IR 05000333/1989020
| ML20011D183 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 11/28/1989 |
| From: | Eapen P, Prividy L, Jimi Yerokun NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20011D182 | List: |
| References | |
| 50-333-89-20, GL-89-04, GL-89-4, IEB-88-004, IEB-88-4, IEIN-86-001, IEIN-86-1, NUDOCS 8912220002 | |
| Download: ML20011D183 (11) | |
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REGION I
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Report No.
50-333/89-20 Docket No.
50-333 s-f License No..-DPR-59 Licensee: New York Power Authority T23 Main Street White Plains, New York 10601 Facility time: James A. FitzPatrick Nuclear Power Plant
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Inspection At: Scriba and White Plains, New York Inspection Conducted: -August 28 September 1, September 5-9, and September 25-29, 1989
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Inspectors:
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( L. J. Prividy, Senior Reactor. Engineer date i
S cial Test Prog ams Section, EB, DRS
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'J. T. 'ferokun, Reactor Engineer, Special date Test Programs Section, EB, DRS
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Approved by:
h. K ' [cb h /gt k /9 9 Dr. P. K. Eapen,I Chief, Special Test Mate '
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Programs Section, EB, DRS Inspection Summary:
Routine Announced Inspection on August 28_- September 1, September 5 - 9 and September 25-29, 1989 (Inspection Report Number 50-333/89-20)
Areas Inspected: The focus of this inspection was design, design changes and modificaticas. Also included in the scope was_the inspection and verification,
ofLitems related to previous open and unresolved items.
Results: The inspectors determined that the modification process was being-controlled by existing administrative procedures. The reviewed modification packages, the installation procedures, work performance, and testing, were determined to be adequate. One licensee identified, non-cited violation and one unresolved item were identified during this inspection concerning design control and these items are discussed in Section 2.2.
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PDR ADOCK 05000333
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Persons Contacted
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Power Authority of the State of New York
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- J.-Brunetti, Manager Civil / Structural NED
J. Ellmers, Supervisor, Nuclear Licensing Engineering - BWR
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- J. Erkan, Project Engineer Supervisor
+* W. Fernandez, Resident Manager
- L. Guaquil, Director Project Engineering
+* T. Herrmann, System Engineering Supervisor
+* J. Hoddy, Project Engineer
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- R. Lauman, Director BWR Operations and Maintenance
M. Licitra, Supervisor __ Project Engineering
+* D. Lindsey, Planning Superintendent
+* R. Liseno, Superintendent of Power
- G. Havrikis, Director Nuclear Engineering and Design
- : E. Neal, Jr., Technical Administrator
- ~ J._Oliveto, Chief Design Supervisor
+* R. Patch, Quality Assurance Superintendent
- D. Ruddy, Plant Engineering Supervisor
+* D. Wallace, Perf. Engineer Supervisor V. Walz, Technical Services Superintendent
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B. Young, Manager NED Mech.
- S. Zulla, Vice President Nuclear Engineering-B.
Nuclear Regulatory Commission
+ 0. LaBarge, Project Manager, NRR
+* W. Schmidt, Senior Resident Inspector
+* R. Plasse, Resident Inspector The inspectors also held discussions with other members of the
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engineering staff.
+ Denotes those present at the mini-exit meeting conducted on site, September 1, 1989.
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- Denotes those present at the mini-exit meeting conducted at
the licensee's corporate office, Septe.mber 8, 1989.
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- Denotes those present at the exit meeting conducted on site,
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September 29, 1989.
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2.0 Design Changes and Modification (57700 and 37838)
The objective of this inspection at the James A. FitzPatrick Nuclear Power Plant (JAFNPP) was to ascertain that design changes and modifications were in conformance with the requirements of the Technical Specifications,10 CFR, the Safety Analysis Report, and the licensee's
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Quality Assurance program.
This objective was accomplished by performing a detailed review of selected modifications listed in Attachment A.
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plant modification packages and installation of these packages were reviewed and the following was verified:
Design changes and modifications were controlled by' approved
procedures.
- Modification packages were reviewed and approved by onsite and
offsite review organizations.
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i Installation procedures were adequately reviewed.
' :t modification test procedures and results were adequately
reviewed.
Installation of modifications conformed with modification packages.
- Temporary modifications were also reviewed during the course of this inspection and the following were verified:
A formal record was maintained of temporary modifications.
- Temporary modifications were promptly identified in the plant.
- Installation and identification of temporary modifications conformed
to the log maintained in control room.
2.1 References
10 CFR 50 (Appetdix B and 50.59)
ANSI N 45.2.11 - 1974
JAFNpP Work Activity Control Procedure (WACP) 10.1.6, Control
of Modification 4 and component Changes Plant.Modificeilon Packages (See Attachment A)
- 1 2.2 Inspection & dings l'
Review and. approval of safety evaluations and modification packages are conducted by the Pl.nt Operations Review Committee (PORC).
Plant modification packages are assembled and reviewed in accordance with WACP 10.1.6 " Control of Modifications and Component Changes",
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Installation procedures for modification (Mod) are prepared by the Site Engineering group and are part of the modification package submitted to PORC for review and approval. The procedures reviewed
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were found to be thorough with adequate detail provided to perform i
the installation of the modification.
Inspection fiadings concerning-each of the modification packages including a review of temporary modifications are given below:
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(1) Mod F1-87-016 (Replacement of Valve 13 MOV-15)
The physical work associated with this modification was completed i
during the 1988 refueling outage. However, the modification package
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had not yet been closed out and the documentation was not placed in the plant's permanent records.
In this modification package the inspector reviewed calculation set No. I which was performed to
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determine the structural adequacy of revised supports needed for conduit relocated per Engineering Change Notice (ECN) #F-1-87016-004 on October 5, 1988.
The inspector noted that this calculation had not been reviewed, verified or approved in accordance with EDP-2.,
Procedure for the Administration of Site Engineering Calculations, and EDP-3, Design Verification Procedure. The inspector discussed this finding with the cognizant Site Engineering personnel and
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determined the following:
Site Engineering becume aware of this problem in July,1989
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and a check calculation DD-87-1223-C-2, Rev. I was completed on August 31, 1989. This calculation fulfilled the design verification requirement for the original calculation set No. 1
and concluded that the conduit supports were adequately designed to meet all design criteria and codes.
Once the problem wac identified, Site Engineering took prompt
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action to identify the causes and corrected them to prevent recurrence. Specifically, all other modification packages
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being closed out were reviewed and no other calculations were determined to be unverified. Site engineering personnel were made asare of the problem by their supervisor. Also, the Vice President-Nuclear Engineering issued a memorand". on October 2, 1989, to all nuclear engineering division personnel emphasizing the need for independent verification of calculations supporting ECNs prior to declaring the system operable.
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Based on this information, the inspector concluded that this licensee identified violation was an isolated occurrence for which timely corrective action was being taken to prevent recurrence.
This violation (NCV 50-333/89-20-01) is not being cited because the criteria specified in Section V. G of 10 CFR Part 2 Appendix C (the Enforcement Policy) were satisfied.
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- c In evaluating contributing causes for the problem concerning the verification of calculation Set No.1, the licensee determined that a possible cause was due to a weakness in another procedure EDP-10, Engineering Change Notices.
Currently, EDP-10 does not contain specific time requirements to process ECNs in accordance with EDP-3, Design Verification Procedure prior to declaring the system operable. The-licensee intends to correct this weakness when issuing Procedure No. MCM-9, Engineering Change Notices, which is a part of the new Modification Control Manual scheduled to be completed by the upcoming refueling outage in March, 1990.
Pending the satisfactory issuance of this procedure, this item reniains unresolved (50-333/89-20-02).
The inspector also reviewed the valve procurement specifications and the installation procedure and no unacceptable items were identified.
(2) Mod F1-89-129 (Nitrogen Supply to the RBCLCW Containment Isolations Valves)
The purpose of this modification was to provide a more reliable source of pneumatic supply to the valve actuators of the Reactor Building Closed Loop Cooling Water System (RBCLCWS) containment isolation valves. This is to be accomplished by replacing the existing non-safety related instrument air supply with a safety-related (QA Category 1, Seismic Class 1) pneumatic supply line from the nitrogen supply system.
The nitrogen supply system is designed to provide adequate flow of nitrogen and to minimize leakage from pneumatic components. The inspector reviewed the modifi':ation package and determined that it was prepared and reviewed according to WACP 10.1.6.
This modification was not installed at the time of this inspection.
However, an installation guideline, in lieu of an installation procedure was attached to the applicable work request (WR) to perform
the modification. This method of installation would result in a lower level review as WRs are not reviewed by PORC, while installation procedures are reviewed by PORC.
The licensee acknow-ledged the concern and agreed to review the matter prior to actual installation of the modification.
Subsequent to the exit interview, the inspector determined in a telcon with the Site Engineering supervisor on October 10, 1989 that (1) an installation procedure was prepared, reviewed and approved by PORC (meeting 89-061) and (2) modification F1-89-129 was satisfactorily completed using this procedure. This satisfactorily resolved the inspector's concern, as the required installation procedure was prepared and approved prior to the installation.
(3). Mod F1-87-127 (Valve Replacement CAD-67, 68, 69 and 70)
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This modification is part of a valve replacement program designed to reduce the containment leakage.
These replacement valves are similar in size to the existing valves but have design improvements to minimize seat leakage.
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The modification package reviewed was in compliance with WACP 10.1.6.
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However, the package contained a valve procurement specification (No.
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JAF-87-01-1) for large bore valves (2h inches and larger) while the valves in this modification are lh inch valves. This was brought to the licensee's attention. The licensee was already aware of this
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discrepancy and the Nuclear Engineering and Design group issued a L
memorandu;a to the modification package stating that the purchase of
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valves CAD 67, 68, 69 and 70 per specification JAF 87-01-1.was
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technically acceptabic.
The inspector performed a walkdown of on going work for this modification. Adequate control of material was observed for both the old and new valves. The replacement valves were inspected for proper flow direction, orientation, damage and adequate supports.
No unacceptable conditions were observed.
The inspector concluded that this modification was being implemented p
satisfactorily.
(4) Mod F1-87-163 (Jet Pump Instrumentation Nozzle Modt fication)
The purpose of this modification is to provide adequate flow through identified stagnant areas of the jet pump instrument nozzles to reduce the dissolved oxygen content in the reactor coolant system water and hence prevent intergranular stress corrosion cracking (IGSCC).
The modification package was prepared in accordance with WACP 10.1.6.
This modification is to add piping connections between the jet pump instrument nozzle drain lines and the recirculation pump suction lines thereby creating flow through the nozzles.
The piping and supports for the modification are to be seismically designed, fabricated, inspected, and installed in accordance with ANSI B31.1 (1967 and 1969 winter Addenda) power piping code.
The engineering aspects of this modification were reviewed and determined to be acceptable.
(5) Temporary Modification (Crescent Area Unit Coolers)
Modification package F1-89-049 has been prepared for the area unit coolers whicn service the crescent areas where the plant Emergency Core Cooling System (ECCS) equipment is located. This modification was prepared in response to a significant lack of cooling experienced by the current area coolers as detailed in NRC Inspection report 50-333/88-23. The temporary modification
implements one of the interim actions discussed by the licensee in a subsequent enforcement conference.
The coolers are air-to-water.
heat er. hangers using service water as their primary cooling medium. The modification is intended to improve the heat transfer of the coolers by accommodating several changes and it is scheduled p
for implementation in the 1990 refueling outage.
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'l Part of the modification will be to change the existing service water supply line fine mesh strainers with a coarser mesh. Several of these strainers had been changed under the control of the plant temporary modification program.
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The inspector reviewed the temporary modifications described in the modification package and in the temporary modification log in the
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control room. A walkdown of the crescent area unit coolers was then
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performed to verify correct installation and location of the
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temporarily installed coarse mesh strainers. These temporary modifications were found properly tagged and in conformance to the control room log description.
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No unacceptable items were identified.
3.0 Other Engineering and Technical Support' Items (37700)
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In addition to the detailed review of certain modifications discussed in
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Section 2.0, the inspector reviewed (1) the engineering organization and work backlog, (2) response to Bulletin 88-04. (3) check valve and IST
program issues, and (4) QA/QC interfaces with engineering activities.
3.1 Engineering Organization and Work Backlog During the initial ite visit the inspector discussed with the resident manager the progress of the evaluation of the site staff in the engineering support area as recommended by the Systematic Assessment of Licensee Performance (SALP) board.
While licensee management has routinely reviewed the progress of its engineering personnel in light of the recent
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reorganization, a formal evaluation hes not been performed.
The inspector discussed with the first line supervisors the
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development of the relatively new Site and System Engineering groups. Acceptance of these groups by existing site working groups is evident.
Licensee supervision expects to be able to better determine the effectiveness of these groups during and shortly after the 1990 refueling outage.
Discussions between the inspector and the Technical Services Superintendent indicated that the licensee does not have a
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comprehensive plan of defining its overall engineering work backlog and a systematic method of prioritizing the work according to safety
significance.
Only recently (April,1989) the maintenance work order control system was inodified to accommodate requests for engineering services which may not be modification oriented. While this change i
may not be the optimum to handle such requests, it is better than the l
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I pervious method. The Technical Services Superintendent recognized this issue of work backlog as one of his top priorities.
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Discussions between the inspector and the Vice President-Nuclear Engineering indicated that a careful and deliberate engineering approach was being taken to ensure an accurate and well controlled design process for the design i
bases reconstitutien effort. This includes the preparation of g
new design and drafting standards, a new design control manual
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and a new modification control manual.
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3.2 Bulletin 88-04 Review The licensee's response to NRC Bulletin 88-04 (Potential Safety-Related Pump Loss) was reviewed. The inspector reviewed Engineering Calculations (JAF Task No. 88-5068) for the NRC Bulletin
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88-04 and held discussions with the licensee's Project Engineering
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group representatives concerning these calculations which referenced I
plant isometric drawings. A walkdown of system piping was then performed to confirm the accuracy of the assumptions made in these calculations. Based on thase reviews, discussions and walkdown, the inspector independently verified that the bases for the licensee's response to Bulletin 88-04 were satisfactory.
3.3 Check Valve / Inservice Test (IST) Program i
The inspector reviewed the status of the licensee's IST program, and in particular, the licensee's positions being formulated in response to Generic Letter 89-04. The licensee's current actions for the check valve program were also reviewed.
While much work remains to be done in the IST Program, definite improvement has been noted during 1989 in its overall direction and administration.
The licensee's contractor, Gilbert Commonwealth, recently completed a study of the JAFNPP check valves in response to addressing the industry check valve concerns identified in INPO SOER 86-03 and NRC Information Notice 86-01. The licensee has included this contractor developed, baseline information in a check valve preventive maintenance program. Critical check valves have been selected to be disassembled and visually l
inspected during the 1990 refueling outage. The licensee's I
approach to addressing industry check valve concerns has been
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quite thorough.
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- 3.4 QA/QC Involvement The inspector discussed the technical services backlog of overdue responses to QA findings with the QA and Technical Services Superintendent. This backlog necessitated involvement of the Vice Presidents of QA and Nuclear Operations in accordance with the escalation requirements of QA Procedure 15.2, Identification, Control and Resolution of Adverse Quality Conditions. Although it was l".
evident that the Technical Services Superintendent was aware of the overdue responses, the inspector was concerned that appropriate priorities were not being established by engineering to promptly resolve these quality issues. However, by the end of the inspection a significant number of overdue responses assigned to the Technical Services Department had been received by QA.
Specifically, by memorandum, dated September 26, 1989, the QA Superintendent noted'
that 30 previously overdue items had responses in the past 45 days such that only eight of the original 41 overdue items were still outstanding. The inspector concluded that the licensee's corrective action controls were working and they are capable of resolving issues in a timely manner.
4.0 Licensee Action on Previously Identified Inspection Findings 4.1 (Closed) Unresolved Item 88-04-02 on IST Program This item had been opened to address several concerns regarding IST program issues. The inspector reviewed the licensee % response to each of these concerns as follows:
Valves 23 HOV-1 and 2, the high pressure coolcra injection
turbine steam stop and throttle valves were not included in the IST Program.
However, a later revision to the IST Program included these valves in the program and are now tested quarterly.
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The IST program assigned a maximum limiting stroke time of
5 seconds to various rapid acting valves which was different from the NRC staff position of 2 seconds. However, a later revision to the IST program assigned a maximum limiting stroke time of 2 seconds to the applicable rapid acting valves.
Certain keep full system check valves had not been included in
the IST program. After evaluating the applicable check valves in the residual heal removal (RHR) and core spray (CSP)
systems, the licensee took the following actions:
(1) Check valves 10 RHR-262 and 277 were added to the IST program while 10 RHR-261 and 276 were not, and S
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(2) Check valves CSP-62A&B, 63A&B and 77A&B had not been included in the IST program and the licensee's subsequent
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evaluation established the bases for this position. While this evaluation appeared to be appropriate, the inspector informed the licensee that sufficient information should
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be included in the licensee', response to Generic Letter 89-04 to permit full evaluation of the licensee position
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Based on this review, the inspector had no further concerns of these IST program issues. This item is now closed.
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4.2 Battery Room Ventilation
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The inspector performed a follow-up review and inspection of the licensee's response to the battery room ventilation concerns described in prior inspection report 50-333/88-15. The concerns addressed the adequacy of the battery room ventilation system to
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meet the design basis requirements for maintaining temperature and eliminating possible hydrogen buildup in the rooms. A review of the licensee's engineering calculations (JAF 88-024-001) and evaluations
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performed was conducted. The licensee's calculations demonstrated that even with the worst case condition (loss of normal ventilation)
it would take over seven days to generate a 2% concentration of l
hydrogen which is_one half of the lower explosive limit of
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The licensee's area monitoring efforts and results of
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such efforts were Jiscussed. Also, the inspector performed an independent walkdown of the battery rooms to verify the assumption made in the calculations and evaluations.
From this review the inspector concluded that the licensee has satisfactorily resolved the battery room ventilation concerns discussed in prior inspection report 50-333/88-15.
5.0 Exit Meeting i
At the conclusion of the inspectior i September 29, 1989, an exit meeting was conducted with the licenwe's representatives (listed in Section 1) to discuss the results anc conclusions of this inspection.
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At no time during this inspection was written material provided to the licensee by the inspectors. Based on the NRC Region I review of this report and discussions held with licensee representatives during this inspection, it was-determined that this report does not contain information subject to 10 CFR 2.790 restrictions.
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5 ATTACHMENT A
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-,g Modification Packages Reviewed
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Modification Number-Tit 1e
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F1-89-129 Nitrogen Supply to the RBCLCW
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Containment Isolation-Valves F1-87.-127 -
Valve Replacement CAD-67, 68, 69 and 70 F1-87-163:
Jet Pump Instrumentation Nozzle
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(Flow) Modification
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F1-89-049 Crescent Area Unit' Coolers
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Improvements-
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.F1-87-016 Valve Replacement 13 MOV-15
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