IR 05000353/1989011

Insp Rept 50-353/89-11 on 890213-0326.No Violations Noted. Major Areas Inspected:Work Activities,Security Implementation,Environ Qualification of Equipment & Proposed Tech Specs

ML20244B432
Person / Time
Site:	Limerick
Issue date:	04/07/1989
From:	Linville J, Williams J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20244B428	List: IR 05000353/1989011 ML20244B425
References
TASK-1.A.2.1, TASK-2.B.4, TASK-3.D.1.1, TASK-TM 50-353-89-11, IEB-83-05, IEB-83-5, IEB-87-002, IEB-87-2, IEB-88-008, IEB-88-8, IEC-77-04, IEC-77-4, NUDOCS 8904190197
Download: ML20244B432 (20)

Text

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Philadelphia ElectricLCompany 3-

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-V. S. NUCLEAR REGULATORY COMMISSION-2 REGION I Report No. 89-11 Docket No.'50-353 License No. CPPR-107- Category A/B

' Licensee: Philadelphia Electric Company-Correspondence Control Desk P.O. Box'7520 Philadelphia, PA 19101-Facility Name: Limerick Generating Station, Unit 2 LInspection Conducted: January. 9 to February 12, 1989 Inspectors: R. A. Gramm, Senior Resident Inspector, Unit 2 R. L. Fuhrmeister, Resident Inspector, Unit 2 G. C. Smith, Safeguards Specialist, SS, NMSB, DRSS Reviewed by: . [r _

F ff 'l J.' H. Willi

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Project Engineer Dats

Reactor Pro s Section12A Approved _by: m[, -([7/fG J s. C. LinvilTe, . Chief / / Dgte /

ctor Projects Sectio a Inspection Summary: Report for Inspection Conducted February 13' to March 26,

'1989 (Report No. 50-353/88-31)

Areas Inspected: Routine inspection by resident and regional-inspectors o work activities, procedures and records relative to startup testing, security implementation, environmental qualification of equipment and proposed Technical Specifications. The inspectors reviewed licensee action on previously identified items and conducted plant inspection tours. The inspection involved 201 hours0.00233 days <br />0.0558 hours <br />3.323413e-4 weeks <br />7.64805e-5 months <br /> by the' inspector Results: Preoperational tests were conducted in a controlled manner and result reviews were performed in accordance with applicable procedure Implementation of security on Unit 2 was found to be well planned and execute Two unresolved items involved Technical 5 specification - FSAR inconsistencies and HPCI-turbine operabilit )

8904190197 890410 PDR ADOCK 05000353 i G PDC

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-Philadelphia Electric Company 4

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TABLE OF CONTENTS

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.PageL 1.0 LSummary . ............................

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5L 2.0 Plant. Tours'. . . . . . . . . . . . . . . . . . . . . . . . . . . 5-i 3.0 Licensr& Action on Previously Identified Items. . . . . . . . . . 6 !

i 4.0. Licensee Action on Bulletins and' Circulars. . . . . . . . . . . . 6

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i 5.0 Three' Mile Island Action Plan Items . . . . . . . . . . . . . . . 8 6.0 'Preoperational Security Inspection. . . . . . . . . . . . . . . . 9 t .!

7.0~ Technical Specification Review. . . . . . . . . . . . . . . . . . 10 8.0' Reactor Pressure Vessel Nozzles . . . . . . . . . . . . . . . . . 12

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9.0' Startup Test Activities . . . . . . . . . . . . . . . . . . . . . 13

'10.0 Appendix.R Program. . . . .................... 14 H 11.0 Diesel Generator - Protective . Trips . . . . . . . . . . . . . . . . 14 12.0 Environmental Qualification Program . .............. 14 13.0 Construction Inspection Program . ................ 15 14.0 High Pressure Coolant Injection System. . ............ 15 l

15.0 Meetings with Licensee. . . . . . ................ 15

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16.0 Assurance of Qualit ...................... 16 Appendix.I Reactor Core Isolation Cooling System-(RCIC) Technical ,

Specification Documents Reviewed I Appendix II Technical Specification changes from Unit I Documents Review I

Appendix III Reactor Protection System (RPS) Technical Specification Documents Reviewed ,

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Philadelphia Electric Company 5 DETAILS

~1.0 . Summary

' Ten NRC open items were reviewed and eight were closed (sections'3, 4~and 5). Implementation of full security for the Unit 2 protected are'a was monitored by a regional specialist (section 6). A review of_the draft

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Technical Specifications was commenced,' verifying conformance to the

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- as-built plant and FSAR requirements (section 7). Conduct.of preo- i perational tests and test results review was inspected (section 9). A '

Region-I Temporary Instruction was closed out (section 11). Problems with ultrasonic examinations of reactor pressure vessel nozzles, fire protection program issues and. equipment qualification were reviewed (sections 8, 10 and 12 respectively)~.

2.0 Plant Inspection Tours (37301, 41301, 51063, 51065, 73053)

The inspector observed in progress. work activities, completed work, and plant status'in several areas during inspection tours. Work was examined

.for' defects and. compliance with regulatory and licensee requirement Particular note was taken of the presence of quality control inspectors and quality control evidence such as inspection records, material ide' notification, nonconforming material identification, housekeeping and equipment preservation. The inspector interviewed craft supervision, craft and quality control personnel in the work areas. Observations are noted below:

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The inspector was informed that incomplete weld penetration and lack of fusion had been found in the longitudinal seam weld on a section of tube steel used for a pipe suppor PECo Finding Report 2P-793 and deportability evaluation L2-89-07 were reviewed. Walkdown inspections were performed by the licensee to observe tube steel sections used on the Emergency and Residual Heat Removal Service Water Systems. The licensee investigation and analysis of the i weld deficiency is ongoin l

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The inspector reviewed the circumstances of a missed cable termination with PEco field engineers. The cable was associated with a Unit 2 tie in to Unit 1 on the Standby Gas Treatment Syste The original instructions to land the cable were issued via Plant Change Notice 8566N and drawing E-2572, the connection list, was revised. Bechtel did not issue a work package since the work area was under PECo jurisdictio PECo has since reviewed all similar cable tie in points and determined this was an isolated case where the cable was not properly lande The inspector observed lapping operations on the B Main Steam Isolation Valve (MSIV). Quality Control personnel were present

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along with Bechtel Field Engineering. The work was properly l controlle h ,,

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The inspector monitored the conduct of General Employee Training Annual Requalification. It was noted that one of the answers in the computer data base was.not consistent with the information provided in,the class. This was pointed out to and verified by the instructor The licensee' stated that question will be reviewed and regraded,.and the automated grading data base will be correcte .0 Licensee- Action on Previously Identified Items (92701, 90711) (Closed) Construction Deficiency Report (84-00-03): Defective relay sockets. This issue was originally identified by General Electric (GE) in Service Information Letter (SIL) 384. The issue was sub-sequently, described in IE Information Notice 82-48. PEco Field Engineering performed an inspection of all the relay sockets not previously replaced or inspected by GE. All those sockets which did

.not pass the test recommended in SIL 384 were replaced (48 of 704).

This item is close (Closed) Construction Deficiency ~ Report (84-00-16): Defective relay connection plugs. The inspector reviewed the following documentation:

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PECo Letter, Kemper to Murley, August 30, 1984 (SDR #153)

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Non-conformance Report'(NCR) 10357 Testing and Laboratories Division Procedure TL-11-04031, Rev. 2, Calibration Testing of Protective Relays and/or Devices Some protective relays for 4kV circuit breakers were supplied with improperly sized connecting plugs. The equipment affected was identified on NCR 10357 and replacement plugs were obtained by Bechtel. The replacement plugs were installed by PECo field ,

engineers during . relay calibration per TL-11-04031. This item is j close i

! (Closed) Unresolved Item 87-11-05: Diesel engine protective trip from fire protection system flow switches. The licensee performed a review of non-Class IE to Class IE interfaces to ensure that the functional association was unique to the flow switch situatio No other cases of non-Class IE device failure were found which would adversely affect the Class IE equipment. The inspector reviewed:

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Drawing E-591, sheet 2

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Work package E29-0192

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Quality control inspection records 20B515-26-11, 20B516-26-11, 20B517-26-11 and 20B518-26-1 The Agastat timer relays were removed from the base socket to disable the trip signal input from the flow switches. This item is close j l

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Philadelphia Electric Company 7 4.0 Licensee Action on Bulletins and Circulars (92703) (Closed) Circular (77-04): Inadequate lock assemblies. The inspector reviewed records related to the installation inspection of security door The doors were properly inspected by the licensee and the deficiencies were correcte This item is close (Closed) Bulletin (83-05): Hayward-Tyler pumps. In October 1981, allegations were received from a journalist that Hayward Tyler Pump Company (HTPC) supplied defective ASME code class and nuclear safety related pumps to plants throughout the world. Subsequent NRC inspection revealed that HTPC had not effectively implemented its Quality Assurance Progra Licensees were requested to supply NRC with information regarding hydrostatic test problems and corrective actions, inservice test requirements and to perform a performance and endurance test for any HTPC supplied equipment. The Unit 2 Safeguard Fill System pumps were performance and endurance tested under Technical Test Procedure (TT) 1.11, which was satisfactorily completed for both pumps. This item is closed, (Closed) Bulletin (87-02): Fastener testing to determine conformance with applicable material specifications. This item had been previously reviewed in NRC inspection reports 87-16 and 88-0 The inspector reviewed the following documents:

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PECo letter dated February 17, 1988 from Gallagher to Russell (NRC)

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PECo letter dated July 26, 1988 from Gallagher to Rossi (NRC)

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Nonconformance Report 12853

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NRC Temporary Instruction 2500/25

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Bechtel QC Instruction R-1.00, " Receiving Inspection"

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Bechtel Construction Procedures, CP-F-2, " Receipt Inspection, Storage and Withdrawal of Materials and Equipment" and CP-0-1,

" Processing of Nonconformance Reports"

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NRC Temporary Instruction TI 2500/26 PECo has responded to both the Bulletin and Supplements 1 and The list of vendors that have supplied safety and non-safety related fasteners to the site has been provided to the NR Construction Procedure (CP) F-2 requires that sampling receipt inspections be performed to verify appropriate fastener sizes, ASTM designations and lengths. QC instruction R-1.00 requires that procurement documentation be reviewed in conjunction with performing the receipt inspection. The QC instruction further requires that material configuration and workmanship be verified on a random basis with the applicable purchase specification requirement The controls over material segregation and control pending QC l inspection were found to have been adequately addressed in the site procedure W

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Philadelphia Electric Company- 8

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.A new pECo QC inspection procedure has been generated for receipt

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inspection of threaded fasteners. .The instruction provides for  !

verification' of: . proper markings, proper size, proper material

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certifications and proper threads. Selected samples will be

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i hold until the test results are reviewe I The out of specification test results for the randomly selected fastener sample were evaluated by engineering and additional _ testing was performe Two heat codes for one inch, and one and a quarter

. inch' diameter stud material' were' identified as unsatisfactory for use on ASME code systems and were-scrapped from the warehous None were found to have been. issued to the field. The two heat codes in question were removed from the site QC heat log to preclude QC acceptance of the items in field installation k Based upon the previously witnessed collection and testing o fasteners and review of site receipt inspection processes, this item is closed, (Closed) Bulletin (88-08): Thermal stresses in piping connected to reactor coolant systems. The licensee reviewed' the Limerick piping design to' identify the potential for cold water leaks into the

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reactor coolant system during normal operation. The licensee review found that no sections of the reactor coolant system piping would be subject to thermal cycling fatigue from cold water leaks. This item is close .0 Three Mile Island Action Plan Items (25401)

" (Closed) I.A.2.1: Immediate upgrading of reactor operator and senior reactor operator training and qualifications. The inspector reviewed the following:

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Limerick SER section 113.2. Limerick procedure A-50

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FSAR section 1.13 The Limerick programs requires the appropriate educational and experience levels for licensed operator (Open) II.B.4: Training for mitigating core damage. The licensee had committed to provide some level of training to personnel from the instrumentation and controls, health physics and chemistry organizations. The inspector asked the licensee to verify that the appropriate training was given. This item remains ope )

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Phila'delphia Electric Company 9

(0 pen) III.D.1.1: Primary coolant outside containment. The inspector reviewed the following documents:

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NUREG 0737

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FSAR section 6.2.8 and 1.13

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SER section 15. PECo letters dated 9/12/84 and 10/12/84 from Kemper to Schwencer(NRC)

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PECo procedures

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ST-1-047-700-2, " Scram Discharge Volume Contamination Piping Inspection"

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ST-1-049-701-2, "RCIC Pump Contaminated Piping Inspection"

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ST-1-049-702-2, "RCIC Turbine Contaminated Piping Inspection"-  !

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ST-1-030-700-2, " Post Accident Sampling System Liquid Sample Loops Contaminated Piping Inspection"

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ST-1-030-701-2, " PASS and Containment Atmospheric Control

Sample Loops' Contaminated Piping Inspection"

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ST-1-052-701-2, "A Core Spray Loop Contaminated Piping Inspection"

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ST-1-052-702-2, "8 Core Spray Loop Contaminated Piping Inspection"

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ST-1-052-705-2, " Safeguard Piping Fill System Contaminated 3 Piping Inspection"  ;

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ST-1-051-701-2, "A RHR Loop Contaminated Piping Inspection"  !

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ST-1-051-702-2, 8 RHR Loop Contaminated Piping Inspection"

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ST-1-051-703-2, "C RHR Loop Contaminated Piping Inspection"

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ST-1-051-704-2, "D RHR Loop Contaminated Piping Inspection" ,

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ST-1-055-701-2, "HPCI Pump Contaminated Piping Inspection"  !

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ST-1-055-7022, "HPCI Turbine Contaminated Piping Inspection"

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ST-1-058-701-2, "A Post LOCA Recombiner Contaminated Piping Inspection"

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ST-1-058-702-2, "B Post LOCA Recombiner Contaminated Piping Inspection" The specified leak check surveillance interval for the surveillance tests was 18 months which is consistent with the FSAR. Procedures have been generated for the appropriate system leak tests. For Unit ,

1, eight surveillance tests were performed prior to fuel load and eight surveillance tests were completed post-fuel loa Pending licensee performance of and evaluation of the surveillance tests for Unit 2, this item remains ope .0 Preoperational Security Inspections (81052, 581700)

On February 25, 1989, a region-based security specialist reviewed the licensee's program to search and secure the Unit 2 buildings and yard f.

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areas prior to establishment of security controls for those area The licensee searched and secured all Unit 2 areas and then incorporated Unit 2 into the existing Unit 1 protected area. The inspector reviewed the

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h Philadelphia Electric Compan licensee's search plan, attended a pre-search briefing.for the search-teams and observed portions of-the searches'being conducted. The inspector's review disclosed that the' licensee had developed a comprehensive search plan, provided a thorough briefing to all search teams and was conducting. thorough searches of the areas observed.-

The search teams consisted of security personnel, instrument and control'

personnel, the required craft-disciplines to ensure all areas were accessible, dog handlers with explosive and drug search dogs and management. personne The inspector determined that the search tea composition was appropriate to ensure that all areas were accessible.and to ensure a thorough searc ~ Technical Specification Review (94300, 71301) 'The inspector performed a review of the first draft Technical Specifications (TS) issued June 3,1988. The inspection purpose was to determine the compatibility between the TS and the as-built plant configuration and operating characteristics.

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The inspection scope included the Reactor Protection System and the Reactor Core' Isolation Cooling systems. These systems were selected

' based on the'results of the Limerick Probabilistic Risk Assessment

' evaluation and the systems reviewed during the Unit 1 TS inspection (50-352/84-52). The as-built configuration and operatin characteristics for the selected systems were ascertained through a review of numerous documents including: the Final Safety Analysis Report (FSAR), Safety Evaluation Report (SER), preoperational test procedures, operating plant procedures, surveillance test-procedures, Piping and Instrument Diagrams (P& ids), electrical drawings and. logic diagrams. .The.as-built plant was examined to assure conformance with the appropriate design criteria. The

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surveillance test procedures were reviewed to assure that the appropriate TS surveillance requirements were properly incorporate The inspection attributes were as follows:

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Verify compatibility of the draft TS with the as-built plant configuratio Verify consistency of the draft TS with respect to the plant procedures and design document Verify adequacy of surveillance tests relative to TS .

surveillance requirement I The inspection was performed during reviewu of plant hardware and I software as follows:

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'1 System walkdowns were performed to verify that as-built !

configurations were consistent with the TS, FSAR, SER and i appropriate design document l i

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Surveillance test procedures were reviewed relative to TS requirements for test scope and frequenc ,

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Operating plant procedures were reviewed to verify consistency '

with the T The Reactor Core Isolation Cooling (RCIC) system was reviewed with respect to the criteria identified in section 7a of this repor ]

The proposed.TS sections identified in Appendix I were compared to '

the other licensing and design documents also listed in Appendix I for compatibilit In addition, selected plant components were examined to ensure consistency with the appropriate document The inspector reviewed the following items: isolation actuation instrumentation setpoints, isolation actuation instrumentation surveillance frequencies, isolation system instrumentation response time, system actuation instrumentation setpoints, system actuation instrumentation surveillance frequencies, system flow verification, system functional testing and isolation valve testing isolation stroke time Based upon the inspector's review, the following questions were discussed with licensee personnel:

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In several cases the surveillance test procedures were not consistent with the Limerick Inservice Test (IST) program. The IST program requires stroke time for opening and closing directions where the surveillance. procedures would time the valve in one direction only. The inspector was informed that these inconsistencies were known and that a program was underway to have compatible surveillance test procedures prior to system turnover to the plant staf Several draft surveillance procedures were reviewed and found to contain the proper valve test requirement Startup Finding Report 2-50A-056 was issued to request a Licensing Document Change Notice to correct a typographical l error in FSAR section 5.4.6.2.4.2.C where reference to valve F022 should be F00 This is a minute change with little significanc The licensee satisfactorily demonstrated that the differences between the current surveillance tests and the licensing documents l are being properly controlled. Based upon the inspector's review, the compatibility of the as-built plant, TSs, licensing documents and operating procedures was confirme Changes from Unit 1 TSs were reviewed to verify consistency with the

, Unit 2 design and as-built plant. Associated documentation is l

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Philadelphia' Electric Company 12

' listed in Appendix II. During comparison of TS curves for. Maximum Average Planar Linear Heat Generation Rate-(MAPLHGR) versus" Average Planar Exposure (APE).to the corresponding FSAR tables, it was note that the MAPLHGR/ APE curve shown on TS figure 3.2.1-4, for BP8CRB094 fuel assemblies shows a limit of 7.8 KW/ foot at an exposure of 45,000 megawatt-days / ton (MWD /t). FSAR table 6.3-4 shows a MAPLHGR/ APE. limit of 7.9 KW/ foot at 45,000 MWD /t. Site personnel

.have referred t!.is issue to the licensing group of PECo's Nuclear-g Engineering Divisio The. Reactor Protection System (RPS) review included those documents

. identified in Appendix II Preoperational testing of.the RPS.had

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not been-completed at the end of the inspection period. This area will: receive further review when testing is completed. During the review, several inconsistencies were noted:

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-TS Figure B 3/ , Reactor Vessel Water Level, shows levels which differ from those given in FSAR table 15.0-2, Input Parameters and Initial Conditions for Transients. This issu and the previously noted MAPLHGR curve issue are considered unresolved (89-11-01).

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'FSAR Section 7.2.1.1.3 describes'th'e power supplies to the RP The normal supply is'an inverter, supplied by a . Class IE battery.

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'.The alternate . supply is a non-Class IE plant auxiliary power source. The Limerick SER describes the Reactor Trip System Power Source in Section 7.2.2.7. This description is in error because its the opposite of the FSAR description. The NRC Limerick Licensing Project Manager has been notified of ~ this erro It will be corrected in a future supplement to'the SE This review will be continued in. subsequent inspection .0 Reactor Pressure Vessel Nozzles (73055)

The inspector was informed that ultrasonic examination of selected nozzles had detected indications that were different from the baseline examinations performed prior to the Mechanical Stress Improvement Program-(MSIP). On four nozzles the indications displayed inside diameter (ID)

connected appearances and'other nozzles had midplane indication The licensee indicated that extensive investigations are ongoing to charac- I terize these indications including:

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digital enhancement of pre-MSIP radiographs

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performance of additional radiography

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performance of additional ultrasonic examinations

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evaluation of effects of geometry, materials and MSIP with a similar nozzle from the Black Fox reactor.

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Philadelphia Electric Company 13 The inspector was informed that the Black Fox nozzle would be shipped to the Electric Power Research Institute (EPRI) for conduct of independent investigatio The effect of the safe end weld geometry and of the MSIP application will be reviewed by EPR Just prior to the end of the inspection period, PECo informed the resident inspectors that two more nozzles had been located (from the cancelled Hartsville reactor) which had the same safe end replacement modification performed. The nozzles are being analyzed to determine similarity to the Limerick nozzle Current plans call for ultrasonic examination (UT), core sampling, performance of Mechanical Stress Improvement (MSIP) and post MSIP UT and core samplin The results of'

this work will be reviewed after they become availabl .0 Startup Test Activities (71302, 35301, 70329, 70435, 70432, 70317, 70315, 70S22)

The inspector reviewed the preoperational test results from the Spent Fuel Pool Cooling system. The test summary accurately reflected that outstanding test completion was controlled by appropriate Test Exceptions. The completed test steps were verified consistent with the acceptance criteria. Test modification by Test Change Notices was found properly controlled. The inspector had no question The inspector witnessed two attempts at performing a scram test for the control rod drive hydraulic preoperational test (2P55.1). The proper initial control rod positions were verified and the slowest four drives were recorde In both attempts witnessed, the zero time signal was not properly recorded due to test instrumentation problems. The inspector was informed that the test was successfully performed on the fourth evolutio The inspector reviewed test results for the Emergency Service Water (ESW)

Loop B. This included review of Flow Balance (2FB54.18) and Preoperational Test (2P54.18) results and the Test Review Board (TRB)

review of the test results. TRB noted that both B and D ESW pumps were below design capacity (10% and 15% respectively) and that this had been documented on a Startup Field Report (SFR). PECo Nuclear Engineering found the pumps acceptable for use based upon the ability to supply all necessary loads for safe shutdown, and the similarity of the data to that obtained during Unit 1 preoperational test TRB also noted that the sum of the flows reported in the flow balance exceeded that shown on the system flow indicator (TI-11-0138). This issue is further addressed in IR 50-352/89-1 The inspector witnessed portions of the Standby Liquid Control System reoperation test (2P53.1). The inspector monitored the vessel injection test and independently calculated pump capacities and flow rate to the vessel to verify that they were within acceptable limit )

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l -The: inspector witnessed portions of the Main Steam System preoperational test (2P83.1). Main. Steam Safety / Relief Valves (MSRVs) connected to the Automatic Depressurization System (ADS) were tested for proper operation of the air actuators and accumulator check valves. The check valves for the accumulators associated with PSU-2F013 H and K did not hol Resolution of'this test exception will be reviewed in a future repor The inspector witnessed portions of the Reactor Protection System preoperational. testing (2P58.1) associated with response time testing of instrument channels. Results of these tests will be evaluated after completion of the test procedur .0 Appendix R Program The inspector met with licensee personnel to discuss the status of the Appendix R assessment program. The inspector was informed that the .

Bechtel self-assessment has been completed and that the PECo  !

self-assessment will be completed in three to four months. The licensee self assessments.have identified some concerns in the program including:

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remote shutdown panel room ventilation

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communication system protection power for remote shutdown panel instrumentation

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RCIC steam inlet valve spurious closure The licensee has evaluated and reported appropriate deficiencies on both Unit l'and 2. The inspector will continue to monitor the implementation lof the licensee efforts to resolve the outstanding areas of concern with the Appendix R program, 11.0 Diesel Generator Protective Trips The inspector reviewed Region I Temporary Instruction 87-04. The Limerick design for bypass of protective trips has been.previously reviewed in NRC inspection reports 50-352/87-21 and 87-24 and 50-353/87-13. Previous inspection has found that the engine will respond to protective trips except when an emergency LOCA start signal is presen The protective trips are developed in a two out of three logi The logic design previously incorporated a trip signal from non-safety related fire suppression flow switches. Those flow switches have been removed from the plant. The licensee analyzed the impact of not bypassing protective trips during dead bus diesel starts and found that false trips were an insignificant contribution to overall engine unavailability and the engine long term unavailability was minimized with the protective trips not bypassed. The Temporary Instruction is considered closed for Limerick _ _- ___ _ _ _ -_-__ _ -_ - - _ _ _ _ _- _-_ - _ - _.__ -____- - _ _ _ - _ _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ . __-_N

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Philadelphia Electric Company 15 12.0 Environmental Qualification Program The inspector met with PECo engineering to discuss the current status of the environmental qualification (EQ) program for Limerick. The inspector was informed that the Bechtel EQ walkdowns have been completed for 100%

of equipment installed inside containment and 50% of equipment installed in the reactor building. Additional walkdowns will be performed prior to fuel load to examine 100% of the equipment in the reactor building. The walkdowns have identified a concern with the lack of EQ conduit seals for instruments supplied by NAMCO, PICO and ASCO which are predominately used in the HVAC system. These were also found deficient on Unit 1 and will be corrected prior to restar The licensee appears to have implemented a comprehensive program to ensure that the plant equipment is properly installed in accordance with

.the EQ progra .0 Construction Inspection Program The' licensee has stated that construction activities are 98.5% complet NRC inspection has proceeded such that the program is 99% complete. The remaining.open construction inspection module is 63050, " structural integrity." All other construction modules are complete with issuance of this repor .0 High Pressure Coolant Injection System (HPCI) (37301)

A HPCI turbine operability issue developed at a BWR 3 where it was identified that on a loss of the HPCI barometric condenser, the resulting steam leakage would cause overheating of the equipment room and subsequent failure of the turbine. The inspector questioned whether the barometric condenser was considered necessary for HPCI operability at Limerick and whether steam leakage resulting from a loss of the barometric condenser was considered in the calculations for sizing the HPCI room coolers. It was determined that the HPCI barometric condenser is considered non-safety-related at Limerick and not necessary for turbine operability. The resulting steam leakage from a loss of the condenser was considered in the radiation dose rate for environmental qualification of the turbine and its control system. It appears, however, that steam leakage was not considered in the calculations for sizing the HPCI room coolers. This issue is considered unresolved pending engineering determination of the time-temperature profile in the HPCI room given that the barometric condenser is inoperative (89-11-02).

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philadelphia Elect'ric Company 16  !

15.0 Meetings with Licensee-(230703B)

~The NRC resi_ dent inspector discussed the issues and_ findings in this report with members of the licensee's staff on a weekly basis, and at an exit meeting held.on March 23, 1989. Based on discussions held with licensee representatives on March 23, 1989, it was determined that this report does not contain information subject to 10 CFR 2.790 restriction The inspector attended a PECo interdepartmental meeting on February 15, 1989, with regard to fuel problems at Unit 1. Strategies were formulated to try to determine the cause of the unexpected. fuel clad failures and if there would be any impact on Unit The: inspector: attended a meeting on February 23,198'), held between PECo personnel and NRC personnel participating in the Maintenance Team Inspection. -Topics discussed included: Unit 1. fuel failures and use of Unit 2 fuel.in the reload strategy, degradation of carbon steel piping in the ESW' system and indications from the ' smart' ultrasonic examination of

. recirculation nozzles of the Unit I reactor vesse The inspectors met with the Startup Manager on March 20, 1989 to discuss the status of test and system tie-in schedules. Also discussed was the -

status of development and approval of surveillance and operating procedures. This information was subsequently presented to NRR by PECo at a meeting on March 22, 198 The inspectors met with pECo representatives on March 22, 1989, to discuss; electrical cable separation. All Unit I and common panels have now been inspected for proper separation (both Unit 1/ Unit 2 and safety related/non-safety related) and all discrepancies resolved. NQA determined that there were several contributing factors: construction procedures permitted installation of non-safety related wiring outside of panels without separation inspection, when a subcontractor had its own qualified QA/QC program (for instance MCC Powers) there was no final inspection by Bechtel QC, separation requirements were not always properly implemented in design drawings, there is no requirement for

' separation review of temporary circuit alterations, many inspections for separation were not properly documented as not having been performed prior to Unit I startup. As a result of NQAs findings, several startup

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administrative procedures, field engineering procedures, testing and laboratories division procedures and a station administrative procedure have been revised. This issue will be the subject of further review

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16.0 Assurance of Quality I Instances were noted of test engineers being inattentive to test-procedure requirements (section 9.0 and IR 50-353/89-23).

I There was evidence of good pre plannirg and coordination of the search of L Unit 2 for the . implementation of full securit Mechanical Stress. Improvement (MSIP) was applied to tri-metallic' welds on

, reacter vessel nozzles. This process had not been previously qualified by EPRI for. tri-metallic welds,. and the reactor vendor had recommended that the process not be applied to reactor pressure vessel nozzle PECo management is now aggressively pursuing resolution of the problems

.. caused by application of MSIP to reactor pressure vessel nozzles, including contingency' planning in the event that rewelding becomes y necessar ;

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Philadelphia Electric Company 18 APPENDIX I Reactor Core Isolation Cooling System

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Technical Specification' review:

3/4.3.2, 3/4.3.5, 3/4.3.7.4.1, 3/4.1.3, 3/4. /4.7 3 Tables 3.3.2-1. 3.3.2-3, 4.3.2.1-1, 3.3.2-2, 3.3.5-1, 3.3.5-2, 4.3.5.1-1, 4.3.7.4-1, 3.6.3-1 FSAR review:

5.4.6, 7.4.1.1, 7.4. Figures 5.4-8, 5.4-9, 5.4-10, 7.4-1 SER review:

7.4.1.1, 7.4.2.2, 7.4.1.4, 6.2.4.1, 15. Drawing review:

P&ID M-49, sheet 2, " Reactor Core Isolation Cooling System" M-50, sheet 2, "RCIC Pump / Turbine" 22A1354, General Electric RCIC Design Specification 761E234AD, General Eelctric RCIC P&ID Limerick Pump and Valve Inservice Test Program 791E421TR, General Electric RCIC Elementary Diagram Procedures review:

ST-2-049-400-2, RCIC-Keep Filled System Calibration Test (LSL-49-219)

ST-2-049-401-2, RCIC-Condensate Storage Tank Level - Low Calibration / Functional Test (LT 3.9-2N035E, LIS-49-2N635E)

ST-2-049-402-2, RCIC-Condensate Storage Tank Level - Low Calibration / Functional Test (LT-49-2NOSSA, LIS-49-2N635A)

ST-2-049-403-2, NSSSS-RCIC Steam Line Differential Pressure-High; Division 1, Channel A, Calibration / Functional Test (PDT-49-2N057A, PDIS-49-2N657A, PDS-49-2N660A, E51A-K15A)

ST-2-049-404-2, NSSSS-RCIC Steam Line Differential Pressure-High; Division 3, Channel C, Calibration / Functional Test (PDT-49-2N057C, PDIS-2N657C, PDS-49-2N660C, E51A-K15C)

ST-2-049-405-2, NSSSS-RCIC Steam Supply Pressure-Low; Division 1, Channel A, Calibration / Functional Test (PDT-49-2N058A, PIS-49-2N658A)

ST-2-049-406-2, NSSSS-RCIC Steam Line Differential Pressure-Low; Division 3, Channel C, Calibration / Functional Test (PDT-49-2N058C, PIS.49-2N658C)

ST-2-049-407-2, NSSSS-RCIC Steam Supply Pressure-Low; Division 1, Channel E, Calibration / Functional Test (PT-49-2N058E, PIS-49-2N658E)

ST-2-049-408-2, NSSSS-RCIC Steam Supply Pressure-Low; Division 3, Channel G, Calibration / Functional Test (PT-49-2N058G, PIS-49-2N658G)

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Philadelphia Electric Company 19-APPENDIX I:(CONT.)

ST-2-049-409-2, NSSSS-RCIC Equipment Room Temperature-High,- RCIC Equipment Room dT-High and RCIC Pipe Routing Area Temperature-High; Division 1 Calibration / Functional Test.(TdTs-49-2N601A, TTS-49-2N602A, TTS-49-2N603A,TTS-49-2N603E,TTS-49-2N603J,TTS-49-2N603N)

ST-2-049-410-2, NSSSS-RCIC Equipment Room Temperature-High, RCIC

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Equipment ~ Room dT-High and RCIC Pipe Routing Area Temperature-High; Division 3 Calibration / Functional Test.(TdTS-49-2N6010, TTS-49-2N602C, TTS-49-2N603C,TTS-49-2N603G,TTS-49-2N603L,-TTS-49-2N603R)

ST-2-049-601-2, RCIC-Condensate Storage Tank Level-Low; Functional Test (LIS-49-2N635E)

ST-2-049-602-2, RCIC-Condensate Storage Tank Level-Low; Functional Test

,(LIS-49-2N635A)

ST-2-049-603-2, NSSSS-P.CIC Steam Line Differential Pressure-High; Division 1, Channel A, Functional Test (PDIS-49-2N657A, POS-49-2N660A, E51A-K15A)

ST-2-049-604-2, NSSSS-RCIC Steam Line Differential Pressure-High; ,

Division 3, Channel C, Functional Test (PDIS-49-2N657C, PDS-49-2N660C, {

E51A-K15L)  !

ST-2-049-605-2, NSSSS-RCIC Steam Supply Pressure-Low; Division 1, Channel ;

A, Functional-Test (PIS-49-2N658A)

ST-2-049-613-2, NSSSS-RCIC Equipment Room Temperature-High, RCIC  :

Equipment Room dT-High and RCIC Pipe Routing Area Temperature-High; 1 Division 1 Functional Test (TdTS-49-2N601A, TTS-49-2N602A, TTS-49-2N603A,TTS-49-2N603E,TTS-49-2N603J,TTS-49-2N603N)

.ST-2-049-614-2, NSSSS-RCIC Equipment Room Temperature-High, RCIC

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Equipment Room.dT-High and RCIC Pipe Routing Area Temperature-High; ;

Division 3 Functional Test (TdTS-49-2N6010, TTS-49-2N602C, f TTS-49-2N603C,TTS-49-2N603G,TTS-49-2N603L,TTS-49-2N603R)  ;

ST-2-049-900-2, NSSSS-RCIC Steam Line Differential Pressure-High l Division 1. Channel A, Response Time Test (PDIS-49-2N657A, j PDS-49-2N660A)  :

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ST-2-050-402-2, NSSSS-RCIC Turbine Exhaust Diaphragm Pressure-High;

- Division 1, Channel A Calibration / Functional Test (PT-50-2N055A, PIS-50-2N655A)

ST-2-050-403-2, NSSSS-RCIC Turbine Exhaust Diaphragm Pressure-High; i Division 3, Channel C (PT-50-2N0550, PIS-50-2N655C) '

ST-2-050-404-2,.NSSSS-RCIC Turbine Exhaust Diaphragm Pressure-High; Division 1, Channel E (PT-50-2N055E, PIS-50-2N655E)

ST-2-050-405-2, NSSSS-RCIC Turbine Exhaust Diaphragm Pressure-High; Division 3, Channel G (PT-50-2N055G, PIS-50-2N655G)  !

ST-2-042-405-2, "ECCS and NSSSS-Reactor Vessel Water Level-Levels 1, 2 and 8; Division 1, Channel A (Core Spray, LPCI, ADS, RCIC and D/G)

Calibration / Functional Test (LT-42-2N091A, LIS-42-2N691A, LIS-42-2N692A, LS-42-2N693A)

ST-2-042-409-2, "ECCS and NSSSS-Reactor Vessel Water Level-Levels 1, 2 and 8; Division 1, Channel E (Core Spray, LPCI, ADS, RCIC and D/G)

Calibration / Functional Test (LT-42-2N091E, LIS-42-2N69E, LS-42-2N692E, LS-42-2N693E)

ST-2-042-435-2, RCIC Reactor Vessel Water Level-Levels 2 and 8; Division

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. APPENDIX I (CONT.)

1, Channel A Calibration / Functional Test (LT-42-2N097A, LIS-42-2N697A,

'LS-42-2N698A)

ST-2-042-436-2, RCIC Reactor Vessel Water Level-Levels 2'and 8; Division

.-l', Channel E Calibration / Functional Te'ts (LT-42-2N097E, LIS-42-2N697E, LS-42-2N698E) .

'ST-2-042-630-2, RCIC Reactor Vessel Water Level-Levels 2 and 8; Division 1, Channel A Functional Test (LIS-42-2N697A, LS-42-2N698A, ST-2-042-631-2,- RCIC Reactor Vessel Water Level-Levels; 2 and 8; Division 1, Channel E Functional Test (LIS-42-2N697E, LS-42-2N698E) .

ST-2-042-665-2, ECCS,-RCIC and NSSSS-Reactor Level and Pressure and Drywell Pressure, Division 1 Channels. A and J (Core Spray, LPCI, ADS, RCIC and D/G) . Functional Test (LIS-42-2N691A, LS-42-2N692A,

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LS-42-2N693A,'PIS-42-2N694A,PIS-42-2N690A,PIS-42-2N690J)

S49.1.D, RCIC' System _ Full Flow Functional Test iS49.1. A, Normal RCIC' Line-up for Automatic Operation 549.9.A, Routine Inspection of RCIC System S49.0.A,. Routine Inspection of RCIC System S49.0.A,1RCIC Test Operation using Auxiliary ~ Steam

$T-1-049-702-2,- RCIC Turbine Contaminated Piping Inspection ST-6-049-320-2, RCIC Operability Verification ST-6-049-230-2, RCIC Pump, Valve and Flow Test ST-1-049-701-2,1 RCIC Pump Contaminated Piping Inspection ST-1-049-202-2, RCIC Cold Shutdown Valve Test ST-6-049-200-2, RCIC Valve Test ST-1-049-100-2,'RCIC. Logic System Functional / Simulated Automatic'

. Actuation ST-1-049-202-2-(draft revision)

ST-1-049-200-2 (draft revision)

ST-6-107-590-2 (draft revision)

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1, ' APPENDIX II-Changes ~from Unit'l 1,. .

Technical Specification Review 2.1.2;- Bases 2.0 ' and 2.1.3;.c3/4.1.4; Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4 and'3.2.1-5;'3/4.2.3,, Figures 3.2.3-la, 3.2.3-1b and 3.2.3-2;

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3/4.2.4:

Final ' Safety Analysis; Report

. Chapter 4 -Table 6.3-4'

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Safety' Evaluation Report ~

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, Section 4; Supplement'2, Sections:4 2 and 4.4; Supplement 3;' Sections

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APPENDIX III

Reactor Protection System

+ C Technical Specification Review

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2.'2, Bases. 2.2.1, 3/4.3.1, Bases' 3/4.3.1 and Figure B3/4.3-1-Final" Safety Analysis Report

. Sections 7.2?and.15.0-

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Safety Evaluation-Report--

, TSection 7.2, Supplement 2,'Section 7.2,-Supplement 3, Section 7.2'

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