IR 05000344/1987028
| ML20237H050 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/30/1987 |
| From: | Mendonca M, Pereira D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20237H022 | List: |
| References | |
| 50-334-87-28, GL-83-28, IEB-84-03, IEB-84-3, NUDOCS 8708240345 | |
| Download: ML20237H050 (7) | |
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U. S. NUCLEAR REGULATORY COMMISSION j
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REGION V
Report No:
50-344/87-28.
l Docket No.
50-344 License No. NPF-1 Licensee:
Portland General Electric Company
~121 S. W. Salmon Street Portland, Oregon 97204
. Facility Name:
Trojan Nuclear Plant
' Inspection at:
Rainier, Oregon Inspection co ucte.d:
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.7, 1987 Inspector:
de/
w as
?30 W
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D'. B. Pereira, Reactor Inspector
'Date Signed Approved by:
7/d o/f 7 M~. M. Mendonca, Chief, Date Signed Reactor Project Section 1 Summary:
Inspection During the Period of July-13-17, 1967 (Report 50-344/87-28)
Areas Inspected:
This routine, unannounced inspection by the Project Inspector involved the areas of Licensed Operater Training, Temporary Instruction 2500/16 (Seismic Interaction between In-Core Flux Mapping and Seal Table), Temporary' Instruction 2515/91.(Inspection Followup to Coneric Letter 83-28, item 4.1), and Information Bulletin 84-03 (Refueling Cavity Water Seal).
During this inspection, inspection modules 30703, 41701, 25591, 25015, and 92703 were used.
Results:
No violations or deviations were identified.
8708240345 070006 PDR ADOCK 0500
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DETAILS
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1.
Persons Contacted a.
Licensee Personnel i
- C. A. Olmstead, Plant Manager.
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- R. P. Schmitt, Manager, Operations and' Maintenance--
- S.'Nichols, Supervisor,' Training-J. D. Reid,5 Manager, Plant Services
- D.' W. Swan,-Manager, Technical Services
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- G. A. Zimmerman, Manager, Nuclear Regulation Branch
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- D. Nordstrom, Engineer, Nuclear Safety and
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Regulation Department
- W. L. Kershul, Engineer, Nuclear Safety and Regulation
. Department
- -N. Dyer, Manager, Radiological Safety Branch R. Reinart, Supervisor, Instrument and Control b.
U.' S. Nuclear Regulatory Commission
- G. Suh c.
Oregon Department of Energy H. Moomey, Oregon Resident Inspector
- l Attended the Exit Meeting on July 17, 1987.
12.
Licensed-Operator Training The licensed operator training program was evaluated for conformance with the requirements of the Technical Specifications (TS), regulatory I
requirements,'and licensee commitments.
The purpose of this interction I
was to determine the effectiveness of the licensed operator training program.
The inspector reviewed records of four senior reacter operators (SR0s) and four reactor operators (R0s) to ensure that requalification training was adequately documented and to determine whether any lessons
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learned from events / activities were effectively factored inco the training program.
J The inspector's review of the above individuals' records indicated that they contained the following:
a.
Copy of the most recent annual written examination and the individual's responses.
b.
Documentation of attendance at all required lectures, c.
Copies of performance evaluations.
d.
Documentation of additional training received in identified deficient areas.
e.
Documentation of procedure reviews and/or self-study had been completed.
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Documentation of the required control manipulations had been completed.
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Documentation that required reading of routed material had been i
completed.
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The inspector attended several training lectures and verified that the technical content of the lectures was generally acceptable.
However, based on the observation of a lecture on Emergency Preparedness procedures,. Emergency Coordinator / Shift Supervisor /STA Responsibilities,
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and the Control Room Staff Responsibilities, additional attention in these areas was considered appropriate.
The observations of this training lecture indicate procedural unfamiliarity and resistance to change of emergency preparedness procedures on the operators' part.
Also, the use of the simulated exercise could have been better organized and presented.
The licensee agreed with these findings and has proceeded to restructure and re present the lecture.
In addition, the inspector felt that the operators were not familiar with the latest proposed emergency preparedness procedure revisions which had not yet been formally approved nor distributed.
The inspector discussed the idea of training on new unapproved procedures prior to approval and issuance with the training supervisor, who informed the inspector that this was the licensee's conscious decision, since the SR0s and R0s would be shortly going to simulator training and it was decided to train on the new EP procedures prior to going to the simulator.
This would eliminate retraining on newly issued EP procedures upon the operator's return from the simulator.
The operations personnel would be trained on the new preliminary procedures and then after training was conducted, the licensee would formally approve and issue them.
It should be noted that there was an operations input and review to the new EP procedures and that only the training class attended by the inspector was considered unfamiliar with the new EP procedureo The inspector attended the Service and Instrument Air System and the Primary and Secondary Chemistry training lectures and verified that they were presented in a clear and concise manner.
The instructors were well qualified to present the technical material and were familiar with the material.
The inspector determined the present pass rate for licensee requalification exams and for initial licensed operator qualification exams over each of the past three calendar years.
The table below
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presents the pass rates for both initial and requalification exams given l
by the NRC and licensee, respectively, over each of the past three calendar years:
1985 1986 1987 Reactor 89.47%
94.44%
100%
I Requalification Operator Exams Senior Reactor 88.23%
100%
100%
Operator
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1 Initial Reactor 66.67%
No exam 40%
Licensed Operator given Exams Senior Reactor
.100%
100%
100%
Operator The inspector reviewed the progress of the licensee with respect to
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accreditation by the Institute of Nuclear Power Operations (INPO) with
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the training supervisor.
The licensed operators and non-licensed operators program was approved and accredited by INP0 in December 1985.
All remaining programs have completed the job analysis, the task analysis, and the self-evaluation phases.
INPO has reviewed the remaining programs, and is presently evaluating the programs for a report
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to the National Nuclear Accrediting Board.
The licensee expects the report to be completed before the end of 1987.
The licensee's licensed operator training program was considered to be in accordance with the licensee commitments, regulatory requirements, and Technical Specifications.
The record review indicated that the licensee's records are complete and that the appropriate training documentation was kept and available.
The inspector's observations of
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the licensed operator training program and the results achieved in NRC
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exams indicate that qualified operators have been acceptably trained.
Finally, training accreditation by INP0 was nearly complete.
No violations of NRC requirements or deviations were identified in this area.
3.
(Closed) Temporary Instruction 2500/16 Seismic Interaction Between In-Cora Flux Mapping And Seal Table.
This temporary instruction (TI) provides guidance for inspecting PWR facilities with movable in-core flux mapping systems to determine whether PWR licensees have performed a system review as a result of IE Information Notice 85-45, " Potential Seismic Interaction Involving The Novable In-Core Flux Mapping System Used in Westinghouse Designed Plants".
Westinghouse designed plants should be specifically inspected to determine if licensees have reviewed the structural adequacy of their system as recommended by Westinghouse in their generic letter.
As a matter of background, on June 22, 1984, Carolina Power and Light Co.
(CP&L) informed the NRC of a potential 10 CFR 21 report concerning a design deficiency at their Shearon Harris facility, which is under construction.
The report defined a potential seismic interaction between the guide tubes at the seal table and the portions of the flux mapping equipment located above the seal table.
The basic problem can be characterized as the classical " class II system over a class I system" situation.
CP&L requested Westinghouse to perform a structural integrity analysis of the in-core flux mapping system, and Westinghouse reported to CP&L in December 1984 with recommendations to correct the problem.
On February 12, 1985, CP&L forwarded an interim report to the NRC indicating that they determined this subject to be reportable under the provisions l
of 10 CFR 50.55(e) and 10 CFR 21 and that a final report would be j
forwarded to the NRC following structural modifications to the system.
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As part of the NRC review of the 10 CFR 21 report, Westinghouse was contacted to determine the generic implications of the problem.
A Westinghouse safety review found the problem to be a potential significant deficiency and a potential unreviewed safety question.
Later, Westinghouse recommended that each utility investigate the adequacy of the restraints provided for the flux mapping system under seismic loadings.
The inspector reviewed the licensee's actions and investigations regarding the adequacy of the seismic restraints for the flux mapping-system and the seal table.
The flux mapping system's moveable frame was restored to its design configuration by adding clips and pins to fix its
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position during operation.
The recommended corrective action was to install two brackets to seismically anchor the movable frame assembly.
These brackets were installed in June 25, 1985 via RDC 84-118, DCP-1, Seal Table Modification.
The licensee conducted their review via Operational Assessment Review (OAR)-85-75, which concluded, based on the above modifications, that the seal table and compc,nents would not present a safety problem as regards a small break LOCA following a seismic event.
The inspector reviewed and inspected the flux mapping system's movable frame assembly and concurred with the licensee's assessment that its frame assembly would not present a safety impact problem upon the seal table.
The new brackets, clips and pins appear to adequately fix its position during power operations.
Based upon this inspector's revicw and inspection of the flux mapping system, Temporary Instruction 2500/16 is closed.
No violations of NRC requirements or deviations were identified.
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Temporary Instruction 2515/91 inspection Followup To Generic Letter 83-28, Item 4.1 (Vendor-Related Modifications For Reactor Trip Breakers The purpose of this Temporary Instruction (TI) is to provide guidance for performing inspection followup to the licensee's response to Generic Letter (GL) 83-28, " Required Actions Based on Generic Implementation of Salem ATWS Events." It is intended to verify the satisfactory completion of the action required in Item 4.1 of GL 83-28 which concerns
vendor-related modifications for reactor trip breakers.
Inspection Report 85-22 performed during the week of July 22-26, 1985, reviewed the requirement of GL 83-28, Item 4.1.
The inspection report reviewed the maintenance requests (MR) for the reactor trip breakers which installed the shunt trip devices as recommended by the latest vendor-issued bulletin.
MR 85-2100 installed the reactor trip breaker shunt trip device so that it would activate from the automatic reactor protection system.
The inspection report adds that an interoffice i
communication dated July 8, 1985 (D. Swan to M. Hoffman), provided formal l
notification that this modification had been completed.
Based upon Inspection Report 85-22 review and results, TI 2515/91 is
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closed.
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No violations of NRC requirements or deviations were identified.
5.
Follow-up on Previous Inspection Findings a.
(Closed) IE Bulletin 84-03 Refueling Cavity Water Seal IE Bulletin 84-03 considers the potential for and consequences of a refueling cavity water seal failure.
The licensee was requested to evaluate (1) gross seal failure; (2) maximum leak rate due to failure of active components such as inflated seals; (3) makeup capacity; (4) time to cladding damage without operator action; (5)
potential effects on stored fuel and fuel in transfer; and (6)
emergency operating procedures.
As a result of this request, the licensee performed an analysis of the potential for and consequences of a refueling cavity water seal failure.
Their response was presented in a letter to NRC dated November 30, 1984, written by then Vice President Bart Withers to Regional Administrator John Martin.
The significant conclusions of the analysis were as follows:
(1) Due to the differences between the Trojan and Haddam Neck Refueling cavity water seal design, material properties, and type of installation, partial or complete failure of the Trojan seal is a very unlikely event.
(2) A failure of the refueling cavity water seal would not result in the uncovery of fuel assemblies in the reactor core, refueling cavity, spent fuel pool, fuel transfer canal, W fuel transfer tube.
The only immediate operator actions required to prevent fuel overheating and/or cladding camage would be to lower.any assemblies in the refueling mast or transfer canal upender to below the vessel flange level.
(3) The only possibility for fuel uncovery as a result of a
refueling cavity water seal failure would be if the rod control
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cluster (RCC) change fixture, located in the refueling cavity, were in use.
In this case, 3.5 inches of active fuel length in the two fuel assemblies in the fixture could be uncovered.
This would not result in overheating of the assemblies to the point where clad damage could result.
High radiation levels inside the Containment from the exposed fuel in the fixture l
would make seal repair operations more difficult.
(4) No plant emergency operating procedures currently exist which specifically address a failure of the refueling cavity water seal, although existing procedures do provide a framework within which correct operator actions would be taken.
As a result of the above findings, the licensee performed the following actions:
Fuel Handling Procedure FHP-13 has been revised to describe actions to be taken by Plant personnel to detect and mitigate
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the consequences of a loss of refueling cavity water.
In addition, FHP-13 contains instructions on cooling system lineups following a loss of refueling cavity water.
Prevention of excessive radiation doses during recovery operations will be accomplished under existing radiation protection and emergency procedures.
The RCC change fixture was modified so that fuel assemblies in the fixture will not be exposed above the water surface in the event of a complete failure of the refueling cavity water seal.
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The modification was completed during the 1987 refueling outage via RDC 85-032, DCP-1 on April-25, 1987.
Based on the above conclusions and actions; it appears that the licensee has performed a satisfactory analysis of the potential for and consequences of a refueling cavity water seal. failure.
Based on
the differences in seal design, material properties and type.of i
installation, the licensee reasons that it is highly unlikely that the Trojan refueling cavity water seal.would fail as did the seal at Haddam Neck.
Information Bulletin 84-03 is closed.
No violations of NRC requirements or deviations were identified.
4.
Exit Interview The inspector met with the licens(a representatives denoted in paragraph l'on July 17, 1987, and summarized the scope and findings of the inspection activities.
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