manager Principal Engineer, QA Programs Nuclear Engineering Service Manager Supervisor, Chemistry Process Civil Mechanical Engineer Manager, Nuclear Mechanical Engineering Lead Engineer, Nuclear Licensing Technical Staff Engineer, Hope Creek Principal Engineer, Nuclear Licensing Licensing and Regulations Manager State of New Jersey R. Pinney Nuclear Engineer U.S. Nuclear Regulatory Commission D. Moy Reactor Engineer, Region I
- Denotes those present at the telephone exit meeting on December 12, 1994.