ML20247M052
| ML20247M052 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 05/26/1989 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | Swetland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 8906020340 | |
| Download: ML20247M052 (7) | |
Text
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May 26, 1989) m L.
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MEMORANDUM FOR:- Paul Swetland, Chief fi Projects Section 1B -
Division of Reactor' Projects Region I b
' 'FROM:
Walter R. Butler, Director.
Project Directorate'I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation'
SUBJECT:
SALP INPUT FOR HOPE CREEK, SAFETY ASSESSMENT / QUALITY VERIFICATION.
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- Enclosed 11s the SALP input for Hope Creek in the' functional area of Safety Assessment / Quality Verification.. Input was received from the NRR technical staff, and'the senior resident at Hope Creek.
For additional information,.
cont'act'Clyde Shiraki at 492-1445..
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Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear. Reactor Regulation
Enclosure:
SALP... Input-cc w/ enclosure:
E. Wenzinger, RI G. Meyer, SRI, Hope Creek
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May 26, 1989 MEMORANDUM FOR: Paul-Swetland, Chief i
Projects Section 28 Division of Reactor Projects Region I FROM:
Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
SUBJECT:
SALP INPUT FOR HOPE CREEK, SAFETY ASSESSMENT / QUALITY VERIFICATION Enclosed is the SALP input for Hope Creek in the functional area of Safety Assessment / Quality Verification.
Input was received from the NRR technical staff, and the senior resident at Hope Creek.
For additional information, contact Clyde Shiraki at 492-1445.
I L
Walter R. Butler, Director Project Directorate I-2
' Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
Enclosure:
SALP Input cc w/ enclosure:
E. Wenzinger, RI G. Meyer, SRI, Hope Creek i
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E_NCLOSURE s.
_ HOPE CREEK GENERATING STATION SALP INPUT T
SAFETY ASSESSMENT / QUALITY VERIFICATION I
_ ANALYSIS:
This new functional area combines the previous functional areas 1
Activities and Assurance of Quality and assesses the effectiveness of the licensee's programs in assuring the safety and quality of plant opera!
activities.
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Category 2 and noted the inconsistent quality of lice l
regarding technical content and timeliness.
functional area was rated as a Category 2 with an improvinThe Assurance of Qu noted that PSE8G had established the programs, procedures,g trend.
The report attention to weak areas such as the engineering departm and working Since it is a relatively new facility, HCGS does not have the same volume licensing actions as most facilities.
During the assessment period sixteen actions (amendments, relief requests, exemptions, etc) were processe,d.
quality of the technical evaluations was generally go The participates in industry groups, and uses acceptable approaches to problem solutions.
l necessary corrections were usually prompt and well handled.T Ventilation System. dealt with a license change request to the Filtration, Recir additional information, and it was April 1989 before it was rece
, and was one instance of supplemental information being req There Specification surveillance test intervals and :llowable outage times for the reactor protection system.
promptly and correctly.
The supplemental information was submitted Eulletins) has been timely and complete.The licensee's response t Frequent communications indicate ethat they commence work on their responses sufficiently in advance that
'are able to meet commitment dates without requesting extensions.
operation; for example, all fourteen safety relief valves tested at power following replacement, not just the required five SRVs the acceptance criteria for High Pressure Coolant Injection (HPCI) System
, and response time testing were reduced for low pressure conditions.
acceptable HPCI overspeed test, the test was repeated to confirm After an acceptability.
When a test engineer raised concerns regardin of isolation valves in primary containment ventilation lines,g the orientation expeditiously raised to the plant management level and corrective actions were the concern was initiated.
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, u PSE&G's' adherence to the concept of personal accountability when observing the Senior Nuclear Shift Supervisors (the "y is most noticeable licensed operators held accountable for plant operations on each o Seniors") the shift. The Seniors ensure that they concur with decisions, such as tech specification interpretations, the acceptability of equipment being. returne to service, and courses of action.
Each morning, the department mana attend a meeting run by the Senior to discuss plant status and plans,gers reinforces the Senior's responsibility and pro which have department managers address his concerns.vides the opportunity for him to The meeting provides ready as well as being a vehicle that quickly involves engine operational problems.
During the evaluation of the feedwater flow mea staff displayed a willingness and ability to analyze data and events Independent of the vendor representatives.
calculations were correct.not accept General Electric (GE) Company GE subsequently acknowledged they had made an error.
Problem identification occurred both from within and from outside organizational element.
resolve plant problems and off-normal events and for tracking actions to completion.. Hope Creek had 170 Incident Reports in Ahich were reportable to the NRC.
1988, 36 of PerformanceEvaluationSystem(HPES),adetailedanalysismethodforP Determining root causes in incidents involving personnel errors.
analysis of personnel errors. analysis technique has the potential for pro This this technique to the personnel errors that occurred during this as period such as those briefly described below.
The facility violated Technical Specifications by operating at nominally 101.2%, worst case 102.2%. Root cause was personnel error in that the established using calculations that were not compen pressure compression.
was not in accordance with NSSS vendor specifications.The omissio (LER88-24)
An isolation of the High Pressure Coolant Injection (HPCI) System occurred during performance of a steam leak detection system surveillance I
procedure.
The I&C technician who performed the surveillance failed to place a bypass switch in the BYPASS position per procedure and when terminal leads were lifted during the course of the surveillance, an isolation of the HPCI primary containment outboard steam supply valve occurred.
Root cause was personnel error in not following an approved procedure.
(LER88-33)
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.( The control room differential pressure was less than technical' specification required values. An engineering review determined that Control Room Emergency Filtration (CREF) system operability had hot been demonstrated.
Two root causes were identified: 1) an inadequate surveillance procedure, and 2) inadequate interface testing following an HVAC design change in the area adjacent to'the control room.
(LER88-25)
A Nuclear Steam Supply System Shutoff Channel "0" isolation occurred when-a fuse was blown on a portion of the Channel "D" isolation logic.
The fuse blew during performance of a Maintenance Department I&C surveillance t
procedure when a meter lead was inadvertently dislodged from test equipment and came in contact with a ground bus-inside the Division 1 Reactor Protection System Lo ReactorWaterCleanup(RWCU)gicCabinet. The isolation caused the System outboard suction isolation valve to auto close, the "A" and "B" RWCU pumps to trip, and isolated Main Steam Line drain valves and a Reactor Recirculation sample valve. -Root Cause-was the lack of accessibility to testing points inside the subject cabinet, which directly led to the meter lead becoming dislodged.
A contributing factor might have been poor work practice or skill.
(LER 88-35)
During the performance of an I&C surveillance test procedure, RHR shutdown cooling was isoi-ted because the procedure did not call for lifting a lead to prevent a valve from closing.
(LER89-04)
Primary Containment Isolation Yalves were declared inoperable due to a missed' surveillance test that resulted from a personnel error.
(LER 88-02)
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A missed surveillance test of the refueling floor exhaust process l
radiation monitor Channel "8" caused by a personnel error resulted in a technical specification violation.
(LER88-04).
A Design Change Package (DCP) inatequacy resulted in the inputs to the primary containment isolation system being inoperable.
(LER88-05)
An isolation of the reactor water clean up system (ESF actuation) resulted from misuse of test equipment, which caused a blown fuse and ESF actuation.
(LER88-18)
A power reduction and ESF actuation (RWCU isolation) were caused by loose l
terminations on a cabinet internal power supply.
(LER88-34)
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AninadequateDesignChangePackage(DCP)causedanoscillationin L
drywell average air temperature measurements. The DCP was inadequately reviewed to determine the impact of its implementation upon the drywell average air temperature measurement. Abnormally oscillating drywell average air temperatures existed for over a month without being detected by operations or engineering personnel, although this parameter is recorded daily to ensure compliance with Technical Specifications.
i (InspectionReport88-24)
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An uncontrolled electrician jumper was installed in the control circuit of the drywell equipment drain pumps and remained undetected for ten months.
(Inspection Report 89-02)
These items were variously caused by technician error, inadequate procedure review, poor work practices, or a loss of control of equipment. They further i
affect the areas of post maintenance testing, workmanship, and management oversight.
In some way, they deal primarily with the I & C area. Since the responsibility of QA is to work toward quality operation of the facility, and these items are clearly nonsupportive of this goal, QA should be heavily involved trying to determine the causes and recommending solutions.
If QA is already involved, their participation is evidently ineffectual.
The Quality Assurance Department, the Onsite Safety Review Group, and the Offsite Safety Review Group are responsible for providing effective, independent review of plant activities.
The station quality assurance (QA)..
organization should be providing day-to-day review in the quality control and in-process review areas and should be integrated into the station's resolution of problems.
In light of the problems experienced in the I & C area, these 1
groups need to reassess their level of involvement and determine if there is more they can do to be of assistance.
The Station Operations Review Committee (SORC) was composed of department managers and provided consistent, effective review of significant plant issues, including design changes, post-trip reviews, reportable events, and station-wide procedures.
During the optical isolator failure, the SORC met during the night to review the course of action before its implementation, a good indicatien of the 50RC's role.
Three areas of the QA program were assessed during this assessment period, procurement, receipt inspection, and audits.
The three areas continue to be effectively documented and administrative 1y controlled and implemented by trained and qualified personnel.
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Station QA involvement in ISI and startup testing was apparent.
In the ISI area OA performed surveillance of in-progress ISI contractor activities, l
in-house reviews of contractor ISI procedures and audits at the contractor l
facilities. QA performed many surveillance activities during the post
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refueling startup testing program. However, QA has not devoted the same level of attention to IST as evident by the NRC review and supported by the NRC identified weaknesses noted in the Maintenance / Surveillance Area.
In addition, a missed surveillance test and an inoperable valve resulted in a technical specification violation.
(LER88-02)
In summary, the performance of the various quality assurance groups has been inconsistent.
Their involvement in solving the problems of personnel errors, inadequate procedure review, and missed surveillance tests is either nonexistent or ineffcetual. High level management attention is necessary to
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.. bring the quality assurance groups together to clearly define or reempha the responsibilities of each group.
responsible managers held accountable for the results. Commitments should be Although the licensee's licensing action submittais were generally of go quality,.in the functional area of Safety Assessment / Quality Verificatio lack of effective quality verification and its corresponding effects on pla safety resulted in a downward trend in this area.
indicated over the last few months of.the assessment period but close An improvement was attention is required to determine if'this is a coincidental or significant
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