IR 05000443/1989007
| ML20246F330 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 08/16/1989 |
| From: | Eapen P, Prividy L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20246F321 | List: |
| References | |
| 50-443-89-07, 50-443-89-7, NUDOCS 8908300261 | |
| Download: ML20246F330 (8) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Report No. - 50-443/89-07 Docket No. 50-443 Li:Ense No.
NPF-56 Licersee: Public Service Comoany of New Hampshire
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1000 Elm Street Manchester, New Hampshire 03105 Facility Name: Seabrook Station, Unit No. 1 Inspection At: Seabrook, Ne d ampshire Inspection Conducted: June 19-23,,1989 Inspector:
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L. J. Prividy, Se'nier-Reactor Engineer date W.N b b P 4/$7 Approved by:
P. K. Espen, Chief,/5pecial Test Pro' grams
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Section Inspection Summary:
Routine Unannounced Inspectivn June 19-23, 1989 (Rv: port No. 50-443/89-07)
Areas Inspected: The focus of this inspection is ' design, design changes, and modifications. Test results associated with the modifications were also reviewed.
Results. The inspector determined that the design change process was being properly controlled by Administrative Procedures. The modification packages including the post modification testing were adequate. Two unresolved items were identified concerning the evaluation of steam leakege through the auxiliary feed pump turbine control valves (Section 2.2) and the evaluation for leakage through reactor coolant system pressure isolation valves (Section 3.3).
There were no violations.
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DETAILS
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- 1.0 persons Cont eted
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Public Service Company of New Hampshire
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- R. Belanger, Lead Compliance Engineer
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' *C. Beverly, Operational' Programs Engineer 4 '
- .S. Burhwald,-QA Supervisor.
- J. Cady, ISEG Supervisor
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J. C6nnally, Program Support. Test Engineer lS. Corcoran,. Technical Suppnrt Engineer
- D. Covill, Surveillan:e Supervisor
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- R. Cyr, Maintenance Manager -
- E. Desmarois, Independent Review Team
'*W. DiProfio, Assistant Station Manager
- R. Faix, Westinghouse Engineer in NHY Engineering
- T. Feigenbaum, Vice-President, Engineering and Quality
- J. Grillo, Operations Manager
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- P. Gurney, Reactor Engineering Supervisor
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~ *R. Gwinn, Operational Programs Engineer
'J. Hanley Training Manager
- G. Kann, Startup Manager
- G. Kline, Technical Support Manager
- J. Marchi, Audit and Evaluations-
- R. Martel, Staff Engineer
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- D.-McLain,. Production Service Manager
- D.-Moody,,. Station Manger.
- D. Nrkins, Operational Programs Engineer
- J. Peterson, A.O. Manager
- R. Sherwin, Planning and Outage' Manager
- E. Sovetsky.T.P. Supervisor
- W.; Temple, Licensing Coordinator J. Tipton, Technical Support Engineer
- J. Vargas, Manager of Engineering
- C Vincer.t, QC-Supervisor U_.S. Nuclear Regulatory Commission
- N. Dudley, Senior Resident Inspector
- D Haverkamp, Reactor Projects Section Chief, Region I
- J. Trapp, Reactor Engineer, Region I
- Denotes those present at exit meeting.
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i 2.0 Design Changes and Modifications _(37700 and 37828)
The primary objective of this inspection was to ascertain that design l
changes and modifications were in conformance with the requirements of the Technical Specifications (TS),10 CFR, the Safety Analysis Report, and the licensee's Quality Assurance program. This objective was accomplished by reviewing the following modifications: (1) Design Coordination Report (DCR)87-422 " Replace Miniflow Valves RH-FCV-610 and
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j 611" and (2) Minor Modification (MM00)89-542 " Modify Pilot Plug Vent
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Holes for MS-V393, V394 and V395." The inspector reviewed these modification packages and the installation of the design changes and verified the following:
The modifications were technically sound and they adequately
addressed the root cause for the change.
The modification packages were reviewed and approved by onsite and
offsite review organizations.
i Decign chcnges and modifications were controlled by approved
procedures.
Post modification test procedures and results were adequately
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Station procedure modifications were made prior to the modification
being declared operable.
Operator training was conducted prior to declaring the modification
Marked up copies of as-built drawings were distributed prior to
declaring the modification operable. Also administrative controls were established to maintain as-built drawings.
Preventive maintenance and inservice inspection and test programs
were properly updated.
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Changes to the FSA.R were properly controlled and updated.
- Installation of modifications conformed with design change package.
- The licensee has been submitting quarterly reports of 10 CFR 50.59 safety l -
evaluations concerning facility and procedure changes which were l
perforri,ed or implemented without prior NRC approval.
The inspector reviewea the last four quarterly reports which covered the period of April 1, 1988 to March 31, 1989. These summaries contained sufficient modification information as required by 10 CFR 50.59.
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'2.1.- References 10 CFR'Part 50,:(50.59, Appendix B and other Sections).
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Reg.. Guide.1.33, Rev. 2,. Quality Assurance Program
. Requirements (Opera +, ion)
ANSI;N45.2 - 199, Quality Assurance Requirements for
Nu' clear Facilities
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. ANSI N45.2.11 - 1974. Quality Assurance Requirements for.
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the Design of Nu: lear Power Plants Seabrook Station Final Safety Analysis Report e
2.2 Inspection Findings Design changes and modifications at the facility are controlled.by-Engineering Procedure 34025, " Scoping, Planning and Scheduling of.
DCRs and Facilities Modifications ano Other Engineering Activities."
This procedure provides the necessary guidance to design, budget, schedule, develop, review and approve proposed design changes and facilities modifications. The key personnel involved in this process are the cognizant' design engineers who are responsible,for the detailed engineering work. After a DCR is completed and approved by the Station Operation Review Committee,'it is implemented in accordance with Procedure No. MT 3.1, "DCR Implementation Plan." At this point the onsite technic;1 support group assumes field engineering ~ resporst-bility. The cognizant implementation engineer in the technical support group is usually the responsible system engineer who inter-faces' closely with the cognizant design engineer to ensure proper DCR implementation.
The inspector verified adequate implementation of the licensee's modification programs using the attributes listed in-Section 2.
Additional observations and concerns are discussed below.
These observations and concerns were besod on an everall review of these modifications with the cognizant design engineer and plant walkdowns with the responsible system engineer.
DCR 87-422
" Replace Miniflow Valves RH-FCV-630 and 611" This modification had been designed and approved but not yet installed. Basically the modification entails changing the miniflow valve from a gate to a globe valve in the residual heat removal (RHR) system recirculation piping and mudifying the piping arrangement. The modified design is intended to im> rove the RHR system performance from a vibration standpoint. Weld failures at instrument connections in this piping had occurred in mid-1987 and subsequently were attributed to flow induced vibration.
In addition to the weld failures, problems were encountered with damage to the disc guides and motor operator for RH-FCV-611. Although all of
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5-these problems occurred somewhat separately, they were all' related
'in that they occurred in the same RHR system recirculation piping
and a comprehensive problem resolution was not apparent until the development of DCR.B7-422.
i The inspector reviewed the calculation for the preliminary sizing of j
the new flow measurement orifice to be installed in the RHR system recirculation piping.
The calculation was its accordance with good engineering practices with clear statement of the p oblem, Issump-tions, details and conclusions. The calculation was independently reviewed per the licensee's design control procedures. The inspector independently reviewed the calculation and verified that tt.e besas of the calculations were technically sound.
HM0f)89-542
" Modify Pilot Plug Vent Holes for MS-V393, V394 and V395"
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This modification concerned the steam supply control v41ves for tlS emergency feedwater (EFW) pump turbine and it had already been implemented.
It consisted of increasing the vent hole area in the pilot plug for these valves since the valves would not open as required under opers'.,ing temperature, pressure and flow conditions.
The post modification testing was performed satisfactorily in accordance with Procedure No. STP-101, " Turbine Driven Emergency Feedwater Start Verification. Test," to demonstrate epertbility.
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results of this acceptance testing indicated that steam leakage i
through these valves while in the closed position was determined to be acceptable. The inspector discussed the basis for this determination with the licensee's program support and engineering managers who indicated that Engineering Evaluation 89-021 dated June 20, :i989 thoroughly evaluated and accepted this valve leakage.
Part of this evaluation considered the impact of steam leakage on the four-inch check valves, MS-V94 and V96 which are located downstream of MS-V393 and V394, respectively. The licensee concluded 'in the avaluation thzt valve leakage should have no detrimental effects or, system operation and thus, safe shutdown capability was not compromised. The licensee recognized that valve leakage should be repaired in accordance with established maintenance practices and that check valves MS-V94 and V96 should be inspected for possible seat wear and hinge pin anti hanger wear.
However, when the inspector conducted a plant walkdown with the system engineer, steam leakage through MS-V394 was sufficient to cause repetitive slamming of the disc (approximately 2-3 cycles per second) of check valve MS-V96.
The it.spector was concerned that this repetitive slamming had the potential to cause fatigue in the internals and eventual valve failure.
The licensee's engineering manager inspected these in plant conditions and acknowledged the inspector's concern.
It appears that Engineering Evaluation 89-021 did not adequately consider the negative impact of increasing steam leakages on system performance.
Further evaluation is rer.uired not only to determine a resolution to the steam leakage
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problem through the cont ml va!ves but a ho to ensure a comprehensive engineering solution for t re11attle steam supply syster for the EFW pump turbine. This item' is unresolved (50-443/89-07-01) pendingLthe completion of maintenance on-these valves and'a reevaluation of.the
. acceptability of EFW pump-steam isolation valve leakage esperia11y-during exter.ded power operation when steam leakages could increase.
3.0 Other Engf oeerintand Tschiral Suppcrt Issues (3'/700 and 737561 l-
-In addition to the d6 tailed review of the modifications discussed in i
k ction'2 above, tre inspector reviewed (1) engineering involvement in l:
procedure develeMet t 'and (2) engineering re9onse to RHR system check
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valve leakage.
3.I' Engineering Ordu,1zatien and Training Observations The _ design and technical wpport engineering staffs ar'e
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experienced with msny engineers h3ving been at the plant
'durmg construction.
Engineering management was genere.11y well informed and abreast
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of technical issues.
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Engineers and engineering management were responsive to NRC r.cece rns. The inspet. tor particularly noted the agineering I
manager's personal, in plant inspecti0n of the MS-V96 check l
valve condftions.
Based or, a discussion and review with the training manager. the
inspOctor concluded that adequate training is available for the engineering staff.
The total engineering work effort was adequately controlled.
- Such work includes response to DCRs, MMODs, engineering evaluas tions, requests for engineering services (RESs), and special engineering activitiec (e.g., task forces). However, the licensee recognized that the number of open RESs of 650-700 was high and the engineering organization was making a cancerted effort to reduce this work backicg.
3.2 Engineering Involvement in Procedure Development The inspector reviewed several pump and valve surveillance test
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procedures to determine if the tests adequately verified component ar.d system requirements. The inspector concluded that the surveillance test procedures selected for review were technically
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adequate.
For example, surveillance test procedure 0X1436.13
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periodically tests the backflow capability of check valves MS-V94 and V96. Also, the EFW pump discharge check valves tre periodically
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tested for backflow. However, upon further review of the feedwater system, the inspector noted that there is no periodic backflow testing i
for the six-inch, normally closed, check valves FW-V215 and V3r,7.
.These valvec are in the startup-feed pump discharge piping whir.h connects into the EFN pump. discharge header. Both valves would be req'uired to remain functional under a seismic event since the. seismic boundary is designated immediately upstream of FW-V216. These valves are tested properly in the forward flow direction. The' program support manager noted that there is ne easy way to individuaHy test'
these velves in the backflow direction. However, he net' d that'it -
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.may be possible without any hardware changes to modify existing test
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procedures and perform a backflow test monitoring for combined leakage through both valves. The program support manager agreed to evaluate the feasibility of such a test. The inspector had no further comment.
T.3 Engineering Response to RHR System Check Valve Leakage l
l On June 19, 1989 while performing tb RHR A pump surveillance test
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durir.g Mode 2 where the RHR pump was operating on reeirculttlon flow, leakage past the RHR to SI cold leg ihjection check valves (RH-VJ5, RH-V29, RH-V30 and RH-V31) was obse.rved by the licensse.
Thu leakage was apparently due to a reduced differential pressure across the check valves, while ruening the RHR pump, allowing flow into tt;u RHR system from the safety injection accumulato'es. During this time, the RHR system was pressurized above the RHR pump suction relief valve set pressure of 450 PSIG. 'This relief valve lifted and
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approximately 500 gallons of water wcs drained from the four SI accumulators.
Pressure in the. system was subsequently vented off through the safety injection test line header to the RWST via'RH-V28, l
SI-V62'and SI-V70.
In order to perform the RHR pump surveillance (0X1413.01), a procedure change was made to maintuin the SI test header vent path to the RWST open foF the duration of the test.
RES89-381 (June 22, 1989) was issued to address concerns expressed by the inspector and the technical support staff. This R$S requested the licensee's design engineering grup to evaluate the ability of the present Peactor coolant system (RCS) pressure isolation valves (PIVs) installed to effectively isolate lower pressure ECCS piping at lower differential pressures. The inspector noted that check valves RH-V15, V29, V30 and V31 had been tested on June 1, 1989 in accordance wit"n the PIV requirements of the Technical Spe~cifications end only RH-V30 had a measurable leakage of 94 ml/ minute. The maximum allow-able leakage for these valves per Technical Specificat ens Table 3.4-1 is 3.0 GPM. However, most of this leak testing is done with higher differential pressures. This item is unresolved (50-443/89-07-02)
pending satisfactory resolution of RES89-381 by the licenses concern" ing an evaluation of the ability of the RCS pIVs installed to ef-L fectively isolate lower pressure ECCS piping at lower differential pressures.
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4.0~ Licensee Action on Previously Identified Open Items (Closed')' Unresolved Item.(50-443/88-11-01):
This item.was npened to address a potential problem of reporting. modification work items as '
complet(. when they actually were incomplete. Specifically, certain work associated wit.h Engineering Change Authorization (ECA) 05/112374 was:
determined to be incomplete when it should have been completed. This item was-left unresolved p.ending the 1itensee's verif f cation that all work associated with ECA 05/112374 was complete as this was an isolated instance.
Subsequent to these developments, additional reviews and plant tialkoowns coricerning field verification were conducted by the licensee's quality group to determine' proper implementation of randomly selected ECAs. The results of these reviews and plant walkdowns are given in licensee memo
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NQG No. 89244 dated May 3, 1989 and they concluded that no generic problem existed in this area. Based on NRC inspection report 502443/88-10 which verified the licensee's action in this regard, the above item is now cloced.
5.0 Unresolved Items Unresolved items are matters about which more information is required to ascertain whether they are acceptable, 6ev'ations or violations. Two unresolved items were identified during thu inspection and they are
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discussed in Sectiont 2.2 and 3.3.
L 6.0. Man,ageme_nt Meetings Licensee management was informed of the scope and purpose of the
. ihspection at the entrance meeting contiucted on June 19, 1989. The fir. dings of the inspection are discussed'with licensee representatives during the course of the inspection. An exit meeting was conducted on June 23, 1989 at the conclusion of the inspection (see Section 1.for attendees) at which time the licensee management was informed of the inspection results.
At no tinies during this inspection was written material provided to the licensee. The licens e did not indicate.that proprietary information was involved within the scope of this inspection.
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