IR 05000219/1986099
| ML20154B261 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 09/30/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20154B212 | List: |
| References | |
| 50-219-86-99, NUDOCS 8805170126 | |
| Download: ML20154B261 (77) | |
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ENCLOSURE 1 U. S. NUCLEAR REGULATORY COMMISSION
REGION I
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SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE INSPECTION REPORT S0-219/86-99 (AMENDED REPORT)
SENERAL PUBLIC UTILITIES NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION s
ASSESSMENT PERIOD:
OCTOBER 16, 1986 - SEPTEMBER 30, 1987 E0ARD MEETING November 17, 1987
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SUMMARY Actual Percent 1.
Plant Operations 1820
2.
Radiological Controls 813
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Maintenance 964
4.
Surveillance 464
5.
6.
Security and Safeguards 165
j 7.
Assurance of Quality N/A N/A 8.
Licensing Activities N/A N/A 9.
Engineering Support 443
10. Training and Qualification N/A N/A Effectiveness
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5089 100
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TABLE 3 ENFORCEMENT ACTIVITY A.
Violations Versus Functional Area By Severity Level Functional No. of Violations in Each Severity Level Area V
IV III II I
Total 1.
Plant Operations
2
5 2.
Radiological Controls
1 3.
Maintenance
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4.
Surveillance
1 5.
Security and Safeguards
1 7.
Assurance of Quality 8.
Licensing Activities 9.
Engineering Support
3
l 10. Training and Qualification Effectiveness Total l-
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Table 3
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SUMMARY Inspection Severity Functional Brief Number Requirements Level Area Oescription 86-37 10 CFR 50, App.B.
IV Engineering Changes to safety-Crit. V, VI Support related electrical systems not docu-mented prior to being implemented.
86-37 Technical IV Plant Three examples Specification Operations of failure to 6.8.1 follow procedures.
87-03 Technical IV Surveillance Failure to prepare Specification a procedure for 6.8.1 a Tech Spec re-quired surveil-lance.
87-12 10 CFR 20.311 IV Radiological Solidified waste (d)(1)
Controls contained exces-sive water.
,87-13 10 CFR 50.55 IV Maintenance Failure to perform a.(g)(4)
hydro after weld
repair.
Technical IV Engineering Failure to perform Specification Support required instru-3.12.I.1 ment surveillances.
10 CFR 50.59 IV Engineering Failure to submit Support repotts required by 10 CFR 50.59.
87-16 10 CFR 50.59 II Plant Tied open suppres-
(a)(1) and Tech-Operations sien chamber to nical Specifica-drywell vacuum catien 3.5.A.3 breakers.
10 CFR 50.72 IV Plant Failure to -ake (b)(1)(ii)
Operations required one hour report.
Technicti III Plant Five examples of Specification Operations failure to follow 6.8 procedures.
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e Table 3
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Inspection Severity Functional Brief
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Number Requirements level Area Description
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87-16 10 CFR 50.59 III Plant Tied open reactor
(Cont.)
(a)(1) and Tech.
Operations building to sup-nical Specifica-pression chamber L
cation 3.5.A.3 vacuum breakers.
87-20 Technical V
Engineering Use of improper Specification Support test gauge during i
4.3.C inservice testing.
87-25 10 CFR 73 V
Security Training
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k TABLE _4 f
LICENSEE EVENT REPORTS A.
LER By Functional Area Number by Cause Code Functional Area A
B C
D
_E X
1.
Plant Operations
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2.
Radiological Controls 3.
Maintenance
4
4.
Surveillance
2
5.
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6.
Security and Safeguards
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Assurance of Quality
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Licensing Activities
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Engineering Support
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Training and Qua*rification Effectiveness Total
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Cause Codes:
A-Personnel Error B-Cesign, Manufacturing, Construction, or Installation Error C-External Cause D-Defective Procedures E-Component Failure X-Other
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Table 4
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B.
LER Synopsis 86-23 SINGLE FAILURE OF CONTAINMENT SPRAY AUTOMATIC B
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INITIATION LOGIC 86-24 POSTULATED HIGH ENERGY LINE BREAK IN ISOLATION B
CONDENSER PENETRATIONS 86-25 GROUNDING OF 4160V ELECTRICAL BUS CAUSED BY A
PERSONNEL ERROR 86-26 REACTOR SCRAM DURING EXCESS FLOW CHECK VALVE A
TESTING 86-27 STANDBY GAS TREATMENT SYSTEM INITIATION CAUSED BY A
PERSONNEL ERROR 86-28 PERSONNEL ERROR DEFEATS AN AUTOMATIC INITIATION A
FUNCTION OF STANDBY GAS TREATMENT SYSTEM 86-29 POTENTIAL INOPERABILITY OF CORE SPRAY EMERGENCY A
SERVICE WATER PUMPS DUE TO INADEQUATE DESIGN AND PROCEDURE REVIEWS
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86-30 ISOLATION CONDENSER "A" ISOLATION ON SPURIOUS
HIGH FLOW SIGNAL 86-31 REACTOR BUILOING CLOSED COOLING WATER TO ORYWELL A
ISOLATION CAUSED BY PERSONNEL ERROR DURING INSTRUMENT FILLING ACTIVITIES 66-32 REACTOR TRIP ON HIGH NEUTRON FLUX CAUSED BY COLD A
FEEDWATER ADDITION DUE TO OPERATOR ERROR 86-33 STANDBY GAS TREATMENT INITIATION CAUSED BY GROUND A
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ON ARM RIBBON CABLE DUE TO PERSONNEL ERROR 86-34 KANUAL SCRAM DUE TO INABILITY TO HAINTAIN CON-E DENSER VACUUM CAUSED BY EQUIPMENT FAILURE 86-35 CONTAINMENT PENETRATION FOUND DEGRADE 0 DUE TO A
ISOLATION VALUES ACTUATOR / VALVE LINKAGE OUT OF
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ADJUSTMENT 87-01 ABSENCE OF NEUTRON FLUX CONTROL ROD BLOCK CLAMPING X CIRCUIT DUE TO INCONSISTENCY BETWEEN TECH SPEC AND PLANT HARDWARE
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i LER hUMBER SUMMARY CAUSE 87-02 MAIN STEAM ISOLATION VALVE CLOSURE CAUSED BY A
OPERATOR ERROR 87-03 STANDBY GAS TREATMENT SYSTEM INITIATION CAUSED E
BY POWER SUPPLY PERTURBATION 87-04 TECHNICAL SPECIFICATION VIOLATION CAUSED BY A
IMPROPER REMOVAL OF EQUIPMENT FROM SERVICE DUE TO PERSONNEL ERROR 87-05 HIGH FLUX SC: OURING RECIRCULATION PUMP START A
DUE TO DISCH: ^E VALVE PARTIALLY OPEN 87-06 TECHNICAL SPECIFICATION VIOLATION CAUSED BY A
IMPROPER STORAGE OF HIGHER ENRICHMENT FUEL DUE TO PERSONNEL ERROR 87-07 BACKUP SAMPLE ANALYSIS INVALIO OUE TO PERSONNEL A
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ERROR 87-03 LIMITING SAFETY SYSTEM SETPOINT FOR TOTAL RECIRCV-B LATION FLOW EXCEEOS TECHNICAL SPECIFICATIONS DUE TO INSTRUMENT OP.IFT 87-09 VOLUNTARY RPT.-OPERATION OF PLANT WITH FLOW BI ASED E SCRAM & R00 BLOCK SETPOINTS OUTSIDE ANALYZED
REGION DUE TO RECIRC LOOP FLOW BACKFLOW 87-10 ELECTRICAL TRANSIENT CAUSES CONTAINMENT ISOLATION X
AND STANCBY GAS TREATMENT INITIATION DUE TO DESIGN CONFIGURATION
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87-11 HIGH RPV LEVEL TURBINE TRIP / SCRAM CAUSED BY LOST
FEEDWATER FLOW SIGNAL DUE TO PROCEDURAL INADEQUACY 87-12 INOPERABLE OFFGAS ORAIN LINE ISOLATION VALVE E
CAUSED BY DEBRIS ACCUMJLATION DUE TO INA0 EQUATE PREVENTIVE MAINTENANCE 87-13 SGTS INITIATION CAUSED BY IMPROPERLY INSTALLE0 A
WIRE CONNECTOR DUE TO PERSONNEL ERROR 87-14 DRrWELL ISOLATION CAUSED BY LIFTING A LEAD A
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87-15 INOPERABLE INTERMEDIATE RANGE MONITOR' DUE TO O
BROKEN FLEXIBLE CONNECTION CAUSE BY
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HAINTENANCE
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LER NUMBER SUMMARY CAUSE 87-16 SETPOINTS FOR THREE OF EIGHT ISOLATION CONDENSER B
PIPE BREAK SENSORS OUT OF SPECIFICATION DUE TO INSTRUMENT DRIFT 87-17 TECH SPEC VIOLATION CAUSED BY INAPPROPRIATE RE-A MOVAL OF SNUBBERS FROM.RVEILLANCE PROGRAM DUE TO PERSONNEL ERROR 87-18 REACTOR BUILDING VENTILATION VALVE INOPERABLE FOR A
MAINTENANCE AND NOT SECURED CLOSED DUE TO PERSONNEL ERROR 87-19 LIMITING SAFETY SYSTEM SETPOINT FOR TOTAL RECIRCU-A LATION FLOW EXCEEDS TECHNICAL SPECIFICATIONS DUE TO FERSONNEL ERROR 87-20 TECHNICAL SPECIFICATION REQUIRED SURVEILLANCE A
OVERDUE DUE TO INADEQUATE SHIFT TURNOVER CAUSED BY PERSONNEL ERROR A7-21 TECHNICAL SPECIFICATION VIOLATION CAUSED BY A
BLOCKING OPEN CONTAIN"ENT VACUUM BREAKERS DUE TO PERSONNEL ERROR 87-22 PLANT SHUTDOWN REQUIRED BY INOPERABLE ACOUSTIC B
MCNITOR DUE TO MARGINAL SPLICE DESIGN RESULTING IN CABLE DAMAGE DURING INSTALLATION 87-23 PARTI AL PRIMARY CONTAINMENT ISOLATION DURIN3 A
TESTING DUE TO PROCEDURAL INADEQUACY 87-24 FAILURE TO POST A FIRE WATCH FOR A NON-FUNCTIONAL A
FIRE BARRIER DUE TO PERSONNEL ERROR IN FAILIN3 TO FOLLOW PROCEDURE 87-25 PRIMARY CONTAINMENT VENT AND PURGE VALVES HAD D
MAXIMUM STROKE IN EXCESS OF DESIGN LIMIT DUE TO INSTALLATION PROCEDURE INADEQUACY 87-26 TEMPORARY VARIATIONS FOUND UNACCEPTABLE DUE TO A
INADEQUATE SAFETY REVIEWS B7-27 ELECTRICAL STORM INDUCED CONTAINMENT ISOLATION B
AND STANDBY GAS TREATMENT SYSTEM INITIATION DUE TO AUTOMATIC BUS TRANSFER TIME EXCEEDING RPS RELAY DROPOUT TIME T-4-4
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LER NUy3ER SuvyARY CAUSE B7-28 HAIN STEAM ISOLATION VALVE CLOSURE CAUSED BY A
DESIGN DEFICIENCY DURING SURVEILLANCE TEST 87-29 HIGH REACTOR PRESSURE SCRAM DUE TO AIR LEAK FRCMA DISLODGED AIR TEST PILOT VALVE CAUSED BY INCORRECT MOUNTING CAP SCREW LENGTH 87-30 LIGHTING ARRESTOR INSULATOR FAILURE INDUCED VOLT-B AGE TRANSIENT CAUSED CONTAINMENT ISOLATION AND SSGTS INITIATION DUE TO AUT0y,ATIC BUS TRANSFER TIME EXCEEDING RPS DELAY DROPOUT TIKE 87-31 VIOLATION OF HIG4 RADIATION AREA TECHNICAL SPECI-A FICATIONS CAUSED SY PERSONNEL ERROR DURING RESPONSE TO FIRE ALARM 87-32 A0G HYDROGEN ANALYZER NOT CALIBRATED IN ACCORD-A ANCE WITH TECH SPEC REQUIREMENTS DUE TO INADEQUATE REVIEW OF RETS AMENDMENT T-4-5
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UNITE 3 STATES ENCLOSURE 2
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NUCLEAR REGULATORY COMMIS$1SN
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REGION I
I 4M ALLENDALE ROAD
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KING OF PRUSSI A. PENN8YLVANIA 19406
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Docket No. 50-219 FEB 121958 GPU Nuclear Corporation ATTN: Mr. P. B. Fiedler Vice President and Director Oyster Creek Nuclear Generating Station P. O. Box 3SS Forked River, NJ 08731 Gentlemen:
Subject:
Systematic Assessment of Licensee Performance (SALP) Report No.
50-219/86-99 The NRC Region I SALP Board conducted a review on November 17, 1987, and evaluated the performance of activities associated with the Oyster Creek Nuclear Generating Station.
The results of this assessment are documented in the enclosed SALP report, which covers the period October 16, 1936 to September 30, 1987. We will contact you shorcly to scheoule a meeting to discuss the report.
At the meeting, you should be prepared to discuss our assessment and any plans you may have to improve performance.
In particular, you should be prepared to discuss the plans you have to upgrade performance in Plant Operations and Technical Support in light of the reducticn in performance in these areas.
Following our meeting and receipt of your response, the enclosed report, your re-sponse, and summary of our findings and planned actions will be placed in the NRC Public Document Room.
Your cooperation is appreciated.
Sincerely, William T. Russell Regional Administrator Enclosure: NRC Region I SALP Report No. 50-219/86-99
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GPU Nuclear Corporation
cc w/ enc 1:
M. Laocart. SWR Licensing Manager Licensing Manager, Oyster Creek Chairman Zech Commissioner Roberts Commissionar Bernthal Commissioner Carr Com.missioner Rogers K. Abraham, PAO, RI (11 copies)
Public Docueent Room (POR)
local Public Document Room (LPOR)
Nuclear Safety Information Center (NSIC)
NRC Resident Inspector State of New Jersey bec w/ encl:
Region I Docket Room (with concurrences)
Manage. ment Assistant, DRMA (w/o enc 1)
J. Taylor, DEDO J. Lieberman, OE W. Russell, RI
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T. Martin, RI W. Johnston, RI D. Holody, RI SALP Board Meeting Attendees R. Brady, RI C. Cowgill, Section Chief, DRP Robert J. Beres, DRSS i
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f ENCLOSURE 3
. LIST OF ATTENDEES SAlp MANAGEMENT MEETING, MARCH 3, 1988 GPU Nuclear Corporation (GPUN)
P. F. Ahern, Senior Staff Specialist, NSCC-TMI G. R. Bond, Director, Systems Engineering R. P. Clark, President, GPUN C. Clawson, Director, Communications D. K. Croneberger, Director, Engineering Projects i
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B. DeMerchant, Oyster Creek Licensing Engineer
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P. B. Fiedler, Vice President and Director, Oyster Creek R. F. Fenti, Manager, Oyster Creek, Mods / Ops I
R. Finfrock, Chairman, GORBs L. Graibian, Civil / Structural Manager D. V. Hassler, TMI-1 Licensing Engineer J. E. Hildebrand, Industrial Safety / Environmental Control Director
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J. D. Kowalski, Oyster Creek Licensing Manager R. L. Long, Vice President and Director, Division of Planning and Nuclear Safety F. F. Manganaro, Vice President and Director, Administration
'i R. S. Markowski, Manager, QA Progran Development and Audit R. J. McGoey, Manager, TMI Licensing
B. T. Meroney, Senior Staff Specialist, NSCC-0C
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O. W. Myers, Vice President / Comptroller K. R. Meddenien, Senior Media Representative i
W. Popow, MCF Production and Technical Director M. B. Roche, Vice President and Director, Division of Quality and Radiological
Controls A. P. Rochino, Manager, Engineering Mechanics A. H. Rone, Plant Engineering Director, Oyster Creek M. O. Sanford, Manager, Mechanical Systems
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J. L. Sullivan, Jr., Plant Operations Director, Oyster Creek J. R. Thorpe, Director, Licensing and Regulatory Af fairs G. E. VonMieda, Chemistry / Materials Director E. G. Wallace, Engineering Services Director P. F. Wells, Safety Review Engineer R. F. Wilson, Vice President and Director, Technical Functions t
K. G. Wolf, Radiological Engineering Manager
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Enclosure 2
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U.S. Nuclear Regulatory Cor. mission (NRC)
L. H. Bettenhausen, Chief, Projects Branch No.1, Division of Reactor Projects (DRP)
R. J. Conte, Senior Resident Inspector, TMI
C. J. Cowgill, Chief, Reactor Projects Section 1A, ORP A. W. Dromerick, Project Manager, Oyster Creek, NRR
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R. Hernan, Project Manager, TMI-1, NRR W. V. Johnston, Director, Division of Reactor Safety (DRS)
W. F. Kane, Director, DRP W. T. Russell, Regional Administrator J. F. Wechselberger, Senior Resident Inspecter, Oyster Creek New Jersey Department of Environmental Protection - Bureau of Nuclear Engineering
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L. H. Hamersky, Nuclear Engineering Supervisor M. Jacobs, Nuclear Engineer
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ENCLOSURE 4
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OPU Nuclear Corporatkwt NQgIgf Post Off.ce Box 388 Aoute 9 Soutn Forked Arver. New Jetsey o67310388 6o9 971 4000 Writer's 06tect Dial Number:
U.S. Nuclear Regulatory Comission Attn: Document Control Desk April 4, 1988 Washington, DC 20555
Dear Sir:
Subj ect: Oyster Creek Nuclear Generating Station Docket No. 50-219 Systematic Assessment of Licensee Performance (SALP) Response As discussed with you at our meeting held in Parsippany on March 3,1988, this letter and its attachments provide our response to the Systematic Assessment of Licensee Performance (SALP) report as requested by your letter of February 12, 1988.
Attachment I provides our response to your two areas of concern which include Plant Operations and Engineering Support. Attachment !! provides additional information and clarification for areas we feel misunderstandings may exist, such as Energency Preparedness and Surveillance. Attachment !!! provides general coments in other areas.
We thank you for the opportunity to share our thoughts with you during the March meeting.
We continue to feel that the SALP is a useful tool in the nuclear industry.
Very truly yours, e
d P. R. Clark President PRC:ded(0454A)
Attachments cc:
Mr. William T. Russell, Administrator c
Region !
U.S. Nuclear Regulatory Comission 475 Allendale Road King of Prussia, PA 19406 Mr. Alexander W. Dromerick, Project Manager U.S. Nuclear Regulatory Comission Washington, DC 20555 WRC Resident Inspector Oyster Creek Nuclear Generating Station
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GFu kves, cac>at.
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orat,on
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ATTACHENT I
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PLANT OPERATIONS
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Introduction
.The most recent NRC Region ! SALP Board review period ended on September 30, 1987.
Strengths were noted in the areas of teamwork and professionalism and in the programs used to improve these areas.
In spite of those strengths, the report cited various performance deficiencies which indicated inconsistencies in the application or appreciation of these programs.
Since September 30, 1987, GpVN has expended considerable resources and measurable progress has been made in correcting deficiencies and further improving its strengths. Emphasis has been placed on providing more consistency in the application of programs.
This sumary provides a discussion of management's focus and of the programs employed.
The discussion is divided into the following areas:
Management Development and Team Building 2.
Elimination of Personnel Errors 3.
Procedural Compliance
Root Cause Management Development and Team Buf1 ding As noted by the NRC, Management has made a special effort to improve professionalism and teawork in Operations.
To achieve this, several programs
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have been implemented. These programs are geared towards improving Operations management capabilities, assuring that individual shif ts function effectively
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as a team, and assuring that 0yster Creek management is completely integrated towards a common goal, and accountable for their role in operating the plant safely and efficiently.
Some of these programs are described as follows:
Successful Shift Management Skills This is a workshop presented by a management consultant to shif t supervisors and selected plant management. The program stresses the shif t supervisor's role as a managee and provides an in-depth analysis of individual management style and personality.
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intended to enhance a supervisor's ability to function as a shif t manager by teaching the following key skills:
comunication
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group development and leadership
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conflict management
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general management and motivation
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action planning
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Five shift supervisors have already attended, and three more are scheduled to attend the next class.
GPUN plans to have all shift supervisors attend this training.
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Supery sor Development Course This is a five day training course presented by GPUN to familiarize the supervisor with current management concepts, methods, practices, and techniques most appropriate for effective and efficient supervisory performance.
Some of the concepts which are taught during the program are as follows:
Role of the Supervisor
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Styles of Supervision
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Comunicating and Listening
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Functions of Management
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Interpersonal Relations
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Human Resources
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Labor Relations
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Giving Recognition
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Problem Solving
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Delegating
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Four shif t supervisors have attended the program and it is intended to have all shift supervisors attend.
Team Skills Training This is a workshop conducted by a consultant - an ex-naval aviator who specializes in control room team building.
The program stresses the existence of team member attitudes and how they may be used to make a team successful.
The key concepts that are taught during the program are as follows:
ef fective comunication
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feedback
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effective influence conflict resolution
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leadership
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This training was originally presented at +he simulator in the fall of 198/ to all licensed operators.
Plant management participated in the I
training and observed sessions with the shif t teems.
Useful feedback was provided to those involved via video tapes and management critiques.
The second phase of this program is continuing at Oyster Creek.
A second round of team skills training at the simulator is l
planned for this spring, l
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Code of Ethics Training
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This is a senior raactor operator workshop presented by INPO to introduce shift personnel tc the "code of ethics" concept.
Two shift personnel and one operations manegement representative attended and will be taking the lead in developing a code of ethics for Oyster Creek operations.
This effort is designed to solidify a professional and fonnal approach to operations.
As the training enhanced specific management skills, a separate assessment was made by management to assure that individuals on each shift were compatible.
With assistance from corporate staff, each shif t member was evaluated regarding personality, behavior, and ability to function without conflict. As a result of this effort, improvements in compatibility were achieved by making several shift team changes, and by a continuing coaching program for various operations personnel.
The work stoppage placed Oyster Creek management in a challenging situation that resulted in developing better team skills.
Management completed the maintenance outage, started up the plant, and in the process, gained valuable experience.
The firsthand knowledge was beneficial and advantageously applied to the work force upon their return.
This experience and the successful integration of the work force back into the plant stand as an example of management conunitment to the success of the plant.
The Director of Oyster Creek held a two day workshop in December 1987 involving managers from the Oyster Creek Division and all on-site supporting Divisions to identify individual expectations that are essential for achieving improved plant perfonnance.
As a follow-up to the workshop, the group has been meeting weekly. A prime goal that resulted was to direct attention to, and improve support of the first line supervisors. To this end a meeting was held with all first line supervisors to discuss their role in the success of Oyster Creek.
Elimination of Personnel Errors It is believed that the potential for personnel errors has been reduced by improving teacrwork, providing training, reducing unnecessary challenges to the
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operators, and improving procedures. Control Room personnel have attended team skills training to reinforce the use of existing skills and knowledge.
This training encourages interaction within the groups and has been discussed in more detail in the previous section.
The result has been improved communication and coordination among shif t personnel which has enhanced the formality and professionalism in the conduct of operations.
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l Various formal training sessions have been provided to all operators on major
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l procedure changes and mishaps in the plant.
Examples include training on the j
eculpment control procedure and on the design and operation of motor operated
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valves.
Procedure training has been useful for providing a uniform interpretation and application of procedures.
Training related to mishaps has provided feedback to correct perfonnance deficiencies and prevent mishaps from recurring.
This approach has been particularly useful at the simulator where hands-on experience has been obtained with a variety of evolutions.
The effectiveness of training of this nature has been measured by written examinations. Additionally, simulator training perfomance was evaluated by Operations and Training management.
Other less formal training has been provided in advance of non-routine evolutions such as implementation of a large freeze seal in the feedwater system.
This has provided a forum for becoming familiar with the evolution and resolving concerns.
The training has consisted of an explanation of the upcoming evolution, required personnel involvement, and potential safety concerns.
Improvements in Equipment Operator (E0) performance have also been addressed.
Considerable effort has been made to assure that EOs increase their attention to non-routine plant conditions.
To achieve this, E0 training has been upgraded to stress their accountabilities especially related to tours.
In addition, shift supervisors have toured the plant with E0s to help identify and correct typical problems.
Management will continue to pursue improving E0 performance in this area.
In addition, either a turnover checksheet or an equipment status board placed in the E0s' room is being evaluated for implementation.
Evaluations of the EOs' performance via Group Operating Supervisor (GOS) observation and Operations QA ;nonitors are also being evaluated for implementation.
Other means for improving training have also been initiated.
A special board has bee.1 formed to provide a conduit for comunication regarding the effectiveness of operator training programs.
Members of the board include instructors and operators whose goal it is to improve the overall quality of operator training. A significant improvement, which provides additional hands on training, is that each licensed operator will now attend two rather than one simulator training session during each requalification cycle, s
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Progress has already been made in alleviating unnecessary challenges to a
operators. Major emphasis has been placed on reducing the engineering and maintenance backlogs and on reducing long standing materiel problems. As a
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result, the materiel condition of the plant has been upgraded and is evident in the significant reduction of temporary variations, control room deficiencies, and control room instruments out of service.
This is the result of increased attention by, and improved communication among maintenance, engineering, and operations personnel.
Additionally significant operating challenges are being addressed by implementing modifications such as the modification to the range switch which will prevent errors in operating it.
Numerous procedure changes have been implemented to further reduce unnecessary challenges to operators. Many of these changes were submitted by Plant Operations personnel with considerable experience in operating the plant.
Most notable are the major changes to Equipment Control Procedure 108 which
included more stringent controls over the temporary variation and safety review processes. Many of the changes were improvements to procedures used by operators to operate equipment, perform tests, and conduct maintenance activities.
Reducing personnel errors will require a commitment to high standards from those most directly involved. For this reason these personnel will participate in developing a code of ethics during 1988 to stand as their statement of this commitment.
Procedural Compliance Management is cognizant of incidents involving procedural non-compliance, and is committed to eliminating contributing causes.
There exists an NRC perception that procedures may not have been followed in order to meet scheduling demands. We agree that there have been instances where GPUN personnel have placed perceived schedule perfonnance above strict procedural compliance.
This has not been in accordance with previous GPUN management direction.
Schedules will continue to be used; however, management will continue to emphasize that at no time should schedules be given priority over plant safety or procedural compliance.
Procedures may exist that do not give clear direction for performing a task at hand. Additionally, in the past, procedure reviews were performed by staff members rather than operators which may have created a situation r.ot conducive to detecting deficiencies.
To correct this, operators will be assigned responsibility for reviewing their own procedures.
This will provide them with the immediate means for improving procedures. Management will continue to emphasize that proceduret must be changed whenever a task cannot be completed as written.
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Root Cause
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Finding the root cause has become a top priority regarding operating problems. Recently this concept has been reinforced by Plant Operations refusing to accept equipment that was out for maintenance before a positive cause has been identified and corrected.
Engineering, operations, and maintenance personnel have been working closely to resolve equipment deficiencies.
Recent resolution of a control rod drive pump motor breaker problem and ongoing investigation into acoustic monitor troubles are examples of this.
Oyster Creek Management has become more sensitive to the importance of finding root causes. As a result, a more inquisitive attitude has developed and problems have been pursued to a permanent resolution.
The operating performance in the past four to five months stands as testimony to the success of this new attitude.
In order to help assure continued success in this effort, Operations Department now chairs the daily 2:30 meeting to coordinate maintenance activities.
The Corporate Safety Review Group has developed. Procedure 1000-ADM-1201.01 (Event Critique and Reporting) which establishes the requirements that each division must comply with when performing a critique at GPUN.
Included in this procedure are detailed guidelines for determining root cause.
Oyster Creek management is in the process of developing a critique procedure that will comply with corporate requirements.
Also, the INP0 sponsored Human Performance Evaluation Systems (HPES) is being implemented at Oyster Creek and THI-1.
The program will be facilitated by the Independent On Site Review Group.
The purpose of HPES is to identify, evaluate, and correct situations that involve human performance errors. A full time HPES coordinator has been selected for each site and will receive special training by INPO in root cause analysis.
_Concl u sion GPUN believes that recent programs to improve management effectiveness, personnel performance, equipment performance, and procedures will be major
inputs to the future success of Oyster Creek. Areas that needed improvement received management attention and were thereby improved.
Emphasis that was
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placed on providing more consistency in the application of programs has produced apparent positive results as shown by Oyster Creek's recent
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ENGINEERING SUPPORT
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GPUN agrees that the performance in the area of Engineering Support has been inconsistent. We are addressing the NRC coments specific to this area in five broad categories (1) engineering staff accountability, (2) vendor control (both suppliers of engineered equipment and Architect / Engineers), (3) schedule insensitivity related to backlog, NRC issues, and modification engineering, (4) communication of design organizations with other functional groups, and (5) technical reviews.
Consistent with the previous SALP response, we have been performing a self assessment of the engineering support area.
Phase I consisted of a structured survey of the engineering staf f and the "user community." The summary of results is completed and the final phase, including conclusions and action plan, is to be complete by May,1988.
We look forward to reviewing the results of this assessment with your staff.
We have been working to improve these areas by the following actions:
Modifications Implement program changes to
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Stop splitting design responsibility on individual modi fica tions.
Move to use only one A/E per plant for design work not perforned by GPUN.
- Enhanced design reviews with stronger and earlier operations /
maintenance input.
- Enforce plant walkdowns by design organization.
Establish schedules to ensure earlier release (target of 6
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months) prior to an outage.
Continually implement corrective actions based upon analysis of
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quality trends.
Backlog - Focus resources to further reduce engineering action items.
Configuration Control Design Document Data Base (CARIRS) improved to further address
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user issues.
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Vendor Manuals - Essential manuals have been reviewed and are
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being controlled.
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Completed Equipment Level Quality Classification List and
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Engineering Data Base. Approximately 25,000 components have been entered into Engineering Data Base.
Owners Group - Increase focus on Owners Group and taking leadership role in them.
Management will continue to emphasize the need for performing thorough and timely responses to technical issues by future actions which include:
(1 )
Prompt implementation of action plans developed as a result of the self assessment.
(2)
Emphasizing the need for operations and maintenance input on plant modifications by adherence to procedural requirement for design reviews.
(3)
Strengthening technical and safety reviewer training.
(4)
Focusing resources on timely reduction of backlog.
(5)
Extending the sound technical support provided by the Start-Up and Test organization to shop testing of engineered vendor hardware.
Certain clarification of examples cited in your report are addressed in Attachment I-A.
The feedback provided in your report will be factored into GPUN's decision making process on these matters.
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ATTACHMENT !!
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EMERGENCY PREPARE 0 NESS
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We believe the substantive aspects of our performance remain strong. We have not noted a performance level change reflected by the lower current SALP rating compared to the prior SALP.
The cited differences include the January 27, 1988, callout.
We concur with the SALP that the problem was corrected by a subsequent call out response.
The most significant performance problem identified by the SALP was that during the 1987 Annual Exercise, the Emergency Support Director (ESD) did not issue a timely protective action recommendation (PAR). Our drill records show that the PAR was timely and was provided to the state ten minutes after the General Emergency was declared.
As for the timeliness of the General Emergency declaration, we believe the ESD was realistic in his declaration.
He did not just "declare the inevitable" because it was the annual exercise.
Rather, he waited until simulated readings showed that emergency action levels had been exceeded.
The timely PAR followed that declaration.
Evacuation Time Estimates (ETE) were known by the ESD.
They are an integral part of the PAR Logic Diagram which the ESD used to make his PAR.
However, the ETE was not important for the situation in the scenario.
At a BWR like OCNGS, the most likely release path for a serious accident is from the primary to secondary containment and then out the stack.
This makes for a very long release (e.g., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).
The integrated dose which requires the PAR is based on this long release.
The worst case ETE is 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />; hence, the nonverbal consideration of ETE's by the ESD.
The FEMA observations were rela ad to a Pinelands High School decontamination
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center not being properly operated and the South Toms River Emergency Management Coordinator not participating in the exercise--both of these issues have been properly dispositioned.
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SURVEILLANCE /IN-SERVICE TESTING
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In this area we would like to comment on the two examples given (Core Spray Pump and Emergency Service Water [ESW]).
We would also like to take exception to your statement on page 34 that "No other troubleshooting of significance was performed to determine the cause of the problem.
Management's willingness to accept the results of a repeat surveillance without a satisfactory explanation as to why the first one failed, demonstrates lack of aggressiveness in root cause detemination."
The following is a description of our troubleshooting efforts:
Unexplained Trip of Core Spray Booster Pump Breaker On February 13, 1987, during the performance of a surveillance the circuit breaker for NZ03A failed to operate properly.
After the breaker tripped, the Operations Department racked the breaker in and out and then re-performed the surveillance test.
This time the breaker closed properly.
The pump was then started manually and the breaker again closed properly.
The Operations Department then issued a short fonn to MCF to investigate the original failure. When MCF and Plant Engineering became involved the breaker was perfonning properly and t'.e failure could not be repeated.
A decision was then made to perform Preventive Maintenance (PM) on the breaker.
This began the same day as the t0ilure occurred.
The PM included:
1.
Long time and instantaneous overload trip settings 2.
Trip torque measurements 3.
Mechanical adjustments of main and arcing contacts 4.
Megger of breaker 5.
Megger of breaker with power leads connected.
All data was reviewed by maintenance and engineering.
Based on no abnormalities being found and the continued proper operation of the breaker, it was returned to service.
Since this incident, regularly scheduled PMs were performed on this breaker on April 26, 1987 and October 29, 1987 with no abnormalities found on either occasion.
No further trips of this breaker have occurred during any of the interim surveillance.
Due to an increase in problems with these breakers even v
before this incident, GPUN started during our last refueling outage, to o'
completely overhaul and upgrade these breakers with solid state trip
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devices.
We are well into this program and expect to be complete with all breakers on site (safety and non-safety related) in 1989 with all but a few before the completion of our next refueling outage.
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ESW Low Flow Incident
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On February 4,1987 at about 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br />, the ESW "B" pump was started for a normal monthly surveillance. The ESW "A" pump surveillance was just completed. According to the operators conducting the test, the following occurred:
The pump was started.
The operator at the intake structure noticed that with the pump running, the discharge pressure was reading 0 psig. (normal pressure is approximately 110 psig). Coincidently, the Control Room.
operator who started the pump noticed that the motor current was reading 20 amps (normal running current is approximately 53 amps).
The pump was shutdown. A cognizant plant engineer proceeded to the intake structure.
At approximately 1030 hours0.0119 days <br />0.286 hours <br />0.0017 weeks <br />3.91915e-4 months <br />, the plant engineer requested Operations to start the pump.
The pressure rose to normal as did motor current.
The surveillance was completed successfully. Assuming the readings reported by Operations to be correct, the following possibilities were considered:
1. The motor became uncoupled from the pump.
2. Obstruction in the piping.
3. Obstruction at the pump suction.
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1.
The pump shaft is coupled to the motor at the top of the motor via a keyed coupling.
If the key had sheared, it is unlikely that the pump would have run when started later.
2.
Obstructions in the piping was the first suspicion.
The obstruction would have been upstream of the pressure gauge and wculd have been basically leak tight t'ecause any leakage downstream of the obstruction would have caused a pressure indication of soce type. An ice plug would have been the only obstruction capable of this because any other material would have had to pass through the pump impellers, which was unlikely.
The ice plug theory was discounted fcr two obvious reasons:
1) the plug could not have celted during the time between pump starts if it was tight enough to deadhead an ESW pump capable of over 200 psig; 2) the temperatures for the previous 3 days were unseasonably high which would not have maintained an ice plug.
Other information which discredits the formation of Ice in the system upstream of the gauge is based upon a strong suspicion (although not proven) that the water from the discharge check valve back to the pump drains out, due to air in-leakage from the pump shaf t packing.
This packing is not leak tight and water can be observed leaking out during testing. Additionally, none of the other pumps showed any indication of a problem.
If ice blockage was the problem, it should J
have af fected all ESW pumps.
There is heat trace and insulation f.
around the ESW piping back to the pump discharge flange.
There is no heat trace or insulation upstream of the flange (pump discharge head
and discharge column) which is where the ice would have had to form to indicate no pressure on the discharge gauge.
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n obstruction at the suction seemed like the most probable cause for the I
condition of low motor current and no discharge pressure.
At the time of
this test, work on the traveling screens was in progress.
It may have been possible that some material (heavy canvas, a heavy rain coat, etc.)
was accidentally dropped during the traveling screen work.
This object could have gotten caught on the pump end bell and when the pump was started, blocked the suction.
When the pump was shutdown, the turbulence could have knocked the obstruction off or the next time the pump was started it could have been sucked into the pump.
Based on this evaluation, the data from the test was analyzed to determine if pump damage had occurred.
If something was drawn through the pump and deposited in the Containment Spray Heat Exchanger, the differential pressure across the exchanger would have shown an increase from previous testing. The data showed no such indications of either pump degradition or heat exchanger differential pressure increase.
Therefore, an obstruction near the pump suction was identified as the most likely cadse of the problem.
Engineerina judgment determined a need for an increase in the testing frequency in the event an obstruction was still in near proximity of the pump suction.
Additional tests were run on February 5,1987 and February 6,1987 then weekly for 4 weeks then monthly as normally scheduled. During April, three tests sere performed on a daily basis due to problems with the ESW System II flow meter.
In all,11 tests were run on ESW System I from February 4,1987 to May 20, 1997 with all data within 2 or 3 percent.
No similar problems with the ESW "B" pump have been noted since.
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ATTACHMENT l-A
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There are certain issues addressed in the SALP Report that require clarification:
1.
Although there have been a number of problems with the Recirculation Pun:ps, problems associated with the pumps themselves have not been of a recurring (i.e., similar) nature.
We are addressing continuing problems wit.h the power and control to these pumps.
2.
Modifications were made to the Trunnion Room Fans which resulted in acceptable fan vibration levels for the past 1-1/2 years.
3.
Problems with the Offgas Sample Pumps (piston chamber flooded with oil)
were corrected by installation of vacuum breaker lines.
4.
We believe the response to the ESW surveillance test failure of February 4,1987 was proper and reflected a sound approach.
Refer to the Surveillance /In-Service Testing section in Attachment II for details.
5.
We believe the response to the Core Spray Pump Breaker (NZO3A) failure on Februa ry 13, 1987 was also proper.
Refer to the Surveillance /In-Service Testing section it: Attachment II for details.
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ATTACHMENT !!!
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RADIOLOGICAL CONTROLS The company agrees that total worker dose should be reduced and has made improvements. We will continue to press this area strongly to improve the general ALARA situation in the company and to reduce collective exposure. We believe the two violations noted were isolated cases; we have cor:ected the deficient administrative procedures, improved training and pre-job briefings for Radiological Controls technicians, strengthened the quality control functions of Radiological Engineering, and have developed a strategy for improving ALARA and reducing exposure. We disagree with the concern raised about the criteria for an ALARA review. We believe that 90 percent of the exposure received during the last refueling outage had the benefit of an ALARA review.
The company has performed a decontamination of a portion of the recirculation loops and is actively planning to conduct another decontamination in a future outage. The 1987 Collective Exposure for Oyster Creek was approximately 520 person-rem. This is the industry average for BWRs.
ASSURANCE OF QUALITY We agree that improvements are needed and are working to make theu in both the performance of the Quality Assurance (QA) Department and more general improvement of work quality at the site.
In the area of the QA Department, we have made improvements in the technical knowledge of inspectors and the training conducted, and are presently performing an inspector-training-needs analysis which is expected to be completed by the second quarter of 1988. We have strengthened the mechanisms for involvement of QA/QC in site activities and have improved inspection techniques. While we agree with the general analysis o" the technical knowledge of the inspectors, we believe some of the exampla presented have errors of information. For example, the alleged impecper QC holdpoints for some valve maintenance w*.s described in a company report, but upon completion of a thorough investigstion, the allegation was found not to be true and this was reviewed with the Senior Resident.
We will review tnis and other examples with the Resident Inspector.
One of the examples cited for QC inspection inadequacy is inappropriate in that the drywell shell thickness readings have extreme variability as a result of the
surface conditions.
The inability to replicate a result is due to the extreme variability of surface condition and instrumentation limitations, but this does not prevent a valid statistical argument being made to support the conclusions.
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ENCLOSURE S SALP BOARD REPORT ERRATA SHEET PAGE LINE NOW READS SHOULD READ
40, 41 No other troubleshooting of sig-The licensee chose to per-nificance was performed to deter-form additional surveillance mine the cause of the problem.
actions as a method of troubleshooting.
Initial actions taken by the lic-i ensee in each case destroyed as found evidence to use in troubleshooting the problem and resulted in not
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determining the cause of
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j the problem.
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Basis: The wording change was made to reflect the additional surveillance actions
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the licensee took in troubleshooting these particular problems.
In addition, the i
wording was changed to clarify the licensee's approach to root cause determination.
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I PAGE LINE NOW READS SHOULD READ
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24-25
Lack of aggressiveness Managements willingness to i
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peat surveillance without
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a satisfactory explanation as to why the first one
failed demonstrates an in-
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compiete approach to root l
ause deternination.
Basis: This wording change was made to more accurately reflect the licensee's i
approach to root cause determination.
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f PAGE LINE NOW READS SHOULD READ
9, 10 A protective action recommenda-A declaration of a general tion (PAR) in a timely manner emergency with the associ-and that evacuation time esti-ated protective action re-mates were not used in reaching commendation in a timely PARS.
manner.
It is not apparent that evacuation times esti-mates were used in reaching P/ R s.
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SALP BOARD REPORT ERRATA SHEET PAGE LINE NOW READS SHOULO READ
44 Trend Nothing (omit word).
3 Trend Nothing (omit word).
36 Trend Nothing (omit word).
29 Trend Nothing (omit word).
41 Trend Nothing (omit word).
26 Trend Nothing (omit word).
3 Trend Nc. thing (omit word).
19 Trend Nothing (omit word).
e Basis: Administrative change s.
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Enclosure 5
t Basis: The wording was changed to reflect the declaration of a general emergency which should immediately precede the PAR recommendation to the State.
This should clarify the actual sequence of events that occurred during the drill. Since the declaration of a general emergency was untimely this delayed the PAR recommendation to the State, which should have been recognized by the licensee if evacuation time estimates (ETE) were effectively employed.
Therefore, the effective use of ETE's was-only apparent as no verbal communication of ETE use was communicated for this particular situation.
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PAGE LINE NOW READS SHOULO READ
44 Trend Nothing (omit word).
3 Trend Nothing (omit word).
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36 Trend Nothing (omit word).
29 Trend Nothing (omit word).
41 Trend Nothing (omit word),
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26 Trend Nothing (omit word).
3 Trend Nothing (omit word).
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19 Trend Nothing (omit word).
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Basis: Administrative changes.
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