ML20101A942

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Review of Limerick Unit 1 Tech Specs
ML20101A942
Person / Time
Site: Limerick Constellation icon.png
Issue date: 10/08/1984
From: Beckman D, Jimenez H, Mills W
PARAMETER, INC.
To:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20101A938 List:
References
CON-NRC-05-82-249, CON-NRC-5-82-249 NUDOCS 8412190447
Download: ML20101A942 (40)
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s. o, INSPECTION REPORT 50-352/84-52 ATTACHEENT A TECHNICAL REVIEW REPORT REVIEW 0F LIMERICK UNIT 1 TECHNICAL SPECIFICATIONS PHILADELPHIA ELECTRIC COMPANY LIHERICK GENERATING STATION UNIT 1 NRC DOCKET NO. 50-352 NRC CONTRACT NO. NRC-05-82-249 PARAMETER, INC. CONTRACT NO. NRC/IE-82/83-70 TASK ORDER N0.70 OCTOBER 8, 1984 Prepared by: Authors:

Parameter, Inc. W.R. Mills 13380 Watertown Plank Road D.A. Beckman Elm Grove, Wisconsin 53122 H.A. Jimenez Prepared for: Lead NRC Engineers:

U.S. Nuclear Regulatory Commission R.M. Gallo Region 1 J.E. Beall 631 Park Avenue N.J. Blumberg King of Prussia, Pennsylvania 19406 o!h!!2PDR

INSFECTION REPORT 50-352/84-52 ATTACHMENT A TABLE OF CONTENTS Section Title P_ age; Executive Summary.......................... 1 1.0 Introduction............................... 2 1.1 Purpose of The Inspection.................. 2 1.2 Background................................. 2 2.0 Evaluation................................. 5 2.1 Residual Heat Removal System............... 5 2.2 Emergency AC and DC Power Systems.......... 9

2. 3 Primary Containment Isolation System....... 12 2.4 Secondary Containment Systems.............. 16 2.5 Primary Containment - Drywell/

Suppression Poo1........................... 18 2.6 Service Water Systems...................... 21 3.0 Conclusion................................. 25 Appendices................................. 26 l

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A EXECUTIVE SUM 4ARY Parameter. Inc. assisted the Nuclear Regulatory Comission (NRC) in performing an inspection to determine if the Limerick Unit 1 nuclear power plant Finai Draft Technical Specifications were compatible with the as-built plant configuration and operating characteristics.

The scope of work for the inspection included the Residual Heat Removal system, emergency AC and DC electrical power systems, service water systems and containment systems. The as-built configention and operating characteristics for portions of these systems were determined by review of documents such as the Final Safety Analysis Report, the NRC's Safety Evaluation Report, Piping and Instrument Drawings, Electrical Drawings, Functional Control Diagrams, Preoperational Test Results and Operating Procedures. The in-situ plant equipment was visually inspected on a sampling basis to verify that actual installations were in accordance with the above documents. Surveillance Test Procedures were reviewed to verify that the Technical Specification surveillance requirements were adequately covered by the surveillance tests and that the as-built plant configuration permitted proper performance of the surveillance tests.

The inspection determined that the Final Draft Technical Specifications issued September 21, 1984 were compatible with the as-built plant configuration and operating characteristics in the areas inspected. The inspection identified several inconsistencies between the Final Draft Technical Specifications and related design documents or plant procedures.

These inconsistencies were presented to the licensee during the exit meeting on September 21, 1984 and were promptly resolved by the licensee.

, These inconsistencies affected only documentation, not the as-built plant configuration or Technical Specification validity.

Based on the validity of the Final Draft Technical Specifications in the areas inspected, the inspection concluded that the Technical Specification i preparation process was functioning appropriately and that the Technical Specifications were compatible with the as-built plant configuration and operating characteristics.

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A

1. INTRODUCTION j 1.1 PURPOSE OF THE INSPECTION The purpose of the inspection was to assist the Nuclear Regulatory Commission (NRC) in determining if the Limerick Unit I nuclear power plant Technical Specifications were compatible with the as-built plant condition and operating characteristics. This technical review report documents the results of that inspection.

1.2 BACKGROUND

During the low-power testing phases at Grand Gulf Unit 1, it was found that discrepancies existed between the Technical Specifications and either the Final Safety Analysis Report (FSAR), the NRC Safety Evaluation Report (SER), or the plant's as-built condition. These discrepancies were eliminated by amending the low-power license and by changing the Technical Specifications.

As a result of the problems found at the Grand Gulf plant, Region I decided to conduct inspections at selected nuclear power plants to provide additional assurance that the plant Technical Specifications were compatible with the as-built plant configuration.

Parameter. Inc. was requested to assist Region I in performing an inspection at Limerick Unit 1 to determine whether or not the Final Draft Technical Specifications (issued September 21, 1984) were compatible with the plant hardware, its operating characteristics and other conditions of the as-built plant for selected safety-related systems, structures and components. The scope of work identified by the NRC for the inspection t

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A included the following systems:

Residual Heat Removal System, Emergency AC and DC Power Systems, Service Water Systems, and Containment Systems.

The inspection plan provided for the following:

Documents Reviewed:

Technical Specifications (TS), FSAR, SER, SurveillanceTest(ST)

Procedures. Piping and Instrument Drawings (P&ID's), Electrical Drawings, Functional Control Diagrams (FCD), Preoperational Tests, and Operating Procedures.

Inspection Criteria:

Compatibility of Final Draft TS with the as-built plant configuration.

Consistency of Final Draft TS with the documents listed above.

Adequacy of STs to provide for implementation of TS surveillance .

requirements.

Inspection Methods:

In addition to the review of the various documents listed above. TS validity was reviewed by the following methods:

System Walkdown:

As-built configurations identified by visual inspection were compared with the TS, FSAR, and SER and other documents listed 3

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A above for selected systems.

Selected STs were walked through in the plant to determine ST validity and adequacy to satisfy TS surveillance requirements.

Surveillance Test Procedure Review:

Plant surveillance test procedures were reviewed against the Technical Specification requirements. Records of test results for surveillances previously performed were reviewed on a selected basis.

Operating Procedures Review:

Selected plant operating procedures were reviewed to verify consistency of plant 0;::rsting modes and characteristics with the Final Draft Technical Specifications.

Program Controls Review:

The licensee's provisions to control the scheduling of surveillance tests were reviewed to determine the adequacy of these provisions to assure that surveillance testing required by the Technical <

Specifications will be performed on the appropriate equipment at the correct frequency.

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A l

2. EVALUATION This section (1) describes the inspections performed and (2) presents the inspection and evaluation results for each system selected for review. The plant Final Draft Technical Specifications were evaluated for compatibility with the as-built plant condition and operating characteristics in the following areas:

Residual Heat Removal System (RHR),

Emergency AC and DC Power Systems.

Primary Containment Isolation Systems, Secondary Containment Systems, Drywell to Suppression Pool Vacuum Breakers, and Emergency Service Water Systems.

This section discusses the inspection performed for each of these areas.

2.1 RESIOUAL HEAT REMOVAL SYSTEM 2.1.1 Scope The RHR system was reviewed with respect to the criteria of Sections 1.1 and 1.2 of this report. Proposed TS 3/4.3.3, 3/4.3.7 3/4.4.9, 3/4.5.1, 3/4.5.2, 3/4.5.3, 3/4.6.3 and 3/4.9.11 were compared to the documents listed in Appendix 2.1 to verify that the proposed TS were compatible with the as-built plant configuration. FSAR and SER. A walkdown of the RHR system was performed to verify that the Final Draft TS and documents listed in Appendix 2.1 were consistent with the in-situ plant configuration. STs were reviewed to verify that the TS surveillance requirements were adequately covered and that the as-built plant configuration permitted performance of the STs. The inspection and evaluation of the RHR focused on the following areas 5

INSPECTION REPORT 50-352/84-52 ATTACHMENT A Low Pressure Coolant Injection (LPCI) mode of RHR, Primary Containment Isolation (PCI) Valves in the RHR system, and Surveillance Test Procedures for the RHR system.

2.1.2 Discussion The as-built design and operating characteristics of the RHR system were determined by review of the documents listed in Appendix 2.1. The in-situ equipment was verified by visual inspection on a sampling basis to be in accordance with the various design documents. Technical Specifications applicable to the RHR system as listed above were reviewed for compatibility with the identified as-built configuration.

Selected Technical Specifications for the LPCI mode of operation were specifically reviewed in detail for compatibility with the as-built plant configuration. The inspection identified the LPCI Emergency Core Cooling System (ECCS) automatic actuation signals, ECCS actuation set points, valve stroke times, and overall LPCI system response time. The Techncal Specification requirements for this equipment were reviewed for compatibtitty with the hardware installed in the plant and described in relevant sections of the FSAR, SER, P&lDs, STs, FCDs and preoperational '

test results.

A walkdown of the RHR system was performed to verify consistency of the as-built condition with information obtained from the above documents. The walkdown included visual inspection of instrumentation and controls and main flow paths for the following modes of the RHR system:

LPCI mode.

Shutdown Cooling mode, Suppression Pool Cooling mode, and Containment Spray mode.

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,. ,o INSPECTION REPORT 50-352/84-52 ATTACHMENT A The flow paths were traced from the suction supply piping through the system discharge piping. The walkdown included inspection of major RHR system components such as RHR pumps, heat exhanges, selected valves, high-point vents, keep-fill system, flood protection and selected motor control centers.

Surveillance Test Procedures listed in Appendix 2.1 were reviewed against the Technical Specifications to verify that the surveillance requirements in the Technical Specifications were satisfied by the Surveillance Test Procedures and to verify that the as-built plant configuration permitted proper performance of the surveillance tests.

The RHR system primary containment isolation valves identified in Technical Specification Table 3.6.3-1 were reviewed. The inspection determined whether or not the Limerick Surveillance Test Procedures adequately provided for testing of these isolation valves.

2.1. 3 Observations This portion of the inspection found that the Final Draft Technical Specifications for the RHR system were compatible with the as-built plant condition and operating characteristics in the areas evaluated. Several inconsistencies as discussed below were identified in the Surveillance Test Procedures which related to the Technical Specifications for the RHR system. These inconsistencies were presented to the licensee in the exit meeting on September 21, 1984. In response, the 11censeo prepared ST revisions which removed the inconsistencies. The inconsistencies and their resolution were as follows:

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A 2.1.3.1 RHR Primary Containment Isolation Valves Four primary containment isolation valves, HV51-142 A (B, C, D), which are located in the one inch lines that bypass the LPCI testable check valves were required to be operable and periodically tested by Technical Specification 3.6.3 and associated Technical Specification Table 3.6.3-1.

These four isolation valves were not included in the Surveillance Test Procedures. The licensee determined that these valves were an addition to the Technical Specifications during the review process and had not been added to the Surveillance Test Procedures at the time the inspection started. The licensee has now added these val'es to the appropriate STs and this item is resolved.

2.1.3.2 RHR Primary Containment Isolation Valves - Closure Times Technical Specification 3.6.3 and associated Table 3.6.3-1 specified a maximum isolation time of 18 seconds for valve HV-C-51-1F103A to isolate an RHR relief valve discharge line to the primary containment. Surveillance test procedure ST-6-051-231-1, Rev.1 had a maximum allowable isolation time of 30 seconds which was less conservative than the proposed Final Draft TS.

The licensee revised the ST to reflect a maximum isolation time of 18 seconds in accordance with the Technical Specifications. This resolution closed this item and did not impact the as-built configuration because the closure time of HV-C-51-1F103A was less than 18 seconds.

2.1.3.3 Surveillance Test Procedures For LPCI Response Tine Surveillance Test Procedures, ST-1-051-801-1 (2, 3, 4) and ST-1-051-851-1 for LPCI responsu time testing permitted the measurement of LPCI pump response time to rated flow separately from the response time to rated 8

INSPECTION REPORT 50-352/84-52 ATTACHMENT A pressure. Simultaneous rated flow and pressure were required to meet the TS. The licensee modified the surveillance test procedures for response time testing to require a measurement to rated 10CI pump flow and pressure simultaneously. This item is resolved.

2.1.3.4 RHR Surveillance Test Procedures Two apparent typographical errors of no safety significance were found in ST-1-051-803-2, Rev.0. This procedure had Div.4 not Div. 3 in the title on page 1 and in the prerequisites on page 3. The licensee has prepared a revision to the ST which resolves this item.

2.1.4 Conclusions The Final Draft Technical Specifications for the RHR system were found to be compatible with the as-built plant configuration in the areas inspected.

The related surveillance test procedures were found to have several inconsistencies which were corrected by the licensee before the end of the inspection.

2.2 EMERGENCY AC & DC POWER SYSTEMS 2.2.1 Scope The emergency AC and DC electrical power systems were reviewed with respect to the criteria of Sections 1.1 and 1.2 of this report. Proposed Final Draft TS 4.8.1. 4.8.2 and 4.8.3 were compared to the documents listed in Appendix 2.2 to verify that the proposed TS were compatible with the as-built plant configuration and operating characteristics. Walkdowns of the emergency diesel generator systems and DC battery systems were performed to verify that actual installed configurations were in accordance with various design documents and single-line electrical diagrams.

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.a . 3 INSpECTICN REPORT 50-352/84-52 ATTACHMENT A The objective of the electrical portion of the inspection was to evaluate whether or not the TS for the emergency AC and DC power systems were compatible with the applicable Surveillance Tests, Sections 8.3 and 9.5 of the FSAR, Preoperational Tests, Single-Line Diagrams and "as-built" configurations.

2.2.2 Discussion The as-built design and operating characteristics of the emergency AC and DC power systems were determined by review of the documents listed in Appendix 2.2. The installed equipment was verified by visual inspection on a sampling basis to be in accordance with the applicable design documents.

TS for the emergency AC and DC power systems were reviewed and compared against the STs and Sections 8.3 and 9.5 of the FSAR to verify that the STs satisfied the TS surveillance requirements and the FSAR commitments.

The preoperational tests identified in Appendix 2.2 were reviewed and compared against the TS to verify that the system operating modes and characteristics were consistent with the TS descriptions and requirements.

The as-built configurations of the emergency diesel generator system and DC battery systems were compared during a field walkdown with the single-line diagrams and the TS surveillance requirements. The field walkdown included the following verifications:

2.2.2.1 DC Battery Systems The single-line diagram was verified to be correct and adequate for the DC battery systems. The equipment and system configuration in the battery rooms were verified to be correct compared to the single-line diagram. The 10

v INSPECTION REPORT 50-352/84-52 ATTACHMENT A cable color codings were verified to be correct for each battery division. l The number and type of fuses, voltmeters, ammeters and selector switches mounted on the instrument panels were verified to be correct. The battery room ventilation system was verified to be operable and in service. The battery cell numbers were properly designated and posted; and a cell was designated " pilot cell" for each 125 VDC battery. The proper battery voltage and current were verified by reading the voltmeter and ammeter mounted on a local panel. Proper battery charger voltage was verified by reading the voltmeter mounted on the charger panel. The batteries were verified to be operable and free from cracks, electrolyte leakage and terminal corrosion. The battery rooms were clean and free of tools and foreign material.

2.2.2.2 Diesel Generation Systems The emergency diesel generator systems and associated protective relay systems were verified to be adequate and in accordance with the single-line diagrams. The number and type of voltmeters, ammeters, synchronizing switches, indicating lights, and other selector switches mounted on the diesel generator panels were verified to be correct. The cable color coding for each diesel generator division was verified to be correct. Even though the diesel generators were not operated during the walkdown, operational readiness was verified based on system condition, various switch positions and status of indicating lights. Adequate day tank fuel level and tank size to hold the required amount of fuel were verified.

2.2.3 Observations l

l The TS descriptions and requirements matched with the associated design

! documents, as-built configurations, STs and preoperational test results.

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D e a e INSPECTION REPORT 50-352/84-52 ATTACHMENT A The audit determined that some TS and STs were being revised and reissued.

For example, Surveillance Test ST-4-095-921-1 agreed with Technical Specification 4.8.2.1.c.4 on battery charger supply values of a minimum of 125 volts for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The values in both documents were expected to change to a minimum of 132 VDC for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. These changes were made when the Final Draft TS was issued and this item was resolved.

2.2.4 Conclusion The information reviewed was adequate and the Techncial Specifications were compatible with the related design documents. STs and as-built condition for the areas inspected.

2. 3 PRIMARY CONTAINMENT ISOLATION SYSTEM 2.3.1 Scope The Primary Containment Isolation System (PCIS) was reviewed with respect to the criteria of Sections 1.1 and 1.2 of this report. Proposed TS 3/4.3.2 and ;!4.3.6 were compared to the documents listed in Appendix 2.3 to verify thst the proposed TS were compatible with the as-butit plant configuration, FSAR and SER. A control room review of the Reactor Water Cleanup System was performed to verify that the Final Draft TS and applicable documents listed in Appendix 2.3 were consistent with the installed plant configuration. Selected STs were reviewed to verify that the associated TS surveillance requirements were satisfied.

Three specific areas of the PCIS were selected for examination:

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l Reactor Water Cleanup System Isolation, Reactor Enclosure Isolation on outside atmosphere to Reactor Enclosure low differential pressure, and Primary Containment Instrument Gas isolation on Primary Containment 12 I

INSPECTION REPORT 50-352/84-52 ATTACHMENT A Instrument Gas to drywell low differential pressure.

-The primary containment isolation valves and trip signals reviewed were listed in FSAR Table 6.2-17 and in Technical Specification Table 3.6.3-1.

l 2.3.2- Discussion The as-built design and operating characteristics of the three areas identified above were determined by review of the documents listed in Appendix 2. 3. This information was compared to the in-situ equipment configuration and characteristics determined by a review of preoperational test results for the installed equipment. The TS and various documents in Appendix 2.3 were compared against each other to determine the consistency and completeness of important design documents.

2.3.2.1 Reactor Water Cleanup System Isolation Three general types of system isolation were reviewed for the Reactor Water ,

Cleanup (RWCU): 1 solation due to external accident signals, isolation due to internal system trouble (e.g., heat exchanger or area high temperature) and isolation due to Standby Liquid Control (SBLC) system initiation. The Technical Specifications were checked against the FSAR, SER, FCD, ,

Elementary Drawings, Preoperational Tests and STs to confirm that the Techncial Specifications reflected the as-built plant configuration.

2.3.2.2 Reactor Enclosure Isolation on Outer Atmosphere To Reactor Enclosure Low Differential Pressure The Technical Specification requirement was checked against the FSAR, SER.

FCD and the STs to confirm that the Technical Specification reflected the as-built plant configuration.

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A f

-2.3.2.3 Primary Containment Instrument Gas Isolation on Primary Containment Instrument Gas To Drywell Low Differential Pressure i d

This isolation provision was a recent proposed change to the Froof and j Review version of the Technical Specification. The FSAR was reviewed to

] determine if the subject isolation signal for the primary containment t

l instrument gas system was appropriately addressed. The STs were reviewed l to verify .that a ST was being perpared to demonstrate that the isolation

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function was correctly installed and periodically demonstrated to be i

operable.

_ Surveillance Test Procedure (ST-2-059-602-1) was in draft form  !

and did contain a provision to verfiy that the isolation function would .

operate upon demand.

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2.3.3 Observations i

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2.3.3.1 Reactor Water Cleanup Isolation The external accident trips and internal system trouble trips' were documented to correspond to the Technical Specifications. The RWCU isolation on SBLC initiation was installed according to the Technical Specification, but there was an inconsistency with the FSAR and the RWCU-

! system FCD. The higher level design documents (the elementary drawings) 1 j showed the same logic as that specified in the Technical Specifications.

This logic was confirmed to be correctly installed by the applicable .

I preoperational test, 1P-53.1(0), which was performed on May 3, 1984. The i

RWCU FC0 and the FSAR needed modification in order to be consistent with j the TS and elementary drawings. Only the above documentation needed l revision, the TS was adequate and consistent with the as-built plant l configuration. Region I incorporated followup of this item into Inspection  !

i l 84-49 for Limerick. The results of Region I followup will be discussed in i a future Inspection Report (352/84-49-02 and 353/84-12) for Limerick. This L

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A item was closed for this audit.

2.3.3.2 Reactor Enclosure Isolation On Outer Atmosphere To Reactor Enclosure Low Differential Pressure The design documents and plant procedures associated with this isolation

, confirmed that the isolation function was installed consistent with the Technical Specification requirements. The approved STs (ST-2-076-600(1)-1)

I had provisions to confirm that the correct logic was in place and operable during performance of each test. The preoperational test (lP-59.1) confirmed that this isolation function had been properly installed.

2.3.3.3 Primary Containnent Instrument Gas Isolation on Containment Instrument Gas To Drywell Low Differential Pressure The FCD and elementry drawings were found to agree with the Technical Specification trip requirements. The draft STs contained provisions to confirm operability of the logic during performance of each test. The preoperational test (1P-25.1) for this isolation confirmed that this trip had been properly installed. The FSAR appropriately discussed this isolation function.

2.3.4 Conclusions The three selected PCIS areas were verified to be installed properly and in conformance with the applicable Technical Specification requirements. The inconsistencies identified during the inspection affected only documentation, not validity of the TS or as-built plant condition. These inconsistencies were resolved prior to close of the inspection.

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l INSPECTION REPORT 50-352/84-52 ATTACHMENT A 2.4 SECONDARY CONTAINMENT SYSTEMS 2.4.1 Scope l

The Standby Gas Treatment System (SGTS), Reactor Enclosure Recirculation System, Secondary Containment Automatic Isolation Valves, and Secondary Containment Integrity Features were reviewed with respect to the criteria of Sections 1.1 and 1.2 of this report. Proposed TS 3/4. 6. 5.1, 3/4. 6. 5. 2,

, 3/4.6.5.3 and 3/4.6.5.4 were compared to the documents listed in Appendix

] 2. 4 to verify that the proposed TS accurately represented the as-built l plant configuration and operating characteristics and agreed with the FSAR r

j and SER. Surveillance Tests were reviewed in detail to assure adequate implementation of the respective TS surveillance requirements and to verify 1 agreement with the as-built conditions.

j. 2.4.2 Discussion

}

} The as-built design and operating characteristics of the above systems and I

equipment were determined by review of the documents in Appendix 2.4 and j the actual installations were verified to be in accordance with the various J

l design documents by visual inspection.

l The documentation review involved comparison of the FSAR, SER, STs, and as-built engineering drawings with TS, including P&ID configurations, FCDs,

{ elementary diagrams and architectural drawings. The Plant Modification 1

Status Printout dated September 17, 1984 was also reviewed to determine the j impact of pending modifications on the TS.

1 The visual inspection included walkdown of principal system flow paths and 1

observation of equipment, instrumentation, and controls; comparison of the proposed Technical Specifications with the in-situ hardware and selected i

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l INSPECTION REPORT 50-352/84-52 ATTACHMENT A design documents; simulated performance (walkthrough) and accuracy verification of issued and approved STs and draft ST revisions; and, detailed discussion of the systems. Technical Specifications, and associated documentation with licensee staff members.

Portions of the following specific systems, subsystems, and equipment were reviewed and observed. Selected portions of the main control room and local instrumentation, controls, alarms, etc. were also confirmed:

Standby Gas Treatment System fans, filters, dampers:

Reactor Enclosure Recirculation System fans, filters, dampers; Secondary Containment isolation valves / dampers, penetrations, temporary penetration closures (as listed in TS Table 3.6.5.2.-l and the Appendix 2.4 STs); and Refueling Area - Reactor Enclosure boundary penetrations and isolation devices.

2.4.3 Observations

,In-Process TS Revisions ST-1-076-310 SGTS Secondary Containment Integrity Verification. Rev.0, implemented the requirements of TS Surveillance Requirement 4.5.6.1.c for Standby Gas Treatment System and Secondary Containment flow, time and vacuum limits.

At the start of the audit, the licensee and the NRC Office of Nuclear Reactor Regulation were processing revisions to these parameters in the Proof and Review version of subject TS thus requiring eventual revision of the ST to provide agreement with the TS. The licensee had identified the i

need to change the ST and was awaiting final TS values to support procedure revision. The correct final values were issued in the Final Draft TS on September 21, 1984 which resolved this item.

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e-INSPECTION REPORT 50-352/84-52 ATTACHMENT A No unresolved differences were identifed between the Final Draft TS and the ,

I documents reviewed or the as built plant configuration. Minor revisions to related documentation resulting from ST reviews, the licensee's ST validation process, and recent plant modifications are discussed in Appendix 2.4.

2.4.4 Conclusions No discrepancies were identified between the Final Draft TS, related design documentation, and the as-built plant configuration and operating characteristics.

2. 5 PRIMARY CONTAINMENT - DRYWELL/ SUPPRESSION P0OL 2.5.1 Scope The Primary Containment was reviewed specifically with regard to the TS surveillance requirements of TS 3/4.6.2.1 for Drywell to Suppression Chamber Bypass Leakage and 3/4.6.4.1 for Suppression Chamber - Drywell Vacuum Breakers. These features were reviewed with respect to the criteria of Sections 1.1 and 1.2 of this report. The proposed TS were compared to the documents listed in Appendix 2.5 to verify that the proposed TS accurately represented the as-built plant configuration and operating characteristics and agreed with the FSAR and SER. Surveillance Tests were reviewed in detail to assure adequate implementation of the respective TS surveillance requirements and to verify agreement with the as-built conditions.

2.5.2 Discussion The as-built design and operating characteristics of the above equipment 18

INSPECTION REPORT 50-352/84-52 ATTACHMENT A were determined by review of the documents in Appendix 2.5 and the actual installations were verified on a sampling basis to be in accordance with the various design documents by visual inspection.

The documentation review involved comparison of the FSAR, SER, STs snd as-built engineering drawings with TS, including P&ID configurations, FCDs, and elementary diagrams. The Plant Modification Status Printout dated September 17, 1984, was also reviewed to determine any potential impact on the TS.

The following specific systems, subsystems and equipment were reviewed and/or observed. Selected protions of the main control room and local instrumentation, controls, alarms, etc. were also confirmed:

Containment Atmosphere Control System.

Suppression Chamber - Drywell Vacuum Breakers, Applicable portions of the:

Nuclear Boiler System.

Reactor Core Isolation Cooling System.

Residual Heat Removal System, Emergency Core Cooling System, Primary Containment Instrument Gas System.

Primary Containment Leak Testing System, and Liquid Radwaste - Collection.

2.5.3 Observations Vacuum Breaker Position Indication Problem ST-2-060-400-1, Containment Systems - Suppression Pool /Drywell Vacuum Breaker Setpoint Check and Channel Calibration, Revision 0, implemented the l requirements of TS 4.6.4.1.b.3 for verification of the vacuum breaker 1

i opening setpoint and valve position indicator (VPI) calibration. The 19

e e e o INSPECTION REPORT 50-352/84-52 ATTACHMENT A

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licensee had been unable to achieve acceptable calibration of the VPIs using the above procedure due to apparent hysteresis in the VPI limit switch linkages. (Although the ST provided apparently acceptable methods and met the TS requirements for the valve opening setpoint.) This problem was similar to those experienced by other licensees with similar valves and surveillance requirements.

Prior to this audit, the licensee had initiated corrective action involving redesign of the VPIs to replace the VPI mechanical arm linkage with a direct acting " plunger" design to eliminate the hysteresis. The design of the modification was in progress during the audit and was not expected to be available until late September, 1984. The new design VPI configurations were discussed with the licensee's headquarters engineering staff and ,

appeared to address the test difficulties.

The licensee's engineering representative indicated that only the outboard valve of each vacuum breaker pair would initially be modified; the inboard valves would be modified a the first opportunity, possibly after fuel load or initial criticality. T3 3/4.6.4 required the VPI channels for both valves to be operable during the power, startup and hot shutdown modes of reactor cperation. The licensee's representative stated that the planned modification schedule would be reviewed in light of the TS and, if necessary, relief from the TS requirements would be requested from NRC.

Region I had previously inspected the adequacy of vacuum breaker position indication at Limerick and had initiated Violation 84-26-05. Region I has incorporated resolution of the above concerns in the required corrective actions to Violation 84-26-05.

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e 0 e Q INSPEcliON REPORT 50-352/84-52 ATTACHMENT A No other differences were found between the Final Draft TS and the documents reviewed or the as-built plant configuration. This item is closed with respect to this inspection.

2.5.4 Conclusions No discrepancies were identified between the TS, related design documentation, and the as-built plant configuration and opera ting characteristics.

C.6 SERVICE WATER SYSTEMS 2.6.1 Scope The Emergency Service Water (ESW; System, the Residual Heat Removal Service Water (RHRSW) System, and the Ultimate Heat Sink (UHS) were reviewed with respect to the criteria of Sections 1 and 2 of this report. The Proposed TS 3/4.7.1.1, 3/4.7.1.2 and 3/4.7.1.3 were compared to the documents listed in Appendix 2.6 to verify that these TS accurately represented the as-built plant configuration and operating characteristics and agreed with the FSAR and SER. Surveillance Tests were reviewed in detail to assure adequate implementation of the respective TS Surveillance Requirements and to verify agreement with the as-built conditions.

2.6.2 Discussion The as-built design and operating characteristics of the above systems and equipment were determined by review of the documents in Appendix 2. 6 and the actual installations were verified to be in accordance with the various design documents by visual inspection.

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A. l The documentation review involved comparison of the FSAR, SER, STs and as-built engineering drawings with TS, including P&ID configurations, FCDs, elementary diagrams and architectural drawings. The Plant Modifications Status Printout dated September 17, 1984 was also reviewed to determine the impact of pending modifications of the TS.

The visual inspection included walkdown of principal system flow paths and observation of equipment, instrumentation, and controls; comparison of the proposed TS with the in-situ hardware and selected design documents; simulated performance (walkthrough) and accuracy verification of issued and approved STs and draft ST revisions; and, detailed dicussion of the systems TS and associated documentation with licensee staff members.

Portions of the above systems and associated equipment were reviewed and observed. Selected portions of the main control room and local instrumentation, controls, alarms, etc. were also confirmed.

2.6.3 Observations 2.6.3.1 ESW System Flow Balance The ESW System was found to con.-ist of two headers (loops) each having two pumps. The system was capr.ble of cross feeding the Turbine Enclosure Cooling Water (TECW) System or the Reactor Enclosure Cooling Water (RECW)

System and all four (4) diesel generators from a single ESW loop. The normal system configuration aligned only two of four diesels to a single header; neither the TECW nor the RECW Systems were normally. aligned to ESW.

The NRC Resident Inspectors had previously reviewed the results of l

Preoperational Test 1FB-54.1, ESW Flow Balance, questioning the ESW Systems capability to provide adequate cooling flow to all required loads in a l

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A single pump, single loop ESW alignment with the additional heat loads above also valved in.

l In response to that question, the licensee conducted flow balance tests in i various single pump, single loop configurations, and determined that the RHR Pump Seal and Motor 011 Coolers and the "B" Core Spray Compartment Unit Coolers would not reveive design flows under the most limiting conditions.

That test data, the TS, and associated procedures were reviewed during this inspection.

The inspection determined that TS 3.7.1.2 and its associated Action Statements did not permit continued reactor operation in the power, startup, or hot shutdown modes with only one ESW Pump and Loop available.

That configuration, however, was permitted for cold shutdown and refueling modes of operation (including irradiated fuel handling) where the analyzed accidents would not involve design basis ESW heat loads.

In order to provide additional assurance that the systems will not be placed in an unauthorized or nonconservative alignment, the licensee was revising the system operating procedures ("S" Procedures) for ESW, RECW, TECW and associated equipment to include cautions and/or prohibitions on such lineups. The draft revisions were reviewed during this audit and, although they appeared to meet the general intent desired, were not finalized and were in preliminary form. Completion of the licensee's actions has been included by Region I in the followup of Unresolved Item 83-19-09. This item was closed with respect to this inspection.

l 2.6.3.2 Ultimate Heat Sink Freeze Protection The Ultimate Heat Sink was designed to be the ESW and RHRSW Spray Pond.

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A

! Section 9.5.2 of the NRC's SER for the facility required addition of a TS i

l Limiting Condition for Operation specifying requirements for drainage of l-piping subject to cold weather freezing.

The draft TS available at the start of this audit did not include such a provision. Discussion with the licensee's Technical Engineering Group established that this omission had been previously identified, discussed with the NRC Office of Nuclear Reactor Regulation, and would be corrected in the next issue of TS. The appropriate TS provisions were included in the Final Draft TS issued September 21, 1984 and this item was closed.

No remaining differences between TS and the documents reviewed or the as-built plant configuration were identified. Minor revisions to related documentation resulting from ST reviews, the li,censee's ST validation process and recent plant modifications are discussed in Appendix 2.6.

2.6.4 Conclusions No discrepancies were idantified between the Final Draft TS, related design documentation and the as-built plant configuration and operating characteristics.

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A

3. CONCLUSIONS The inspection determined that the Final Draft Technical Specifications issued September 21, 1984 were compatible with the as-built plant configuration and operating characteristics in the areas inspected. The inspection identified several inconsistencies between the Final Draft Technical Specifications and related design documents or plant procedures.

These inconsistencies were presented to the licensee during the exit meeting on September 21, 1984 and were promptly resolved by the licensee.

These inconsistencies affected only documentation, not the as-built plant configuration or Technical Specification validity.

Based on the validity of the Final Draft Technical Specifications in the areas inspected, the inspection concluded that the Technical Specification preparation process was functioning appropriately and that the Technical Specifications were compatible with the as-built plant configuration and operating characteristics.

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A APPENDIX 1 LICENSEE PERSONNEL CONTACTED At Limerick Site By Telecon. Sept. 25, 1984 P. Duca T. Shannon V. Cwietniewicz J. Goldman T. Shea R. George J. Muntz W. Sproat J. Krais G. Beck D. Kelsey B. Smith L. Flopkins K. Meck J. Armstrong W. Rekito J. FitzGerald G. Gibson K. Kemper J. Corcoran K. Meck G. Gilbody l

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A APPENDIX 2 DOCUMENTS REVIEWED l

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A APPENDIX 2.1 System / Equipment: Residual Heat Removal System Technical Specifications Reviewed: 3/4.3.3, 3/4.3.7, 3/4.4.9, 3/4.5.1, 3/4.5.2, 3/4.5.3, 3/4.6.3, 3/4.9.11.

FSAR Sections Reviewed: 5.4, 6.2, 6.3, 7.3.

SER Sections Reviewed: 5.4, 6.2, 6.3.

Procedures Reviewed:

ST-6-051-201-1, "A" Loop RHR Refueling Valve Test, Rev.0 ST-6-051-204-1, "B" Loop RHR Refueling Valve Test, Rev.0 ST-1-051-801-1 Div.1 LPCI System Response Time Testing, Rev.0 ST-1-051-802-1 Div. 2 LPCI System Response Time Testing, Rev.0 ST-1-051-803-1, Div. 4 LPCI System Response Time Testing, Rev.0 1

iT-1-051-804-1. Div. 4 LPCI System Response Time Testing, Rev.0

.T-3-051-231-1, "A" RHR Pump, Valve and Flow Test, Rev.1 ST-6-051-232-1, "B" RHR Pump, Valve and Flow Test, Rev.1 ST-6-051-233-1, "C" RHR Pump, Valve and Flow Test, Rev.1 ST-6-051-234-1, "D" RHR Pump, Valve and Flow Test, Rev.1 ST-1-051-851-1. A LPCI Initiation Response Time Summation, Rev.0 ST-1-051-852-1, B LPCI Initiation Response Time Summation, Rev.0 ST-1-051-853-1, C LPCI Initiation Response Time Summation, Rev.0 ST-1-051-854-1, D LPCI Initiation Response Time Summation, Rev.0 l

l Drawings and Other References Reviewed:

Residual Heat Removal System, P&ID 8031-M51, Sheets 1&2 Residual Heat Removal System, Functional Description 8031-M51FD, Sheets 1-4 Residual Heat Removal Ssytem, FCD 8031-M-1-E11-1030-F-1.7 i l i RHR Schematic Diagram, 8031-E-366 Rev.8, Sheets 1, 2 RHR Schematic Diagram, 8031-E-360, Rev.9, Sheets 1-3 l 28

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INSPECTION-REPORT 50-352/84-52-ATTACHMENT A APPENDIX 2.1 (CONT'D)

RHR Schematic Diagram, 8031-E-353. Rev.6 RHR Schematic Diagram, 8031-E-102, Rev.16 Sheets 1, 2 l

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A APPENDIX 2.2 l

System / Equipment: Emergency AC and DC Electrical Power Systems Technical Specifications Reviewed 4.8.1.1 AC Sources - Offsite. Power to Class 1E Distribution Systems

a. Determination of operability (7 days)
b. Determination of operability (18 days) 4.8.1.2 Diesel Generator Operability
a. Staggered Testing / Verification
1. Day Fuel Tank
2. Fuel Storage Tank
3. Fuel Transfer Pump System ,

' 4. Diesel starting to at least 882 r.p.m. in less than 10 seconds

5. Diesel Generator Synchronized
6. Diesel Generator Breaker Line-up 1
7. Air Starting Pressure
b. Purge Water from Fuel Oil Day Tank
c. Purge Water from Storage Tank
d. Fuel Oil Sample Analysis
e. Diesel Generator Inspection and Testing (18 months - shutdown)
1. Manufacturer's recommended inspection
2. Load Rejection Test - Partial 3 Load Rejection Test - Full
4. Simulated Loss of Offsite Power
5. Simulated E.C.C.S. initiation without L.0.0.P.

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6. Simulated E.C.C.S. Initiation with L.0.0.P.

' 7. Verification of automatic D.G. Trip

8. D.G. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Load Test 9 Verify Maximum loads vs. rating
10. Verifying D.G. capability to synchronize with the offsite power and load transfer

, 11. Auto initiation override of test signals

12. Load sequence timer verification
13. D.G. starting lockout verification
f. A.C. Sources Shutdown
g. D.C. Power Source - Operability 4.8.1.2 A.C. Sources Shutdown 4.8.2.1 D.C. Power Source - Operability
a. Battery & Charger Testing - 7 days
b. Battery & Charger Testing - 92 days 30

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A l l

APPENDIX 2.2 (CONT'D) l

c. Battery & Charger Testing - 18 months Battery Discharge Testing - 18 months
d.
e. Battery Discharge Testing - 60 months
f. Battery Discharge Test - Life Test 4.8.2.2 D.C. Power Source -Demonstrate Operability Cross Reference to Section 4.8.2.1 4.8.3.1 Safeguard Divisions Power Distributions Systems - Operating -

Verify correct breaker alignment and busses voltage (weekly) 4.8.3.2 Safeguard Divisions Power Distributions Systems - Shutdown -

Verify correct breaker alignment and busses voltage (weekly)

FSAR Sections Reviewed:

8.3 Onsite Power System 8.3.1 A.C. Power System 8.3.2 D.C. Power System 9.5.4 Diesel Generator Fuel Oil System 9.5.5 Diesel Generator Cooling Water System 9.5.6 Diesel Generator Starting System 9.5.7 Diesel Generator Lubrication System 9.5.8 Diesel Generator Combustion Air Intake Procedures Reviewed ST-1-092-111 1, Rev. 0 - D11 Diesel Generator 4 kV Safeguard Loss of Power ST-6-092-312-1, Rev. 0 - D12 Diesel Generator Operability Test Run ST-1-092-780-1, Rev. 0 - Unit 1 Diesel Generators Simultaneous Startup Test ST-1-092-990-1, Rev. 0 - Unit 1 Diesel Generator Failure Report ST-6-092-311-1, Rev. 0 - D11 Diesel Generator Operability Test Run ,

ST-4-092-911-1, Rev. 0 - D11 Diesel Generator Outage Inspection ST-4-092-631-1, Rev. 0 - D11 Diesel Generator Fuel Tank Cleaning and Pressure Check ST-4-095-951-1, Rev. 0 - Division I 125/250 VDC Safeguard Battery 18 Months Service Test (Shutdown)

ST-6-095-901-1, Rev. 0 - Division I 125/250 VDC Safeguard Battery Weekly Inspection 31

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A APPENDIX 2.2 (CONT'D)

ST-6-095-911-1, Rev. 0 - Division I 125/250 VDC Safeguard Battery Quarterly (3 Months) Inspection ST-4-095-921-1, Rev. 0 - Division I 125/250 VDC Safeguard Battery 18 Months Inspection ST-4-095-931-1, Rev. 0 - Division I 125/250 VDC Safeguard Battery 60 Months Performance Test ST-6-095-450-1, Rev. 0 - Divisions I thru IV 125/250 VDC Safeguard Power Distribution Alignment and Voltage Check ST-6-094-450-1, Rev. 0 - 120 VAC Safeguard Power Distribution Alignment and Voltage Check Drawings and Other References Reviewed:

E-1, Rev. 12 - Single Line Diagram E-15. Rev. 14 - Single Line Meter and Pelay Diagram 4 kV Safeguard Power System - 1 Unit E-16. Rev. 13 - Single Line Meter and Relay Diagram 4 kV safeguard Power System - 2 Unit E-17, Rev. 9 - Single Line Meter and Relay Diagram 114A and 124A Generator Area 440V Load Centers and 114B and 124B Reactor Area 440V Load Centers - 1 Unit l E-20. Rev. 5 - Single Line Meter and Relay Diagram Diesel Generators -

1 and 2 Units i E-28, Rev. 10 - Single Line Meter and Relay Diagram D114, D124, D134,

! D144 Safeguards Load Centers 440 Volt - 1 Unit E-30, Sheet 1, Rev. 17 - Single Line Diagram - Instrumentation Sheet 2, Rev. 9 - AC System - 1 Unit Sheet 3, Rev. 8 E-33, Sheet 1, Rev. 20 - Single Line Meter and Relay Diagram Sheet 2, Rev. 21 - 125/250 VDC System - 1 Unit Sheet 3, Rev. 19 i

Preoperational Test Report IP-2.1, Rev. 0 - 125 Volt Safeguard DC Power Systems Division III and IV Preoperational Test Report IP-2.2, Rev. 0 - 125/250 Volt Safeguard DC Power Systems Divisions I and II Preoperational Test Report 1P-24.1, Rev. 0 - Standby Diesel Generator System 32

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A APPENDIX 2.3 System / Equipment: Primary Containment Isolation System Technical Specifications Reviewed:

3/4.3.2 Isolation Actuation Instrumentation 3/4.6.3 Primary Containment Isolation Values FSAR Sections Reviewed: 5.4, 6.2, 7.3, 7.6, 7.7. 9.3 SER Sections Reviewed: 5.4, 6.2, 7.4, 7.6, 9.3 Procedures Reviewed:

ST-2-076-600(1)-1 ST-2-059-602(3)-1 Preoperational Test 1P-53.1 Preoperational Test 1P-59.1 Preoperational Test IP-25.1 Drawinas and Other References Reviewed:

FCD, 8031-M-1-631-1020-F1.10 Elementary Diagram, B21-1090 E 12.19 Elementary Diagram, C41-1030 F 4.2 P&ID, M-41 Nuclear Boiler System P&ID, M-44 Reactor Water Cleanup System P&ID, M-59FD Primary Containment Instrument Gas System l

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A APPENDIX 2.4 System / Equipment: Standby Gas Treatment System. Reactor Enclosure Recirculation System. Secondary Containrrent Isolation, Valves and Integrity Features Technical Specifications Reviewed:

3/4.6.5.1, 4. 6. 5.1. b and c; 3/4.6.5.2, 4.6.5.2.a. b& c; 3/4.6.5.3, 4.6.5.3.a. b.1; b.3, e and ft 3/4.6.5.4,4.6.5.4.a FSAR Sections Reviewed: 6.1, 6.2, 6.5, 7.3, 9.4, 11.5, 13.5, 15 SER Sections Reviewed: *1.8 *1. 9. 6. 2. 3, 6. 2. 4, 6. 5, 9. 4. 2, 9. 4. 5, 11. 5,

  • 13 *15 Procedures Reviewed:

ST-1-076-321-0 "A" SBGTS Charcoal Absorber & HEPA Filter Test, Rev.0 ST-1-076-200-1 Secondary Containment Isolation Valve Timing Test, Rev.0 ST-1-076-310-1 SGTS Reactor Enclosure Secondary Containment Integrity Test, Rev. 1 ST-6-076-310-1, Reactor Enclosure Secondary Containment Integrity Verification, Rev. O ST-6-076-360-1, Refuel Floor Secondary Containment Verification ST-1-076-250-1, SGTS and RERS Flow Test, Rev.0 ST-2-026-618-1, Reactor Enclosure Exhaust Division 2A Functional Test, Rev.0 Drawinas and Other References Reviewed:

l Reactor Enclosure HVAC P&ID 8031-M76, Sh 1-3 Containment Atmosphere Control P&ID 8031-M57 8h 1, 2 Primary Containment Instrument Gas 8031-M59 Elementary Diagrams 8031-E-341, 343, 357, 365, 470, 472, 474, 475, 478, 482, 483, 484, 609, 610, 624 Architectual Drawings 8031-A-205 thru A-211, A-114 34 l

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.u ,r INSPECTION REPORT 50-352/84-52 ATTACHMENT A l

APPENDIX 2.4 (CONT'D) l HVAC Plan Drawing 8031-M-1059 Functional Control Diagrams - 8031-M76FD,' shl-16, M59FD (including CON 001, PCR 363)

The following observations made during the inspection involved minor procedure discrepancies or pending procedure revisions resulting from the licensee's reviews, ST validation process, or recent plant modifications.

In each case the licensee's actions were either sufficiently planned or complete to be considered acceptable.

ST-1-076-321-0, "A" Charcoal Absorber and HEPA Filter Test, Rev. O, required correction of typographical valve numbering errors.

ST-1-076-200-1 Secondary Containment Automatic Isolation Valve Timing Test, Rev. O, required revision to reflect a recent design change in the operating logic of Containment Instrument Gas System isolation valves.

ST-1-076-360-1, Reactor Enclosure Secondary Containment Integrity Verification, Rev. O, required revision to clarify Reactor Enclosure penetration identification, identify temporary boundary penetration closures, and effect minor editorial corrections.

  • Denotes Supplemental SER Sections also reviewed.

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INSPECTION REPORT 50-352/8'4-52 ATTACHMENT A APPENDIX 2.5 System / Equipment: Primary Containment Suppression Chamber, Drywell and Vacuum Breakers Technical Specifications Reviewed:

, 3/4.6.4.1, 4.6.4.1.b.1, 2 and 3; 3/4.6.2.1, 4.6.2.1.d FSAR Sections Reviewed: 6. 2, 6A, 7. 3, 9. 4, 13. 5, 15 SER Sections Reviewed: *1.8, *1.9, 3.0, 6.2.1, 6.2.2, 6.2.4, 6.5, 7.3, 7.4,

  • 13, *15 Procedures Reviewed:

ST-1-060-730-1 Drywell to Suppression Pool Bypass Leakage Test, Rev. O ST-6-060-760-1 Suppression Pool /Drywell Vacuum Breaker Valve Cycling Test, Rev. 1 ST-2-060-400-1 Containment Systems - Suppression Pool /Drywell Vacuum Breaker Setpoint Check and Channel Calibration, Rev. 0

, ST-6-057-200-1 CAC Valve Test, Rev.0 Drawinas and Other References Reviewed:

Containment Atmosphere Control P&ID 8031-M57, sh 1&2 Functional Control Diagrams 8031-M57Fd, sh 1-4 Elementary Diagrams 8031-E-370, -62d2

  • Denotes Supplemental SER Sections also reviewed i

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INSPECTION REPORT 50-352/84-52 ATTACHMENT A APPENDIX 2.6 System / Equipment: Emergency Service Water, RHR Service Water, Ultimate Heat Sink Technical Specifications Reviewed:

3/4.7.1, 4.7.1.la & b; 3/4.7.1.2, 4.7.1.2.a & b; 3/4.7.1.3 FSAR Sections Reviewed: 3.9, 6.3, 7.3, 7.4, 9.2.2, 9.2.6, 13.5, 15 SER Sections Reviewed: *1. 8, *1. 9, 6.2.2, 7.4, 9.2.1, 9.2.2, 9.2.5, *13,

  • 15 Procedures Reviewed:

ST-6-011-231-0, "A" LOOP RHRSW Pump Valve and Flow Test, Rev. O ST-2-012-600-0 Radiation Monitoring - RHRSW RM Channel "A" Functional Test, Rev.0 ST-6-012-451-0, "A" LOOP RHRSW Lineup Verification, Rev. O ST-6-012-401-0, "A" RHRSW Auto Isolation Test, Rev. O ST-6-011-451-0, "A" LOOP ESW Lineup Verification, Rev. O ST-6-011-401-0, "A" LOOP ESW Valve Auto Actuation Test, Rev. O ST-6-011-231-0, "A" LOOP ESW Pump, Value and Flow Test, Rev. O ST-1-092-111-1 D11 Diesel Generator 4KV Safeguard Loss of Power Test, Rev. 0 1FB-54.1, ESW Flow Balance Test (Data Sheets for Partial Flow Test Only)

S11.1.A (Col-1), Equipment Alignment of ESW Loop "A" System, Rev. 1 (DRAFT)

! S11.1.A (Col-2), Equipment Alignment of ESW Loop "B" System, Rev. 1 l (DRAFT)

S11.1.A (Col-3), Equipment Alignment of ESW control Switches, Rev. 1 (DRAFT)

S11.1. A Alignment of ESW Ssytem, Rev.1 (DRAFT) l S11.8.A Alternate Cooling of RECW Heat Exchangers, Rev. 1 (DRAFT)

S11.8.B. Alternate Cooling of TECW Heat Exhangers, Rev.1 (DRAFT) 37

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INSPECTION REPORT N0. 50-352/84-52 ATTACHMENT A l

1 APPENDIX 2.6 (CONT'D)

S11.8.C. Alternate Cooling of Reactor Recirculation Pump Seal and Motor Coolers Drawings and Other References Reviewed:

ESW P&ID 8031-M-11, Sh 1-3 RHRSW P&ID 8031-M-12 RHR P&ID 8031-M-51 Functional Control Diagrams, 8031-M-12 FD, 8h 1-5, -M11FD, sh 1-4 Elementary Diagrams 8031-E-321 thru 328. E-373 thru E-377 Miscellaneous Observations and Findings:

Several observations made during the inspection involved minor procedure discrepancies or pending procedure revisions resulting from the licensee's reviews, ST validation process, or recent plant modifications.

The licensee's actions were either sufficiently planned or complete to be considered acceptable and questions identified were resolved during the inspection. As noted in Section 2.6 of this report the following observation has been included by Region I in the followup to Unresolved Item 83-19-09:

ST-6-011-451-0, "A" Loop ESW Lineup Verification, Rev. O, is being revised to improve the presentation of Diesel Generator ESW supply and return valve positions to prevent possible valving errors.

This item is closed with respect to this inspection.

  • - Denotes Supplemental SER Sections Also Reviewed l

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