IR 05000440/1985088

Safety Insp Repts 50-440/85-88 & 50-441/85-27 on 851202-06 & 09-13.No Violation,Deviation or Significant Issue Noted. Major Areas Inspected:Action on Previous Insp Findings & Evaluation of Action Re IE Bulletin & Confirmatory Items

ML20138R234
Person / Time
Site:	Perry
Issue date:	12/26/1985
From:	Knop R, Scheibelhut C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20138R228	List: IR 05000440/1985088 ML20138R227
References
50-440-85-88, 50-441-85-27, IEB-85-001, IEB-85-002, IEB-85-1, IEB-85-2, NUDOCS 8512310075
Download: ML20138R234 (9)

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U. S. NUCLEAR REGULATORY COMMISSION e

REGION III

Report N /85088 (DRP); 50-441/85027 (DRP)

Docket N ; 50-441 License N CPPR-148; CPPR-149 Licensee: Cleveland Electric Illuminating Company Post Office Box 5000 Cleveland, OH 44101 Facility Name: Perry Nuclear Power Plant, Units 1 and 2 Inspection at: Perry Site, Perry, OH Inspection Conducted: December 2-6 and December 9-13, 1985 Inspector:

apH.See%i elhut wf /3 7/6'/

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Approved by:

RFukJA fe R. C. Knop, Chief / 2 8[

Reactor Projects, Section 1A Date !

Inspection Summary Inspection on December 2-6 and December F -13,1985 (Repcrts No. 50-440/85088 (DRP); 50-441/85027 (DRP))

Areas Inspected: Routine safety inspection by a Regional Inspector of licen-see actions on previous inspection findings and evaluation of licensee action with regard to IE Bulletins and confirmatory items called for by a licensing branch site review. The inspection involved a total of 72 inspector-hours onsite by one NRC inspector and includes 0 inspector-hours during off-shift s Results: No violations, deviations, or safety significant issues were identi-fle , ,

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Details Persons Contacted Cleveland Electric Illuminating Company

• M. D. Lyster, Manager, Ferry Plant Operations Department (PPOD)

• P. R. Stead, Manager, Nuclear Engineering Department (NED)

• C. M. Shuster, Manager, Nuclear Quality Assurance Department (NQAD)

• R. J. Tadych, General Supervisor, PPOD

• B. D. Walrath, General Supervising Engineer, NQAD

• V. J. Concel, Operations Engineer, Perry Plant Technical Department (PPTD)

• B. B. Liddell, Operations Engineer, PPTD

• P.,A. Russ, Compliance Engineer, PPTD

• T. X. Heatherly, Compliance Engineer, PPTD

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• W. J. Colvin, Operations Engineer, PPOD

) *B. S. Perrell, Licensing Engineer, NED

• Denotes attendance at the December 13 exit meetin The inspector also contacted other licensee and contractor personnel during the course of the inspectio . Licensee Actions on Previously Identified Items (92701, 92702) (Closed) Violation (440/84015-02 (DRP)): " Instrument air system cleanliness not verified during testing". In Inspection Report 50-440/85066 (DRP), 50-441/85023 (DRP), the inspector was satisfied j with the corrective actions taken by the licensee. However, the NRC Office of Nuclear Reactor Regulation (NRR) had questions concerning an amendment to the PSAR that changed the maximum allowable size of particulate contamination in the air from three micrometers to 40 micrometers. Accordingly, the item was lef t open until the questions were resolve Supplement 7 to the Safety Evaluation Report for the Perry Nuclear Power Plant (NUREG 0887) accepted the change in the maximum allow-able size of the particulate contamination. Accordingly, this item is considered close (Closed) Open Item (440/85010-03 (DRP)): "Preoperational tests to satisfy surveillance requirements". The inspector reviewed Perry Administrative Procedure (PAP)-1105, " Surveillance Test Control",

Revision 1, dated May 1, 1985. A major purpose of the revision was to provide a controlled means of evaluating to what extent the l results of the preoperational testing program could be used to satisfy surveillance program requirements. The inspector concluded that the program appeared adequate. However, the inspector wanted ,

to review a sample of the preoperation test sections approved for surveillance procedure credit during a future inspection to verify proper implementation of the progra . - . , _ - . - _ _ _ _ - - ._ . - _ _ _ . - _ _ - . . - _ _ . _ - . . _ - _ _ -_ -. -

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The inspector obtained a current listing of all SVIs (Surveillance Instructions) performed for credit. The listing indicated whether the particular surveillance was performed in accordance with the pertinent SV1 or performed during preoperational testing. A review

, of the list showed a total of 326 SVIs performed for credit. Of these,186 were performed in accordance with the SVIs. A total of 101 Type B and Type C local leak rate tests depended on preopera-tional testing results to fulfill the surveillance requirement Nine other miscellaneous tests relied on preoperational test re-sults. A total of 30 safety / relief valve tests depended on the manufacturer's certified tests of set pressure and reset pressur Further investigation showed that there are a total of 78 Class 1, 2, and 3 safety / relief valves. Of the 19 Class 1 valves, all were retested in 1985. Of the 61 Class 2 and 3 valves, 29 were retested in 1985. The balance of 30 valves will rely on the manufacturer's test result The inspector compared a sampling of the preoperational test pro-cedures with the surveillance instructions for those instances where preoperational test results were used for surveillance credit. In all cases, the comparison showed that the preoperational test pro-cedures were equivalent to the surveillance instructions. In the case of the local leak rate testing, the preoperational test pro-cedures called for the use of the pertinent SVI to perform the tes In most instcnces, the operability of a Class 2 or 3 system does not depend on the operability of a safety / relief valve installed in the cystem. Therefore, the use of manufacturer's test data was con-sidered appropriat In an interview with SVI test personnel, it was found that preopera-tional test data will not be used for required surveillance of active valves and pump Based on the above, the inspector concludes that PAP-1105 is being properly implemente (Closed) Open Item (440/85010-05 (DRP)): " Security Plan In-structions approved by PORC being written to an unapproved security plan". During attendance at a Plant Operating Review Committee (PORC) meeting, the inspector noted that three security implementing procedures (instructions) were being recommended for approval even though the security plan had not received the required NRC Office of Nuclear Reactor Regulation (NRR) approval. The inspector was con-cerned that the plan finally approved may differ from the submitted plan thereby possibly invalidating the implementing procedures. The inspector was also concerned that the situation may occur with l respect to other plans requiring NRR approva When NRR approval has been given for a plan, the licensee reviews the plan and the implementing procedures to assure that they are in agreemen If they are not, the procedures are revised to bring them into agreement with the plan.

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s Inspection Report 50-440/85065 delineates the results of an inspection conducted by Region Ill inspectors of the agreement between the NRR approved security plan and the implementing pro-cedures. They were found to be in agreemen The Emergency Plan and the Process Control Program have received NRR approval and review of the approved plans and implementing pro-cedures to assure agreement has taken place. These reviews are contained in memorandum, J. D. Anderson to S. F. Kensicki, dated November 11, 1985. The remaining plan requiring NRR approval, the Offsite Dose Calculation Manual (ODCM) has not been submitted to NRR for approval. In memorandum, S. J. Woj ton to S. F. Kensicki, dated December 9, 1985, a commitment was made to review all applicable procedures / instructions to verify agreement with the approved ODF The inspector reviewed the Inspection Report and memoranda and concludes that the implementing procedures required by NRR approved plans are in agreement and that the implementing procedures for the ODCM will be in agreement with the pla (closed) Open Item (440/85022-48 (DRP)): "SRIA and RHR as-built discrepancies". During a walkdown of systems to compare the plant as-built condition with site controlled P& ids, the inspector noted several discrepancies in the Safety Related Instrument Air (SRIA)

l system and in the Residual Heat Removal (RHR) system. These dis-crepancies included missing pipe caps, missing permanent valve identification tags, and missing relief valve discharge pipe stubs.

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' The licensee determined that the relief valve discharge pipe stubs were never intended to be installed and were not shown on the piping drawings. They were shown on the P&ID as a convention, since the

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I valves discharge to the atmosphere. The operations section had a program in progress to identify all plant system valves shown on the P& ids and assure that they were correctly tagged. An independent verification program was then implemented to assure that all valves were properly tagged. The verification process is essentially

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~ complete tem for all systems and is complete for safety-related sys- l The licensee issued Work Authorization (WA)-NTS-85-1021 to walkdown the three RHR loops from suction to discharge and install any missing pipe caps. This WA was also entered into the Master Deficiency List (MDL). The missing pipe caps were installed and the installation witnessed by quality control inspectors. In the SRIA system, the pipe caps had been removed to perform a preoperational tes i They had not been reinstalled because the system had to be !

retested.

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Their removal had been entered into the MDL. They have ;

' subsequently quality controlbeen reinstalled with the installation witnessed by inspector '

The inspector made a plant inspection and by sampling areas of the RHR and SRIA systems, found all pipe caps and permanent valve identification tags installed. Based on the verification programs in progress and the evidence, the inspector considers the item closed.

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e. (Closed) Open Item (440/85033-17 (DRP)): " Battery maintenance practices prior to service and performance testing in confitet with IEEE-450-1975". The inspector found that the licensee's procedures for performing service and performance tests on the safety-related batteries required the performance of an equalizing charge and routine preventive maintenance prior to the tests. IEEE Standard 450-1975 (and -1980), " Recommended Practice for Maintenance, Testing and Replacement of Large Lead Storage Batteries for Generating Stations and Substations", specifically prohibits the performance of an equalizing charge and routine preventive maintenance prior to running performance and service test The licensee revised the following surveillance instructions:

SVI-R42-T5211, " Service Test of Battery Capacity - Division 1".

SVI-R42-T5212, " Service Test of Battery Capacity - Division 2".

SVI-E22-T5213, " Service Test of Battery Capacity - Division 3".

SVI-R42-T5215, " Performance Test of Battery Capacity - Division 1".

SVI-R42-T5216, " Performance Test of Battery Capacity - Division 2".

SVI-E22-T5217, " Performance Test of Battery Capacity - Division 3".

The revised SVIs delete the requirement for performance of the equalizing charge and routine preventive maintenance prior to running the test The inspector reviewed the revised SVIs listed above and found they met the service and performance tests requirements found in IEEE-450-1975 and -198 (Closed) Unresolved Item (440/85046-01 (DRP)): "A. Variance between ANSI and PNPP definition of Plant Manager. B. Four examples of qualifications discrepancies for operations staff."

During a detailed review of the operational staffing plan delineated in the FSAR, the inspector noted two areas of discrepancy between the plan and the requirements of ANSI Standard N18.1-1971,

" Selection and Training of Nuclear Power Plant Personnel". The structural location of the Radiation Protection and Techni-cal Sections in the organization structure resulted in opera-tional autonomy of these sections from the individual classi-fied in the FSAR as " Plant Manager". The qualifications of four individuals designated in the FSAR and staffing plan appeared to be at variance with the require-ments of N18.1-1971 and Regulatory Guide 1.8, Revision 1,

" Personnel Selection and Training".

The licensee took the following actions to resolve these discrepancies: The licensee amended (Amendment 21) the FSAR to update and clarify the plant organization. It also furnished two oper-ating procedures (POP-0105 and PAP-0104) which showed that the Manager of the Perry Plant Operations Department has the

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authority to fulfill the responsibilities of " Plant Manager" with the authority at any time to direct any and all resources or other elements within the Nuclear Operations Division for the safe operation of the plan The licensee replaced one of the people that did not have appropriate qualifications with one that did have the qualifi-cations. The licensee also provided additional information on the other three people to show that they were qualified. This information was submitted to NRR along with the amendment indicated in "A" abov In Supplement 7 to the Perry Plant Safety Evaluation Report (NUREG 0887), the staff accepted Amendment 21 to the FSAR and the supple-mental information contained in the two operations procedures and staff qualifications. Therefore, this item is close g. (Closed) Open Item (440/85053-04 (DRS)): " Awaiting NRR action on licensee planned submittal to delete startup testing requirements for Main Steam Isolation Valve Leakage Control System". During the test program review, the inspector noted that there was no existing or planned startup test procedure for the Main Steam Isolation Valve

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Leakage Control System (MSIVLCS) to correspor.i to the testing com-mitments contained in the Final Safety Analysis Report (FSAR).

Since the MSIVLCS was tested during the preoperation testing program (F32 System Preoperational Test 1E32-P001, Rev. 1, results approved April 29, 1985) the licensee amended the FSAR (Amendment 21) to delete the startup testing requirement. Surveillance Instruction (SVI)-E32-T0397, "MSIV Leakage Control System Function Test", was written and approved to provide periodic testing of the MSIVLCS during the life of the plant to assure operabilit '

Supplement 7 to the Safety Evaluation Report for the Perry Nuclear Power Plant (NUREG 0887) accepted the deletion of the MSIVLCS test from the startup test program because the test objectives were completed during the preoperational test program. This' acceptance was predicated on routine performance of surveillance testing of the syste The inspector determined that the preoperational test did demon-strate operability of the MSIVLCS and a review of the SVI showed that performance of the SVI would continue to demonstrate operability in a timely manne (Closed) Open Item (440/85056-02 (DRP)): "Nonconformance Tracking for Augmented Quality Systems". During a review of the licensee's corrective action system documentation, the inspector noted three Deficiency Reports (DRs) addressing fire protection system hard-ware. The licensee's quality assurance program indicates that the fire protection system, although classed as a nonsafety-related system, would be covered under an augmented quality assurance pro-gram commensurate to the system's importance to nuclear safety of the plant. Under the augmented program it appeared that those fire

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protection system deficiencies should have been resolved through the Nonconf ormance Report (NR) system instead of the DR syste The licensee agreed that NRs should have been used instead of DRs during the installation of the fire protection system. To ensure the use of NRs for handling fire protection deficiencies during construction of Unit 2, Plant Administration Procedure (PA)-0202,

" Fire Protection Quality Assurance", has been revised. The revision clearly requires the use of NRs to handle construction defici-encies. During operations Perry Operations Procedure (POP)-1501,

" Identification and Control of Deficient Items", requires the use of the NR system to cover deficiencies in all systems. The use of DRs is not recognized. To determine if the deficiencies identified by DRs during the construction of the Unit 1 fire protection system were reportable to the NRC under 10 CFR 50.55(e) criteria, the licensee reviewed all of the DRs issued against fire protection systems, structures, and components. Of the 158 DRs reviewed, none were determined to be reportabl The inspector reviewed PA-0202 and determined that the use of NRs is required to handle construction deficiencies during the installation 2 of the fire protection system in Unit 2. The inspector also re-viewed P0P-1501 and determined that NRs will be used to cover all deficiencies discovered during the operation of Unit 1. A sample review of fire protection system DRs based on DR titles indicated that none was reportable under 10 CFR 50.55(e) criteria. The pri-mary difference between the DR and NR system of deficiency reporting and resolution was the lack of review for 10 CFR 50.55(e) applic-ability under the DR system. Since their review has been made and no reportable instances found, it is concluded that the instellation of the fire protection system in Unit 1 is satisfactory, (Closed) Open Item (440/85073-01 (DRP)): " Diesel Generator Control

! Circuit Wiring Changes Incomplete". An allegation had been made concerning diesel generator control circuit design deficiencie The allegation was closed (see Inspection Report 50-440/85073) hased on a review of an Engineering Change Notice (ECN) that corrected the design deficiencies. An item was opened to track the completion of the work described in the EC Work Order (W.O.) number 85000-9193 was written to perform the work. The work and circuit testing have been complete The inspector reviewed the quality control records associated with the circuit testing performed at the completion of the work and- l

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concludes that the circuit deficiencies have been correcte No violations or deviations were identifie . Licensing Review Follow-up Inspection Items (92701)

The Office of Nuclear Reactor Regulation, Division of Systems Integra-tion, Instrumentation and Controls Systems Branch (ICSB), has requested that Region III inspectors follow up on the applicant's activities in

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certain areas identified in a report transmitted on May 31, 1985, docu-menting a site review by ICSB personnel. One such item is verified ;

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(Closed) Open Item (440/85033-07 (DRP)): " Verify that heat tracing has been installed on the safety-related sensing lines located in the outdoor bunker that is adjacent to the condensate storage tank (ICSB trip report, Section 4.a)".

Work order numbers 85000-5362, -5536, -5537, -5538, -5820, -5822,and-6728 covered the installation of heat tracing on the sensing lines and sensors located in the outdoor bunker adjacent to the condensate storage tank. Work order numbers 85000-5871, -5873, -5882, -5890, and -5902 covered the insulation of the above heat tracin The inspector reviewed the above complet;d work orders and found them satisfactory. The inspector also visited the work site after the heat tracing was installed and before the bulk of the insulation was applied and found the heat tracing installed in accordance with the drawings and specification No violations or deviations were identifie . Evaluation of Licensee Action with Regard to IE Bulletins (92703)

For the IE Bulletins listed below, the inspector verified that the Bulletin was received by licensee management and reviewed for its applic-ability to the facility. If the Bulletin was applicable the inspector verified that the written response was within the time period stated in the Bulletin, that the written response included the information required to be recorded, that the written response included adequate corrective action commitments based on information presented in the Bulletin and the licensee's response, that the licensee's management forwarded copies of the written response to the appropriate on-site management representa-tives, that information discussed in the licensee's written response was accurate, and that corrective action taken by the licensee was as des-cribed in the written respons (Closed) IE Bulletin 85-01 (440/85001-BB, 441/85001-BB): " Steam Binding of Auxiliary Feedwater Pumps". The Bulletin referenced problems in operating Pressurized Water Reactors (PWRs) wherein high pressure hot feedwater leaked back through check valves into auxili-ary feedwater pumps. This rendered the pumps inoperable because of steam bindin While the Perry Plant utilizes a Boiling Water Reactor (BWR) that does not utilize auxiliary feedwater pumps, the Emergency Core Cooling Systems (ECCS) are arranged to pump cooling water into the reactor vessel through check valves. However, each ECCS system also contains a normally closed, motor operated gate valve in the in-jection line. Because of the proven integrity of the gate valves, and the fact that the backflow problem has not been found in operat-ing BWRs, the licensee considers the Bulletin to be not applicabl .

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The inspector agrees with the licensee's decisio (Closed) IE Bulletin 85-02 (440/85002-BB, 441/85002-BB): "Under-voltage trip attachments of Westinghouse DB-50 type reactor trip breakers". The Bulletin references recent reactor trip breaker reliability probleme with Westinghouse type DB-50 breakers. The breakers and similar breakers with mechanical undervoltge trip attachments have been the subject of earlier IE Bulletins (83-01, 83-04, and 83-08).

The licensee reiterated that there are no breakers installed any-where in the Perry Plant that have mechanical undervoltage trip attachments. Therefore the Bulletin is not applicable. The in-spector agrees with this assessmen 'No violations or deviations were identifie . Exit Interview This inspector met with licensee representatives (denoted in paragraph 1)

at the conclusion of the inspection on December 13, 1985. The inspector summarized the scope and findings of the inspection. The licensee acknowledged the inspector's findings. The licensee did not indicate that any of the information disclosed during the inspection could be considered prioprietary in natur