IR 05000311/1985008
| ML18092A610 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/22/1985 |
| From: | Cerne A, Cook W, Hillman B, Kenny T, Norrholm L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18092A608 | List: |
| References | |
| 50-311-85-08, NUDOCS 8505070218 | |
| Download: ML18092A610 (16) | |
Text
Report N Docket N U.S. NUCLEAR REGULATORY COMMISSION
REGION I
50-311/85-08 50-311 License N DPR-75 Priority Category __
Licensee:
Public Service Electric and Gas Company P. 0. Box 236 Hancocks Bridge, New Jersey 08038 Facility Name:
Salem Generating Station Inspection At:
Hancocks Bridge, New Jersey Inspection Conducted:
March 21-22, 25-29, 1985 In specters:
iJkt~l-UeU~
~T. Ke
, Senior Resident Inspector, Team
Leader Inspection Summary:
Inspection on March 21-22, 25-29, 1985 (Report No. 50-311/85-08)
4-19-85 date 4* /q-~5 date 4_zz.gs *
date A special team inspection composed of Region I Senior Resident Inspector from Indian Point Unit 2, Senior Resident Inspector from Seabrook, Resident Inspec-tor from Ginna and a Reactor Engineer from Region I conducted a review of se-lected Design Change Requests (OCR) to ascertain that they were implemented in accordance with NRC regulations and site approved procedure The inspection involved 116 hours0.00134 days <br />0.0322 hours <br />1.917989e-4 weeks <br />4.4138e-5 months <br /> on site by the team.
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8505070218 ~~86§311
~DR ADOCK PDR
Results:
A violation was identified in the program with regard to construction work being performed by the in-house maintenance organization. Also identified was missed communications in the transfer of information from engineering to the construction forces in the field, and an isolated instance of field changes not being documente These results are described in detail within the body of the report.
DETAILS Introduction and Inspection Objectives A special inspection team composed of resident inspectors and a reactor engineer with expertise in construction (mechanical and electrical),
start-up, and operations performed reviews of selected modifications that had been installed at Salem Generation Station Unit 2 during the forced six-month outage, to document that they were installed to NRC regulations and industry standard The team primarily focused on the areas of pro-curement, welding, welder qualification, cable installation, quality assurance coverage, Q.C. control, operations upgrade to the new or modif-ied systems, and overall training on the newly modified or installed sys-tems. The team did not assess the Operations Department in the operations of the unit because the remaining unit at Salem had been in operation throughout the duration of the outage. Operators on the units at Sa 1 em
- _rotate frequently from one unit to the othe.
Identified Strong Points The inspEo:ctors collectively agreed that the below-listed areas of the licensee's design change program were above average for an operating sta-tion:
The bi-annual review of operating, abnormal and emergency procedure Good communications between the operations, engineering and training departments to assure station operators and support personnel receive essential station modification information in a timely manne QA audits program is a comprehensive and well-structured program pro-viding a valuable management feedback mechanis An agressive training program is conducted for plant operators in the area of plant modification A computerized system to monitor taggirig and lineup of plant systems is in us The licensee is rewriting the welding program to obtain an 11 R 11 stam The completeness of contractor weld packages were found to be above averag.
Identified Weak Points The inspectors agreed that the below-listed weak points were areas that could lead to violations or installation in an incorrect manne However, in all* the cases examined by the inspectors there were no violations or incorrect installation *
The use of weld symbols between the engineering department and the field construction crews allowed for misinterpreted result Changes to welding procedures in the field with regard to non-essen-tial variables are not always controlle.
Review of Design Change/Modification Program The inspector reviewed the implementation of administrative procedures and controls for station design changes and modifications as prescribed by the licensee 1 s current programs. It should be noted that the station adminis-trative procedure, AP-8, was recently revised and that engineering pro-cedure EDD-1 is presently being revise The licensee maintains good communication between the engineering and operations staff in the development and tracking of Design Change Requests (DCR 1 s). Engineering personnel are assigned to specific DCR 1 s in accord-ance with their system specialty or engineering disciplin Additional engineering staff members are specifically assigned to provide the neces-sary coordination and liaison with construction contractors and opera-tions staf Engineering staff OCR workloads appear to be evenly dis-tribute Th2 inspector interviewed the Station Quality Assurance Engineer (SQAE) to discuss the Quality Assurance (QA) organization arid its interface with DCR 1 Due to the acceptance of the station 1 s major contractor 1 s (CATA-LYTIC) QA program by the licensee, station QA conducts only spot check survei 11 ances of routine OCR modification and Contra 11 ed Work Package (CWP) activities. The SQAE stated that additional station QA surveillance is conducted when the OCR involves more sensitive safety-related systems or component Station QA does review OCR administrative development and final acceptance documentation of all DCR 1 The inspector reviewed OCR surveillance checklists and Technical Specification surveillance schedules and found no discrepancie The identification of a lapse in the OCR station/contractor QA coverage in the maintenance area is discussed in paragraph 6 of this repor.
Audits Review The following Quality Assurance audits in the areas of engineering, design change requests, and training were reviewed by the inspector:
Audit No. S-81-2, Training Program (NRB)
Audit No. S-81-25, Quality Assurance Audit of Design Division Audit No. S-83-2, Training Program Audit No. S-84-2, Training Program (NRB)
Audit No. S-84-4, Nuclear Engineering Department Audit No. S-84-30, Design Change Request
The inspector determined that the reviewed audits and 1985-1986 Audit Schedule, Revision 2, January 28, 1985 are consistent with Technical Specification requirements and FSAR commitment Audit checklists were determined to be satisfactory and of adequate scop Findings were we 11 documented and appear to receive appropriate management di stri buti o A sampling of Corrective Action Reports and Quality Action Reports and their respective responses were reviewe Responses are appropriate and timel The QA audits program is a comprehensive and well structured program pro-viding a valuable management feedback mechanis No violations were identifie.
Design Change/Modification Training The inspector reviewed procedures and conducted interviews with the licen-see1 s operations and training staffs to assess the adequacy of the licen-see1s training program regarding station modifications and design change The inspector determined that timely notification of design changes is accomplished through the Operations Department via the operation's daily newsletter, The Rising Su In addition, Operations Department Informa-tion Directives (IDs) have been established to summarize station design changes, equipment modifications, procedural changes and Technical Spec-ification change Although not required reading, operators are i:eld responsible for station changes and pertinent information promulgated in The Rising Suno Operators acknowledge, in writing, having read and understood the ID 1 In addition to receiving input from the above documentation, the training staff receives direct input from the Engineering Department on design changes and modifications via the monthly OCR summary report This in-formation, in total, is used to upgrade the licensed operator requalifica-tion program, new license operator training programs and required simu-lator modifications. Training supervisors are responsible for the incor-poration of these changes into the lesson plan The simulator training staff is currently implementing a computer-based Configuration Management System to assist the Simulator Change Control Board in tracking and pri-oritizing necessary simulator modification Inspector discussions with the training staff regarding proposed changes to the OCR engineering control process governed by EDD-1 indicate that Training Department input and earlier awareness of station modifications will be incorporated into the OCR development and implementation phase of the progra There appears to be good communication between the Operations, Engineer-ing, and Training Departments to assure station operators and support personne 1 receive essent i a 1 station modification information in a ti me ly manner.
No violations were identifie *
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6 Design Changes and Modifications The inspectors reviewed selected portions of the following Design Change Request (OCR) packages, to verify that the specifications of the appropri-ate design criteria for the requested modifications and for adherence to the requirements of Station Administrative Procedure (AP) No. 8, which controls the design change, test and experiment progra EC-1305 (Changes to Charging System)
2EC-1615 (Pressurizer Safety Valve Discharge Piping Modifications)
2EC-1916 through 1919 (RCS RTD Bypass Loop Isolation Valve Replace-ment)
2EC-1991 (Restoration of the Emergency Makeup Connection From the RWST to the Spent Fuel Pit)
2EC-2003 (Replacement of the 2C Station Battery)
2EC-2081 (Auxiliary Feedwater Pump Replacement)
OCR ~EC-1638 - Reactor Protection (shunt trip)
OCR 2EC-1677A - RCS (installation of RTD's)
DC:R 2EC-1501 -
DG Control System (black out bypass)
OCR 2EC-1559 - Safe shutdown NI installation OCR 2EC-1967 - Upgraded RVLIS System OCR 2EC-1406 - Replacement of 28 V Undervoltage Relays The following design changes were selected for additional followup, to include verification of the design criteria against the appropriate code requirements, work implementation, appropriate QA/QC inspection coverage, and specific field installation wor A field inspection of the area around the Unit 2 Refueling Water Storage Tank (RWST) revealed that work -
on OCR 2EC-1991 had not yet been initiate Interviews with a licensee outage coordinator revealed that this modification could be accomplished with the plant in operation, and therefore, was not a required modifica-tion prior to Unit 2 restart.
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7 EC-1305 The inspector examined twelve (12) valves which had been modified to move them closer to the CVCS header discharge pipin Reattachment of these valves to the nozzle bosses on the header required weld preparation of the boss sockolet and the valve socket body for vee-groove full penetration weldin The inspector examined the as-welded conditions and checked the records for documentation of the required liquid penetrant testing and informational radiographic examination For three (3) of the examined valves, the inspector noted that the existence of a reducer socket fitting precluded achieving the vee-groove weld specified in the design criteri Instead, a modified single bevel groove weld had been installe After questioning by the inspector, a Field Questionnaire (FQ 032936) was generated to document acceptability of these condition The inspector verified the qua 1 i ty of these we 1 ds and confirmed that appropriate NOE had been performe The inspector then had discussions with the licensee with respect to specified design requirements being deviated from without prior authorization. The licensee's engineering personnel stated that the programmatic controls, along with proper initiation of field questionnaires to account for different field conditions, are intended to prevent such deviation The inspector deter*;:iined that, since no other examples of a variance were identified, this is considered to be an isolated concer However, he suggested that additionai licensee attention to adequate design control of encoun-tered field conditions is warrante The inspector also examined the applicable Welding Procedure Specif-ication (WPS) (Catalytic Procedure M13A-7) and the supporting Proced-ure Qua 1 i fi cation Records to verify conformance to the criteria of the ASME Boil er and Pressure Vesse 1 Code,Section I Whi 1 e no vi o-1 at ions or deviations were identified, the inspector noted that three nonessential variables, listed in Section IX as applicable to the Gas Tungsten Arc Welding (GTAW) process, were specified on the WPS, but were not followed in the subject valve weldin Failure to adhere to these nonessential variables (QW 402.5, 404.3, and 408.5) does not compromise the quality of the welds or qualification of the WPS, but*
does de vi ate from the guidance for WPS contra 1 in Section I The licensee indicated that a new PSE&G welding/brazing manual is cur-rently in review and scheduled for implementation in conjunction with the PSE&G application for 11 R
and "NR 11 Certificate The inspector noted that the proposed welding program addresses nonessential vari-ables in greater detail and considers licensee initiative in this area to be a strength. However, with respect to the deviation from the variables identified in the current program, inspector followup is warranted to verify the adequacy of existing controls, as well as the future program changes (85-08-01).
- In addition to the valve welds discussed above, the inspector exam-ined one pipe support (2A-CVCG-449) installed under this OC Con-figuration, expansion bolting, member sizes and welds were checke While no violations were identified, the inspector did note the in-stalled weld lengths deviated from the design details delineated on drawing D-4279-10, as modified by DCR-2EC-1305, M27/l. The cause of this deviation was traced to a communication problem between the engineer and construction personnel because of misinterpretation of welding symbol Standard AISC/AWS weld symbols are utilized at Salem and should have precluded the identified proble Thus, this communication problem appears to be an example of possible program weaknes The licensee evaluated the subject support and determined that it was structurally sound, despite the missing weld The licensee also conducted a sample inspection of other pipe supports to determine if the subject symbol interpretation, regarding intermittent fillet welds, represented a generic concer No other concerns were iden-tifie The inspector examined additional intermittent fillet weld-ing in connection with another modification and found it acceptabl The inspector concluded that the subject support welding concern is an isolated case. With greater li~ensee attention both to the stand-ard welding symbols in use at Salem and to better communication be-tween engineering and construction personnel, furthe*
problems, potentially affecting the quality of installed hardware, can be eliminate Overall, with respect to OCR 2EC-1305, program strengths wer :' i den-t ifi ed in the areas of the completeness of the contractor welding records and the 1 icensee welding manual and program changes in con-junction with their 11 R 11 and 11 NR 11 certificate applications. A program weakness, as exemplified above in the change of the valve vee-groove configuration, the deviation of field welding conditions from listed Section IX nonessential variables, and the support welding discrep-ancies, appears to exist in the area of communications between en-gineering and construction personnel, as it affects field modifica-tion EC-2003 The inspector reviewed the OCR Package to include any Operational Design Change Notices, and the following Allied C&D Power Systems documents:
C&D Stationary Battery Installation and Operating* Instructions (PSBP-301960)
Drawing M-8683 (PSBP-304333)
Battery Discharge Characteristics (PSBP-304331)
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The inspector examined the new 2C station battery installation cable terminations; intercell, intertier, and interrack connections; bat-tery rack bolting and mounting; and seismic foundation detail Doc-umentation for the battery 2-hour discharge and 8-hour service tests were reviewed and QA coverage was verifie Selected technical de-tails (eg: intercell connector torques and material requirements)
from the C&D installation manual were spot-checked against the record of proper installation. With regard to the battery installation it-self, no violations were identifie However, with regard to the battery rack installation process, the licensee could produce no objective evidence that the seismic design details had been followed or that QA/QC coverage of the installation process had been provide The battery rack installation included special processes, such as Hilti drop-in anchorages and high-strength bolting for the rack structural stee This work was contracted by the PSE&G Maintenance Department to a site contracto The contrac-tor did not provide a record of in-process work and apparehtly no QC coverage of this work was provide The contractor did not have its own QA program and the PSE&G QA Department handled the rack installa-tion as a routine maintenance activity, for which QC coverage was not necessarily require However, GAP-5. 3, 1 Inspection Program 1,
of \\he Nuclear Operations Quality Assurance Manual, requires certain minimum Inspection Hold Points (IHP).
One of the IHP 1 s, as delineated in Attachment 1 to the QAP requires verification of bolt torque on foundations and seismic restraint Since no waiver of this IHP was docurr~nted, the lack of QC coverage of the seismic battery rack installation, represents a violation of the PSE&G QA program and also of 10 CFR 50, Appendix B, Criterion II. (85-08-02)
Since this example appeared to be related to the fact that the Work Order (84-12-20-063-2) for this work referred only to the OCR and not to any existing Maintenance Work Procedure with established IHP 1 s, a concern was raised regarding a potentially generic defic-iency in the QA coverage of maintenance work, where new construction, and not typical maintenance, is encountere QA signatures on both the OCR package and the Work Order apparently represent recognition of the work implementation and the records review, and not necessar-ily of the need for QA/QC coverage, where require This appears to be a program weakness since its impact could extend beyond the bat-tery rack example cited in the above violatio With respect to the battery racks specifically, the licensee conduc-ted a re-inspection of the racks, checking for high-strength bolt torque and proper Hilti anchorag No improper installation condi-tions were identified. - Thus, while the licensee has confirmed the
adequacy of the existing installation, the programmatic controls which allowed the lack of in-process records and QC coverage to go undetected remains a concer With the exception of the battery rack installation, the other con-trols, design, and work in regard to the new battery installation in accordance with OCR 2EC-2003 appeared to be adequate and well docu-mente EC-2081 The inspector reviewed parts of the Design Change Package for the in-process work of replacing the existing Byron-Jackson auxiliary feedwater pump #22 with an equivalent Ingersoll-Rand pum Specific design details on the pump foundation modification were inspected and field work examined which included verification of proper member sizes, weld sizes and lengths, configuration and material require-ment Involvement of the Authorized Nuclear Inspection Agency for this work was note The specification of design criteria for piping rework, to include h1drostatic testing, was verified. Weld rod material con-trol for some pipe fitting installation was observed in-process and found acceptable.
With regard to the sampled documentation and field work done in ac-cordance with this OCR, no violations were identifie EC-1638,1559,1501,1677A The inspector verified by package review and system walkdowns that the subject design changes were installed using the standards set forth in IEEE Electrical Standards, the FSAR, and approved station procedure The inspector a 1 so verified that the post maintenance testing as de 1 i neated in the design documents was performed in ac-cordance with approved procedures by qualified personne With respect to these DCR's, no violations were identifie.
Operational Verification of Design Changes The inspector selected the following design changes to verify that the control room annunciator windows, control buttons and physical changes required by the design changes were in place:
2EC-1559 - Nuclear Instrumentation System - installation of two new detectors to meet Appendix R 2EC-1991 - Safety Injection System - installation of 3-inch T
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2EC-1638 -
Reactor Protection -
addition of Reactor Trip Breaker Shunt Trip 2EC-1789 -
Valve Motor Upgrade for Various Valves -
environmental qualification 2EC-1967 - RVLIS - seismic upgrade 2EC-1835, 1677A - Reactor Coolant System - seismic and environmental upgrade 2EC-1724, 1305 - Chemical and Volume Control -
new globe valves and deletion of vent and drain valve 2EC-1501 -
Diesel Generator Control System - provide remote diesel start 2SC-1014 -
Steam Generator Feed Pump - power supply change to pre-clude voltage spikes 2SC-1412A -
Annunciator System -
delete redundant condenser low vacuum alarm 2~C-1108A - Control Room Console - replace push button tags The inspector also reviewed:
Cont re: l room drawings to verify that t'.1e contro 11 ed drawings were annota~ed with the current system desig Operating procedures to verify that the necessary changes had been incorporate Valve line up checkoff lists to verify that the affected valving had been incorporated into the lis In a 11 the above cases, the proper changes had been made in accordance with station approved procedure The inspector reviewed the valve lineups and the verification valve line-ups of the Safety Injection System, Chemical and Volume Control System and Reactor Coolant System to confirm that the lineups had been updated by the operations staf The inspector verified that the valve lineup was per-formed properl One finding, with regard to valve lineups, was identified that can be considered an isolated case, but is nevertheless, an issue that should be addressed by the licensee. A design change, which added a valve to the
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RWST within the Safety Injection System, was scheduled to be performed during the outage; however, the change was not implemented and the check-list was updated to show the new valv The valve lineup checklist per-formed prior to startup did not have the valve on it, but the next revis-ion of the valve checklist di No other documents within the control of the operations staff had been change Through discussions with licensed and non-licensed operators, the inspec-tor verified that all those interviewed had received training and were aware of the plant modifications that had been installed during the forced outag No violations were identifie.
Management Interview An entrance interview was conducted on March 25, 1985 to apprise the plant staff of the purpose and scope of the inspectio At times during the inspection various members of the plant staff were contacted covering initial findings. (See Attachment 1 for persons contacted).
An exit in-terview was conducted on March 29, 1985 to discuss the team 1 s findings (Entrance and Exit Attendees may be found on Attachments 3 and 4).
No information provided to the inspectors was identified as proprietary tit the exit meeting.
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Attachment 1 Persons Contacted During Inspection 85-08 J. Barger, Quality Assurance Engineer S. Cornman, Engineer, Operations Department J. Cortez, Senior Staff Engineer C. Frew, Welding Engineer V. Getsinger, Senior Maintenance Supervisor J. Hannracker, Electrical Engineer (Stone & Webster)
R. Heaton, Engineer, Instrumentation & Controls Dep R. Hettmannsperger, Engineer, Maintenance Department S. Hooks, Engineer, Maintenance Department C. Hassler, Lead Engineer, Maintenance Department S. Hassler, Electrician, Maintenance Department R. Jorgenson, Quality Assurance Auditor G. Kapp, Principal Engineer R. Kazunas, Engineer, Operational Test Group G. Lipsey, Outage Coordinator (MAC)
A. Parmelee, Engineer, Nuclear Construction Support C. Perez, Field Welding Engineer (Catalytic Inc)
D. Perkins, Station Quality Assurance Engineer G. Ruane, NCS Engineer W. Schultz, Programs & Audits Engineer, Q. W. Straubmuller, Chief Designer, Mechanical T. Taylor, Manager, Nuclear Engineering Control J
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Attachment 2 Documents Reviewed During Inspection 85-08 Administrative Procedures:
No. 3, "Document Control Program" No. 4, "Station Operations Review Committee" No. 5, "Operating Practices Program" No. 8, "Design Change, Test and Equipment Program" No. 14, "Station Training Program" No. 18, "Vendor Manual Upgrade Program" Nuclear Quality Assurance Manual, GM-9 Quality Assurance Procedure 2. 2, "Design Change Request/Package Review Quality Assurance Procedure 7-1, "Corrective Action" SALEM FSAR Engineering Department Directive #1 (EDD-1), "Operational Design Change Control, Salem Nuclear Generat*::1g Station" Operations Department Administrative Directives Training Procedures:
No. 305, 11 NRC Licensed Operator Requal ification Program" No. 306, "Plant Design Change Review Program" Technical Specifications CFR IEEE Electrical Standards
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Attachment 3 Entrance Meeting Held March 25, 1985 - Attendance For Inspection 85-08 Name A. c. Cerne w. A. Cook B. M. Hillman T. J. Kenny R. J. Summers 0. J. Jagt E. A. Li den G. E. Lipsey 0. A. Perkins B. W. _Smith, J. M. Zupk J J Title Sr. Resident Inspector-Seabrook Resident Inspector-Ginna Reactor Engineer Senior Resident Inspector Resident Inspector-Salem Nuclear Engineer Manager-Nuclear Licensing Regulation Outage Manager-Salem Site Quality Assurance Engineer Technical Staff-Associate Engineer General Manager-Salem Operations
Attachment 4 Exit Meeting Held March 29, 1985 - Attendance For Inspection 85-08 Name A. C. Cerne W. A. Cook B. M. Hillman T. J. Kenny L. Norrholm R. J. Summers M. 0. Bandeira D. W. Dodson L. Fink J. E. Gallagher J. N. Leech G. Lipsey L. K. Miller W. O'Brien D. A. Perkins B. W. Smith, J P. White J. M. Zupko, J Title Senior Resident Inspector-Seabrook Resident Inspector-Ginna Reactor Engineer-Region I Senior Resident Inspector Chief, RPS 28, Region I Resident Inspector-Salem Assistant Manager,Nuclear Systems En Engineer, Nuclear Licensing & Re Site Representative (Atlantic Electric)
Maintenance Manager,Salem Operations Principal Engineer, Nuclear Licensing Outage Services Assistant Engineer, Salem Operations Engineer, Salem Operations Site Quality Assurance Engineer Associate Engineer, Technical Staff Outage Services General Manager,Salem Operations