ML20140D575
| ML20140D575 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/29/1986 |
| From: | Beckman D, Imagawa A, Kaufman S PARAMETER, INC. |
| To: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| Shared Package | |
| ML20140D564 | List: |
| References | |
| 50-354-85-64, NUDOCS 8602030059 | |
| Download: ML20140D575 (71) | |
Text
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ENCLOSURE 1 TECHNICAL REVIEW REPORT
, NRC REGION I INSPECTION NO. 50-354/85-64 REVIEW OF HOPE CREEK TECHNICAL SPECIFICATIONS PUBLIC SERVICE ELECTRIC AND GAS NRC DOCKET NO. 50-354 NRC CONTRACT NO. NRC-157-01-003 ONSITE ACTIVITIES CONDUCTED December 2 - 13, 1985 4
Prepared for: Prepared by:
U.S.N.R.C. Region I Parameter, Inc.
631 Park Avenue 13380 Watertown Plank Rd.
King of Prussia, PA 19406 Elm Grove, WI 53122 i
NRC Liaison Personnel: Atitho r s :
J.R. Strosnider D.A. Beckman ,
P.W. Esselgroth A . II Imagawa R.W. Borchardt S. Kaufman I
8602030059 860129 PDR ADOCK05000g4 G
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TABLE OF CONTENTS E
SECTION TITLE PAGE EXECUTIVE
SUMMARY
............................ 1
1.0 INTRODUCTION
................................. 4 1.1 PURPOSE
1.2 BACKGROUND
& GENERAL SCOPE I
1.3 GENERAL EVALUATION CRITERIA a
1.4 GENERAL EVALUATION METHODS 2.0 EVALUATION................................... 7 i 2.1 REACTOR PROTECTION SYSTEM.............. 7 2.2 PRIMARY CONTAINMENT................... 10 INTEGRITY & LEAKAGE ~
~ DRYWELL SUPPRESSION CHABEER 2.3 PRIMARY CONTAINMENT ISOLATION i SYSTEM............................, . 13 2.4 SECONDARY CONTAINMENT...............,<. 15 SECONDARY CONTAINMENT INTEGRITY FITLRATION, RECIRCULATION &
VENTILATION SYSTEM 2.5 SERVICE WATER SYSTEMS................. 17 2.6 REACTOR CORE ISOLATION C00 LING........ 19 2.7 AC POWER SOURCES...................... 22
, . 2.8 DC POWER SOURCES...................... 24 2.9 ONSITE POWER DISTRIBUTION............. 26 2.10 HIGH PRESSURE COOLANT INJECTION AND-AUTOMATIC DEPRESSURIZATION SYSTEMS.... 29 2.11 LOW PRESSURE COOLANT INJECTION........ 32 2.12 CORE SPRAY SYSTEM..................... 35 l 2.13 STANDBY LIQUID CONTROL SYSTEM......... 37 3.0 GENERAL CONCLUSIONS......................... 39 4.0 APPENDICES.................................. 41 PERSONS CONTACTED
- INSPECTION PLAN INSPECTION DATA SHEETS l
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Executive Summary EXECUTIVE
SUMMARY
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Parameter, Inc., under the direction of the Nuclear Regulatory Commission, conducted an inspection at the Hope Creek Nuclear Power Station:
to determine whether the draft Technical Specifications (TS) and the Final Safety Analysis Report (FSAR) are compatible with the as-built plant configuration and operating characteristics; and, to determine whether the draft Technical Specification Requirements are definitively measurable.
The inspection was concentrated on plant systems, structures and components having particular significance with respect to minimizing the severity of potential accidents and accident consequences. The systems evaluated included: the reactor protection and safeguards actuation systems, standby liquid control system, primary and secondary containments and related support systems, emergency core cooling systems, and electrical power systems.
The inspection involved about 300 inspector hours onsite during the period December 2- 13, 1985.
The facility descriptions and operating characteristics for the systems, structures and components found in the PSAR, the NRC Safety Evaluation Report (SER) and the " proof and review" (draft) TS were compared to licensee drawings, procedures and actual plant hardware to establish whether the as-built configuration of the systems, structures and components is compatible with the safety analyses and proposed TS.
Licensee documents reviewed included: Piping and Instrumentation Drawings, Logic Diagrams, Electrical Schematics and One Line Diagrams, Operating and Emergency Procedures, Surveillance and Inservice Test Procedures, Calibration Procedures and data, Maintenance Procedures, Preoperational Test Procedures and data, Administrative Procedures, System Design Specifications and data sheets calculations, and correspondence. In situ plant equipment was visually inspected on a sampling basis to verify that actual installations agreed with the various documents.
Surveillance Procedures were also reviewed to verify that the surveillance methods planned by the licensee were consistent with the requirements of the draft TS and that the proposed TS requirements were definitively measurable or determinable.
At the time of the inspection the draft TS were att11 under development by the licensee in conjunction with the NRC Office Page 1
Executive Summary of Nuclear Reactor Regulation (NRR). The inspection was
, conducted using draft TS promulgated by NRC on September 30, 1985, and using pending TS revisions (T. Bush (PSE&G) Memo, Promulgation of Proof and Review Technicial Specifications -
Draft Revision 1, dated November 26, 1985). These revisions were awaiting submittal to NRR.
The inspection determined that these Technical Specifications were compatible with the as-built systems, structures, and components in the areas inspected and that compliance with the Technical Specifications could be definitively measured or determined.
Because both the TS and the licensee's implementing procedures were still under development, many plant configuration, operating characteristic, and parameter details remained to be firmly established. The licensee's programs for accomplishing this appeared to be functioning satisfactorily.
Several isolated inconsistencies and discrepancies areas were identified with respect to these activities and were presented to the licensee during the inspection. These discrepancies involved incorrect translation of TS requirements into surveillance procedures. Examples include Emergency Diesel Generator Loading requirements, ECCS pump test conditions, and ESF actuation and interlock setpoints. In all cases, the errors were a result of changes in TS since procedure issuance.
In this regard, the licensee's plans included a " final" review of all operating phase procedures to identify and correct such discrepancies. The licensee maintains a computerized system which lists all baseline references used in procedure preparation and which is used to identify procedures affected by changes to their input / reference baseline data. The licensee expected to initiate this review within the next month and complete it prior to receipt of the Operating License.
Additional minor discrepancies involved the efficacy of individual procedures, the omission of precaution or initial condition requirements, and inappropriate instruction steps.
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All discrepancies were either resolved during- the _ inspection or appropriate resolution was identified and initiated by the licensee. No programmatic breakdowns or systemic problems were identified CONCLUSION The Technical Specification preparation process appears to be functioning properly. The licensee is maintaining adequate management control over the process.
Page 2
Executive Summary
. The Technical Specifications and implementing procedures reviewed appear to be compatible with the as-built plant configuration. That information which is still under development for incorporation into the Technical Specifications and implementing procedures appears to be subject to sufficient management control to assure adequate completion of the process.
VM
, %e hem e g W Page 3
INTRODUCTION 1.0 - INTRODUCTION 1.1 - PURPOSE The purpose of this inspection was to assist the Nuclear Re8ulatory Commission in determining that the Hope Creek Nuclear Power Station Technical Specifications were compatible with the as-built configuration of plant systems, structures and components and that the Technical Specification requirements were definitively measurable or determinable.
1.2 - BACKGROUND AND GENERAL SCOPE Startup testing and subsequent plant operation at commercial nuclear power plants has demonstrated that discrepancies sometimes exist between the plant's Technical Specifications (TS), Final Safety Analysis Report (FSAR), Safety Evaluation Report (SER), and as-built plant configuration. During low power physics testing at the Grand Gulf Nuclear Station, Unit 1, -, *
- si8nificant discrepancies of this nature were identified and subsequently corrected.
This inspection was conducted to gain additional assurance that the proposed llope Creek TS are compatible with the assumptions and requirements of the safety evaluations performed and the as-built plant configuration. Parameter, Inc. was requested to assist NRC Region I in performing this inspection at the Hope Creek Site. The inspection involved about 300 inspector hours onsite durinC the period December 2- 13, 1985.
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The general scope of the inspection included:
Report Section 2.1 REACTOR PROTECTION SYSTEM 2.2 PRIMARY CONTAINMENT INTEGRITY & LEAKAGE DRYWELL SUPPRESSION CIIAMBER MSIV SEALING SYSTEM 2.3 PRIMARY CONTAINHENT ISOLATION VALVES 2.4 SECONDARY CONTAINMENT SECONDARY CON.TAINMENT INTEGRITY FILTRATION, RECIRCULATION, VENTILATION SYSTEM 2.5 SERVICE WATER SYSTEMS 2.6 REACTOR CORE ISOLATION COOLING 2.7 AC POWER SOURCES 2.8 DC POWER SOURCES 2.9 ONSITE POWER DISTRIBUTION Page 4
INTRODUCTION 2.10 llIGil PRESSURE COOLANT INJECTION SYSTEM
. AND AUTOMATIC DEPRESSURIZATION SYSTEM 2.11 LOW PRESSURE COOLANT INJECTION SYSTEM 2.12 CORE SPRAY SYSTEM 2.13 STANDBY LIQUID CONTROL SYSTEM The following 8eneral categories of documents were reviewed:
Technical Specifications Final Safety Analysis Report NRC Safety Evaluation Report (with Supplement 3)
Piping and Instrumentation Diagrams (P& ids)
Instrumentation and Control Logic Diagrams Electrical One Line Diagrams Electrical Schematic Diagrams Instrument Loop Drawings Plant General Arrangement & Layout Drawings Preoperational Test Procedures and test data Surveillance Test Procedures Maintenance Procedures Operating Procedures Emergency Operating Procedures Inservice Test Procedures Administrative Procedures Setpoint Calculations Loop Calibration Procedures and data 1.3 - GENERAL EVALUATION CRITERIA The above systems and documentation were reviewed with respect to:
The compatibility of the draft TS with the as-built configuration of the systems, structures and components; The consistency of the draft TS with the documents listed in 1.2 above; The capability to definitively measure or determine compliance with the TS requirements considering both the software and hardware available; and, The adequacy of the licensee's suryeillance and inservice test programs to provide for the implementation of the TS Surveillance Requirements.
1.4 - GENERAL EVALUATION METl!0DS Prior to the onsite inspection activities, the proof and review Page 5
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INTRODUCTION
! Technical Specificactions were reviewed to identify those
- , systems, structures and components which were particularly
, significant with respect to preventing or mitigating the l consequences of analyzed accidents.
( During the onsite inspection activities, the facility
{ descriptions, operating characteristics and related information found in the TS, the FSAR, and the SER were compared to the l licensee documents lisJ,ed in Section 1.2. Concurrently, the TS were evaluated to confirm that the performance criteria and requirement established therein were definitvely measureable or determinable.
Particular emphasis was given to the efficacy of surveillance tests and inservice tests established by the licensee to demonstrate conformance with TS and the requirements of y 10CFR50.55a The detailed inspection plan used to conduct the onsite j activities is provided as Appendix 1 to this report. Key
! evaluation items included:
- 3. Plant drawings were reviewed to establish that the plant j design and construction documents were compatible with the
Preoperational and functional tests were reviewed to verify
!' that the "as tested" system configura tions were consistent with the FSAR, TS, and SER.
Surveillance Tests were reviewed where available to verify
- their conformance with the TS and to establish that the TS requirements could be definitively measured.
] Operating, Emergency, Maintenance, and Inservice Test 1 procedures were reviewed where available to establish their 4
conformance with the TS and accuracy with respect to the l design and construction documents and with the as built j plant.
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EVALUATION 1
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! 2.0 EVALUATION i
GENERAL i In addition to the specific inspection and review items below, the administrative and Emergency Operating Procoedures listed in l Appendix 2.0 were used throughout the inspection for evalution
- of the licensee's programs.
i (NOTE: All procedures listed herein are suffixed by
"(Q)" signifying their applicability to nuclear safety .
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related activities or activities important to safety.)
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h 2.1 -- REACTOR PROTECTION SYSTEM 2.1.1 - Evaluation Criteria and Scope The Reactor Protection System (RPS) monitors the plant parameters below and functions to automatically shutdown the
, reactor by inserting the control rods whenever predetermined
! setpoints are reached:
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a) Neutron flux *
, b) Reactor Vessel (RV) high pressure i c) RV low water level
) d) Main Stop Valve closure j e) Turbine Control Valve fast closure l f) Main Steam Isolation Valve (MSIV) closure g)
Scram Discharge Volume high level c h) Drywell high pressure
- i) Main Steamline high radiation
! Note: Neutron Monitoring System not separately nor
- specifically reviewed.
l In addition to the above, the Reactor Mode Selector Switch
, (SIIUTDOWN position) and the manuni scram pushbutton switches
- will provide manual scram capability. The manual scram 1
pushbutton circuitry is independent from the automatic RPS
. signals.
j The RPS consists of two independent systems, A and B; each l system has two independent reactor shutdown ic channels.
Reactor shutdown logic channels Al and A2 for the lo9'A" system and i
B1 and B2 for the "B" system. Each logic channel receives, as a l minimum, one input signal from the RPS monitored parameters.
i These parameters are measured by at least four independent l Page 7
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- EVALUATION instrument channels.
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) Each shutdown logic channel is arranged in a "one out of two" logic and each reactor shutdown system is arranged in a "one out I of two twice" logic. This arrangement provides testing j capability during reactor operation without shutting down the
! reactor.
The RPS instrumentation and equipment were reviewed'with respect
] to the criteria and methods of Section 1.3 and'1.4 of this
- report. See Appendix 2.1 for a listing of documents reviewed, i
I Proposed TS 3/4.3.1, 3/4.3.2, 3/4.3.3 and 2.2.1 were compared to the documents listed in Appendix 2.1 to verify that the proposed l TS accurately represented the as-built plant configurations and i operating characteristics and were in agreement with the j information in the FSAR and SER.
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{ RPS and Engineering Safety Features (ESP) actuation
! instrumentation and controls were also ~ reviewed in conjunction -
j with the other plant systems as. discussed throughout this j report.
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- 2.1.2 Discussion i
The review of the RPS instrumentation included normal, abnormal, j and emer8ency operations described by the FSAR and the j licensee's draft and approved procedures.
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- The system configuration drawings, operating logic diagrams,
} system operating parameters and limits, surveillance and j preoperational test procedures, operating procedures, and 1 statistical setpoint calculations and calculation methodology l were reviewed on a sampling basis to ensure that the design i features were accurately reflected by the test and operating i
i methods and that these methods were consistent with the '-
requirements of the proposed TS.
The following system features and instrument setpoint j calculation methodology were specifically reviewed:
RPS power supply and distribution system alignments and operations t
Instrumentation elementary and logic diagrams
- Instrument channel operability l Calibration
! Response time
{ Functional Testing i
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EVALUATION Testability of RPS Instrumentation System
. During Operation During Shutdown Instrument setpoint calculation methodology Setpoints Margins & Limits The selective review of the RPS instrumentation drawings, surveillance and operating procedures, and verification of RPS instrumentation setpoints and limits were performed on instrument channel "A". The review indicates that the design features were in agreement with the proposed TS requirements.
The selective review of RPS instrument setpoint calculations and methodology indicates that the actual vendors' data were used where available; and conservative values were used where vendors' data (environmental and seismic) were not available.
The reviewed calculations indicate that there are sufficient margins to meet the requirement setpoints and allowable limits.
Visual inspection of the RPS instrumentation system per se was not performed. However, the independent instrumentation channel separation requirements and the RPS instrumentation and controls in the control room and other plant areas were inspected in
, conjunction with other system observations, e.g. electrical power systems, c o n t a 4.n me n t systems, ECCS systems, etc. These inspections established that the design features were in agreement with the proposed TS.
2.1.3 Observations The following minor inconsistencies were identified. In each case the licensee provided a resolution as noted:
- 1. In OP-SO.SB-001, Reactor Protection System Operation, the listing of RPS scram functions omitted the APRM Downscale Trip of f,5% rated thermal power from the tabulation.
The licensee issued On the Spot Change #P-1 acceptably incorporating the data; a permanent change will be incorporated in the next procedure revision.
- 2. The drywell high pressure setpoint in Table 7.2-1 of the FSAR (1.5 psig) did not agree with the 1.68 psig setpoint of TS Table 2.2.1-1 and Table SB-001 of OP-SO.SB-001, RPS Operation.
i The data in Table 7.2-1 was extraneous. The licensee has issued FSAR Change Notice (CN) #1271 which will delete the Page 9
. EVALUATION l l
setpoint, acceptably correcting the Table.
- 3. The licensee's administrative procedures require appending a
" List of 'Laters'"'to each procedure and annotating the procedure cover with a caution for each procedure in which key information such as setpoints is not yet available. In IC-CC.SB-009, Reactor Protection System Turbine Main Stop Valve Position Switches Calibration, Revision 0, a list of "laters" was not provided for such missing setpoints.
The licensee iIsued On the Spot Change #P-1 which added a standard "later" statement to the procedure cover sheet and an Attachment 2- " List of Laters" in the next revision.
The licensee's actions in this regard were found acceptable.
2.1.4 Conclusions No inconsistencies between the TS, FSAR, SER and the as-built plant were noted during the visual inspection. The as-built system was found in agreement with the various documents reviewed. The TS requirements were found to be definitively measurable.
2.2 _ PRIMARY CONTAINMENT 2.2.1 - Evaluation Criteria and Sc6pe The containment systems include a Mark I primary containment and a domed secondary containment surrounding the primary and housing equipment essential for a safe shutdown in the event of a design basis Loss of Coolant Accident (LOCA) and thereby preventing the release of fission products to the environment in excess of that specified in 10CFR100.
The primary containment is of the pressure suppression type and consists of the drywell, enclosing the reactor vessel, the pressure suppression chamber or torus, and the connecting vent system between the drywell and torus.
The torus also serves as water storage chamber which condenses the steam released as a result of a breach in the Reactor Coolant Pressure Boundary (RCPB) and as the water source for the ECCS Systems.
The primary containment and its related equipment were reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 j of this report. See Appendix 2.2 for a listing of documents Pa8e 10
EVALUATION reviewed.
The review included:
Primary Containment Integrity and Leaka8e Drywell TS and Design Features Suppression Chamber TS and Design Features Primary Containment Isolation System (PCIS) is discussed in Section'2.3 of this report and the Secondary Containment in Section 2.4.
For the primary containment system review, proposed TS 3/4.3.6, 3/4.5.3, and 3/4.6.1 were compared to the documents listed in Appendix 2.2 to verify that the proposed TS accurately represented the as-built plant configuration and operating characteristics and were in agreement with the information in the FSAR and SER.
2.2.2 Discussion The features.of these systems reviewed included normal, abnormal and emergency operations as described by the FSAR, Section 6.2, Containment Systems, and the licensee's draft and approved procedures.
The as-built configuration portion of the review included a sampling based overview of system piping configuration, instrument and control setpoints and operating logic, system operating parameters and electrical control design. The circuits and logic functions of the systems were included in the review.
Operating Procedures, Surveillance and Inservice Tests, and Preoperational Tests were reviewed on a sampling basis to determine that the design features were accurately reflected by the test and operating methods and that these methods were consistent with the requirements of the proposed TS.
Test methods and results of the preoperational tests were used on n samplin8 basis to verify that the system functioned within the parameters of the design drawings and requirements.
Typically, where a preoperational test proved the functions of a logic circuit, the detailed procedures and results were compared with the logic and elementary diagrams to verify that the test accurately reflected the circuits and that the circuits were consistent with the design bases reflected in the FSAR, SER, and TS.
Specifically, the system features and operations involving the following were reviewed:
Normal system alignments and operations Emergency system alignments and operations Page 11
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EVALUATION I
Pressure Suppression Operations 1
, System testing alignments and methods I&C Functional Tests Flow Path Valve Lineups and Operability Testing System Operational Readiness Testin8 A visual inspection of portions of the systems and selected equipment established that the design features were accurately translated into the as-built systems. The visual inspection included verification of system piping, fluid system flow path and component configuration, main control station instrumentation, simulated partial performance of systems ali8 nments and tests, and general comparison of the systems with the proposed TS.
2.2.3 Observations Two minor inconsistencies were identified. In each case, the licensee either provided a resolution or demonstrated that the matter had been previously identified and was in the process of resolution.
- 1. The label on control room instrument SS-PR-4960A3 (10C650E) states " Suppression Chamber Pressure" but should read "Drywell Pressure" according to P&ID M-57-1, Revision 13.
The licensee acknowledged the above and submitted Field Questionnaire (FQ) #0P-598 to correct the labeling.
Correction of the control room labeling of the Drywell pressure instrument is recommended for NRC:RI confirmation.
- 2. The main control board label for Reactor Building - Torus Vacuum Relief Valve HV-5029 (Containment Atmosphere Control System) pushbutton is engraved "HV-502B". The licensee demonstrated that Design Change Package 221 had previously identified this error and that corrective action would be accomplished.
Correction of the labeling error is recommended for NRC:RI confirmation in conjunction with Item 1 above.
2 2.2.4 Conclusions Except as noted above, no discrepancies were identified. The as-built configuration of the systems, structures, and components compared satisfactorily with the documents reviewed.
The Technical Specification requirements were definitively measurable.
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EVALUATION
. 2.3 -- PRIMARY CONTAINNENT ISOLATION SYSTEM 2.3.1 - Evaluation Criteria and Scope Primary containment isolation is initiated when sensors monitoring accident diagnostic parameters trip to initiate closure of the primary containment isolation valves and other isolation functions. There are typically two isolation valves per line. The control circuits are arranged in dual isolation channels so that trip must occur in both logic channels to close an isolation valve. Each logic channel contains at least two independent tripping sensors from each measured variable, only one of which is required to trip a logic channel.
The Primary Containment Isolation System (PCIS) and its related equipment were reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 of this report. See Appendix 2.3 for a listin8 of documents reviewed.
Proposed TS 3/4.6.1 and 3/4.6.3 were compared to the documents listed in Appendix 2.3 to verify that the proposed TS accurately represented the as-built plant configuration and operating characteristics and were in agreement with the information in the FSAR and SER. Particular emphasis was placed on the valve closing times for the automatic isolation valves listed in TS Table 3.6.3.-l.
2.3.2 Discussion The features of these systems reviewed included normal, abnormal and emergency operations as described by the FSAR, Section 6.2, Containment Systems, Section 6.7, Main Steam Isolation Valve (MSIV) Sealing System, and the licensee's draft and approved procedures.
The as-built configuration portion of the review included a sampling based overview of system piping configuration,
, instrument and control setpoints and operating logic, system operating parameters and electrical control design. The circuits and logic functions of the Nuclear Steam Supply Shutoff System and the MSIV Sealing System were included in the review.
Operating Procedures, Survell1ance and Inservice Tests, and Preoperational Tests were reviewed on a sampling basis to determine that the design features were accurately reflected by the test and operating methods and that these methods were consistent with the requirements of the proposed TS.
Test methods and results of the preoperational tests were used Page 13
EVALUATION on a sampling basis to verify that the system functioned within
, the parameters of the design drawings and requirements.
Specifically, the system features and operations involvin8 the following were reviewed:
Normal system ali8nments and operations Emergency system ali 8nments and operations System testing ali 8 nments and methods I&C Functional Tests Flow Path Valve Lineups and Operability Testing System Operational Readiness Testing A visual inspection of portions of the systems and selected equipment established that the design features were accurately translated into the as-built systems. The visual inspection included verification of system piping, fluid system flow path and component configuration, main control station instrumentation, simulated partial performance of system ali8 nments and tests, and general comparison of the systems with the proposed TS.
2.3.3 Observations A -number of minor inconsistencies were identified. In each case, the licensee either provided a resolution or demonstrated that the matter had been previously identified and was in the process of resolution.
- 1. Valve closing times for PCIS automatic isolation valves listed in TS Table 3.6.3-1 and FSAR Table 6.2-16 differed for the values stated for a number of valves.
The licensee acknowledged the above and provided FSAR Change Notice #1040 demonstrating that the TS and FSAR would be brought into agreement. Change Notice #1040 i s ,,t o be incorporated in the FSAR as Amendment 14 and is expected to be issued to NRC in January, 1986. The Change Notice had been formally approved by licensee management and had been forwarded to the facility architect engineer (AE) for processing and printing.
The pending FSAR amendment is subje$t~to detailed review by NRC:NRR. The inspector had no further questions on this matter.
2.3.4 Conclusions Except as noted above, no discrepancies were identified. The as-built configuration of the systems, structures, and components compared satisfactorily with the documents reviewed.
The Technical Specification requirements were definitively Page 14
EVALUATION i
1 measureable.
i 2.4 - SECONDARY CONTAINMENT 2.4.1 - Evaluation Criteria and Scope The secondary containment function is provided by the reactor building and functions to minimize the ground level release of radioactive material during normal and accident operations.
Through the Filtration, Recirculation and Ventilation System (FRVS), it provides the controlled, elevated release of the building atmosphere. The secondary containment also functions as the primary containment / confinement when the drywell is open during refueling or maintenance operations.
The reactor buildin8 encloses the primary containment system and provides fuel storage facilities and other reactor auxiliary and service equipment. The building structure is designed for an internal pressure of 2 psig and can be maintained at a negative pressure of 0.25 inches of water by FRVS to minimize the ground level release of fission products.
3 The secondary containment and its related equipment were reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 of this report. See Appendix 2.4 for a listing of documents reviewed.
Primary Containment Isolation System (PCIS) is discussed in Section 2.3 of this' report and the Primary Containment in Section 2.2.
For the secondary containment system review, proposed TS 3/4.6.5.1, .2, and .3 were compared to the documents listed in Appendix 2.4 to verify that the proposed TS accurately represented the as-built plant confi8uration and operating characteristics and were in agreement with the information in the FSAR and SER.
2.4.2 Discussion The features of these systems reviewed included normal, abnormal and emergency operations as described by the FSAR, Section 6.2, Containment Systems, Section 6.8, Filtration, Recirculation, and -
Ventilation System, and the licensee's draft and approved procedures.
The as-built configuration portion of the review included a sampling based overview of the Reactor Building and containment system features including ductwork configuration, blowout Page 15 l
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- EVALUATION panels, instrument and control setpoints and operating logic, ;
. system operating parameters and electrical control design. The !
circuits and logic functions of the ECCS Actuation System were included in the review.
Operating Procedures, Surveillance and Inservice Tests, and
- Preoperational Tests were reviewed on a sampling basis to determine that the design features were accurately reflected by the test and operating methods and that these methods were consistent with the requirements of the proposed TS.
Test methods and results of the preoperational tests were used on a sampling basis to verify that the system functioned within the parameters of the design drawings and requirements.
Typically, where a preoperational test proved the functions of a logic circuit, the detailed procedures and results were compared with the logic and elementary diagrams to verify that the test accurately reflected the circuits and that the circuits were consistent with the design bases reflected in the FSAR, SER, and TS.
Specifically, the system features and operations involving the following were reviewed:
Normal system alignments and operations Emergency system alignments and operations Controlled, filtered Reactor Building exhaust Post-LOCA Operations System testing alignments and methods I&C Functional Tests Flow Path Lineups and Operability Testing System Operational Readiness Testing A visual inspection of portions of the systems and s e l'd c t e.d equipment established that the design features were accurately translated into the as-built systems. The visual inspection fincluded verification of system piping and ductwork, system flow paTli and component configuration, main control station instrumentation, simulated partial performance of system i
alignments and tests, and general comparison of the systems with the proposed TS.
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2.4.3 Observations A number of minor inconsistencies were identified. In each case, the licensee either provided a resolution or demonstrated that the matter had been previously identified and was in the process of resolution.
- 1. Surveillance Log Procedure OP-DL.XX-026, Revision 0 (Draft),
Attachment la - Form 1, indicates that a number of Drywell Volumetric Average Temperature Readings are recorded to j Page 16
EVALUATION
. obtain an overall average per TS. This data is to be routinely obtained via the plant process computer but is not
. so annotated on the log sheets.
The licensee acknowledged the above and indicated that a suitable editorial change would be made to the log sheets.
- 2. OP-ST.GR-001, Reactor Building Ventilation System Operation, Revision 0, Section 5.3.4, requires three supply and exhaust fans to be in service during refueling operations per TS.
No documentation provisions for implementation of this requirement could be found.
The licensee committed to review and issue documentation requirements as appropriate. NRC:RI confirmation of the licensee's actions in this regard is recommended.
2.4.4 Conclusions Except as noted above, no discrepancies were identified. The as-built configuration of the systems, structures, and components compared satisfactorily with the documents reviewed.
The Technical Specification requirements were definitively measurable.
2.5 - STATION SERVICE WATER AND SAFETY AUIILIARIES COOLING SYSTEMS 2.5.1 - Evaluation Criteria and Scope i
The Station Service Water System (SSWS) removes heat from the Safety and Auxiliaries Cooling System (SACS) and other systems by providing river water to the SACS (and other) heat exchangers and discharging the heated water to the cooling tower discharge canal.
The system consists of two loops with independent pump trains, traveling water screens, screen wash subsystems, strainers, and associated piping, instrumentation and valves.
The SACS system is a closed loop cooling system consisting of two trains. The system is a complement of a connected non-safety related Turbine Auxiliaries Cooling System (TACS).
SACS provides cooling to ESF equipment including the RHR heat exchangers.
Both systems were reviewed per the criteria and methods of Sections 1.3 and 1.4 of this report. See Appendix 2.5 for a listing of documents reviewed.
Page 17
EVALUATION
. Proposed TS 3/4.7.1.1, .2, .3, and .4 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 were in agreement with the information in the FSAR and SER.
2.5.2 - Discussion The features of these systems reviewed included normal, abnormal, and emergency operations described by the FSAR, Section 15, Accident Analysis, and the licensee's draft and approved procedures.
The as-built configuration portion of the review included a sampling based review of system piping configuration, instrumentation and control setpoints and operating logic, system operating parameters and limits.
Operating Procedures, Surveillance and Inservice Tests were reviewed determine that the design features were accurately reflected by the test and operating methods and that these methods were consistent with the requirements of the proposed TS.
The Preoperational Test for the Safety & Turbine Auxiliaries Cooling was reviewed on a sampling on a sampling basis to establish that the system functioned as portrayed by the design drawings and requirements.
3 Specifically, the system features and operations involving the following were reviewed:
Normal system alignments and operations Emergency system alignments and operations System testing alignments and methods Flow Path Valve Lineups and Operability Testing System Operational Readiness Testing Pump and Valve Inservice and Operability Testing 2.5.3 - Observations Minor inconsistencies were identified and resolved by the licensee during the inspection. All inspector questions were acceptably resolved.
Page 18
EVALUATION 2.5.4 Conclusions No discrepancies were identified. The as-built configuration of the system, structures, and components compared satisfactorily with the documents reviewed. The Technical Specification requirements reviewed were definitively measurable.
2.6 -- REACTOR CORE ISOLATION COOLING SYSTEM 2.6.1 - Evaluation Criteria and Scope The Reactor Core Isolation Cooling (RCIC) System consists of a
, turbine, pump, piping and valves, and instrumentation designed to maintain sufficient reactor water level inventory to ensure i
the continuity of core cooling.
System provide the means to inject water to the core when the reactor is isolated or during a small break Loss of Coolant Accident (LOCA).
This system and related equipment were reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 of this report.
See Appendix 2.6 for a listing of documents reviewed.
Proposed TS 3/4.7.4, 3/4.3.2, 3/4.3.5, and 3/4.7.3.7.4 were compared to the documents listed in Appendix 2.6 to verify that the proposed TS accurately represented the as-built plant
, configuration and operating characteristics and were in j agreement with the information in the FSAR and SER.
- 2.6.2 - Discussion The features of these systems reviewed included normal, abnormal, and emergency operations described by the FSAR, Section 15, Accident Analysis, and the licensee's draft and approved procedures. -
The as-built configuration portion of the review included a sampling based review of system piping configuration, instrumentation and control setpoints and operating logic, j system operating parameters and limits, and electrical controls design. The circuits and logic functions of the-RCIC Actuation Instrumentation and RCIC Isolation Instrumentation were included in the review.
Operating Procedures, Surveillance and Inservice Tests, and Preoperational Tests were reviewed on a sampling basis to
' determine that the design features were accurately reflected by the test and operating methods and that these methods were Page 19
EVALUATION consistent with the requirements of the proposed TS.
The test methods and results of the preoperational tests were used on a sampling basis to establish that the system functioned as portrayed by the design drawings and requirements. For example, where a preoperational test verified the functions of a logic element, the detailed test methods and results were compared with the logic and elementary diagrams to establish that the test accurately reflected the circuits and that the circuits were consistent with the design basis reflected in the FSAR, SER, and TS.
The licensee's planned startup (post-criticality) testing was also reviewed to establish that the design requirements requiring reactor steam for verification will be tested.
Specifically, the system features and operations involving the following were, reviewed:
Normal sybtem alignments and operations Emergency system alignments and operations System testing alignments and methods I&C Calibrations I&C Functional Tests Flow Path Valve Lineups and Operability Testing System Operational Readiness Testing Pump and Valve Inservice and Operability Testing A visual inspection of portions of the systems and selected equipment established that the design features were accurately translated into the as-built systems. The visual inspection included verification of system piping and fluid system flowpath and component configuration, main and auxiliary control station instrumentation and controls, simulated partial performance of partial system alignments, and general comparison of the systems and equipment with the proposed TS.
2.6.3 - Observations A number of minor inconsistencies were identified. In each case, the licensee either provided a resolution or demonstrated that the matter had been previously identified and was in the process of resolution.
- 1. Preoperational Test PTP-BD-1, RCIC System Preoperational Test, Revision 0, verified that the High Steam Flow Isolation Time Delay relay was set to 2.9 - 3.1 seconds instead of the 3 - 13 second criteria of TS Table 3.3.3-2.
One of the two timers was reset from an acceptable 5.09 seconds to an unacceptable 2.9 seconds during PTP-BD-1 to meet the test specification. The second timer had an as found setting of 3.1 seconds and was not reset during the Page 20
e EVALUATION test. (See Section 2.10 of this report for a similar
. finding with regard to the High Pressure Coolant Injection System.)
The philosophy of the time delay function and the desirability to have the time delay set higher in the acceptable range to avoid spurious RCIC pump trips on startup steam flow surges was discussed with the licensee.
The licensee acknowledged the above and demonstrated that the PTP had been written in accordance with the NSSS preoperational test specification and that the criterion had changed as the TS were developed. The licensee advised that the time delay relay would be reset in accordance with the surveillance procedures for calibration and functional testing of the logic. This will be accomplished prior to the equipment being required to be operable per TS.
Resetting of the HPCI and RCIC High Steam Flow Isolation Time Delay relay is recommended for confirmation by NRC during a future inspection. (
- 2. TS 4.7.4.b requirew the RCIC pump functional test and flow verification to be performed with a reactor steam supply pressure and pump back pressure of 1000, +20, -0 psig.
OP-ST.BD-001, RCIC Pump OP203 Inservice Test, Revision 0, did not address this test condition in the body of the procedure or acceptance criteria.
The licensee advised the the acceptance criteria and the above test condition would be included in the IST baseline and test data sheet which would be appended _to the procedure when baseline data was eventually taken. 'This data sheet is intended to " stand alone" for each procedure, permitting the data sheet to be revised for new data or limits without requiring revision of the procedure body.
This finding and proposed licensee resolution is similar to that discussed in Sections 2.11 (Low Pressure Coolant Injection (LPCI)) and 2.12 (Core Spray System (CSS)) of this report.
Verification of the licensee's incorporation of the above TS requirement into the RCIC, LPCI and CSS pump test procedures is recommended for fallowup by NRC:RI.
- 3. FSAR Section 5.4.6.2.2.2.1, Valve Operation Requirements, specifies valve stroke rates (inches / minute) for key system valves under full system pressure or flow conditions.
Examples include the pump steam supply and exhaust valves, pump discharge valves, pump mini-flow valves, etc.
Page 21
EVALUATION 4
These requirements were not verified during preoperational
, testing and would not be specifically determined during startup testing. Similar examples were found for the HPCI system. Such test was done, however, for the RHR system.
The licensee advised that the test procedures were developed in accordance with NSSS Vendor preoperational and startup test specifications (See listing, Section 2, this report) which did not specify individual valve testing. Further, the overall system response time, including the aggregate valve operating times were tested during the startup phase, verifying overall system acceptability under design operating conditions.
This licensee approach was discussed with NRC:RI management and found acceptable.
2.6.4 Conclusions Except as noted above no discrepancies were identified. The as-built configuration of the system, structures, and components
- compared satisfactorily with the documents reviewed. The s Tec.hnical Specification requirements reviewed were definitively measurable.
2.7 - AC POWER SOURCES
%. 2.7.1 - Evaluation Criteria and Scope The AC Power Sources consist of two sources of 500 KV offsite
{ power to the 13.8 KV yard ring bus, which supplies the Class 1E I AC power distribution system at 4.16 KV, 480 V and 208/120 V.
independent channels. Each channel has 4.16 KV normal and alternate power sources and a dedicated 4.16 KV emergency diesel generator (EDG) with complete auxiliary systems such as fuel and lube oil, starting air, and cooling systems. The EDG serves as the emergency electric power sources in case both the normal and alternate power sources to the channel are lost.
The AC Power Sources and their associated equipment were reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 of this report. See Appendix 2'. 7 for a listing of documents reviewed.
Proposed TS 3/4.8.1.1. and 3/4.8.1.2 were compared to the documents listed in Appendix 2.7 to verify that the TS accurately represent the as-built plant configuration and operating characteristics and were in agreement with the Page 22
EVALUATION information in the FSAR and SER.
2.7.2 Discussion The review of the systems and equipment included the normal, abnormal and emergency operations described by the FSAR and the licensee's draft and approved procedures.
The system confi8 uration drawings, operating logic diagrams, system operating parameters and limits, surveillance and preoperational test procedures, operating procedures and equipment technical manuals were reviewed on a sampling basis to ensure that the design features were accurately reflected by the test and operating methods and that these methods were consistent with the requirements of the proposed TS.
- The folTowing system features and operations were specifically reviewed:
Normal and abnormal system ali 8nments and operations Emergency System alignments and operations During loss of offsite power (LOP)
During LOCA without LOP During LOCA with LOP System testing alignments and methods System equipment operability testing Load Sequencer Emergency Diesel Generators EDG Auxiliary Systems A selective visual inspection of the systems and equipment established that the design features were accurately translated into as-built systems. The visual inspection also verified that the system configuration, equipment and bus arrangement (four independent channels), main control room and local station instrumentation and controls, and system operability maintenance were in agreement with the proposed TS requirements.
2.7.3 Observations l
t The following minor discrepancies were identified. In each case, the licensee provided a resolution as noted.
~~
- 1. In TS 3/4.8.1 - AC Sources, Section 4.8.1.1.2.h.9, the continuous rating of the EDG was shown as 4737 KW. The l correct value is 4430 KW.
The licensee initiated TS Change Request (CR) #143 which, as reviewed, would acceptably correct the above in the next revision of the TS. The next planned TS revision is expected during January, 1986; NRC:RI followup of this item Page 23 I _ _ _ _ _ . . _ _ _ _
---an. --,A-a- 3 .M - m a-+--e -N= n* A 4*- e +-=m- -.- A ----ma'e4 +*e- un d 2 wh s EVALUATION i
i i
t is recommended.
- 2. In OP-ST.KJ-005, Integrated Emergency Diesel Generator Test 1AG400 - 18 Months, Revisio'n 0:
j a) The EDG 10%-2 hour overload limit (4873 KW) was omitted j from the procedure Limitations, Section 3.2.2.
b) ~
The erroneous 4737 KW continuous rating (Item 1 above) i was included throughout the procedure.
i The licensee issued OSC #P-2 acceptably correcting the
, above; permanent procedure chan8es will be made during the
- next procedure revision.
- 2.7.4 Conclusions
^ ~
No in c o'n's i s t e n c i e s were noted during the visual inspection.
1 Except as noted above, the as-built system is in agreement with 4
the documents reviewed and the TS requirement"S 'were definitively measurable 2.8 ,DC POWER. SOURCES 2.8.1 - Evaluation Criteria and Scope The Class 1E DC Power Sources consist of six Class 1E 125 VDC station batteries, two Class 1E 250 VDC station batteries and j their respective battery chargers. The 125 VDC station batteries 1AD411, IBD411, ICD 411, and IDD411 with their associated chargers provide four independent power sources to
. their corresponding Class 1E 125 VDC distribution system which j also supply the Class 1E 120 VAC distribution system via j inverters.
The 125 VDC station batteries ICD 447 and IDD447 provide power 9
sources for the fire protection and security systems. The 250 i VDC station batteries 10D421 and 10D431 provide power sources for the High Pressure Coolant Injection (HPCI) System and for the Reactor Core Isolation Cooling (RCIC) System, respectively.
, The DC Power Sources and their associated equipment were reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 of this report. S~ee Appendix 2.8 for a listing of documents reviewed. '
! Proposed TS 3/4.8.2.1, and 3/4.8.2.2 were compared to the 1
documents listed in Appendix 2.8 to verify that the TS accurately represent the as-built plant configuration and operatin8 characteristics and were la agreement with the j Page 24 s
1
- - - --._. .,,-.. _ . - . _ - - ~ - ~ , - . _ - . . , - - , - . . . - - - - - , , _ - - , . , . - - - - - _- - - - ._ , _ . . _ . _ . . - - . . - - , - . , .
EVALUATION information in the FSAR and SER.
O 2.8.2 Discussion The review of the systems and equipment included the normal, j abnormal and emergency operations described by the FSAR and the licensee's draft and approved procedures.
l The system configuration drawings, operatin8 logic diagrams, system operating parameters and limits, surveillance and j preoperational test procedures, operating procedures and equipment technical manuals were reviewed on a sampling basis to ensure that the design features were accurately reflected by the test and operating methods and that these methods were consistent with the requirements of the proposed TS.
The following system features and operations were specifically reviewed: -
Equipment Ratings Independence of redundant power sources Normal, abnormal and emergency system a alignments and operations System testing alignments and methods System equipment operability testing Batteries
- Battery Chargers i
System ventilat*nn requirements A selective visual inspection of the systems and equipment i established that the design features were accurately translated into as-built systems. The visual inspection also verified that the system configuration, equipment and bus arran8ement (four independent channels), main control room and local station instrumentation and controls, and system operability maintenance were in agreement with the proposed TS requirements.
2.8.3 Observations j
. The following minor discrepancy was identified. The licensee provided a resolution as noted.
- 1. In MD-ST.PK-002, 125 VDC Quarterly Battery Surveillance Test, Revision 0,: '
a) In Section 1.0, a typo 8raphical error identified the system as 250 VDC instead of 125 VDC.
b) In Section 3.0, Precautions and Limitations, the battery room ventilation requirements and hydrogen accumulation Page 25 1
,-. ._~ - ,. - . _ _ _ _ - - . - _ _ _ _ _ . - - _ - _ = _ - - - . _ - - - , _ . _ _ - . - . _ - _ _ . . - _ _ _ - _ _ _
i EVALUATION i limits included in the system operating procedure and
, other surveillance procedures was omitted.
The licensee committed to review the above and make appropriate corrections in the next routine procedure i
revision. NRC:RI confirmation of this action is
- recommended.
2.8.4 Conclusions No inconsistencies were noted durin8 the visual inspection. The i
as-built system is in agreement with the documents reviewed and the TS requirements were definitively measurable 2.9 -- ONSITE POWER DISTRIBUTION SYSTEM
, .c 2.9.1 - Evaluation Criteria and Scope The onsite Class 1E electrical power distribution system consists of four independent channels of 4.16 KVAC, 480 VAC, 208/120 VAC, 250 VDC, and 125 VDC distribution systems. The i normal and alternate power sources for the AC, system are supplied from the 13.8 KV yard ring bus to the 4.16 KV busses (See Section 2.7).
l The station batteries supply power to the DC system which also supplies the Class 1E 120 VAC distribution system via inverters (See Section 2.8).
l The Class IE power system supplies all Class 1E loads that are a needed for safe and orderly shutdown of the reactor, maintaining the plant in a safe shutdown condition, and mitigatin8 the consequences of an accident 4
, The system and associated equipment were reviewed with respect l to the criteria and methods of Sections 1.3 and 1.4 of this report. See Appendix 2.9 for a listin8 of documents reviewed.
I '
, Proposed TS 3/4.8.3 was compared to the documents-listed in j Appendix 2.9 to verify that the TS accurately represent the j as-built plant confi8 uration and operating characteristics and were in agreement with the information in the FSAR and SER.
l 2.9.2 Discussion i
i The review of the systems and equipment included the normal,
- abnormal and emergency operations described by the FSAR and the
! Page 26 ;
i
EVALUATION licensee's draft and approved procedures.
The system configuration drawings, operating and alarm logic diagrams, circuit breaker schematic diagrams, system operating parameters and limits, surveillance and preoperational test procedures, operating procedures and equipment technical manuals were reviewed on a sampling basis to ensure that the design features were accurately reflected by the test and operating methods and that these methods were consistent with the requirements of the proposed TS.
The following system features and operations were specifically reviewed:
Normal and abnormal ali 8 nments and operations
_ Emergency system ali8 nments and operations During loss of offsite power (LOP)
During LOCA without LOP During LOCA with LOP System testing alignments and methods System operability testing System overload protection schemes Penetration conductors
,,, MOV thermal overloads Bypassed Not Bypassed Circuit Breakers Tripped by LOCA signals Primary and Backup Reactor Protection System (RPS) Power Supply System A selective visual inspection of the systems and equipment established that the desi8n features were accurately translated into as-built systems. The visual inspection also verified that the system configuration, equipment and bus arrangement (four independent channels), main control room and local station instrumentation and controls, and system operability maintenance were in agreement with the proposed TS requirements.
2.9.3 Observations The following minor discrepancies were identified. The licensee provided resolutions as noted.
- 1. The voltage trip setpoints for the RPS power supplied identified in FSAR Section 8.3.1.5.2 do not agree with the voltage trip setpoints identified in TS 4.8.4.4. The FSAR trip setpoints are based upon the RPS scram solenoid valve voltage ratings (115 VAC 110%) while the TS trip setpoints are based on 120 VAC 110%.
Page 27
EVALUATION This inspection confirmed that the actual nameplate rating
, for the scram solenoid valves is 115 VAC. Operation of the system at the TS limits may exceed the range of satisfactory operation for the scram solenoids, i.e. result in overheating at the high limit voltage or chattering at the low limit voltage.
The licensee stated that the actual voltage trip setpoints would be established based on the results of preoperational test, PTP-SB-1, Revision 0. That test will determine the actual voltage on the RPS bus based on the actual voltage drops for all loads on the RPS bus. Appropriate FSAR, TS and procedure changes would then be made.
This item was also reviewed and will be tracked for licensee completion by a concurrent NRC inspection.
- 2. In OP-SO.PB-001, 4.16 KV System Operation, Revision 0:
a) Sections 5.8.3.2, 5.8.3.3, 5.9.4, and 5.9.5 provide instructions for removing breaker / controller control power fuses but instruct the operator to place the operating switch in off after the fuses are removed.
The switches should be placed in off prior to fuse removal to prevent inadvertent automatic or remote operation during fuse removal to avoid high interruption currents and potential personnel hazards, b) Typographical errors in both Sections 2.1.2 and 2.4.2 incorrectly refering the operator to Attachment 2 should read Attachment 1.~
c) Typographical errors in Section 2.1.5 identifying the non-Class 1E lockout relays list as Attachment 3 should read Attachment 2.
The licensee advised that the above changes would be incorporated into the next routine revision of the procedure. NRC:RI followup of the licensee actions is
~
recommended.
2.9.4 Conclusions No inconsistencies were noted during the visual- in s pec tio n .
Except as noted above, the as-built system is in agreement with the documents reviewed and the TS requirements were definitively measurable Page 28
EVALUATION 2.10 _ HIGH PRESSURE COOLANT INJECTION SYSTEM AND AUTOMATIC
. DEPRESSURIZATION SYSTEM 2.10.1 - Evaluation Criteria and Scope The High Pressure Coolant Injection (HPCI) System provide the means to inject water to the core during a Loss of Coolant Accident (LOCA). The system consists of a turbine driven pump powered by reactor steam and taking suction from either the Condensate Storage Tank or Suppression Pool.
The Automatic Depressurization System will reduce reactor pressure upon indication of a design basis accident and failure of HPCI to permit injection to the reactor core by the low pressure ECCS systems (Core Spray and Low Pressure Coolant Injection).
These systems and their related equipment were reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 of this report. See Appendix 2.10 for a listing of documents reviewed.
Proposed TS 3/4.5.1., 3/4.5.2, 3/4.3.2, 3/4.3.2, were compared to the documents listed in Appendix 2.10 to verify that the proposed TS accurately represented the as-built plant configuration and operating characteristics and were in agreement with the information in the FSAR and SER.
2.10.2 - Discussion The features of these systems reviewed included normal, abnormal,-and emergency operations described by the FSAR, Section 15, Accident Analysis, and the licensee's draft and approved procedures.
The as-built configuration portion of the review included a sampling based review of system piping configuration, instrumentation and control setpoints and operating logic, system operating parameters and limits, and electrical controls
-design. The circuits and logic functions of the ECCS Actuation Instrumentation and HPCI Isolation Instrumentation were included in the review.
Operating Procedures, Surveillance and Inservice Tests, and Preoperational Tests were reviewed on a sampling basis to determine that the design features were accurately reflected by the test and operating methods and that these methods were consistent with the requirements of the proposed TS.
The test methods and results of the preoperational tests were Page 29
EVALUATION used on a sampling basis to establish that the system functioned
. as portrayed by the design drawings and requirements. For example, where a preoperational test verified the functions of a logic element, the detailed test methods and results were compared with the logic and elementary diagrams to establish that the test accurately reflected the circuits and that the circuits were consistent with the design basis reflected in the FSAR, SER, and TS.
Specifically, the system features and operations involving the following were reviewed:
Normal system ali8nments and operations Emergency system alignments and operations ECCS Injection Phase Operations ECCS Recirculation Phase Operations System testing alignments and methods I&C Calibrations I&C Functional Tests Flow Path Valve Lineups and Operability Testing System Operational Readiness Testing Pump and Valve Inservice and Operability Testing A visual inspection of portions of the systems and selected equipment established that the design features were accurately translated into the as-built systems. The visual inspection included verification of system piping and fluid system flowpath and component configuration, main and auxiliary control station instrumentation and controls, simulated partial performance of partial system alignments and tests, and general comparison of the systems and equipment with the proposed TS.
2.10.3 - Observations A number of minor inconsistencies were identified. In each case, the licensee either provided a r e s o lit t i o n or demonstrated that the matter had been previously identified and was in the process of resolution.
l 1. Preoperational Test PTP-BJ-1, ilPCI System Test, Revision 0, l
yerified that the Iligh Steam Flow Isolation Time Delay relay was set to 2.9 - 3.1 seconds instead of the 3 - 13 second criteria of TS 3.1.3. See Section 2.6 of this report for a similar finding with regard to the Reactor Core Isolation Cooling System.
The licensee acknowledged the above and demonstrated that the PTP had been written in accordance with the NSSS preoperational test specification and that the criterion had changed as the TS were developed. The licensee advised that the time delay relay would be reset in accordance with the l
Page 30
EVALUATION surveillance procedures for calibration and functional
, testing of the logic. This will be accomplished prior to the equipment being required to be operable per TS.
Resetting of the IIPCI and RCIC High Steam Flow Isolation Time Delay relay is recommended for confirmation by NRC durin8 a future inspection.
- 2. I&C Functional Test IC-FT.BJ-001, HPCI Division I Functional Test, Low CST Suction Switchover, contained acceptance criteria for CST level inconsistent with TS 4.3.1. The TS provided an nominal setpoint of 2.76% and a maximum allowable setpoint of 3.6% of level span. The procedure's acceptance criteria were 2.96 inches of water (column) and 3.6 inches of water (column).
The licensee provided Form SA-AP.XX-042-1, Field Questionnaire dated 11/7/85 which had been forwarded to site engineering identifying and requestin8 resolution of the discrepancy. Accompanying meetin8 notes of the Setpoint Calculation Group / Operations Interface Meeting of 11/21/85 3 indicated that a TS change would be processed to correct the units to " inches" and that the procedure would be changed to correct the actual value.
Correction of TS 4.3.1 units for Low CST Level and correction of IC-FT.BJ-001 setpoin t limits are recommended for NRC verification.
- 3. OP-ST.BJ-002, llPCI System Functional Test, implementa TS 4.5.1.c.2.a for head / flow testing. The TS requires a reactor steam pressure and pump back pressure of 200-215 psig for the test.
The procedure also performs valve actuation verification testing, a'nd IIPCI sys t em response time testing. Although this criteria for reactor steam pressure is included in the initial conditions, the action steps applicable to the head / flow test begin about halfway through the procedure.
Neither the action steps nor the data sheets and included the requirements.
Similarly, the procedure was silent with respect to the pump back pressure requirements.
On the Spot Change #P-3 was issued on December 12, 1985 to acceptably revise the procedure, include the acceptance criteria and make it consistent with the TS.
- 4. The ADS lii 8h Drywell Pressure Bypass Timer provides for actuation of the ADS system after a time delay in the absence of a high drywell pressure si8nal. OP-SO.SN-001, Page 31
EVALUATION Nuclear Pressure Relief and Automatic Depressurization
. System Operation, Section 3.3, included a setpoint of 8 minutes for the ADS High Drywell Pressure Bypass Timer. FSAR Table 6.3-2 showed the same setpoint as 6 minutes. TSs include a setpoint of 5.0-5.5 minutes.
The licensee demonstrated that the TS value was correct, that the FSAR value was NSSS vendor " analytical" value which represented a maximum value; and that the procedure value required correction.
On the Spot Chan8e # P-1 was issued on December 11, 1985 to acceptably correct the above procedure. Review of related surveillance, functional test and calibration procedures found the limits acceptably reflected.
2.10.4 Conclusions Except as noted above no discrepancies were identified. The as-built configuration of the system, structures, and components compared satisfactorily with the documents reviewed. The Technical Specification requirements reviewed were definitively measurable.
2.11 - LOW PRESSURE COOLANT INJECTION MODE OF THE RESIDUAL HEAT REMOVAL SYSTEM 2.11.1 - Evaluation Criteria and Scope The Low Pressure Coolant Injection (LPCI) System is an operating mode of the Residual Heat Removal (RHR) system and provides the means to inject high volume, low pressure water to the core during a Loss of Coolant Accident (LOCA). The system consists of four motor driven pumps drawing suction from the Suppression Pool and discharging to the core via four RPV nozzles.
This system and the related equipment was reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 of this report. See Appendix 2.11 for a listing of documents reviewed.
Proposed TS 3/4.5.1., 3/4.5.2, 3/4.3.2, 3/4.3.3, were compared to the documents listed in Appendix 2.11 to verify that the proposed TS accurately represented the as-built plant configuration and operating characteristics and were in agreement with the information in the FSAR and SER.
2.11.2 - Discussion Page 32
EVALUATION The features of these systems reviewed included normal,
, abnormal, and emergency operations described by the FSAR, Section 15, Accident Analysis, and the licensee's draft and approved procedures.
The as-built configuration portion of the review included a sampling based review of system piping configuration, instrumentation and control setpoints and operating logic, system operating parameters and limits, and electrical controls design. The circuits and logic functions of the ECCS Actuation Instrumentation were included in the review; a specific review of reactor vessel level trips and LPCI actuation signals was performed.
Operating Procedures, Surveillance and Inservice Tests, and Preoperational Tests listed in Appendix 2.11 were reviewed on a
_ sampling basis to determine that the design features were accurately reflected by the test and operating methods and that y these methods were consistent with the requirements of the proposed TS.
The test methods and results of the preoperational tests were used on a sampling basis to establish that the system functioned as portrayed by the design drawings and requirements, a
Specifically, the system features and opera tions involving the
- following were reviewed
- Normal system alignments and operations Emergency system alignments and operations ECCS Injection Phase Operations
, ECCS Recirculation Phase Operations System testing alignments and methods I&C Calibrations System Functional Tests Flow Path Valve Lineups and Operability Testing Pump and Valve Inservice and Operability Testing A visual inspection of portions of the systems and selected equipment established that the desi8n features were accurately translated into the as-built systems. The visual inspection
- included verification of system piping and fluid system flowpath and component configuration, main and auxiliary control station instrumentation and controls, and general comparison of the l systems and equipnent with the proposed TS.
2.11.3 - Observations t
A number of minor inconsistencies were identified and were resolved by the licensee during the inspection. Only the following item required corrective action involving document Page 33
1 1
EVALUATION i d
revision.
i
- 1. OP-SO.BC-001, RHR Operating Procedure, Section 5.2.1.3, LPCI Initiation Observation, included an incorrect value for the RHR Loop Injection MOV (HV-F017A, B, C, and D) valve openin8 Permissive reactor pressure interlock setpoint value.
TS Table 3.3.3-2, ECCS Actuation Instrumentation Setpoints, requires the above valve opening permissive setpoints to be 1450 psig (decreasing). The operating procedure stated that the valves would open on permissive at <460 psig.
When the licensee was advised of the above, On the Spot Change #P-1 was issued on December 13, 1985 to acceptably correct.
- 2. "TS 4.5.1.b.2 requires the LPCI pump functional test and flow verification to be performed against a line pressure corresponding to a pump back pressure of 120 psid.
OP-IS.BC-003 and -004, RHR Pump Inservice Tests, Revision 0, did not address this test condition in the body of the procedure or acceptance criteria.
The licensee advised the the acceptance criteria and the above test condition would be included in the IST baseline and test data sheet which would be appended to the procedure when baseline data was eventually taken. This data sheet is intended to " stand alone" for each procedure, permitting the data sheet to be revised for new data or limits without requiring revision of the procedure body.
- This finding and proposed licensee resolution is similar to that discussed in Secticas 2.6 (RCIC) and 2.12 (Core Spray System (CSS)) of this report.
Incorporation of the above limits into the RCIC, LPCI and CSS pump test procedures is recommended for followup b 3 NRC:RI.
2.11.4 Conclusions
- No significant discrepancies were identified. The as-built
! configuration of the system, structures, and components compared satisfactorily with the documents reviewed. The Technical i Specification requirements reviewed were definitively I
measurable.
l
! Page 34 4
- , m . , - , ---,,_,n- , _ - -
,r,--,.- - - -,+--e--
EVALUATION
. l 2.12 _ CORE SPRAY SYSTEM 2.12.1 - Evaluation Criteria and Scope The Core Spray (CSS) System functions to spray water from the Suppression Chamber at high volume and low pressure directly to the core during a Loss of Coolant Accident (LOCA). The system consists of four motor driven pumps drawing suction from the
- Suppression Pool and discharging to the reactor vessel via j sparger nozzles immediately above the core.
- This system and the related equipment was reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 of this report. See Appendix 2.12 for a listing of documents reviewed.
, Proposed TS 3/4.5.1., 3/4.5.2, 3/4.3.2, 3/4.3.3, were compared
! to the documents listed in Appendix 2.12 to verify that the j proposed TS accurately represented the as-built plant configuration and operating characteristics and were in agreement with the information in the FSAR and SER.
2.12.2 - Discussion The features of these systems reviewed included normal, abnormal, and emergency operations described by the FSAR, Section 15, Accident Analysis, and the licensee's draft and approved procedures.
4 The as-built configuration portion of the review included a sampling based review of system piping configuration, instrumentation and control setpoints and operating logic, system operating parameters and limits, and electrical controls design. The circuits and logic functions of the ECCS Actuation Instrumentation were included in the review; a specific review of reactor vessel level trips, CSS actuation signals, individual pumps and valve actuation logics, and system time response was performed.
! Operating Procedures, Surveillance and Inservice Tests, and Preoperational Tents listed in Appendix 2.12 were reviewed on a 4 sampling basis to determine that the design features were accurately reflected by the test and operating methods and that these methods were consistent with the requirements of the 2
proposed TS.
I i The test methods and results of the preoperational tests were used on a sampling basis to establish that the system functioned i as portrayed by the design drawings and requirements.
I Specifically, the system features and operations involving the
, Page 35 i
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EVALUATION following were reviewed:
Normal system alignments and operations 4
Emergency system alignments and operations ECCS Injection Phase Operations ECCS Recirculation Phase Operations System testing alignments and methods System Functional Tests Flow Path Valve Lineups and Operability Testing Pump and Valve Inservice and Operability Testin8 I
A visual inspection of portions of the systems and selected equipment established that the design features were accurately translated into the as-built systems. The visual inspection included verification of system piping and fluid system flowpath and component configuration, main and auxiliary control station instrumentation and controls, and general comparison of the systems and equipment with the proposed TS.
2.12.3 - Observations Minor inconsistencies were identified. All inspector questions were resolved by the licensee during the inspection except as noted below.
- 1. TS 4.5.1.b.1 requires the CSS pump functional test and flow verification to be performed against a test line pressure corresponding to a pump back pressure of 1105 psid.
OP-IS.BE-001 and -002, CSS Pump Inservice Tests, Revision 0, did not address this test condition in the body of the procedure or acceptance criteria.
l The licensee advised the the acceptance criteria and the above test condition would be included in the IST baseline and test data sheet which would be appended to the procedure j uhen baseline data was eventually taken. This data sheet is intended to " stand alone" for each procedure, permitting the data sheet to be revised for new data or limits without requiring revision of the procedure body.
This finding and proposed licensee resolution is similar to that discussed in Sections 2.11 (Low Pressure Coolant Injection (LPCI)) and 2.6 (RCIC) of this report.
t
! Verification of the licensee's incorporation of the above TS l requirement into the RCIC, LPCI and CSS pump test procedures is recommended for followup by NRC:RI.
Page 36 l
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EVALUATION 2.12.4 Conclusions No significant discrepancies were identified. The as-built confi 8 uration of the system, structures, and components compared
, satisfactorily with the documents reviewed. The Technical Specification requirements reviewed were definitively measurable.
2.13 -- STANDBY LIQUID CONTROL SYSTEM 2.13.1 - Evaluation Criteria and Scope The Standby Liquid Control (SBLC) System provides the means to manually (or automatically in conjunction with the Redundant Reactivity Control System) inject borated water into the reactor core to terminate critical reactor operation.
The system consists of two pump trains, a storage tank, and 4
test / flushing tank and accessories.
This system and its related equipment were reviewed with respect to the criteria and methods of Sections 1.3 and 1.4 of this report. See Appendix 2.13 for a listing of documents reviewed.
Proposed TS 3/4.6.5 was compared to the documents listed in Appendix 2.13 to verify that the proposed TS accurately represented the as-built plant configuration and operating characteristics and were in agreement with the information in the FSAR and SER.
2.13.2 - Discussion The features of these systems reviewed included normal, abnormal, and emergency operations described by the FSAR, Section 15, Accident Analysis, and the licensee's draft and approved procedures. : --
The as-built configuration portion of the review included a i sampling based review of system piping configuration, instrumentation and control setpoints and operating logic, system operating parameters and limits.
Operating Procedures, Surveillance and Inservice Tests were reviewed determine that the design features were accurately reflected by the test and operatin8 methods and that these methods were consistent with the requirements of the proposed TS. Preoperational tests were not reviewed.
Pa8e 37
EVALUATION Specifically, the system features and operations involving the
. following were reviewed:
Normal system alignments and operations Emergency system alignments and operations System testing alignments and methods Flow Path Valve Lineups and Operability Testin8 System Operational Readiness Testing Pump and Valve Inservice and Operability Testing 2.13.3 - Observations Minor questions were identified and resolved'by the licensee during the inspection. All inspector questions were acceptably resolved.
2.13.4 Conclusions No discrepancies were identified. The as-built cenfiguration of the system, structures, and components compared satisfactorily with the documents reviewed. The Technical Specification requirements reviewed were definitively measurable.
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Pa8e 38
General Conclusions 3.0 -- GENERAL CONCLUSIONS 4
The inspection found that the proof and review Technical Specifications (TS) were compatible with the Final Safety Analysis Report (FSAR), Safety Evaluation Report (SER), the facility's procedures, and the as-built plant as reflected by the engineering drawin8s, data and in situ hardware.
The TS and FSAR are reasonably complete and in agreement for the project status (Operating License projected for February, 1986).
Similarly, only several of the operating and surveillance procedures used or requested during the inspection were in preliminary form. Overall, the licensee appears to have the TS surveillance program largely in place.
The licensee's programs for TS development and implementation appear to be functioning satisfactorily.
Only three exceptions were noted. These exceptions are not considered to be significant detractors from the licensee's programs:
- 1) Pump and valve inservice testing procedures pursuant to TS 4.0.5 and the ASME Boiler and Pressure Vessel Code,Section XI, are well along in development but the baseline data and corresponding procedure input is in the process of development. Baseline data is currently being developed by the test program and does not appear to represent a major factor in preparations for licensed operations.
- 2) The containment isolation valve and ptnetration testing programs per 10CFR50, Apendix J, was under development. No assessment of progress was made except that few of the procedures requested were available during the inpsection.
- 3) Most TS surveillance procedures were available in issued form. Those containing "laters" for completion of information were being effectively tracked and appeared to be a minority of the procedures reviewed. The licensee plans to implement the TS surveillance program on each system as soon as it is turned over for operations.
Although a detailed review of turnover status and d
surveillance impimentation was not reviewed, inspector observations and discussions with licensee personnel indicated that the rate of system turnovers and surveillance implementation may impact the licensee's ability to complete implementation of the surveillance program by their projected license date.
The specific findings of this inspection indicate no major systemic or programmatic problems in TS implementation. The Page 39 i
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General Conclusions procedures reviewed were, in nearly all cases, complete and appeared effective in definitively determining or assuring compliance with TS. No substantive disagreements between the plant and the documentation were identified. The specific findings of the inspection fall into the following general categories:
- 1) Setpoints and parametric limits were found to be available and consistent with the design and safety analysis requirements in nearly all cases. The specific discrepancies found were typically the result of revisions to baseline documents (TS, FSAR, etc.) which followed development of the document containing the discrepancy.
The licensee presented his plans for performing a review of implementing documents during final preparations for license receipt to correct such discrepancies and replace procedure "laters" with correct, current information. The licensee maintains a computerized system which lists each information or data input for each procedure and which can provide a listing of affected documents when an input reference
~ undergoes change.
The licensee also maintains an extensive computerized commitment control system for control of licensing commitments contained in the plant procedures. Although this program and its mechanisms were not reviewed in detail, the output of the system was reviewed as FSAR and SER action items were verified to be included in plant procedures. This licensee system was found to be effective and provides additional assurance of acceptable incorpororation of licensing requirement into plant procedures.
- 2) Similar inconsistencies were found between the FSAR and TS, e.g. EDG load limits, drywell high pressure trip setpoints, etc. The review and revision of these documents is also continuing with the licensee main tainirq another computerized system to track, correct, and verify FSAR-TS discrepancies. Maintenance of the current level of licensee effort appears appropriate to eliminate such discrepancies.
None of the specific discrepancies identified would have had a material negative impact on safety of operation and did not indicate programmatic nor systemic problems.
Page 40
SECTION 4.0 APPENDICES 1.0.......... INSPECTION PLAN 1.1.......... PERSONS CONTACTED 2.0 - 2.13... INSPECTION DATA SHEETS Page 41
Hope Creek Inspection Plan Page 1
. APPENDIX 1.0 HOPE CREEK - INSPECTION PLAN VERIFICATION OF AS-BUILT CONDITIONS TO TECHNICAL SPECIFICATIONS AND FSAR/SER OBJECTIVES:
Conduct, on a sampling basis, reviews and inspections of as-built safety related systems, structures, and components in order to:
determine whether the Technical Specifications and FSAR/SER are compatible with the Hope Creek as-built plant, and to determine whether Technical Specification requirements are definitively measurable.
General Scope ,
The facility descriptions, operating characteristics, and related information found in the FSAR, SER and the proposed Technical Specifications (TS) will be compared to corresponding licensee drawings, procedures, and actual plant hardware to establish whether the as-built configuration of the systems, structures and components is compatible with the safety analyses and proposed (TS).
Conmurrent with the above, the TS will be evaluated to confirm that the performance criteria and requirements established by the TS cap be definitively measured or determined, i.e. that the means and methods to establish conformance with the TS requirements are responsive, sensitive, and sufficiently definitive to actually establish the required level of conformance.
Particular emphasis will be given to the efficacy of surveillance tests and inservice tests established by the licensee to demonstrate conformance with TS and the requirements of ASME B&PV Section XI and 10CFR50.55a.
In general, the systems, structures, and compone'nts to be reviewed will include a sample of the following-Containment & Support Systems Primary Containment Secondary Containment Filtration, Recirculation & Ventilation System Containment Isolation Systems & Valves
Hope Creek
, Inspection Plan Pa8e 2
. Suppression Pool Spray Radiation Monitorin8 i High Pressure Safety Injection & RCIC Low Pressure Safety Injection Core Spray
, Standby Liquid Control Automatic Depressurization Vital AC Power (4160/480/120 VAC) ;
Emergency Diesel Generators Vital DC Power Emergency Service Water / Safety Auxiliaries Cooling General Instrumentation & Controls Inspection Items Documents:
Technical Specifications Final Safety Analysis Report Safety Evaluation Report and Supplements Surveillance / Test Procedures Preoperational Test Procedures Inservice Test Procedures Normal, Abnormal and Emergency Operating Procedures Process & Instrumentation Diagrams Elementary, Logic, and Loop Drawings Fabrication and Installation Drawings Equipment Technical Manuals Inspection Tasks:
- l. Identify the TS applicable to the subject systems and select i
a sample of requirements (Limiting Conditions'for Operation, Surveillance Requirements, etc.) for inspection. Review the corresponding sections of the FSAR and SER.
- 2. Obtain applicable as built (or Approved for Construction) i P& ids, Elementary Diagrams, Loop and Logic Diagrams, etc.
. for the subject systems. Select areas of inspection by identifyin8 (red lining) portions of each drawing. Develop a listing of specific equipment items within the system area which are subject to the TSs. J
- 3. Verify for selected portions of each system that:
l
- 1) the proposed TS adequately reflect the systen configuration depicted by the drawings,
i
Hope Creek Inspection Plan Page 3
. 3) the proposed TS are consistent with the FSAR commitments and SER conclusions.
Confirm that the system configuration and equipment will support definitive measurement or determination of conformance with TS performance criteria and requirements
- 4. Develop a checklist of items for field verification during system and procedure walkdowns.
- 5. identify and obtain the operating, surveillance and other pertinent licensee procedures applicable to the system areas and TS being reviewed. Working from the drawings and TSs to the procedures, confirm that:
- 1) the procedure (s) adequately address the selected equipment and TS requirements identified in the FSAR and SERs,
- 2) procedures accurately reflect the installed (as-built) hardware configuration and condition, and
- 3) the test and or operating methods meet the TS or FSAR/SER requirements, commitments and analyses (review actual performance data where practical).
- 6. Include procedure field verification items in checklist for system and procedure walkdowns.
- 7. Conduct an in plant walkdown of subject systems to verify the results of the document review; confirm that:
- 1) the as built hardware configuration matches the information obtained from the document review,
- 2) the installed hardware is adequately addressed in the procedures and TS,
- 3) the licensee's test and operating methods are appropriate to the actual equipment, and
- u. y e.
- 4) the equipment configuration and features provide for 1
definitive determination or measurement of conformance with the TS. ,
1
- 8. Review the licensee's program for correlating TS requirements to procedures and procedure revision needs, design change impact upon TS and TS implementing procedures, planning and scheduling of surveillance testing, etc.
l 1
1
.. ,-_ _ .._ _ .._.... . - - . _ . - ~ . . . _ , . , _ _ _ - . _ . . _ _ _ _ _ _ _ , _ - _ . . , , . _ . _
APPENDIX 1.1 REVIEW OF HOPE CREEK TECHNNICAL SPECIFICATIONS
, LICENSEE PERSONNEL CONTACTED DURING INSPECTION The inspection team met held discussions with and inspected plant systems with numerous licensee personnel. Listed below are the licensee contacts who materially participated in the inspection and entrance or exit meetings.
NAME TITLE
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C. Allen Operations Consultant P. Ayers Shift Test Engineer M. Azzaro Operations Department B. Binz Engineer R. Burticelli General Mgr, Engrg & Plant Betterment F. Cielo Principal Engineer G. Connor Operations Manager J. Cox Startup Engineer C. Churchman Site Engineering Manager J. Das Setpoint Program, Lead Engr.
W. Denardi Engineer D. Distel Principal Licensin8 Engineer S. Funsten I&C Engineer A. Giardino Manager, Station QA P. Hopkins I&C Engineer J. Isaacs Setpoint Program Manager J. Johnson Operations Department B. Jolly Engineer M. Kobran Tech Spec Coordinator D. Lamastra Engineer
,B. Markowitz Project Manager - Bechtel C. McNeil Vice President - Nuclear W. Merritt Lead Engineer - Operations R. Monger Procedure Engineer (contractor)
M. Moran Engineer M. Mortar-*lo Senior Staff En8 1neer S. Mukhopadhyay Setpoint Engineer G. Nayler Operations Department
, D. Newman Operations Department J. Nichols Technical Mgr, Operations W. Pavincich Principal Engineer E. Pfister Technical Reviewer, I&C B. Preston Manager, Licensin8 & Regulation T. Ram Supervisin8 Engineer J. Ruckie ,
Maintenance Engineer R. Salveson General Mgr, Hope Creek Operations V. Tanenbaum Setpoint Program, Lead Engr.
S. Tanner I&C Engineer (contractor)
,4 APPENDII 2.0 0
GENERAL REFERENCES In addition to the specific inspection and review items discussed elsewhwere herein the adminsitrative procedures and Emergency Operating Procedures listed below were reviewed and used throughout the inspection for the evaluation of the licensee's various programs for TS implementation:
GE STARTUP TEST SPECIFICATIONS, REVISION O GE PREOPERATIONAL TEST SPECIFICATIONS, REVISION 0 ST-TE.ZZ-012, VERIFICATION AND DOCUMENTATION OF SURVEILLANCE REQUIREMENTS BY OTHER THAN SURVEILLANCE TEST, REVISION 0
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TECHNICAL SPECIFICATION DEFICIENCY TRACKING SYSTEM REPORT, 11/29/85 STATION ADMINISTRATIVE PROCEDURES (SA-AP.ZZ- (Q)):
12 TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS 13 CONTROL OF TEMPORARY MODIFICATIONS 14 STATION PERSONNEL QUALIFICATION AND TRAINING 24 RADIOLOGICAL PROTECTION PROGRAM 27 STATION INSERVICE INSPECTION PROGRAM 30 STATION RESPONSE AND COMMITMENT CONTROL PROGRAM 32 REVISIONS AND CHANGES TO STATION PROCEDURES EMERGENCY OPERATING PROCEDURES (OP-EO.ZZ- (Q)), REVISION 0:
099 POST SCRAM RECOVERY 100 REACTOR SCRAM 101 RPV CONTROL 2
102 CONTAINMENT CONTROL 103 REACTOR BUILDING CONTROL 201 LEVEL RESTORATION .
203 STEAM COOLING 204 SPRAY COOLING ~"m 205 ALTERNATE SHUTDOWN COOLING 206 REACTOR FLOODING 207 LEVEL / POWER CONTROL 301 BYPASSING MSIV INTERLOCKS 308 ALTERNATE INJECTION USING SERVICE WATER 312 SUPPRESSION CHAMBER MAKEUP USING HPCI 313 SUPPRESSION CHAMBER MAKEUP USING RCIC 314 SUPPRESSION CHAMBER MAKEUP USING SERVICE WATER 315 SUPPRESSION CHAMBER MAKEUP USING CORE SPRAY 316 SUPPRESSION CHAMBER LEVEL REDUCTION - HPCI 317 SUPPRESSION CHAMBER LEVEL REDUCTION - RCIC
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Data Sheets Page 1 APPENDIX 2.1 INSPECTION REPORT DATA SHEET REACTOR PROTECTION SYSTEM TECHNICAL SPECIFICATIONS:
2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION FSAR
REFERENCES:
7.2.1.1, 7.2.1.2, 7.2.1.3 NRC SER
REFERENCES:
7.2.1., 7.2.2.4, 7.2.2.5, 7.2.2.7 DOCUMENTS REVIEWED:
NUMBER TITLE REVISION E-0033-1 AUX BLDG - CONTROL AND DG AREA 9 480 V MCC TABULATION 10B446, 10B491 E-118-5-29 FEEDER BREAKER UNIT WITH GROUND FAULT DEVICE 6 M-41-1 NUCLEAR BOILER P&ID, SH 1 & 2 7/6 115D6002AC RPS MG SET CONTROL DIAGRAMS 6 729E611AC RPS BLOCK DIAGRAMS 6 791E414AC RPS ELEMENTARY DIAGRAMS 9 AB-0062-1 CALCULATION - MSIV SEALING SYSTEM A HI-LO 1 PRESSURE SETTINGS BB-0165-1 CALCULATION - REACTOR VESSEL PRESSURE O BB-0175-0 CALCULATION - REACTOR VESSEL PRESSURE O BB-0192-1 CALCULATION - REACTOR VESSEL LEVEL 0 IC-CC.BB-035 NUCLEAR BOILER, DIV 1 CHANNEL A, N21-H678A, O HIGH REACTOR PRESSURE CALIBRATION 0 IC-CC.SB-009 RPS - TURBINE MAIN STOP VALVE POSITION SWITCH 0 CALIBRATION IC-FT.BB-001 NUCLEAR BOILER - DIV 1, CHANNEL A, B21-N691A, 0 REACTOR VESSEL LEVEL TRIPS, 1,2,8 (CS, RHR, HPCI) FUNCTIONAL TEST IC-FT.BB-023 NUCLEAR BOILER - DIV 1, CHANNEL A, B21-678A 0 HIGH REACTOR PRESSURE FUNCTIONAL TEST i IC-TR.BB-103 NUCLEAR BOILER - DIV 1 CHANNEL A, B21-N078A 0 REACTOR VESSEL PRESSURE SENSOR IC-TR.BB-104 NUCLEAR BOILER - DIV 1, CHANNEL A, B21-N678A 0 HIGH REACTOR PRESSURE CHANNEL W OP-SO.SB-001 REACTOR PROTECTION SYSTEM OPERATION O OP-ST.SB-002 RPS SIMULATED OPERATION - 18 MONTH TEST 0
Data Sheets Page 2 APPENDIX 2.2 INSPECTION REPORT DATA SHEET PRIMARY CONTAINMENT & SUPPORT SYSTEMS TECHNICAL SPECIFICATIONS:
3/4.3.6.1 PRIMARY CONTAINMENT INTEGRITY 3/4.3.6.2 PRIMARY CONTAINMENT LEAKAGE 3/4.5.3.1 SUPPRESSION CHAMBER 3/4.6.1.6 D/W & S/C INTERNAL PRESSURE 3/4.6.1.7 DRYWELL AVERAGE AIR PRESSURE 3/4.6.1.8 D/W & S/C PURGE SYSTEM 3/4.6.2.3 SUPPRESSION POOL COOLING FSAR
REFERENCES:
6.2.6.1, 6.2.6.2, 6.2.6.5 NRC SER
REFERENCES:
6.2.6, DOCUMENTS REVIEWED:
NUMBER TITLE REVISION M9-ILP-302 PRIMARY CONTAINMENT LEAKAGE TEST 'O PTP-GP-2 PRIMARY CONTAINMENT LEAK RATE TEST 0 NI-B21-1050 STEAM LEAK DETECTION SYSTEM ELEMENTARY VARIOUS DIAGRAMS (10 SHEETS)
NI-B21-1090 NUCLEAR STEAM SUPPLY SHUT 0FF SYSTEM 7 ELEMENTARY DIAGRAMS NI-C71c1020 REACTOR PROTECTION SYSTEM ELEMENTARY DIAGRAMS 5 J-102-0 HI RADN & LOCA/ISOLN SIGNALS FAN 0UT (LOGIC) 6 J-108-0 MISC ALARM SYSTEMS LOGIC DIAGRAM 0 OP-DL.ZZ-026 SURVEILLANCE LOGS /ATT 1 - CONDITIONS 1,2 & 3 0 M9-1AP-102 INSERVICE INSPECTION PROGRAM (SEC. 2) DRAFT OP-ST ZZ-002 PRIMARY CONTAINMENT INTEGRITY VERIFICATION - 0 MONTHLY OP-ST.ZZ-004 PRIMARY CONTAINMENT AIRLOCK OPERATIONAL TEST 0 M-60-1 P&ID PRIMARY CONTAINMENT LEAKAGE RATE TESTING 1 J-60-1 PRIMARY CONTAINMENT LEAKAGE RATE TESTING 6 LOGIC DIAGRAM J-59-0, SH 8 PRIMARY CONTAINMENT INSTRUMENT GAS LOGIC 6 E-0000-0 ELECTRICAL PLAN DRAWING INDEX 6 PTP-SM-2 NUCLEAR STEAM SUPPLY SHUT 0FF & PRIMARY 0 CONTAINMENT ISOLATION SYSTEMS EQUIPMENT ACTUATION SA-AP.ZZ-24 RADIOLOGICAL PROTECTION PROGRAM 2 OP-DL.ZZ-001 SHIFT NARRATIVE LOGS 0 M-57-1 CONTAINMENT ATMOS CONTROL P&ID 13 J-57-1 CONTAINMENT ATMOS CONTROL LOGIC DIAGRAM 7 OP-ST.GS-003 REACTOR BUILDING / SUPPRESSION CHANBER VACUUM 0 BREAKER OPERABILITY TEST - MONTHLY
Data Sheets
, Page 3 !
f ,
OP-ST.GS-002 DRAYWELL AND SUPPRESSION CHAMBER PURGE O
. SYSTEM VALVE TEST MD-ST.BB-001 SUPPR CHAMBER EXTERNAL VISUAL EXAMINATION 0 M9-ILP-20601 SUPPR CHAMBER VISUAL INSPECTION 0 M-57-0, SH 8 CONTAINMENT ATMOS CONTROL - TORUS TO DRYWELL 7 l VACUUM RELIEF VALVES
, J-101-0, SH 12 OUT OF SERVICE STATUS DISPLAY - FRVS/ECCS & 4 SACS PUMP ROOM COOLERS l
)
i i
l 1
1 9
e.
l 4
I
)
Data Sheets Page 4 APPENDIX 2.3 INSPECTION REPORT DATA SHEET PRIMARY CONTAINMENT ISOLATION VALVES TECHNICAL SPECIFICATIONS:
3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES TABLE 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES 3/4.6.1.4 MSIV SEALING SYSTEM FSAR
REFERENCES:
5.4.5, 6.2.3 NRC SER
REFERENCES:
6.2.4, 6.2.6, 6.7 DOCUMENTS REVIEWED:
NUMBER TITLE REVISION OP-ST.SM-001 PCIS/NSSSS ISOLATION FUNCTIONAL TEST - 18 MOS 0 PTP-GD-2 LOCAL LEAK RATE TEST 0 PTP-KP-1 MSIV SEALING SYSTEM TEST 1 OP-ST.KP-001 MSIV SEALING SYSTEM FUNCTIONAL TEST - 18 MOS 0 OP-ST-SE-001 TIP SHEAR CONTINUITY VERIFICATIN - 18 MOS 0 M-72-1 MAIN STEAM ISOLATION SEALING SYSTEM P&ID 5 J-72-0, SH 1 MAIN STEAM ISOLATION SEALING SYSTEM LOGIC 7 J-4072-0, SH 2 RPV PRESSURE MONITOR. LOOP DIAGRAM 5 J-102-0, SH 5 HI RAD & LOCA ISOLATION SIGNALS FANOUT LOOP 5 NI-B21-1090-62 NUCLEAR STEAM SUPPLY SHUT 0FF SYSTEM ELEMENTARY 11 NI-C71-1020 RPS ELEMENTARY DIAGRAM 5
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Data Sheets Page 5 APPENDIX 2.4 -
INSPECTION REPORT DATA SHEET SECONDARY CONTAINMENT & SUPPORT SYSTEMS TECIINICAL SPECIFICATIONS:
3/4.6.5 SECONDARY CONTAINMENT FSAR
REFERENCES:
6.2.3, 6.8, 9.4.2, 14.2 NRC SER
REFERENCES:
6.2.4, 6.5.12, 6.5.3 DOCUMENTS REVIEWED:
NUMBER TITLE REVISION OP-ST.ZZ-003 REACTOR BUILDING / SECONDARY CONTAINMENT 0 VERIFICATION - MONTHLY A-4642-1 REACTOR BUILDING PLAN VIEW DRAWINGS 5 PTP-GS-1 CONTAINMENT ATMOSPIIERE CONTROL 0 M-83-1 REACTOR BUILDING SUPPLY CONTROL DIAGRAM 11
. 11 - 8 3 - 1 .. REACTOR BUILDING SUPELY -LOGIC DI AGRAM 7 M-84-1 REACTOR BUILDING EXHAUST CONTROL DIAGRAM 12 11-84-1 REACTOR BUILDING EXHAU5T LOGIC DIAGRAM 7 M-57-1 REACTOR ATMOSPilERE CONTROL P&ID 13 J-57-0, Sil 8 REACTOR BUILDING TO TORUS VACUUM RELIEF 7 VALVE LOGIC DIAGRAM PTP-GU-1 FILTRATION, RECIRCULATION & VENTILATION O SYSTEM (FRVS) TEST SP-ST.GU-001 FRVS OPERABILITY TEST - MONT LY 0 OP-SO.GR-001 REACTOR BUILDING VENTILATION SYSTEM OPERATION 0 SP-SO.GU-001 FRVS SYSTEM OPERATION 0 OP-ST.GU-004 REACTOR BUILDING INTEGRITY FUNCTIONAL TEST 0 J-101-0, Sil 12 FRVS/ECCS & SACS ROOM COOLER LOGIC DIAGRAM 4 Il-83-0 REACTOR BUILDING SUPPLY - FRVS FANS LOGIC 8 II-84-0 REACTOR BUILDING EXilAUST - FRVS DAMPERS LOGIC 7 C152(0)-2763-1 INSTRUCTION MANUAL - PRIMARY CONTAINMENT 1 PERSONNEL ACCESS AIR LOCK
l Data Sheets Page 6 i
. l APPENDIX 2.5 INSPECTION REPORT DATA SHEET STATION SERVICE WATER SYSTEMS TECHNICAL SPECIFICATIONS:
3/4.7.1.1 SAFETY AUXILIARIES COOLING (SACS) 3/4.7.1.2 STATION SERVICE WATER (SSWS) 3/4.7.1.3 ULTIMATE HEAT SINK 3/4.7.3.4 REMOTE SHUTDOWN INSTRUMENTATION FSAR
REFERENCES:
9.2, 7.3.1.2, 14.2.12 NRC SER
REFERENCES:
9.2.1, 9.2.2 DOCUMENTS REVIEWED:
NUMBER TITLE REVISION M-11 SACS P&ID 11 OP-ST.EG-002 SACS FUNCTIONAL TEST 0 OP-ST.EG-001 SACS FLOWPATH VERIFICATION - MONTHLY OP-SO.EG-001 Ogg._
SACS /TACS OPERATION O ICI-CC.EG-001 SACS CH b DISCHARGE FLOW - REMOTE SilUTDOWN O PANEL CALIBRATION OP-IS.EG-001 A SACS PUMP INSERVICE TEST 0 OP-IS.EG-101 SACS SUBSYSTEM A VALVES INSERVICE TEST 0 OP-IS.EA-001 SSWS SUBSYSTEM A VALVES INSERVICE TEST 0 OP-IS.EA-101 A SSWS PUMP INSERVICE TEST 0 MM-10-1 SSWS P&ID VARIOUS J-10-1 SSWS PUMPS A&C LOGIC DIAGRAM, SH 2 9 J-10-1 SAC $2HX OUTLET VALVE LOGIC DIAGRAM Sil 6 9 J-11-1 SACS PUMP A&D LOGIC DIAGRAM Sil 2 7 J-11-1 SACS /TACS VALVE LOGIC Sil 7 6 J-11-1 RIIR/ SACS OUTLET VALVE SII 16 7 PTP-EG-1 SAFETY & TURBINE AUXILIARIES COOLING TEST 1
.. y
. i-
Data Sheets Page 7 APPENDIX 2.6 INSPECTION REPORT DATA SHEET REACTOR CORE ISOLATION COOLING SYSTEM TECHNICAL SPECIFICATIONS:
2 3/4.7.4 REACTOR CORE ISOLATION COOLING (RCIC) 3/4.3.2 RCIC ISOLATION ACUTATION INSTRUMENTATION 3/4.3.5 RCIC ACTUATION INSTRUMENTATION -
3/4.3.7.4 REMOTE SHUTDOWN INSTRUMENTATION FSAR
REFERENCES:
5.1.9, 5.4.6, 6.2.4, 7.4.1.1, 7.6.1.3, 15 NRC SER
REFERENCES:
5.1, 5.2.1.1, 5.4.6, 7.4.1.4, 7.6.1.3, 15.9.3, 14.2.12.3.12 DOCUMENTS REVIEWED:
- NUMBER TITLE REVISION M-50-1 RCIC P&ID - PUMP & TURBINE 13 M-49-1 RCIC P&ID 10 J-49-0 RCIC LOGIC DIAGRAMS (20 SHEETS) VARIOUS PTP-BD-1 RCIC PREOPERATIONAL TEST PROCEDURE . 0 OP-ST.BD-001 RCIC PIPING AND FLOW PATH VERIFICATION 0 OP-ST.BD-003 RCIC FUNCTIONAL AND FLOW VERIFICATION - 18 MOS 0 OP-IS.BD-001 RCIC PUMP OP 203 INSERVICE TEST 0 TE-SU.BD-145 RCIC SURVEILLANCE TEST DEMO STARTUP TEST 1 TE-SU.BD-144 RCIC COLD QUICK START TO RPV STARTUP TEST 1 TE-SU.BD-143 RCIC STEP CHANGES FROM REMOTE S/D PANEL TEST 0 TE-FT.BD-005 RCIC CONDENSATE STORAGE TANK LOW LEVEL TEST 0 TE-SU-BD-142 RCIC RPV INJECTION TEST 1 TE-SU-BD.141 RCIC CST INJECTION TEST 1 IC-FT.BB-003 NUCLEAR BOILER RV LEVEL TRIP FUNCTIONAL TEST 0 IC-CC.BD-006 CST LO-LO LEVEL CHANNEL CALIBRATION O IC-CC.FC-002 RCIC DIV 4 STEAM LINE FLOW ISOLATION CALIBRATIN O
Data Sheets Pa8e 8 APPENDIX 2.7 INSPECTION REPORT DATA SHEET AC SOURCES & STANDBY DIESEL GENERATOR TECHNICAL SPECIFICATIONS:
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 AC SOURCES FSAR
REFERENCES:
8.3.1.1.3, 9.5.4., 9.5.5., 9.5.7 NRC SER
REFERENCES:
8.1.3.2, 8.3.1.3., 8.3.1.4, 8.3.1.5, 8.3.1.7, 8.3.1.8, 9.5.4.2, 9.5.6, 9.5.7 DOCUMENTS REVIEWED: '
NUMBER TITLE REVISION i
E-0001-0 STATION SINGLE LINE DIAGRAM 6 E-0006-1 4.16 KV CLASS 1E POWER SYSTEM SINGLE LINE 4 METER & RELAY DIAGRAM E-0007-1 SYNCHRONIZING SINGLE LINE DIAGRAM 5 E-0040-1 SYNCHRONIZING SCHEMATIC METER & RELAY DIAGRAM 6 E-0084-0 CLASS 1E 4.16 KV STA PWR SYS SWGR D-G CKT BKR 8 (52-40107) ELECTRICAL SCHEMATIC DIAGRAM E-3080-0 CLASS 1E STA PWR SWGR - 4.16 KV SYSTEM D-G 9/7 CET BKR LOGIC DIAGRAM SH 1 & 2 E-3081-0 DIESEL GENERATOR CONTROL LOGIC DIAGRAM SH 1&2 6/7 IC-FT.PE-001 EMERGENCY LOAD SEQUENCER SYSTEM FUNCTIONAL TEST 0 OP-SO.KJ-001 EMERGENCY DIESEL GENERATORS OPERATION 0 OP-ST.KJ-001 EMERGENCY DIESEL GENERATOR 1AG400 OPERABILITY 0 TEST - MONTHLY OP-ST-KJ-005 INTEGRATED EDG 1AG400 TEST - 18 MONTHS 0 OP-ST.KJ-010 EDG LUBE OIL CHECKS - MONTHLY 0 OP-ST.KJ-012 EDG FUEL OIL STORAGE TANK SAMPLE & WATER CHECKS 0 QUARTERLY TM-9N90 VENDOR'S LOAD SEQUENCER OPERATION & MAINTENANCE -
INSTRUCTIONS 9
6
Data Sheets Page 9 4
APPENDIX 2.8 INSPECTION REPORT DATA SHEET
! DC POWER SYSTEMS TECHNICAL SPECIFICATIONS:
REFERENCES:
8.3.2.1.2, 8.3.2.1.2.1, 8.3.3.2.1.2.2., 8.3.2.1.2.3, 8.3.2.1.2.5, 8.3.2.2., TABLE 8.3.7 NRC SER
REFERENCES:
8.3.2.1, 8.3.2.2, 8.3.2.3, 8.3.2.5, 8.3.2.6, 8.3.2.7, 8.3.2.8 DOCUMENTS REVIEWED:
NUMBER_
TITLE REVISION E-0009-1 125 VDC SYSTEM - CHANNELS A&C SINGLE LINE 6
~
METER & RELAY DIAGRAM SH 1 4 E-0011-1 250 VDC SYSTEM - UNIT 1 SINGLE LINE METER 6 l
AND RELAY DIAGRAM SH 1 -
- E-0119-0 125 VDC SYSTEM SCHEMATIC METER & RELAY DIAGRAM 8 SH 1 3 E-3090-0 125 VDC SYSTEM LOGIC DIAGRAM SH 1 6 MD-ST,PK-002 125 VDC QUARTERLY BATTERY SURVEILLANCE 1 MD-ST .'PK-005 125 VDC BATTER CHARGER SERVICE TEST 1 OP-SO.PK-001 125 VDC ELECTRICAL DISTRIBUTION SYSTEM 0 OPERATION, CHANNEL A (ONLY)
OP-ST.ZZ-001 POWER DISTRIBUTION LINEUP - WEEKLY 0 PTP-PK-1 125 VDC CLASS IE POWER SYSTEM PREOP TEST 1 4.1 CALCULATIONS FOR SIZING BATTERY 11AD411 (CH A) 4 as
(
Data Sheets Page 10 APPENDIX 2.9 INSPECTION REPORT DATA SHEET ONSITE POWER DISTRIBUTION TECHNICAL SPECIFICATIONS:
3/4.8.3 ONSITE POWER DISTRIBUTION FSAR
REFERENCES:
8.3.1.1.2, 8.3.1.1.4, 8.3.1.1.6, 8.3.1.2, 8.3.1.5, 8.3.2.2., TABLE 8.3.7 NRC SER
REFERENCES:
8.3.1.3, 8.3.3.1.2, 8.3.3.3.5, 8.3.3.3.6, 8.3.3.5.1, 8.3.3.5.2, 8.3.3.5.3, 8.3.3.5.5., 8.3.3.5.6 DOCUMENTS REVIEWED:
NUMBER TITLE REVISION g.:
E-0001-0 STATION SINGLE LINE DIAGRAM 6 E-0006-1 4.16 KV CLASS lE POWER SYSTEM SINGLE LINE 4 METER AND RELAY DIAGRAM SH 1 E-0021-1 120 VAC INSTRUMENTATIN & MISC SYSTEMS SINGLE 6 LINE METER AND RELAY DIAGRAM, SH 1 E-0018-1 480 VOLT CLASS 1E UNIT SUBSTA 10B410, 10B450 8 SINGLE LINE METER AND RELAY DIAGRAM SH 1 '
E-0019-1 CLASS 1E AUX BLDG D/G AREA (10B411) 7 480 %QLT MCC TABULATION ~SH 1 *-
E-0022-1
~
CLASS 1E MCC INTAKE STRUCTURE (10B553) 6 480 VOLT MCC TABULATION E-0046-1 4.16 KV CLASS 1E STA PWR SYS SWGR SCHEMATIC 6 METER AND RELAY DIAGRAM E-0068-0 CLASS 1E 4.16 KV STA PWR SYS SWGR MAIN 9 CKT BKR 52-40108 ELEC SCHEMATIC DIAGRAM E-0069-0 CLASS 1E 4/16 KV STA PWR"SYS SWGR MAIN 7 CKT BER 52-40101 ELEC SCHEMATAIC DIAGRAM i E-0097-0 450 V SYSTEM UNIT SUBSTATION MCC & PANEL 6 FDR CKT BKRS ELECTRICAL SCHEMATIC DIAGRAM E-3060-0 CLASS 1E STA PWR SWGR - 4.16 KV SYSTEM MAIN 13 CKT BKR LOGIC DIAGRAM i E-3120-0 120 VAC UPS SYSTEM ALARM LOGIC DIAGRAM 5 E-3132-0 480V SYSTEM UNIT SUBSTATION MCC & PANEL 5 FDR SKT BKR LOGIC DIAGRAM E-3134-0 480 V SYSTEM UNIT SUBSTATION FDR BKRS 8 i ALARM INPUTS LOGIC DIAGRAM E-6024-0 CORE SPRAY SYSTEM - CORE SPRAY MIN FLOW 3 VALVES ELEC SCHEMATIC DIAGRAM E-6231-0 RHR SYSTEM - RHR SHELL SIDE BYPASS VALVE 7 ELEC SCHEMATIC DIAGRAM SH 2 E-6075-0 HPCI PUMP SUCTION VALVE F042 SCHEMATIC SH 5 3 F-6084-0 RCIC MAIN STM SUPPY VALVE SCHEMATIC SH 7 7
Data Sheets Page 11 E-6085-0 RCIC WARMUP VALVE SCHEMATIC Sil 3 6
. MD-ST.ZZ-009 MOTOR OPERATED VALVE SURVEILLANCE DRAFT OP-SO.PB-001 4.16 KV SYSTEM OPERATION 0 OP-SO.PN-001 480 V DISTRIBUTION SYSTEM OPERATION O OP-SO.PU-001 120 VAC ELECTRICAL DISTRIBUTION OPERATION 0 OP-ST.22-001 POWER DISTRIBUTION LINEUP - WEEKLY 0
... sn
1 l
Data Sheets l Page 12 '
APPENDIX 2.10 INSPECTION REPORT DATA SHEET HIGH PRESSURE COOLANT INJECTION AUTOMATIC DEPRESSURIZATION SYSTEM TECHNICAL SPECIFICATIONS:
3/4.5.1 ECCS SYSTEMS - OPERATING 3/4.5.2 ECCS SYSTEMS - SHUTDOWN 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION -
3/4.3.3 ECCS ACTUATION INSTRUMENTATION 3/4.4.2.1 SAFETY RELIEF VALVES FSAP
REFERENCES:
5.2.4.1, 6.3 (ALL), 7.3.1.1.1 NRC SER
REFERENCES:
5.2.2, 6.3, 7.3.1.1, 7.6.1.3, 7.3.2.9 DOCUMENTS REVIEWED:
NUMBER TITLE REVISION M-55-1 . HPCI P&ID 15 M-56-1 HPCI PUMP & TURBINE P&ID 12 E-6074-0 HPCI TURBINE AUX OIL PUMP SCllEMATIC 4 E-6074-1 HPCI VACUUM TANK CONDENSATE PUMP SCHEMATIC 4 E-6075-0 HPCI PUMP AND VLV SCHEMATIC (9 SHEETS) 5 OP-ST.BJ-001 HPCI PIPING AND FLOW PATH VERIFICATION 0 OP-SO.BJ-001 HPCI SYSTEM OPERATION 0 OP-ST.BJ-002 IIPCI SYSTEM FUNCTIONAL TEST 0 PTP-BJ-1 HPCI PREOPERATIONAL TEST 0 IC-FT.BJ-001 HPCI DIV 1 LOW CST LEVEL SUCTION TRANSFER 0 FUNCTIONAL TEST IC-PT.FJ-002 IIPCI DIV 1 HIGli SUPPR CHBR LEVEL SUCTION O TRANSFER FUNCTIONAL TEST OP-SO.SN-001 NUCLEAR PRESSURE RELIEF & ADS OPERATION O 22A2919AF GE SYSTEM DESIGN SPECIFICATION - ADS 14 OP-ST.SN-002 ADS LOGIC FUNCTIONAL TEST 0 OP-ST.SN-001 ADS AND SRV MANUAL OPERABILP'Y TEST 0 l M-41-1 ANUCLEALR BOILER P&ID Sil 1 & 2 12/10
! M-42-1 NUCLEAR BOILER INSTRUMENT DIAGRAM SH 1 & 2 7/6 J-41-0 NUCLEAR BOILER LOGIC DIAGRAM - ADS Sil 1-3 VARIOUS IC-FT.BB-003 NUCLEAR BOILER RV LEVEL TRIP CALIBRATION 0 IC-FT.BB-004 NUCLEAR BOILER RV LEVEL TRIP CALIBRATION O IC-FT.SN-007 ADS DIV 2 LOGIC B - ADS TIMER FUNCTIONAL TEST 0 IC-FT.SN-008 ADS DIV 4 LOGIC D - ADS TIMER FUNCTIONAL TEST 0 IC-FT.BB-007 NUCLEAR BOILER RV LEVEL TRIP FUNCTIONAL TEST 0 IC-FT.BB-008 NUCLEAR BOILER RV LEVEL TRIP FUNCTIONAL TEST 0
Data Sheets Page 13 APPENDIX 2.11 INSPECTION REPORT DATA SHEET LOW PRESSURE COOLANT INJECTION MODE OF THE RESIDUAL HEAT REMOVAL SYSTEM TECilNICAL SPECIFICATIONS:
3/4.5.1 ECCS SYSTEMS - OPERATING 3/4.5.2 ECCS SYSTEMS - SilUTDOWN 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION 3/4.3.3 ECCS ACTUATION INSTRUMENTATION
,, FSAR
REFERENCES:
5.1.10, 5.2.5, 6.3, 7.3.1.1.1.4, 7.4.1.4, 7.6.1.2, 14.2.12.1.5 NRC SER
REFERENCES:
5.2.5, 5.4.7, 6.3.3, 7.3.1, 7.4.1, 7.6.1 DOCUMENTS REVIEWED:
NUMBER TITLE REVISION M-51-1 RESIDUAL IIEAT REMOVAL (RilR) P&ID 14 729630AC RIIR FUNCTIONAL CONTROL DIAGRAMS Sil 1-4 VARIOUS J-51-0 . RIIR LOGIC DIAGRAMS Sil 1- VARIOUS OP-SO.BC-001 RIIR OPERATING PROCEDURI* 12A OP-ST.BC-001 LPCI PIPING AND FLOWPATil VERIFICATION O OP-ST.BC-004 LPCI SUBSYSTEM A ECCS TIME RESPONSE TEST 0 OP-ST.BC-002 SUPPRESSION dilAMBER SPRAY FLOW PATl! VERIFICATION 0 OP-IS.BC-004 D RIIR PUMP INSERVICE TEST 0 OP-IS.BC-003 B RIIR PUMP INSERVICE TEST 0 PT-BC-1 RilR PREOPERATIONAL TEST 0 IC-CC.BB-004 NUCLEAR BOILER RV LEVEL TRIPS CALIBRATION O IC-CC.BB-003 NUCLEAR BOILER RV LEVEL TRIPS CALIBRATION 0 ,
Data Sheets Page 14 APPENDII 2.12 INSPECTION REPORT DATA SHEET CORE SPRAY SYSTEM TECHNICAL SPECIFICATIONS:
3/4.5.1 ECCS SYSTEMS - OPERATING 3/4.5.2 ECCS SYSTEMS - SHUTDOWN 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION 3/4.3.3 ECCS ACTUATION INSTRUMENTATION FSAR
REFERENCES:
6.1.1.2, 6.2.3.2.1.2, 6.3, 7.3.1.1.1.3, 14.2.12.1.7 l NRC SER
REFERENCES:
6.3
- DOCUMENTS REVIEWED:
NUMBER TITLE REVISION M-52-1 CSS P&ID . 12 J-52 CSS LOGIC DIAGRAMS, SH 1-10 '"#"' 2 OP-ST.BE-001 CSS PIPING & FLOW PATH VERIFICATION - MONTHLY 0 OP-ST.BE-002 CSS LOOP A ECCS TIME RESPONSE FUNCTIONAL TEST 0 OP-SO.BE-001 CSS OPERATING PROCEDURE O OP-IS.BE-001 A&C CSS PUMPS INSERVICE TEST 0 OP-IS.BE-002 B&D CSS PUMPS INSERVICE TEST 0 E-6022-0 CSS PUMP SUCTION VLV SCHEMATIC DIAGRAM 3 E-6023 CSS PUMP ISOLATION VLV SCHEMATIC DIAGRAM 3 E-6024 CSS PUMP MINIFLOW VLV SCHEMATIC DIAGRAM #"- 3 I IC-FT.BB-004 NUCLEAR BOILER RV LEVEL TRIPS CALIBRATION O
%" ei 4 4
4
)
i i
---.m.= _ _ _ __ _ ,_.._ __ , _.._ _._ _ _. m , _ _ _ _ _ , _ . , _ _ _ _ _ _ . _ _ _ _ , _ _ . _ , _ - _ _ . . . . . - _ _ _ _ . _ - , _ . . . _ ~ _ _ _ , . . , _ _ _ _ - _ _
.-. .- _ - - - . _ _. . _ _ = _ _ . .
- Data Sheets Page 15 APPENDIX 2.13 INSPECTION REPORT DATA SHEET
, STANDBY LIQUID CONTROL SYSTEM 4
TECHNICAL SPECIFICATIONS:
j 3/4.6.5 STANDBY LIQUID CONTROL SYSTEM FSAR
REFERENCES:
3.6.1.2.1.10, 6.2.4.3.1.11, 7.4.1.2, 7.6.1.7.2, t 9.3.5 NRC SER
REFERENCES:
9.3.5, 7.6.1.7 DOCUMENTS REVIEWED:
NUMBER TITLE REVISION M-48-1 SBLC P&ID 8 l OP-SO.BII-001 SBLC SYSTEM OPERATION 0 OP-ST.BII-001 SBLC VALVE OPERABILITY TEST - MONTHLY 0 OP-ST.BII-002 SBLC FLOW TEST - 18 MONTIIS 0 OP-ST.BII-003 SBLC SYSTEM TANK FLOW TEST - 18 MONTilS 0 OP-ST.BII-004 SBLC STORAGE TANK IIEATER OPERABILITY TEST 0
- OP-DL.ZZ-026 SURVEILLANCE LOGS DRAFT OP-IS . Bil-001 SBCC PUMP INSERVICE TEST 0 I OP-IS.BII-002 SBLC MINIMUM FLOW TEST 0 I
i
.=
i l
ENCLOSURE 2 NRC Open Items Resulting From Inspection Report No. 50-354/85-64 The following inspector follow items were developed from a review of Inspection Number 50-354/85-64 " Review of Hope Creek Technical Specifications" prepared under contract by Parameters, Inc.
I -
The label on control room instrument SS-PR-4960A3 (10C650E) states
" Suppression Chamber Pressure" but should read "Drywell Pressure" according to P&ID M-57-1, Revision 13.
The licensee acknowledged the above and submitted Field Questionnaire (FQ) #0P-598 to correct the labeling. See Section i 2.2.3 of Inspection Report 85-64. (85-64-01)
The mair. control board label for Reactor Building - Torus Vacuum Relief Valve HV-5029 (Containment Atmosphere Control System) pushbutton is engraved "HV-5028". The licensee demonstrated that Design Change Package 221 had previously identified this error and that corrective action would be accomplished. See Section 2.2.3 of Inspection Report 85-64. (85-64-02)
OP-ST.GR-001, Reactor Building Ventilation System Operation, Revision 0, Section 5.3.4, requires three supply and exhaust fans to be in service during refueling operations per Technical Specification. No documentation provisions for implementation of this requirement could be found. See Section 2.4.3 of Inspection Report 85-64. (85-64-03)
Preoperational Test PTP-BD-1, RCIC System Preoperational Test, Revision 0, verified that the High Steam Flow Isolation Time Delay relay was set to 2.9 - 3.1 seconds instead of the 3 - 13 second 1
criteria of TS Table 3.3.3.-2. One of the two timers was reset from an acceptable 5.09 seconds to an unacceptable 2.9 seconds during PTP-BD-1 to meet the test specification. The second timer had an as found setting of 3.1 seconds and was not reset during the test.
l The licensee acknowledged that the PTP had been written in accordance with the NSSS preoperational test specification and that the criterion had changed as the TS were developed. The licensee advised that the time delay relay would be reset in accordance with i the surveillance procedures for calibration and functional testing of the logic. This will be accomplished prior to the equipment j being required to be made operable per TS. See Section 2.6.3 of Inspection Report 85-64. (85-64-04)
2 TS 4.7.4.b requires tne RCIC pump functional test and flow verification to be performed with a reactor steam supply pressure and pump back pressure of 1000, +20, -0 psig. OP-ST .BD-001, RCIC Pump OP203 Inservice Test, Revision 0, did not address this test condition in the body of the procedure or acceptance criteria.
The licensee advised that the acceptance criteria and the above test condition would be included in the IST baseline and test data sheet which would be appended to the procedure when baseline data was eventually taken. This data sheet is intended to " stand alone" for each procedure, permitting the data sheet to be revised for new data or limits without requiring revision of the procedure body. See Section 2.63. of Inspection Report 85-64. (85-64-05)
FSAR Section 5.4.6.2.2.2.1, Valve Operation Requirements, specifies valve stroke rates (inches / minute) for key system valves under full system pressure or flow conditions. Examples include the pump steam supply and exhaust valves, pump discharge valves, pump mini-flow valves, etc.
In TS 3/4.8.1 - AC Sources, Section 4.8.1.1.2.h.9, the continuous rating the EDG was shown as 4737 KW. The correct value is 4430 KW.
The licensee initiated TS Change Request (CR) #143 which, as reviewed, would acceptably correct the above in the next revision of the TS. See Section 2.7.3 of Inspection Report 85-64. (85-64-06)
In MD-ST.PK-002, 125 VDC Quarterly Battery Surveillance Test, Revision 0,:
a) In Section 1.0, a typographical error identified the system as 250 VDC instead of 125 VDC.
b) In Section 3.0, Precautions and Limitations, the battery room ventilation requirements and hydrogen accumulation limits included in the system operating procedure and other surveillance procedures was omitted. See Section 2.8.3 of
- Inspection Report 85-64. (85-64-07)
I i -
In OP-SO.PB-001, 4.16 KV System Operation, Revision 0:
a) Sections 5.8.3.2, 5.8.3.3, 5.9.4, and 5.9.5 provide instructions for removing breaker / controller control power I fuses but instruct the operator to place the operating switch in off after the fuses are removed. The switches should be j placed in off prior to fuse removal to prevent inadvertent I
automatic or remote operation during fuse removal to avoid high
! interruption currents and potential personnel hazards.
b) Typographical errors in both Sections 2.1.2 and 2.4.2 incorrectly referring the ope'ator to Attachment 2 should read Attachment 1.
I'
l 3
c) Typographical errors in Section 2.1.5 identifying the non-Class 1E lockout relays list as Attachment 3 should read Attachment
- 2. See Section 2.9.3 of Inspection Report 85-64. (85-64-08)
I&C Functional Test IC-FT.BJ-001, HPCI Division I Functional Test, Low CST Suction Switchover, contained acceptance criteria for CST level inconsistent with TS 4.3.1. The TS provided an nominal setpoint of 2.76% and a maximum allowable setpoint of 3.6% of level span. The procedure's acceptance criteria were 2.96 inches of water (column) and 3.6 inches of water (column).
The licensee provided Form SA-AP .XX-042.1, Field Questionnaire dated November 7, 1985 which had been forwarded to site engineering identifying and requesting resolution of the discrepancy.
Accompanying meeting notes of the Setpoint Calculation Group / Operations Interface Meeting of November 21, 1985 indicated that a TS change would be processed to correct the units to " inches" and that the procedure would be changed to correct the actual value.
See Section 2.10.3 of I.R. 85-64. (85-64-09)
TS 4.5.1.b.2 requires the LPCI pump functional test and flow 1
verification to be performed against a line pressure corresponding
- to a pump back pressure of greater than or equal to 20 psid.
! OP-IS.BC-003 and _004, RHR Pump Inservice Tests, Revision 0, did not address this test condition in the body of the procedure or accep-tance criteria, l
j The licensee advised that the acceptance criteria and the above test l condition would be included in the IST baseline and test data sheet which would be appended to the procedure when baseline data was eventually taken. This data sheet is intended to " stand alone" for each procedure, permitting the data sheet to be revised for new data or limits without requiring revision of the procedure body.
This finding and proposed licensee resolution is similar to that discussed in Sections 2.6 (RCIC) and 2.12 (Core Spray System (CSS) of this report. See Sections 2.11.3 and 2.12.3 of Inspection Report 85-64. (85-64-10) l
NRC Form 6, Rev. Oct. 80 OUTSTANDING ITEMS FILE SINGLE DOCKET ENTRY FORM Transaction Type
~-X New Item
___ Modify Delete Docket Number Borchardt Strosnider 50-354 Originator Reviewing Supervisor Item No. Type Module # Area Resp. Action Due Date Updt/Close Date 85-64-01 IFI 371301 PSC Z/U 4/17/86 Originator Modifier / Closer Borchardt
Description:
Correct control room labeling of drywell pressure instrument (page 12)
Item No. Type Module # Area Resp. Action Due Date Updt/Close Date 85-64-02 IFI 371301 PSC Z/U 4/17/86 Originator Modifier / Closer Borchardt
Description:
Correct labeling for Reactor Building - Torus Vacuum Relief Valve H (Page 12)
Item No. Type Module # Area Resp. Action Due Date Updt/Close Date 85-64-03 IFI 371301 PSC Z/U 4/17/86 l
Originator Modifier / Closer Borchardt
Description:
Provide dccumentation provisions to implement OP-ST.GR-001 (page 17)
, Item No. Type Module # Area Resp. Action Due Date Updt/Close Date l 85-64-04 IFI 371301 PSC Z/U 4/17/86 l Originator Modifier / Closer Borchardt
Description:
Reset HPCI and RCIC High Steam Flow Isolation Time Delay relay (Page 21) l l
2
- Item No. Type Module # Area Resp. Action Due Date Updt/Close Date 85-64-05 -l FI 371301 PSC Z/U 4/17/86 Originator Modifier / Closer Borchardt
Description:
Incorporate T.S. requirement into RCIC, LPCI and CS pump test procedures (page 21)
Item No. Type Module # Area Resp. Action Due Date Updt/Close Date 85-64-06 IFI 371301 PSC Z/U 4/17/86 Originator Modifier / Closer Borchardt
Description:
Correct diesel generator rating in T.S. (page 23)
Item No. Type Module # Area Resp. Action Due Date Updt/Close Date 85-64-07 IFI 371301 PSC Z/U 4/17/86 Originator Modi fier/ Closer Borchardt
Description:
Make corrections to MD-ST.PK-002 (page 25)
Item No. Type Module # Area Resp. Action Due Date Updt/Close Date 85-64-08 IFI 371301 PSC Z/U 4/17/86 Originator Modifier / Closer Borchardt
Description:
Make corrections to OP-SO.PB-001 4.16 KV System Operator (page 28)
Item No. Type Module # Area Resp. Action Due Date Updt/Close Date 85-64-09 IFI 371301 PSC Z/U 4/17/86 Originator Modifier / Closer Borchardt
Description:
Correct T.S. 4.3.1 units and IC-FT.BJ-001 setpoint limits (page 31)
Item No. Type Module # Area Resp. Action Due Date Updt/Close Date 85-64-10 IFI 371301 PSC Z/U 4/17/86
3 o Originator Modifier / Closer Borchardt
Description:
Incorporate limits and requirements into RCIC, LPCI and CS pump test procedures (page 34 and 36)
O NRC Form 766 U.S. NUCLEAR REGULATORY COMMISSION Principal Inspector:
Beckman, Donald A.
Reviewer: J. Strosnider INSPECTOR'S REPORT Office of Inspection and Enforcement Inspectors: Transaction Type Docket #s/ Inspect #s/ Seq #s Parameter Inc. I - Insert 050354 85-64 D. A. Beckman, M - Modify A. H. Imagana D - Delete R - Replace Licensee / Vendor:
Public Service Electric and Gas Company P. O. Box 236 Hancocks Bridge, N.J. 08038 Period of Inspection: Inspection Performed By: Organization Code of From To Region:
12/2/85 12/13/86 1 - Regional Office Staff Region Division Branch 2 - Resident Inspectors RI B A 3 - Performance Appr. Team
- 4 - Other - Parameter, Inc.
Regional Action: Type of Activity Conducted (*oie only):
1 - NRC Form 591 *02-Safety 07-Special 12-Shipment / Export
- 2 - Regional Office 03-Incident 08-Vendor 13-Import Letter 04-En f o rcua.ent 09-Mat. Acct. 14-Inqui ry 05-Mgmt. Audit 10-Plant Sec. 15-Investigation 06-Mgmt. Visit 11-Invent. Ver.
Inspection Findings: Total No. of Enforcement Report Contains Violations Conf. Held 2.790 Information A-0 ABCD B-1 - Clear C-2 - Violations 0-3 - Deviation 4 - Viol. & Dev.
Letter or Report Transmittal Oatt Letter Issued:
Letter or Report Transmittal Date Sent to HQ for Action:
Y i
~ j l
0 0
MODULE INFORMATION:
Rec- Module Direct Percentage Module Req.
ord No, Insp. Hrs. Complete Status Followup 371301 300 100 C