ML020250504
| ML020250504 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 01/11/2002 |
| From: | Bergendahl H FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2751, TAC M7958 | |
| Download: ML020250504 (145) | |
Text
FENOC Davis-Besse Nuclear Power Station 5501 North State Route 2 S~
Oak Harbor, Ohio 43449-9760 FirstEnergy Nuclear Operating Company Howard W Bergendahl 419-321-8588 Vice President - Nuclear Fax: 419-321-8337 Docket Number 50-346 License Number NPF-3 Serial Number 2751 January 11, 2002 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Third Ten-Year Interval Inservice Testing Program for the Davis-Besse Nuclear Power Station, Unit 1 Ladies and Gentlemen:
In accordance with 10 CFR 50.55a(f)(5)(i), the FirstEnergy Nuclear Operating Company (FENOC) herewith submits the revised Davis-Besse Nuclear Power Station (DBNPS),
Unit 1 Third Ten-Year Interval Inservice Testing (IST) Program. The DBNPS IST Program is based on the requirements of 10 CFR 50.55a(f) and the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance (OM) of Nuclear Power Plants, 1995 Edition with 1996 Addenda. The DBNPS has received approval from the NRC to extend the Second Ten-Year Interval and defer the start of the Third Ten-Year Interval until February 1, 2002 (TAC No. M7958).
Included as Sections 6.0, 7.0 and 8.0 of the DBNPS Third Ten-Year Interval IST Program are several relief requests. Similar relief requests have been previously approved for the DBNPS Second Ten-Year interval, for other plants, or has been proposed for acceptance by the NRC in the proposed rule change for 10 CFR 50.55a. A table of the relief requests and supporting cross-references is provided in Attachment 1.
NRC approval of these relief requests is requested by June 13, 2002 for the duration of the Third Ten-Year Interval IST Program.
Docket Number 50-346 License Number NPF-3 Serial Number 2751 Page 2 If you have any questions or comments, please contact David H. Lockwood, Manager Regulatory Affairs, at (419) 321-8450.
Attachments cc: J. E. Dyer, Regional Administrator, NRC Region III S. P. Sands, DB-l NRC/NRR Project Manager C. S. Thomas, DB-1 Senior Resident Inspector Utility Radiological Safety Board
Docket Number 50-346 License Number NPF-3 Serial Number 2751 Page 1 of 1 Cross-Reference Table for DBNPS Third Ten-Year Interval IST Program Relief Requests and Corresponding Supporting Reference DBNPS 3rd Ten-Year Interval IST Program Relief Request Number/Subject Supporting Reference RG-1 Seabrook Station NRC Safety Evaluation Authorized Nuclear Inservice Inpector Report (SER), TAC No. MA8532, Section 3.4 RP-1 Perry Unit 1, 2 d Ten-Year IST Interval, NRC Digital Instrument Calibration Range SER, March 31, 1999 (TAC No. MA3328),
Section 3.2.
RP-2 DBNPS 2nd Ten-Year IST Interval, NRC SER, Component Cooling Water Pumps TAC No. M89034, Section 3.0; Beaver Valley Unit 2 NRC SER, TAC No.
M98909, Section 5.0 RP-3 DBNPS 2 nd Ten-Year IST Interval, EDG Fuel Oil Transfer Pumps TAC No. M84151, Section 3.0 RP-4 DBNPS 2 nd Ten-Year IST Interval, NRC SER, Service Water Pumps TAC No. M76025, Section 2.2.3; Beaver Valley Unit 2 NRC SER TAC No.
M98909, Section 6.0 RV-1 Perry Unit 1 2 nd Ten-Year IST Interval NRC Containment Vacuum Relief Check Valves SER, August 9, 1999, TAC No. MA3328, Section 3.7; Sequoyah, Units 1 and 2 NRC SER, TAC Nos.
MB 1502 and MB 1503, Section 2.1 RV-2 Proposed Rule 10 CFR 50.55a, Industry Codes SW Header to CCW Isolation Valves and Standards, RIN 3150-AG61, Aug. 3, 2001, Federal Register, (Vol. 66, No. 150, 40640)
Docket Number 50-346 License Number NPF-3 Serial Number 2751 Page 1 Davis-Besse Nuclear Power Station, Unit 1 Third Ten-Year Interval Inservice Testing Program (105 Pages, with 35 Drawings, Follow)
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Third Ten Year Interval Inservice Testing Program Plan REVIEW/APPROVAL PREPARED BY:
DATE: /.2 /!7 A/
REVIEWED BY:
DATE: _
1__"_/\\\\_
APPROVED BY:
APPROVED BY:
REVIEWED BY:
Supervisor, Test/Performance Engineering DATE: i Z. 1) 0J DATE: ___-"___
Manager, Plant Engineering Au rzeNclar DATE:nr Authorized Nuclear InservAinpco 1 of 105 I
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Fr*ep ucerCo7.,7.
Revision 0 TABLE OF CONTENTS Page
1.0 INTRODUCTION
3 2.0 PUM P TESTING...............................................................................
. 6 3.0 VALVE TESTING............................................................................
6 4.0 TEST TABLE LEGENDS AND DEFINITIONS.............................
7
5.0 REFERENCES
13 6.0 GENERAL RELIEF REQUESTS.....................................................
15 7.0 PUMP RELIEF REQUESTS...........................................................
17 8.0 VALVE RELIEF REQUESTS...........................................................
26 9.0 COLD SHUTDOWN JUSTFICATIONS..........................................
29 10.0 REFUELING JUSTIFICATIONS.....................................................
77 11.0 SAMPLE DISASSEMBLY JUSTIFICATIONS..............................
85 12.0 PUMP TEST TABLE.....................................................................
87 13.0 VALVE TEST TABLE.....................................................................
88 2 of 105
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 F, stEnerozy Nuc!ear Oparatino Company
1.0 INTRODUCTION
1.1 General This document describes the Davis-Besse Nuclear Power Station (DBNPS) Inservice Testing (IST) Program for pumps and valves for the third ten year interval. The IST Program has been prepared to comply with the requirements of 10 CFR 50.55a(f),
Inservice testing requirements and the ASME OM Code-1995 Edition, 1996 addenda.
DBNPS has received NRC approval to extend the 2nd 120-month interval and defer the start of the 3rd 120-month interval to February 1, 2002 (TAC NO. M7958). On September 11, 2000 a request for an alternative pursuant to 10 CFR 50.55a(a)(3)(i) was submitted to the NRC (Serial Number 2668) to implement a Risk-Informed Inservice Testing Program for Air Operated Valves (AOV). Currently, this request is not approved, subsequently all components meeting the scoping criteria for inclusion in the IST Program will meet the requirements of the ASME OM Code-1995 Edition, 1996 addenda.
1.2 Background
The Construction Permit for Davis-Besse Nuclear Power Station (DBNPS) was issued on March 24, 1971 therefore 10 CFR 50.55a(f)(2) applies for design and accessibility.
DBNPS is currently in the 3rd 120-month interval, which began on September 12, 2000.
Full implementation of the new plan will begin on February 1, 2002. The 3rd Ten-Year Interval will end on September 12, 2010.
1.3 IST Program Scope Pursuant to 10 CFR 50.55a(f)(4) pumps and valves which are classified as ASME Code Class 1, Class 2 and Class 3 must meet the inservice testing requirements set forth in the ASME OM Code and addenda to the extent practical within the limitations of design, geometry and materials of construction of the components.
ISTA 1.4, Owners Responsibility, states that the Owner shall determine the appropriate Code Class for each component of the power plant using the classification criteria specified in 10 CFR 50.55a. DBNPS defines 'ASME Class 1, 2, or 3' components as those components that are within the ASME XI Inservice Inspection Program boundaries for Class 1, 2, or 3, or IWE as shown on the Inservice Inspection Drawings (ISID).
These classifications were developed using, Regulatory Guide 1.26, 1976 and NUREG-0800 Section 3.2.2 (10 CFR 50.55a, Footnote 9).
Certain components not meeting the DBNPS definition of 'ASME Class 1, 2, or 3' have been added to the IST Program. These components are tested in accordance with the ASME OM Code and have been given a NC (Not Classified) designation in the test tables.
3 of 105
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FrstEnergy Nuclear Operating Company Components classified as Class 1, 2 or 3 that are required to perform a specific safety function in support of any of the following (ISTB 1.1 and ISTC 1.1) are included in the IST Program:
- a.
Shutting down the reactor to the safe shutdown condition.
- b.
Maintaining the reactor in the safe shutdown condition.
- c.
Mitigating the consequences of an accident.
1.4 Safe Shutdown DBNPS defines the criteria for safe shutdown as that station condition in which the reactor is 1.0 percent sub-critical, and the Reactor Coolant System temperature and pressure are in the normal operating range, as defined in Section 7.4.1 of the DBNPS Updated Safety Analysis Report (USAR). The DBNPS Technical Specifications define this condition as HOT STANDBY (MODE 3). The DBNPS licensing basis, safe shutdown condition, does not require COLD SHUTDOWN (MODE 5). Therefore pumps and valves needed to achieve Cold Shutdown from the Hot Standby condition are not required to be included in the IST Program.
1.5 Mitigating the Consequences of an Accident DBNPS defines equipment that 'mitigates the consequences of an accident' as equipment whose malfunction could increase offsite dose beyond the licensing limits described in the USAR or the Davis-Besse Safety Evaluation Report (NUREG-0 136).
1.6 Exclusions and Exemptions The following components are excluded and/or exempted from testing under the IST Program.
- 1.
ASME Class 1, 2, or 3 pumps not provided with an emergency power source.
- 2.
ASME Class 1, 2, or 3 pumps provided with an emergency power source solely for operating convenience.
- 3.
Pump drivers except where the pump and driver form an integral unit and the pump bearings are in the driver.
- 4.
Valves that have no specific safety function as defined in Section 1.3, and are used solely for:
- a.
Operating convenience (such as manual vent, drain, instrument, or test valves)
- b.
System control (such as pressure regulating valves)
- c.
Maintenance
- 5.
External control and protection systems responsible for sensing plant conditions and providing signals for valve operation.
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Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FirstEnargy Nuclear OperaitrQ Co7 np 7
- 6.
Skid mounted components tested as part of the major component.
1.7 Augmented Testing Program The Augmented Testing Program provides for periodic testing of pump and valve functions that are outside the scope of the IST Program. Component functions are included in the Augmented Testing Program based on their importance, commitments, or DBNPS management decision. The Augmented Testing Program is prepared and maintained separately from the IST Program and is not subject to the requirements of 10 CFR 50.55a. It is not intended that all component testing outside the IST program be included in the Augmented Testing Program.
1.8 Program Changes All IST Program changes will be documented. Changes relating to the following items may be performed without obtaining prior NRC approval.
- 1.
Addition or removal of components from the IST Program or alteration of component testing requirements or testing frequency based upon the Code of Record.
- 2.
Rescinding (temporarily or permanently) NRC approved Relief Requests, provided the requirements of the Code of Record are met.
- 3.
Incorporating guidance contained in Generic Letter 89-04, Guidance on Developing Acceptable Inservice Testing Programs.
- 4.
Incorporating guidance contained in NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, as allowed by Generic Letter 89-04, Supplement 1.
- 4.
Incorporating later editions of the ASME Code, entirely or in part, as allowed by 10 CFR 50.55a.
- 5.
Adoption of approved Code Cases referenced in Regulatory Guide 1.147 and/or 10 CFR 50.55a.
1.9 Code Cases No Code Cases are adopted as part of this update.
1.10 General Relief Requests General Relief Request RG-1 eliminates the duties of the Authorized Nuclear Inservice Inspector.
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Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program
,Revision 0
FirstEnergy Nuclear Operating Company 2.0 PUMP TESTING 2.1 General Pump Relief Requests Pump Relief Request RP-1 asks for alternate ranges for digital test instruments.
2.2 Pump Relief Requests Pump Relief Request RP-2 asks for permission to use reference delta pressure verses flow curves in-lieu of a fixed reference point test for the Component Cooling Water Pumps.
Pump Relief Request RP-3 asks for alternate testing requirements for the Diesel Fuel Oil Transfer Pumps.
Pump Relief Request RP-4 asks for permission to use reference delta pressure verses flow curves in-lieu of a fixed reference point test for the Service Water Pumps.
3.0 VALVE TESTING 3.1 General Valve Relief Requests None 3.2 Valve Relief Requests Valve Relief Request RV-1 asks for a leakage test frequency in accordance with 10 CFR 50 Appendix J Option B for the Containment Vacuum Relief Check Valves.
Valve Relief Request RV-2 asks for a 2-year frequency for exercising the Service Water manual valves.
3.3 Cold Shutdown Justifications There are 48 Cold Shutdown Justifications allowed by ISTC 4.2.2(c) and 4.5.2(b),
Exercising Requirements.
3.4 Refueling Justifications There are 8 Refueling Justifications allowed by ISTC 4.2.2(d) and 4.5.2(c), Exercising Requirements.
3.5 Sample Disassembly Justifications There are 2 Sample Disassembly Justifications allowed by ISTC 4.5.4(c), Valve Obturator Movement.
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FENOC FirstErargy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 4.0 TEST TABLE LEGENDS AND DEFINITIONS
4.1 Systems
11 16 18 20 24 37 38 45 49 50 51 52 60 61 64 65 74 83 4.2 Pump Type:
CHP CVL
4.3 Parameter
NM NV Service Water Component Cooling Water Station Air/ Instrument Air Station Drainage Diesel Air Demineralized Water Sampling Feedwater Borated Water Storage/Decay Heat Auxiliary Feedwater Core Flood High Pressure Injection Containment Ventilation Containment Spray Reactor Coolant System Makeup System Nitrogen System Main Steam System Centrifugal, Horizontally Mounted Pump Centrifugal, Vertical Line Shaft Pump Not Measurable, See the Relief Request Not a Variable Speed Drive 7 of 105
FENOC FirsmEnergy Nuclear Operatinq Company 4.4 Code Classification:
1 2
3 NC Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Components classified by DBNPS as Class 1 Components classified by DBNPS as Class 2 or 1WE Components classified by DBNPS as Class 3 Components Not Classified by DBNPS as 1, 2, or 3 but are tested in accordance with the ASME OM CODE.
4.5 Valve Categories:
A B
C Valves (other than self actuated) which require a seat leakage measurement Valves (other than self actuated) which do not require a seat leakage measurement Valves that are self-actuated Category C valve which also requires a seat leakage measurement A special classification for power operated stop check valves which do not require a seat leakage measurement AC BC 4.6 Valve Function:
At Active Passive P
4.7 Valve Types:
AN BF BL BS CK DA GL GT RL Angle Valve Butterfly Valve Ball Valve Balanced Stop Valve Check Valve Diaphragm Valve Globe Valve Gate Valve Safety/Relief Valve (Continued) 8 of 105
FENOC FrslEnergy Nuclear Operating Cmpany 4.7 Valve Types:
(Continued)
SC SV TW VR Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Stop Check Valve Solenoid Valve Three Way Valve Vacuum Relief Check Valve 4.8 Actuator Types:
AO MA MO SA SO Air Operated Manually Operated Motor Operated Self Actuated Solenoid Operated 4.9 Valve Positions:
AI C
0 O/C As-is Closed Valves whose normal position may be either open or closed, depending on plant configuration.
Locked closed Locked open No Fail Position Open Valves that have safety functions in both the open and closed position.
Throttled T
4.10 Justifications:
CS RJ SDJ Cold Shutdown Justification Refueling Justification Sample Disassembly Justification 9 of 105
FENOC FwrslEnergy Nuclear Operatg Company 4.11 Relief Requests:
RG RP RV 4.12 Valve Test Codes:
EX FO FC FF LU LK TO TC PF PV RF SC SD SO SR WT 4.13 Test Frequencies:
10y 2Y 5Y Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 General Relief Request Pump Relief Request Valve Relief Request Manual exercise test open and closed Open fail-safe test Close fail-safe test Full forward flow test Leak test per Appendix J Leak test (for reasons other than Appendix J)
Stroke time open Stroke time closed Partial forward flow Valve position verification Reverse flow closure test Stroke closed, timing not required Sample disassembly Stroke open, timing not required Safety relief valve tests Weight Test Every ten years (also in accordance with group sampling requirements for safety/relief valves)
Every two years (or once per cycle)
Every five years (also in accordance with group sampling requirements for safety/relief valves)
As required by 10 CFR 50, Appendix J, Option B B
(Continued) 10 of 105
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FirstEnergy Nuclaar Operating Com*pany 4.13 Test Frequencies:
(Continued)
C Cold shutdown Q
Quarterly R
Every refueling outage RE Every even numbered refueling outage RO Every odd numbered refueling outage Q/2Y Quarterly to support Group A or B pump tests and every 2 years for comprehensive pump testing SI Testing in one direction is performed during shutdown and testing in the other direction is performed at the same interval.
4.14 Miscellaneous:
AC Alternating Current AFP Auxiliary Feedwater Pump AFPT Auxiliary Feedwater Pump Turbine AFW Auxiliary Feedwater ALARA As Low As Reasonably Achievable BWST Borated Water Storage Tank CCW Component Cooling Water CFT Core Flood Tank CIV Containment Isolation Valve CRDC Control Rod Drive Cooling CS Containment Spray CST Condensate Storage Tank CTMT Containment D/P Differential Pressure DBNPS Davis-Besse Nuclear Power Station DC Direct Current DH Decay Heat DW Demineralized Water (Continued) 11 of 105
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FirsrErmerqy Nuclear Operating Company 4.14 Miscellaneous Acronyms: (Continued)
ECCS Emergency Core Cooling System EDG Emergency Diesel Generator FW Feedwater H2 Hydrogen HPI High Pressure Injection HVAC Heating Ventilation and Air Conditioning HX Heat Exchanger IA Instrument Air ISID Inservice Inspection Drawings Showing the Owner Specified Code Classifications (Class 1, 2, 3 or IWE)
Piping and Components LOCA Loss of Coolant Accident LPI Low Pressure Injection MDFP Motor Driven Feedwater Pump MS Main Steam MU Makeup NA Not Applicable OTSG Once Through Steam Generator RCP Reactor Coolant Pump RCS Reactor Coolant System RPS Reactor Protection System SA Station Air SFAS Safety Features Actuation System TRN Train 12 of 105
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FirstEnerqy Nuclear Operatin 2
Company
5.0 REFERENCES
Title 10 Code of Federal Regulations, Energy, March 26, 2001 Proposed Rule 10 CFR 50.55a, Industry Codes and Standards, RIN 3150-AG61, Aug. 3, 2001, Federal Register, (Vol. 66, No. 150, 40640)
Regulatory Guide 1.26, Quality Group Classifications And Standards For Water-, Steam-, And Radioactive-Waste-Containing Components Of Nuclear Power Plants, February, 1976 Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program NUREG-0800, Standard Review Plan, Section 3.2.2, System Quality Group Classification, Generic Letter 89-04, Guidance On Developing Acceptable Inservice Testing Programs, April 3,1989 Generic Letter 89-04, Supplement 1: Guidance On Developing Acceptable Inservice Testing Programs, April 4,1995 NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, April 1995 Summary of Public Workshop Held in NRC Regions On Inspection Procedure 73756, "Inservice Testing of Pumps and Valves," And Answers to Panel Questions on Inservice Testing Issues, July 18, 1997 TAC NO. MA7958, Authorization of an Alternative to 10 CFR 50.55a(f)(4)(ii) Regarding Schedule for Submitting an Inservice Testing Program Plan Update - Davis Besse Nuclear Power Station, February 10, 2000 TAC NO. MB0520, Request for Additional Information (RAI) Regarding Relief Request to Implement a Risk Informed Inservice Testing Program for Air-Operated Valves at Davis-Besse Nuclear Power Station, June 4, 2001 TAC NO. M76025, Second 10-Year IST Program at DBNPS, December 2, 1991 TAC NO. M89034, Relief from Certain ASME Code Requirements for Inservice Testing for Davis-Besse Unit 1 Nuclear Power Station, October 7, 1994 TAC NO. M84151, Davis-Besse Inservice Testing of Pumps and Valves, April, 23, 1993 TAC NO. M98909, Safety Evaluation of Relief Requests for Second 10-Year Interval for Pumps and Valves Inservice Testing (IST) Program - Beaver Valley Power Station, Unit No. 2, November 18, 1997 TAC NO. MA8532, Safety Evaluation of Relief Requests for Second 10-Year Interval Inservice Testing Program Plan, Seabrook Station, Unit No. 1, November 1, 2000 TAC NO. MB 1502 and MB 1503, Sequoyah Nuclear Plant, Units 1 and 2 - Approval of Inservice Testing Relief Request, May 22, 2001 (Continued) 13 of 105
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FirsiEnerqy Nuclear OperatirQ Compar'y
5.0 REFERENCES
(Continued)
TAC NO. MA3328, Safety Evaluation of the Inservice Testing (IST) Program Relief Requests for Second 10-Year Interval - Perry Nuclear Power Plant, Unit 1, March 31, 1999 and August 9, 1999.
ASME OM CODE-1995 Edition, 1996 Addenda Davis-Besse Nuclear Power Station Technical Specifications Davis-Besse Nuclear Power Station Updated Safety Analysis Report Inservice Inspection Boundary Diagrams, ISID2-001 through ISID2-046 (35 Drawings) 14 of 105
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FirslEnergy Nuclear Operating C paov 6.0 GENERAL RELIEF REQUEST General Relief Request RG-1 Code Requirement:
ASME OM Code-1995 Edition, 1996 Addenda, ISTA 1.4(f), Possession of an arrangement with an Authorized Inspection Agency, ISTA 1.5, Accessibility of the Inspector and ISTA 2.1, Inspection.
Basis for Relief:
In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives would provide an acceptable level of quality and safety.
In the ASME OMb Code-1997 Addenda to the ASME OM Code-1995 Edition for Operation and Maintenance of Nuclear Power Plants, ISTA 1.4, Owner's Responsibility was rewritten deleting the requirement for possession of an arrangement with an Authorized Inspection Agency.
ISTA 1.5 was written to eliminate reference to access provisions for the Inspector, but the requirements for access provisions for examination personnel and equipment remain. ISTA 2.1, which detailed specific requirements for access for the Inspector, qualification of the Authorized Inspection Agencies, Inspectors and Supervisors and the duties of the Inspector, has been deleted in its entirety. The above is also true for ASME OM Code-1998 Edition, through the 2000 addenda.
On August 3, 2001 the NRC published in the Federal Register the proposed rule change for 10 CFR 50.55a, RIN 3150-AF61. The proposed revision to 10 CFR 50.55a endorses ASME OMb Code 1997 Addenda, ASME OM Code-1998 Edition, through the 2000 addenda without exception, as it pertains to the deletion of the activities to the ANII.
ANSI Part N626.1 describes the qualifications and duties for ANIls, which are applicable to Section X1. This part specifically addresses the duties to verify nondestructive tests, welding, heat treatment, and repairs and replacements; but is silent on the responsibilities concerning IST. Furthermore, ANII review of inservice testing programs is far less comprehensive than the reviews performed on inservice inspection activities.
The ANII inspection of inservice test programs consists of a review of the inservice test plan and a records review of tests performed.
FENOC Quality Assurance Program also performs these inspections and oversight functions. These inspection activities are being duplicated by the two separate organizations. There is no added safety or quality-related benefit in this duplication.
(Continued) 15 of 105
FENOC FirsiEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 General Relief Request RG-1 (Continued)
Alternative:
Specific requirements for Access for the Inspector, Qualification of the Authorized Inspection Agencies, Inspectors and the Duties of the Inspector (ANII) will not be addressed in the Inservice Testing Program.
FENOC's Quality Assurance Program processes provide an equivalent, or greater, level of quality and safety.
This Relief Request was previously approved for the Safety Evaluation of Relief Requests For The Second 10-Year Interval Inservice Test Program Plan, Seabrook Station, Unit No. 1 (TAC NO. MA8532).
Note:
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FENOC F stEnergy Nuclear Op erating Company 7.0 PUMP RELIEF REQUESTS Pump Relief Request RP-1 Applicability:
Code Requirement:
Basis for Relief:
Alternative:
Note:
Class 2 and 3 Pumps ISTB 4.7.1(b)(2) - Digital instruments shall be selected such that the reference value shall not exceed 70% of the calibrated range of the instrument.
In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives would provide an acceptable level of quality and safety.
Plant process computer points are used for instrumentation in inservice testing of pumps. The computer points are used in lieu of the associated analog indicators in order to meet ASME Code instrument accuracy requirements. In addition to using computer points, temporary digital instruments (M&TE) are also used in pump testing. In many cases the reference values exceed 70% of the computer point or temporary digital instrument range. The basis for the 70% originated from ASME Section Xl (IWA 5264), which provided requirements for pressure testing instrumentation ranges and to ensure readings in the required action range would be on scale. Since the computer points use permanent plant instrumentation as input, the ranges, by design, are selected to account for all expected operating and testing conditions. Surveillance tests are written such that the temporary instrumentation (digital or analog) is not over-ranged. In addition, digital instrumentation is significantly less susceptible to damage from over ranging and the accuracy of a digital instrument is precise throughout its full-calibrated range.
Tables ISTB 5.2.1-2 and 5.2.2-1,which list the acceptance criteria for quarterly testing, state that the maximum acceptable value of the measured parameter is 110% of the reference value. Table ISTB 5.2.3-1, which list the acceptance criteria for comprehensive testing, states that the maximum acceptable value of the measured parameter is 103% of the reference value.
Digital instruments used to verify the required action levels of Tables ISTB 5.2.1-2, and 5.2.2-1 will be selected such that the reference value shall not exceed 90% of the calibrated range.
Digital instruments used to verify the required action levels of Tables ISTB 5.2.3-1 will be selected such that the reference value shall not exceed 97% of the calibrated range.
These proposed alternatives will provide an acceptable level of quality and safety.
A similar Relief Request was previously approved in the Safety Evaluation of the Inservice Testing (IST) Program Relief Requests for the Second Ten Year Interval - Perry Nuclear Plant, Unit 1 (TAC No. MA3328, March 31, 1999, Section 3.2.)
17 of 105 Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0
FENOC FirstEnergy Nuclear Operating Company System:
Pumps:
Safety Function:
Code Requirements:
Basis for Relief:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Pump Relief Request RP-2 Component Cooling Water P43-1, P43-2 & P43-3, Component Cooling Water Pumps, Class 3, Group A Provides cooling water to transfer heat from safety-related equipment to the Service Water System.
ISTB 5.2. 1(b) - For centrifugal and vertical line shaft pump, the resistance of the system shall be varied until the flow rate equals the reference point. The differential pressure shall then be determined and compared to the reference value.
ISTB 5.2. 1(d) - Vibration (displacement or velocity) shall be determined and compared with the reference value.
ISTB 5.2. 1(e) - All deviations from the reference values shall be compared with the ranges of tables ISTB 5.2.1-1 and 5.2.1-2.
In accordance with 10 CFR 50.55a(f)(6)(i), relief is requested on the basis that the compliance with the Code requirements is impractical and that the proposed alternatives will provide reasonable assurance the components are operationally ready.
The CCW system was not designed with installed pump test lines. To achieve the same operating point for each test manual butterfly valves, which are not designed to throttle flow, would be used. In addition repeatability using these valves to throttle is poor. Depending on plant operating and climatic conditions, the cooling requirements range from minimum cooling loads (=3000 GPM) to 100 percent (=8000 GPM).
System operating conditions do not allow adjusting system resistance without significant impact on the plant's thermal stability.
A fixed flow rate through the pump aligned to the essential and non essential loads cannot be accomplished because system resistances are continuously varying and flows to parallel loads are dependent on each other. Spent Fuel Cooling and Boric Acid Evaporators have temperature control valves, which vary demand on the CCW system according to heat load. Component cooling water flow to the Reactor Coolant Pump coolers varies dependent on the throttle valve positions on the supply lines for the four pumps. Component cooling water flow to the Control Rod Drives pass through filters whose flow will change dependent on filter loading.
(Continued) 18 of 105
FENOC FifstEnerpy Nuclear Operating Compaoy Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Pump Relief Request RP-2 (Continued)
Alternate Testing:
Note:
ISTB 4.5, "Establishment of Additional Set of Reference Values",
provides for multiple sets of reference values. A pump curve is merely a graphical representation of the fixed response of the pump to an infinite number of flow conditions, which are based on some finite number of reference values verified by measurement.
Based on the lack of designed throttling capability, damage to the plant's equipment or a plant transient/trip could occur if the resistance of the system is varied to achieve a single reference point for testing, it is impractical to perform testing in accordance with the Code requirements.
Pump reference curves (developed per the guidelines in NUREG-1482, Section 5.2, "Use of Variable Reference Values for Flow Rate and Differential Pressure During Pump Testing") will be used to compare flow rate with developed pump head at the flow conditions dictated by Component Cooling Water System loads each quarter. Baseline vibration data obtained at various flow points on the pump curve will be used to develop a vibration verses flow curve.
All deviations from the reference curves shall be compared with the ranges of tables ISTB 5.2.1-1 and 5.2.1-2.
This proposed alternative testing will provide reasonable assurance that the Component Cooling Water Pumps are operationally ready.
This Relief Request was previously approved for the Second Ten Year Inservice Test Program (TAC NO. M89034 Section 3.2) and Safety Evaluation of Relief Requests For The Second 10-Year Interval Inservice Test Program Plan, Beaver Valley Power Station Unit 2 (TAC NO.
M98909 Section 6.0) 19 of 105 Basis for Relief:
FENOC FirstEnerqy Nuclear Operating Company System:
Pumps:
Safety Function:
Code Requirement:
Basis for Relief:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Pump Relief Request RP-3 Diesel Fuel Oil Transfer P195-1 & P195-2, EDG Fuel Oil Transfer Pumps, Class 3, Group A Transfer diesel fuel oil from the Emergency Diesel Generator (EDG)
Fuel Oil Storage Tanks to the EDG Day Tanks.
ISTB 5.2. 1(b) - For centrifugal and vertical line shaft pump, the resistance of the system shall be varied until the flow rate equals the reference point. The differential pressure shall then be determined and compared to the reference value.
ISTB 5.2. 1(d) - Vibration (displacement or velocity) shall be determined and compared to the reference value.
Table ISTB 5.2.1 Group A Test Hydraulic Acceptance Criteria In accordance with 10 CFR 50.55a(f)(6)(i), relief is requested on the basis that the compliance with the Code requirements is impractical and that the proposed alternatives will provide reasonable assurance the components are operationally ready.
10 CFR 50.55a(f)(2) requires that Class 1 and 2 components be designed and be provided with access to enable testing if the construction permit was issue between January 1, 1971 and July 1, 1974. The EDG Fuel Oil Transfer System is Class 3, therefore are not designed to permit performance of Code inservice testing. These are canned rotor pumps, submerged inside the underground EDG Fuel Oil Storage Tank. There are no installed flow instrumentation, pressure instrumentation, valve test connections, or recirculation line.
Typically at other plants, the diesel generator fuel oil system for each diesel generator is safety related (i.e. seismic, Q) and consists of a large storage tank with a capacity sufficient for seven days of diesel generator operation, a fuel oil transfer system, and a day tank local to the diesel generator with a capacity on the order of four hours of diesel generator operation. In contrast, the original DBNPS fuel oil system consisted of a single large non-safety related above ground storage tank serving both diesel generators with seven days capacity, and a large safety related day tank for each diesel generator each with capacity for approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> of diesel generator operation. The day tanks were each provided with safety related fill connections to accommodate refilling the day tanks within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following an event if refill from the normal storage tank was unavailable.
(Continued) 20 of 105
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FirsEnergy Nudear Operating Company Pump Relief Request RP-3 Basis for Relief:
(Continued)
This design was modified to its current configuration at the request of the NRC during plant licensing. The current configuration consists of a safety related seven-day capacity underground storage tank for each diesel generator. Each of the seven-day underground storage tanks has an internally mounted submerged EDG Fuel Oil Transfer Pump normally supplying the corresponding 20-hour capacity day tank. The large 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> capacity day tanks and the safety related fill connections and the non-safety related above ground storage tank has been retained from the original design. Because of the large capacity of the day tanks, and the three diverse methods of replenishing the day tanks during diesel generator operation, the DBNPS diesel fuel oil transfer pumps are of lower safety significance than in typical fuel oil transfer systems with relatively small day tanks.
The EDG Fuel Oil Transfer Pumps are low flow, rated at 10 gpm. They automatically start on low EDG Day Tank level of seven feet, approximately 5050 gallons, then automatically shut off at seven and one-half feet; this corresponds to approximately 250 gallons pumped.
This safety feature maintains a minimum level as required per Technical Specification 3.8.1. An EDG Fuel Oil Storage Tank has a capacity of approximately 40,000 gallons. The EDG Day Tanks have a capacity of 6000 gallons.
EDG fuel design flow is = 4.5 gpm; therefore each day tank can last approximately 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br />. This time period is sufficient to allow offsite fuel oil delivery directly into the day tanks.
The only possible flow measurement is by measuring EDG Day Tank volume change over time. Error in measuring this volume is dependent on fuel oil temperature and a limited change in level indication because the EDG Day Tank has a large upper circular section. Flow rate is dependent upon EDG Fuel Oil Storage Tank level and fuel oil viscosity, which varies with environmental temperature conditions.
It is estimated that modification of the fuel oil transfer system to accommodate Code flow, differential pressure and vibration measurements would cost approximately $500,000. This modification would involve replacement of the existing pumps and their relocation external to the tanks, installation of flow test loops and installation of flow and pressure instrumentation. DBNPS considers an expenditure of this magnitude unwarranted considering the reduced safety significance of the DBNPS fuel oil transfer system as compared to typical designs.
(Continued) 21 of 105
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 FirstEnorgy Nuc1ea-7 Oeraig Coxnpeny Pump Relief Request RP-3 Basis for Relief:
(Continued)
It is impractical to take vibration measurements on these pumps. The pumps and motors are located inside the EDG fuel oil storage tank, are not accessible during operation, and are submersed in the fuel oil being pumped.
To date no maintenance has been required for these pumps. The pumps have successfully started and delivered fuel oil upon demand. The latest flow test indicates pump design flow rates are being met.
To perform the Code testing would require extensive plant modifications.
Performance of Code testing requirements without major modification to plant structure is impractical.
Alternate Testing:
No vibration monitoring will be performed.
Pump flow functional testing is performed each month as required per Technical Specification 4.8.1.1.2. The pumps are observed to automatically startwith a corresponding increase level in the EDG Day Tank.
Pump flow rate tests are performed each cycle. A predetermined oil level above the transfer pump will be set prior to testing. The flow rate is obtained by measuring a change in EDG Day Tank level over time. A EDG Day Tank level change of approximately 150 gallons or more shall be timed to determine flow rate.
Flow rate will be calculated from the known increase in EDG Day Tank level. Pump suction pressure shall be preset by fuel oil level adjustment.
Pump discharge is consistent since there are no throttle valves. Based upon these conditions pump flow rates should be repeatable and capable of predicting pump degradation.
A low required action range of less than 6 gpm will be used in lieu Table ISTB 5.2.1-2. This range will ensure EDG transfer pumps do not degrade below required design system flow requirements. Pump flow rates will be trended for degradation.
No alert levels will be specified hence required action will be performed if pump flow rate is determined to be outside the acceptable range.
(Continued) 22 of 105
FENOC fstEnerqy Nuclear Operatinq Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Pump Relief Request RP-3 (Continued)
Periodically, the EDG Fuel Oil Storage Tanks are drained, cleaned, and filled with fresh oil. The EDG Day Tanks are also drained, cleaned and inspected. At these times a long term pump duration test is possible.
The transfer pump will be required to consecutively pump 1000 gallons of fuel from the Emergency Diesel Generator Fuel Oil Storage Tank to the EDG Day Tank. Flow rate will be measured and evaluated for degradation.
These proposed alternative tests will provide reasonable assurance that the EDG Fuel Oil Transfer Pumps are operationally ready.
This Relief Request was previously approved for the Second Ten Year Inservice Test Program (TAC NO. M84151 Section 3.0) 23 of 105 Alternate Testing:
Note:
FENOC Firs Enerqy Nuclear GparatinQ Company System:
Pumps:
Safety Function:
Code Requirement:
Basis for Relief:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Pump Relief Request RP-4 Service Water P3-1, P3-2 & P3-3, Service Water Pumps, ASME Class 3, Group A These pumps provide cooling water from the Ultimate Heat Sink to safety-related equipment.
ISTB 5.2. 1(b) - For centrifugal and vertical line shaft pumps, the resistance of the system shall be varied until the flow rate equals the reference point. The differential pressure shall then be determined and compared to the reference value.
ISTB 5.2. l(d) - Vibration (displacement or velocity) shall be determined and compared with the reference value.
In accordance with 10 CFR 50.55a(f)(6)(i), relief is requested on the basis that the compliance with the Code requirements is impractical and that the proposed alternatives will provide reasonable assurance the components are operationally ready.
The Service Water System is in continuous operation during all modes of plant operation whose flow varies with the temperature requirement of the various safety and non-safety related loads. The system was not designed with installed pump test lines. System operating conditions do not allow adjusting system resistance without significant impact on the plants thermal stability. Depending on plant operating and climatic conditions, the cooling requirements range from minimum cooling loads
(=6000 GPM) to 100 percent (=10000 GPM) with many of the loads automatically placed in operation in response to local temperature requirements. Operating experience has shown that plant conditions due to heat loads requiring cooling by the Service Water System preclude setting the Service Water Pumps to the exact flow rate for a specific reference value.
ISTB 4.5, "Establishment of Additional Set of Reference Values",
provides for multiple sets of reference values. A pump curve is merely a graphical representation of the fixed response of the pump to an infinite number of flow conditions, which are based on some finite number of reference values verified by measurement.
(Continued) 24 of 105
FENOC FirsrEnerpy Nuclear Operating Compary Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Pump Relief Request RP-4 (Continued)
Alternate Testing:
Note:
Pump reference curves (developed per the guidelines in NUREG-1482, Section 5.2, "Use of Variable Reference Values for Flow Rate and Differential Pressure During Pump Testing") will be used to compare flow rate with developed pump head at the flow conditions dictated by Service Water System loads each quarter. Baseline vibration data obtained at various flow points on the pump curve will be used to develop a vibration verse flow curve.
All deviations from the reference curves shall be compared with the ranges of tables ISTB 5.2.1-1 and 5.2.1-2.
These proposed alternative tests will provide reasonable assurance that the Service Water Pumps are operationally ready.
This Relief Request was previously approved for the Second Ten Year Inservice Test Program (TAC NO. M76025 Section 2.2.3) and Safety Evaluation of Relief Requests For The Second 10-Year Interval Inservice Test Program Plan, Beaver Valley Power Station Unit 2 (TAC NO.
M98909 Section 6.0) 25 of 105
FENOC FirsfEnar~w Nuclear Operating Company 8.0 VALVE RELIEF REQUESTS Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Valve Relief Request RV-1 System:
Valve(s):
Safety Function:
Code Requirement:
Basis for Relief:
Containment Ventilation CV5080, CV5081, CV5082, CV5083, CV5084, CV5085, CV5086, CV5087, CV5088, & CV5089, Containment Vacuum Relief Check Valves, Class 2, Category AC These valves must open to prevent the Containment Vessel from exceeding its external design pressure (0.5 psid between the Containment Vessel and the Annulus). These valves must close for containment isolation.
ASME OM Code-1995, 1996 Addenda, Appendix 1, 11.3.7(b) - Leak tests shall be performed on all Class 2 and 3 containment vacuum relief valves at each refueling or every 2 years, whichever is sooner, unless historical data requires more frequent testing.
In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.
These Vacuum Relief Check Valves are functionally tested every refueling outage in accordance with Appendix I, 1 1.3.7(a) and 1 7.3.8(a).
This testing ensures that the Vacuum Relief Check Valves prevent the Containment Vessel from exceeding its external design pressure.
In addition to their primary function of protecting the Containment from an under-pressure condition these valves also serve as primary containment isolation valves and are required to be tested for leakage on a periodic basis. 10 CFR 50 Appendix J, which sets forth the rules and conditions for containment leakage rate testing, has a section designated Option B-Performance Based Requirements. This section permits leakage rate testing to be performed at intervals of up to 5 years, based on the valve's performance history. Option B eliminated the prescriptive requirements that were deemed marginal to safety, and allowed a components past performance to be the determining factor for the testing interval.
ASME OM Code-1998, 1999 Addenda, Appendix I, Section 1-1380
[formerly 11.3.7], Test Frequency, Class 2 and 3 Primary Containment Vacuum Relief Valves, was modified to accept 10 CFR 50 Appendix J for leak testing requirements.
(Continued) 26 of 105
FENOC FisrEnerqy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Valve Relief Request RV-1 (Continued)
Alternate Testing:
Note:
On August 3, 2001 the NRC published in the Federal Register (Vol. 66 No. 150) the proposed rule change for 10 CFR 50.55a, RIN 3150-AF61.
The proposed revision to 10 CFR 50.55a endorses these revisions to the ASME OM Code, without exception, as it pertains to relief valve leak testing.
The leakage rate testing of the Containment Vacuum Relief Check Valves will be performed in accordance with the requirements of 10 CFR 50 Appendix J, Option B (Performance-Based Requirements).
This proposed alternative testing will provide an acceptable level of quality and safety.
This Relief Request was previously approved for Perry Nuclear Power Plant Second Ten Year Inservice Test Program (TAC NO. MA3328 Section 3.7) and Safety Evaluation of Relief Requests For The Inservice Test Program, Sequoyah Nuclear Plant Units 1 and 2 (TAC NO.
MB 1502 and MB 1503, Section 2.1) 27 of 105 Basis for Relief:
FENOC FirstEnerqy NuclerOperating Company System:
Valve(s):
Safety Function:
Code Requirement:
Basis for Relief:
Alternate Testing:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Valve Relief Request RV-2 Service Water SW232, SW233, SW234 & SW236, SW Header to CCW Isolation Valves, Class 3, Category B These valves must be capable of being manually opened to provide a safety grade backup makeup water source to the CCW Surge Tank in the event that the normal non-safety grade makeup water source is unavailable.
ISTC 4.2.1, Exercising Test frequency. Active Category A and B valve shall be tested nominally every 3 months.
In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.
The affected manual valves are 1" globe valves that are normally closed and not normally operated during plant operation. These valves are only required if the normal non-safety related makeup source becomes unavailable and the seismically qualified Component Cooling Water System has lost inventory needing makeup, during a Seismic Event.
The predominant degradation and failure mechanisms (motor failures, electrical failures, switch settings, etc.) associated with power operated valves do not exist for manual valves. These valves have been tested every two years since November 1994 and every quarter since December 2000 with no failures or observable degradation.
These valves are located in a non-harsh environment. Testing has demonstrated that a 2-year exercising frequency is adequate to ensure the safety-related function of these valves has not degraded.
The 1999 addenda to 1998 Edition of the ASME OM Code, added ISTC 3540, Manual Valves, to the Code allowing a 5 year frequency for exercising manual valves located in a non-harsh environment.
On August 3, 2001 the NRC published in the Federal Register (Vol. 66, No. 150, 40640) the proposed rule change for 10 CFR 50.55a, RIN 3150 AF61. In this proposed rule change the NRC took exception to ISTC 3540 and required a 2-year exercising frequency for manual valves in 10 CFR 50.55a(b)(3)(vi).
SW232, SW233, SW234 and SW236 will be exercised every 2 years.
This proposed alternative testing will provide an acceptable level of quality and safety.
28 of 105
FENOC FirsEneqpy Nuclear Operatiq Cn oy 9.0 COLD SHUTDOWN JUSTIFICATIONS Cold Shutdown Justification CS-1 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Auxiliary Feedwater AF1 & AF2, Auxiliary Feedwater Pumps 1 & 2 Suction Line Check Valves Prevents reverse flow to the Condensate Storage Tank from the safety related Auxiliary Feedwater Pump suction supply source.
Verify forward flow and reverse flow closure quarterly.
The valve lineup to perform reverse flow testing, isolates one Auxiliary Feedwater Pump and the Condensate Storage Tanks suction source to both Auxiliary Feedwater Pumps. Service Water would be the only suction source to the remaining Auxiliary Feedwater Train. If the remaining Auxiliary Feedwater Pump were needed to feed the steam generators, Service Water (which is raw, untreated lake water) would be injected into the Steam Generators. This would cause chemical contamination of the steam generators, which could result in tube degradation, tube leakage, and reduced life expectancy of the Steam Generators. Service water provides an emergency safety related water supply to the Auxiliary Feedwater Pumps, to protect against a seismic event, which could result in the loss of the non-seismic normal water supply from the Condensate Storage Tanks.
AF1 and AF2 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
29 of 105 Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0
FENOC FirstEnergy Nudcar Oper7ating Ccoipany Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-2 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Auxiliary Feedwater AF15 & AF16, Auxiliary Feedwater Pump 1 & 2 Minimum Flow Check Valve These valves must open to pass forward flow to provide minimum flow protection for the Auxiliary Feedwater Pumps These valves must close to prevent over pressurization of the idle Auxiliary Feedwater Pump seals, suction piping, and bearing and governor cooling water piping.
Verify forward flow and reverse flow closure quarterly.
The valve lineup to perform reverse flow testing, isolates both Auxiliary Feedwater Pump's minimum recirculation lines rendering both Auxiliary Feedwater Trains inoperable. This would place the plant in Tech. Spec.
3.0.3 and require entry into a one-hour action statement.
AF15 and AF16 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
30 of 105
FENOC FirstEnergy Nuclear Operatin Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-3 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Auxiliary Feedwater AF39 & AF43, Auxiliary Feedwater to Steam Generators 1 & 2 Supply Check Valves Forward flow to the Steam Generators is required to ensure adequate decay heat removal capability exists.
Reverse flow closure is required to prevent steam binding of the Auxiliary Feedwater Pumps, and over-pressurization of the pump seals, suction piping, and bearing/governor cooling piping.
Verify forward flow and reverse flow closure quarterly.
Forward flow through these check valves to the Auxiliary Feedwater nozzles, spray water directly onto the tubes at the upper end of the Steam Generators. Injecting this relatively cold water during plant operation causes severe thermal stresses, which could damage the Steam Generators and may also lead to moisture carry-over in the steam, which could damage the main turbine. In addition, the auxiliary feedwater flow to the steam generators will cause an unacceptable plant transient.
AF39 and AF43 will be forward flow tested during cold shutdown and reverse flow closure tested at the same interval.
31 of 105
FENOC FirsEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-4 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Auxiliary Feedwater AF49 & AF52, MDFP to Auxiliary Feedwater Line 1 & 2 Discharge Check Valves.
AF49 and AF52, must prevent reverse flow to ensure auxiliary feedwater is directed to the Steam Generators, and does not back flow into the Motor Driven Feedwater System.
Verify forward flow and reverse flow closure quarterly.
The valve lineup to reverse flow test these check valves results in the inability to feed a Steam Generator from either Auxiliary Feedwater Pump and/or the Motor Driven Feedwater Pump. This lineup requires the closing of AF599 or AF608 which renders both trains of Auxiliary Feedwater and the Motor Driven Feedwater Pump System inoperable requiring plant shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in accordance with Technical Specification 3.7.1.2c.
AF49 and AF52 will be reverse flow tested during cold shutdown and forward flow tested at the same interval.
32 of 105
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-5 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Auxiliary Feedwater AF72 & AF73, Auxiliary Feedwater Pump 1 to Steam Generator 1 & 2 Supply Check Valves AF72 and AF73 must open to allow a forward flow to the Steam Generators to ensure adequate decay heat removal capability exists.
These valves also must close to prevent reverse flow to the opposite auxiliary feedwater train, if its associated Auxiliary Feedwater Pump is not operating or if insufficient discharge pressure exists.
Verify forward flow and reverse flow closure quarterly.
The valve lineup to reverse flow test any of these check valves results in the inability to feed a Steam Generator from either Auxiliary Feedwater Pump and/or the Motor Driven Feedwater Pump. This lineup requires the closing of AF599 or AF608 which renders both trains of Auxiliary Feedwater and the Motor Driven Feedwater Pump System inoperable requiring plant shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in accordance with Technical Specification 3.7.1.2c.
Forward flow through these check valves to the Auxiliary Feedwater nozzles, spray water directly onto the tubes at the upper end of the Steam Generators. Injecting this relatively cold water during plant operation causes severe thermal stresses, which could damage the Steam Generators and may also lead to moisture carry-over in the steam, which could damage the main turbine. In addition, the auxiliary feedwater flow to the steam generators will cause an unacceptable plant transient.
AF72 and AF73 will be forward flow and reverse flow closure tested during cold shutdown.
33 of 105
FENOC FirslEnergy Nuclear Operatinq CoMp&aM Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-6 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Auxiliary Feedwater AF74 & AF75, Auxiliary Feedwater Pump 2 to Steam Generator 1 & 2 Supply Check Valves AF74 and AF75 must open to allow a forward flow to the Steam Generators to ensure adequate decay heat removal capability exists.
Verify forward flow and reverse flow closure quarterly.
The valve lineup to reverse flow test any of these check valves results in the inability to feed a Steam Generator from either Auxiliary Feedwater Pump and/or the Motor Driven Feedwater Pump. This lineup requires the closing of AF599 or AF608 which renders both trains of Auxiliary Feedwater and the Motor Driven Feedwater Pump System inoperable requiring plant shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in accordance with Technical Specification 3.7.1.2c.
Forward flow through these check valves to the Auxiliary Feedwater nozzles, spray water directly onto the tubes at the upper end of the Steam Generators. Injecting this relatively cold water during plant operation causes severe thermal stresses in the Steam Generator and may also lead to moisture carry-over in the steam, which could damage the main turbine. The consequences of these effects make it undesirable to forward flow test these check valves quarterly.
AF74 and AF75 will be forward flow and reverse flow closure tested during cold shutdown.
34 of 105
FENOC FirstEnerqy Nucloar Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-7 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Auxiliary Feedwater AF599 & AF608, Auxiliary Feedwater to Steam Generator 1 & 2 Line Stop Valves These valves must close to isolate flow to their respective Steam Generator in the event of a steam line or feedwater line break.
Exercise and time closed quarterly.
Technical Specification 3.7.1.2c requires entry into a one-hour action statement if either AF599 or AF608 is closed. During a postulated feedwater or main steam line break affecting the opposite steam generator, a complete loss of auxiliary feedwater flow to both Steam Generators could occur. The control circuitry for these valves does not allow partial stroking.
AF599 and AF608 will be exercised and timed closed during cold shutdown.
35 of 105
FENOC FirslEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-8 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Auxiliary Steam AS274, Auxiliary Steam Header to Auxiliary Feedwater Turbine Check Valve Reverse flow closure of this valve provides the pressure boundary between the Auxiliary Feedwater Pump Turbine safety related main steam supply, and the non-safety related auxiliary steam system. This valve prevents the diversion of steam from the Auxiliary Feedwater Pump Turbines in the event of a failure of the non-safety related auxiliary steam system. AS274 is the ISI Class 3 boundary between the Auxiliary Feedwater System and the non-class Auxiliary Steam System, which is rated at 300 PSIG.
Verify forward flow and reverse flow closure quarterly.
During normal operation, the upstream manual isolation valve AS273 is maintained closed and is only opened in MODES 4, 5, and 6 for Auxiliary Feed Pump testing when the main steam lines are depressurized. There is no upstream test connection between AS274 and AS273. Testing AS274 with main steam pressure during normal operation would require opening AS273 potentially exposing low pressure piping to high-pressure steam. This poses a personnel safety concern and could result in over-pressurizing the lower pressure Auxiliary Steam System if leakage were to occur.
AS274 Will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
36 of 105
FENOC FistEnergy Nuclear Operating Corotany Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-9 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Component Cooling Water CC 17, CC18 & CC 19, Component Cooling Water Pump 1, 2 & 3 Discharge Line Check Valves These valves must open and be capable of passing sufficient forward flow to provide cooling of essential header loads.
These valves must close to prevent diverting CCW flow through an idle pump when two pumps are mechanically aligned to the same train.
Verify forward flow and reverse flow closure quarterly.
To test these valves, a forward flow rate of 7860 GPM must be achieved.
This flow rate is higher than the normal operating flow rates through the heat exchangers, and would disturb the normal thermal equilibrium of the non-essential header loads.
Testing of these valve at this flow rate during normal operation requires each pump to be aligned to the non-essential loads and increasing total CCW flow by manually throttling flow through the associated Decay Heat Removal Heat Exchanger. This alignment will result in the diverting CCW flow from the containment header, hence reducing flow to the Reactor Coolant Pump Seals and Motor coolers, Control Rod Drive Coolers, and the Letdown Coolers. This would increase RCP bearing and seal, CRD Stator, and letdown temperatures affecting plant operation and possibly result in equipment damage and/or reactor trip.
CC17, CC18, and CC19 will be forward flow tested during cold shutdown and reverse flow closure tested in same interval.
37 of 105
FENOC FirslEnergy Nuclear Operating Ccmpany Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-IO System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Component Cooling Water CC1407A & CC1407B, CCW Return Containment Isolation Valves These valves must close for containment isolation.
Exercise and time closed quarterly.
These valves are normally open to provide a component cooling water return flow path from the Reactor Coolant Pump seals and motors, the Letdown Coolers, and the Control Rod Drive Mechanisms. Disrupting cooling water flow to these components during normal operation would result in component damage and/or a plant trip or transient. The control circuitry for these valves does not allow partial stroking.
CC1407A and CC1407B will be exercised and timed closed during cold shutdown.
38 of 105
FENOC Fir~s1Erer~y NuciearOprtn op y
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-11 Component Cooling Water CC 1407C, CCW Containment Penetration 4 Thermal Expansion Check Valve Safety Function:
Code Testing:
Justification:
Alternate Testing:
CC1407C must close for containment isolation.
CC1407C must open to relieve pressure that could occur between the penetration isolation valves due to thermal expansion following a design basis LOCA Verify forward flow and reverse flow closure quarterly.
The valve lineup required to perform forward flow or reverse flow testing of CC1407C isolates the Component Cooling Water return flow path from the Reactor Coolant Pump Seals and Motors, the Letdown Coolers, and the Control Rod Drive Mechanisms. Disrupting cooling water flow through these components during normal operation would result in component damage and/or a plant trip or transient. In addition, CC1407C and other valves to be manipulated to perform this test are located within Containment.
CC1407C will be forward flow and reverse flow closure tested during cold shutdown.
39 of 105 System:
Valve(s):
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-12 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Component Cooling Water CC141 IA & CC14113B, Component Cooling Water Inlet Containment Isolation Valves These valves must close for containment isolation.
Exercise and time closed quarterly.
These valves are normally open to supply Component Cooling water to the Reactor Coolant Pump Seals and Motors, and the Letdown Coolers.
Disrupting cooling water flow to these components during normal operation would result in component damage and/or a plant trip or transient. The control circuitry for these valves does not allow partial stroking.
CC1411 A and CC 14111B will be exercised and timed closed during cold shutdown.
40 of 105
FENOC FirstEnerqy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-13 Component Cooling Water CC 1411 C, CCW Containment Penetration 3 Thermal Expansion Check Valve Safety Function:
Code Testing:
Justification:
Alternate Testing:
CC1411 C must close for containment isolation.
CC1411 C must open to relieve pressure that could occur between the penetration isolation valves due to thermal expansion following a design basis LOCA.
Verify forward flow and reverse flow closure quarterly.
The valve lineup required for forward flow or reverse flow closure testing of CC 1411 C isolates the Component Cooling Water supply to the Reactor Coolant Pump Seals and Motors, and the Letdown Coolers.
Disrupting cooling water flow through these components during normal operation would result in component damage and/or a plant trip or transient. In addition, CC1411C and other valves to be manipulated to perform this test are located within Containment.
These valves will be forward flow tested and reverse flow closure tested during cold shutdown.
41 of 105 System:
Valve(s):
FENOC FirslEnergy Nuclear Gpr77n 0 777py Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-14 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Component Cooling Water CC1567A & CC1567B, CCW Containment Isolation Valves to the Control Rod Drive Mechanisms.
These valves must close for containment isolation.
Exercise and time closed quarterly.
These valves are normally open to supply Component Cooling Water to the Control Rod Drive Mechanisms. Disrupting cooling water flow to these components during normal operation would result in CRD stator overheating and a reactor trip. The control circuitry for these valves does not allow partial stroking.
CC1567A and CC1567B will be exercised and timed closed during cold shutdown.
42 of 105
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-15 Component Cooling Water CC1568, CCW Containment Penetration 12 Thermal Expansion Check Valve Safety Function:
Code Testing:
Justification:
Alternate Testing:
CC 1568 must close for Containment isolation.
CC 1568 must open to relieve pressure that could occur between the penetration isolation valves due to thermal expansion following a design basis LOCA.
Verify forward flow and reverse flow closure quarterly.
The valve lineup required for forward flow or reverse flow testing of CC 1568 isolates the Component Cooling Water header supplying cooling water to the Control Rod Drive Mechanisms. Disrupting the cooling water supply to these components could result in stator damage and/or a plant trip or transient. In addition, CC1568 and other valves to be manipulated to perform this test are located within the Containment.
CC 1568 will be forward flow tested and reverse flow closure tested during cold shutdown.
43 of 105 System:
Valve(s):
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-16 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Component Cooling Water CC4100, CC4200, CC4300 & CC4400, Reactor Coolant Pump Seal Return Line Isolation Valves These valves must close if a tube rupture occurs, or excessive leakage develops, in the Reactor Coolant Pump Seal Cooler. If this occurs, these valves automatically close when the Component Cooling Water system pressure, sensed just upstream of the valves, reaches 150 psig.
Exercise and time closed quarterly.
Exercising any of these valves closed during normal operation will isolate the cooling water return flow path through the associated Reactor Coolant Pump Seal. Disrupting cooling water flow would result in damage to the pump seal and/or a plant trip or transient. The control circuitry for these valves does not allow partial stroking.
CC4100, CC4200, CC4300 and CC4400 will be exercised and timed closed during cold shutdown.
44 of 105
FENOC FrstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-17 Core Flood CF2C, Core Flood Penetration 47A Thermal Expansion Check Valve Safety Function:
Code Testing:
Justification:
Alternate Testing:
CF2C must close for containment isolation.
CF2C must open to relieve pressure that could occur between the penetration isolation valves due to thermal expansion following a design basis LOCA.
Verify forward flow and reverse flow closure quarterly.
Verification of CF2C forward flow during normal operation, without a Containment entry, would require borated water to be added to Core Flood Tank I through a 3/8" sample line, with CF2B closed. Flow would be indicated by a Core Flood Tank level change, or by an external flow instrument, which would rely on the absence of boundary valve leakage for accuracy. The Core Flood System is not designed to add borated water via the sample line.
To establish flow from Core Flood Tank 2 would require an approximate CFT pressure differential of 75 PSID, the forward flow set-point of CF2C. This would exceed the 50 pound maximum pressure differential between Core Flood Tanks allowed by TS 3.5.1. Additionally a Containment entry and entry into a locked high radiation area would be required to perform this testing.
CF2C will be forward flow tested and reverse flow closure tested during cold shutdown.
45 of 105 System:
Valve(s):
FENOC FirsrEnlerqy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-18 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Core Flood CF15 & CF16, CFT 1 & 2 Fill and Pressurizing Stop Check Valves These valves must close for containment isolation.
Verify forward flow and reverse flow closure quarterly.
CF15 and CF16 are inboard Containment Isolation Valves for Penetrations P44A and P71 C respectively. Forward flow testing these valves during normal operation would require the venting and re pressurizing the Core Flood Tanks. The Nitrogen vented during this process would enter the Radioactive Waste Gas system increasing the amount of gaseous radwaste. In addition, the venting challenges the ability to maintain the Core Flood Tanks between 575 and 625 PSIG in accordance with Tech Spec 3.5. ld which has a one hour action statement.
CF15 and CF16 will be forward flow tested during cold shutdown and reverse flow closure tested at the same interval.
46 of 105
FENOC FirstEnergy Nuclear Operating Comny System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-19 Ventilation and Atmospheric Monitoring CV 124 & CV 125, Containment Hydrogen Analyzer and Radiation Monitor Return Line Containment Isolation Check Valves These valves must close for containment isolation.
Verify forward flow and reverse flow closure quarterly.
Testing these valves during normal plant operation would require a Containment entry.
CV 124 and CV 125 will be reverse flow closure tested during cold shutdown and will be forward flow tested at the same interval.
47 of 105
FENOC FirtEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-20 System:
Valve(s):
Safety Function:
Ventilation and Atmospheric Monitoring CV209 & CV210, Containment Hydrogen Dilution Blower Discharge Containment Isolation Check Valves These valves must open to permit dilution of the post-LOCA containment atmosphere.
These valves must close for containment isolation.
Code Testing:
Justification:
Alternate Testing:
Verify forward flow and reverse flow closure testing quarterly.
Testing these valves during normal plant operation would require a Containment entry.
CV209 and CV210 will be reverse flow closure tested during cold shutdown and will be forward flow tested at the same interval.
48 of 105
FENOC FRstEneqy Nuc]oar Operath. Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-21 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Containment Purge CV5005, CV5006, CV5007, & CV5008, Containment Purge Inlet and Exhaust Containment Isolation Valves These valves must close for containment isolation and to meet the requirements of TS 3.9.4 during core alterations or movement of irradiated fuel within Containment Exercise, time closed and fail closed quarterly.
Technical Specification 3.6.1.7 requires these valves to be closed with control power removed in Modes 1-4. These penetrations are of special concern for maintaining containment integrity, due to their large size, the type of valve used, and direct path to atmosphere. They have had an industry history of poor leakage performance following valve movement, which is the reason the unique post-use testing requirements are imposed by Technical Specification Surveillance requirement 4.6.1.2.2. In this case, the preservation of containment integrity under TS 3.6.1.7 takes precedence over monitoring the valves for potential degradation per the code. The control circuitry for these valves does not allow partial stroking.
CV5005, CV5006, CV5007, and CV5008 will be exercised, timed closed, and failed closed during cold shutdown.
49 of 105
FENOC FirsfEnergy Nuclear Operat 1otr 011t*
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-22 Decay Heat DH7A & DH7B, BWST Suction Isolation Valves Safety Function:
Code Testing:
Justification:
Alternate Testing:
DH7A and DH7B receive an open command on an SFAS Level 2 actuation, provided DH9A and DH9B are fully closed and BWST level exceeds 8 feet. These valves must close upon an SFAS Level 5 (low BWST level) actuation when an operator initiates opening DH9A and DH9B.
Exercise, time open and closed quarterly.
DH7A and DH7B are open and de-powered during normal operation to prevent possible repositioning to address 10CFR 50 Appendix R fire protection concerns. Operator action is required to close the 480VAC supply breakers before any valve movement can occur. Closing these valves during normal operation isolates the normal suction supply for the Decay Heat/LPI Pumps, High Pressure Injection Pumps, Containment Spray Pumps and the alternate suction for the Makeup Pumps. This necessitates the opening of this equipment's power supplies to prevent damage due to pump start resulting in a complete loss of an Emergency Core Cooling Train, placing the plant in an unacceptable risk category.
The control circuitry for these valves does not allow partial stroking.
DH7A and DH7B will be exercised and timed open and closed during cold shutdown.
50 of 105 System:
Valve(s):
FENOC FirsrEnergy Nuclear Operating Conpany Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-23 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Decay Heat DH1 1 & DH12, RCS to Decay Heat System Isolation Valves These valves are required to open to achieve a post-LOCA flow path for boron dilution, and are required to be open for Low Temperature Overpressure (LTOP) concerns.
These valves must close for post-LOCA system alignment.
Exercise, time open and closed quarterly.
DH1 1 and DH12 isolate the low pressure Decay Heat System from the high pressure Reactor Coolant System. They are interlocked to automatically close, and prevent opening, when RCS pressure reaches or exceeds 301/266 PSIG, respectively. Opening these valves during normal operation would require defeating protective interlocks and increase the risk of over pressurizing the Decay Heat system. The control circuitry for these valves does not allow partial stroking.
DH1 1 and DH12 will be exercised and timed open and closed during cold shutdown.
51 of 105
FENOC FirstEnergy Nuclear Operating7 Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-24 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Decay Heat DH76 & DH77, Decay Heat to Reactor Coolant System Stop Check Valves DH76 and DH77 must allow 3000 gpm forward flow to meet decay heat removal requirements during a LOCA.
These valves are pressure isolation valves and must close to meet the seat leakage requirements of TS 3.4.6.2.
Verify forward flow and reverse flow closure testing quarterly.
During normal plant operation, reverse flow testing of these valves requires entry into Containment. The Decay Heat Pumps must inject water into the Reactor Coolant System in order to verify forward flow.
The Decay Heat pumps develop insufficient head to pump water into the RCS during normal plant operation.
DH76 and DH77 will be forward flow and reverse flow closure tested during cold shutdown.
52 of 105
FENOC Firs(Energy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-25 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Decay Heat DH8 1 & DH82, DH Pump Suction Line Check Valves from the BWST DH81 and DH82 must pass forward flow to the suctions of the Decay Heat Pumps and the Containment Spray Pumps.
Reverse flow closure is required to prevent water in the Containment Emergency Sump from flowing back to the BWST during a large break LOCA.
Verify forward flow and reverse flow closure quarterly.
DH81 and DH82 are the BWST suction check valves for both the Decay Heat Pumps and the Containment Spray Pumps, so a flow rate of 4300 GPM must be established to verify full forward flow. In order to obtain a flow rate of 4300 GPM through these valves, it is necessary to run the Decay Heat Pump and the Containment Spray Pump simultaneously, which requires entry into multiple Technical Specification action statements.
Reverse flow testing of these valves requires isolation of the respective BWST header. This isolates the normal suction supply for the Decay Heat/LPI Pumps, High Pressure Injection Pumps, Containment Spray Pumps and the alternate suction for the Makeup Pumps. This necessitates the opening of this equipment's power supplies to prevent damage due to pump start resulting in a complete loss of an Emergency Core Cooling Train, placing the plant in an unacceptable risk category.
The benefits of quarterly testing are outweighed by the risk incurred due to the multiple systems being inoperable and the amount of time and operator actions required to restore the equipment to an operable status.
DH81 and DH82 will be forward flow and reverse flow closure tested during cold shutdown.
53 of 105
FENOC FirstEnergy Nu*cer Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-26 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Decay Heat DH125, DH126, DH127 & DH128, Decay Heat System Train Cross-Connect Line Check Valves These valves must prevent reverse flow to provide Decay Heat System Train separation.
Verify forward flow and reverse flow closure quarterly.
Both Decay Heat System Trains are rendered inoperable during reverse flow testing. The test lineup requires one Decay Heat Pump suction line to be isolated and the other Decay Heat Pump lined up to recirculate to the BWST. Since the Decay Heat Pump is inoperable when lined up to the BWST, testing would require entry into a one hour action statement in accordance with TS 3.0.3.
DH125, DH126, DH127, and DH128 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
54 of 105
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-27 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Feedwater FW601 & FW612, Steam Generator 1 & 2 Main Feedwater Stop Valves These valves must close to isolate the steam generators on a SFRCS initiation.
Exercise and time closed quarterly.
Closing these valves during normal operation would result in a loss of main feedwater to the associated Steam Generator and would require a power reduction, and could result in a plant transient or trip. The control circuitry for these valves does not allow partial stroking.
FW601 and FW612 will be exercised and timed closed during cold shutdown.
55 of 105
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-28 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
High Pressure Injection HP32, HPI Pump 1 Flow Test Isolation Stop Check Valve This valve must open to provide a minimum recirculation flow path for HPI Pump 1.
Verify forward flow and verify reverse flow closure quarterly.
There are no test connection valves between this check valve and the next downstream isolation valve. Testing the reverse flow closure for this valve requires the partial disassembly and insertion of specialized testing equipment into the downstream piping of the check valve.
HP32 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
56 of 105
FENOC FRrStEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-29 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
High Pressure Injection HP33, HPI Pumps Recirculation Line to BWST Check Valve This valve must open to provide a minimum recirculation flow path for HPI Pump 2.
Verify forward flow and verify reverse flow closure quarterly.
Testing the reverse flow closure for this valve requires the partial disassembly and insertion of specialized testing equipment into the downstream piping of the check valve. There are no isolation valves between this location and the BWST necessitating the isolation of all inputs to the tank, resulting in inoperability of both trains of the ECCS.
HP33 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
57 of 105
FENOC Firs(Energy Nuclear Operfing Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-30 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Instrument Air IA501, Instrument Air Containment Isolation Check Valve This valve must close for containment isolation.
Verify forward flow and reverse flow closure quarterly.
IA501 is the inboard Containment Isolation Valve for Penetration P43A.
Testing this valve during normal operation would require a Containment entry.
IA501 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
58 of 105
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-31 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Station Air SA502, Station Air Header to Containment Isolation Check Valve This valve is required to close for containment isolation.
Verify forward flow and reverse flow closure quarterly.
SA502 is the inboard Containment Isolation Valve for Penetration 42A.
Testing this valve during normal operation would require a Containment entry. Additionally, the outboard Containment Isolation Valve for Penetration 42A is maintained closed during normal operation so no flow exists through this line.
SA502 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
59 of 105
FENOC FirstEnergy Nucleat Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-32 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Main Steam ICS 1 A & ICS 1 B, Main Steam Atmospheric Vent Valves These valves must close in response to an SFRCS actuation signal.
These valves are required to open under various USAR scenarios to provide a method of removing decay heat or depressurizing a steam generator.
Exercise, time open and closed, and fail closed quarterly.
During power operations, these valves must be isolated prior to opening to prevent a plant transient. Manual isolation of these valves for testing is impractical due to the 900 PSID, which would be developed across the isolation valve following testing. Partial stroking during normal operation would vent main steam to atmosphere causing an unacceptable plant transient.
ICS 1lA and ICS11B will be exercised and timed open and closed, and failed closed during cold shutdown.
60 of 105
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-33 Main Steam MS 100 & MS101, Main Steam Isolation Valves Safety Function:
Code Testing:
Justification:
Alternate Testing:
These valves must close upon an SFRCS actuation to isolate the Steam Generators to mitigate the consequences of high-energy line breaks and to provide containment isolation.
Exercise, time closed, and fail closed quarterly.
Exercising these valves during normal operation would isolate 50% of the steam flow to the main turbine and would cause a transient that would result in a turbine and reactor trip. It is undesirable to intentionally place the plant through the transient caused by abruptly interrupting steam flow through one of these valves. The control circuitry for these valves does not allow partial stroking.
MS 100 and MS 101 will be exercised, timed closed, and fail closed tested during cold shutdown.
61 of 105 System:
Valve(s):
FENOC F~rstEnergy Nu~Iaa¢ OperatinQ opn Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-34 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Main Steam MS 145 & MS 146, Auxiliary Feed Pump Turbines 1 and 2 Main Steam Minimum Flow Line Check Valves These valves must close to provide Auxiliary Feedwater Train separation.
Verify forward flow and reverse flow closure quarterly.
MS 145 and MS 146 were installed to pass sufficient steam flow through check valves MS734 and MS735 to prevent the discs from banging.
Disc banging had been a chronic failure mechanism for MS734 and MS735. The resultant disc/seat damage was determined to be the cause of numerous surveillance test failures.
Reverse flow closure testing of MS 145 and MS 146 requires isolation of the steam flow through MS734 and MS735 and running a high pressure and temperature drain hose for venting of high-pressure steam to atmosphere. These valves are located in the highly congested Auxiliary Feed Pump rooms where personnel egress is restrictive. When testing requires venting of high-pressure steam, significant safety precautions and contingency plans are required to ensure personnel safety.
If these check valves or their isolation valves leak during testing it would necessitate the need to defeat the automatic isolation features and depressurize the entire steam header to an AFPT to determine if the check valve was the source of failure. This would make a train of Auxiliary Feedwater inoperable for significant time and increase plant risk (CDF).
Because of the increased risk to personnel and plant safety, significant contingency plans are required making it burdensome to perform this testing with no compensating increase of plant reliability or safety.
MS 145 and MS 146 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
62 of 105
FENOC F fsrEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-35 System:
Valve(s):
Safety Function:
Makeup MUIA, MUIB, MU2A, MU2B, & MU3, RCS Letdown Cooler Isolation Valves MU1A, MUIB and MU2B must provide automatic isolation of their respective Letdown Cooler if a tube leak occurs (high pressure detected on the CCW side of the cooler) to isolate the affected Letdown Cooler and maintain RCS pressure boundary integrity.
MU2A and MU3 must close for containment isolation.
Code Testing:
Exercise and time closed quarterly (MUlA, MUIB, MU2A, and MU2B).
Exercise, time closed and fail closed quarterly (MU3).
Justification:
Alternate Testing:
These valves are in the normal flow path for the RCS letdown line.
Exercising these valves during normal operation would disrupt the letdown flow, which would upset the RCS feed and bleed balance causing plant instability, and could result in Letdown Cooler damage due to thermal cycling. The control circuitry for these valves does not allow partial stroking.
MU1A, MUIB, MU2A, MU2B, and MU3 will be exercised and timed closed, and MU3 will be failed closed tested, during cold shutdown.
63 of 105
FENOC Firr Energy Nuclea Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-36 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Makeup MU38, MU59A, MU59B, MU59C, MU59D, MU66A, MU66B, MU66C, & MU66D, RC Pump Seal Supply and Line Containment Isolation Valves.
These valves must close for containment isolation.
Exercise and timed closed quarterly (MU59A/B/C/D).
Exercise, time closed, and fail closed quarterly (MU38, MU66A/B/C/D).
Exercising these valves, during normal operation or at cold shutdown when the RCS is pressurized results in a loss of normal seal water flow to the RCP seals. This would disrupt the normal flow path within the seal and cause reactor coolant to flow into the seals causing seal damage.
The control circuitry for these valves does not allow partial stroking.
MU38, MU59A, MU59B, MU59C, MU59D, MU66A, MU66B, MU66C, & MU66D will be exercised and timed closed, and MU38, MU66A/B/C/D will be fail-close tested during cold shutdown.
64 of 105
FENOC FirstEnerqy Nuclear Operativ Co m p any Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-37 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Makeup MU169, Normal Makeup to RCS Line Check Valve This valve must open to provide a boration flow path.
This valve must close to prevent reverse flow ensuring HPI flow path integrity.
Verify forward flow and reverse flow closure quarterly.
This valve is in the normal makeup line to the Reactor Coolant System.
To verify reverse flow closure requires termination of the normal makeup flow. Termination of normal makeup flow requires using the alternate makeup line, resulting in manual pressurizer level control. In addition, isolation of all makeup flow would be required, causing thermal shock to high pressure injection nozzle thermal sleeve. Additionally, a Containment entry would be required to establish the test boundary.
MU169 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
65 of 105
FENOC FirstEnerQy Nuciear Operating Copa n
y Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-38 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Makeup MU242, MU243, MU244 & MU245, RCP Seal Injection Line Containment Isolation Stop Check Valves These valves must close for containment isolation.
Verify forward flow and verify reverse flow closure quarterly.
Reverse flow testing these valves during normal operation or at cold shutdown when the RCS is pressurized results in a loss of normal seal water flow to the RCP seals. This would disrupt the normal flow path within the seal and cause reactor coolant to flow into the seals causing seal damage. Additionally a Containment entry would be required to establish the test boundary.
MU242, MU243, MU244 and MU245 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
66 of 105
FENOC FArs)Ener*jy Nuclear Operatinq opn Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-39 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Makeup MU800, Alternate Makeup to RCS/Feed and Bleed Check Valve This valve must open to provide an alternate boration injection flow path.
This valve must close to prevent reverse flow ensuring HPI flow path integrity.
Verify forward flow and verify reverse flow closure quarterly.
Forward flow testing MU800 during normal operation would require shifting RCS makeup to the alternate line. This would inject cold Makeup water into the BPI RCS nozzle causing a thermal cycle.
Thermal cycles on the HPI/MU thermal sleeves are limited and shall be minimized to preclude thermal transient induced crack initiation.
MU800 will be forward flow tested during cold shutdown and reverse flow closure tested at the same interval.
67 of 105
FENOC FirstErergy Nuclear Operaling y
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-40 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Makeup MU3971 & MU6405, MU Pump Suction Three Way Valves These valves must close to isolate non-safety-related piping from safety related piping.
Exercise and time closed quarterly.
MU3971 and MU6405 are three-way suction valves. Their normal position is closed with Makeup pump suction lined up to the Makeup Tank. When these valves are open, pump suction swaps to the Borated Water Storage Tank (BWST).
When cycling, these 3-way valves are open to both the Makeup Tank and the BWST. Exercising the valves during normal plant operation would cause highly borated water, from the BWST, to enter as normal makeup flow to the RCS, resulting in undesirable RCS boration and a uncontrolled reactivity event. The control circuitry for these valves does not allow partial stroking.
MU3971 and MU6405 will be exercised and timed closed during cold shutdown.
68 of 105
FENOC FirstEnerqy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-41 Makeup MU6422, Normal Makeup to RCS Isolation Valve Safety Function:
Code Testing:
Justification:
Alternate Testing:
MU6422 must close for containment isolation.
Exercise and time close quarterly Exercising MU6422 would terminate normal makeup flow to the RCS which would result in undesirable pressurizer level transients. In addition, closing MU6422 would isolate all makeup flow causing thermal shock on the HPI nozzle and thermal sleeves when the makeup flow is restored. The control circuitry for these valves does not allow partial stroking.
MU6422 will be exercised and timed closed during cold shutdown.
69 of 105 System:
Valve(s):
FENOC FirsfEnerQy Nu*lear Operating Connp7ny Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-42 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Nitrogen NN58, Nitrogen Penetration 44B Containment Isolation Check Valve This valve must close for containment isolation.
Verify forward flow and reverse flow closure quarterly.
NN58 is the inboard Containment Isolation Valve for Penetration 44B.
Testing this valve during normal operation would require a Containment entry.
NN58 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
70 of 105
FENOC Firs Ese rgy Nuclear Operating oon Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-43 Reactor Coolant RC 10, Pressurizer Spray Line Isolation Valve Safety Function:
Code Testing:
Justification:
Alternate Testing:
RC1O must close to isolate the normal spray line, minimizing the RCS pressure drop that would occur if spray control valve RC2 stuck open.
RC10 must also close to ensure that auxiliary spray flow is directed to the Pressurizer.
Exercise and time closed quarterly.
Closing this valve during normal operation isolates pressurizer spray flow. The pressurizer operating procedure Limits and Precautions state, "Minimum pressurizer spray flow shall be maintained to ensure Pressurizer delta T and RCS boron concentration equalization requirements". The control circuitry for these valves does not allow partial stroking.
RC10 will be exercised and timed closed during cold shutdown.
71 of 105 System:
Valve(s):
FENOC FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-44 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Reactor Coolant RC 113, Quench Tank Penetration 41 Containment Isolation Check Valve This valve must close for containment isolation.
Verify forward flow and reverse flow closure quarterly.
RC 113 is the inboard Containment Isolation Valve for Penetration 41.
Testing this valve during normal operation would require a Containment entry.
RC 113 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
72 of 105
FENOC Fi S Ener qy Nuclear Operait opa n
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-45 System:
Valve(s):
Safety Function:
Reactor Coolant RC229C, RC Penetration 48 Thermal Expansion Check Valve This valve must open to relieve pressure that could occur between the Penetration 48 isolation valves due to thermal expansion following a LOCA.
This valve must close for Containment Isolation.
Code Testing:
Justification:
Alternate Testing:
Verify forward flow and reverse flow closure quarterly.
RC229C is the inboard Containment Isolation Valve for Penetration 48.
Testing this valve during normal operation would require a Containment entry.
RC229C will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
73 of 105
FENOC Firs)Enerqy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-46 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Reactor Coolant RC4608A, RC4608B, RC461OA & RC46 10B, RCS Loop High Point Vent Valves These valves must open to vent non-condensable gases and steam from the RCS high points, which could disrupt natural circulation of the RCS during an emergency event.
These valves must close to maintain system integrity after opening.
Exercise and time open and closed, and fail closed quarterly.
Opening these valves during normal operation vents reactor coolant to the floor of Containment causing considerable contamination and possible boron induced corrosion on containment equipment. Failure of these valves to close would result in a small break loss of coolant accident. The control circuitry for these valves does not allow partial stroking.
RC4608A, RC4608B, RC4610A and RC4610B will be exercised and timed open and closed, and failed closed during cold shutdown.
74 of 105
FENOC FirstEnEargy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-47 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Service Water SW17, SW18 & SW19, Service Water Pump Discharge Check Valves These valves must open to pass design forward flow when their respective pump is running to provide cooling water flow to safety related components. These valves must close to prevent short-circuiting of cooling water through an idle pump.
Verify forward flow and reverse flow closure quarterly.
Full forward flow testing could require the system flow to be increased beyond the normal temperature demand of the system, the resulting flow could cause excessive heat removal from equipment being serviced.
System operating conditions will not allow adjusting system resistance without significant impact on the plant equipment thermal equilibrium.
This is a temperature variable system that is in continuous operation during all modes of plant operation. Depending on plant operating conditions and climatic conditions, the cooling requirements range from minimum cooling loads to 100 percent with many of the loads automatically placed in operation in response to local temperature requirements. Because of these operating requirements, it is not practical to adjust flow to achieve the required full forward flow testing conditions.
SW17, SW18 and SW19 will be forward flow tested during cold shutdown and reverse flow closure tested at the same interval.
75 of 105
FENOC Firs(Energy Nuclear Operating Company Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Cold Shutdown Justification CS-48 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Service Water SW57, SW Header from TPCW Heat Exchanger Check Valve SW57 must close to isolate the essential piping from non-essential piping to prevent flooding following a seismic event.
Verify forward flow and reverse flow closure quarterly.
Reverse flow testing this valve during normal operation would require securing cooling water flow through the Turbine Plant Cooling Water (TPCW) Heat Exchangers, which would interrupt cooling to turbine loads. This would result in component overheating, potential equipment damage, and a plant trip.
SW57 will be reverse flow closure tested during cold shutdown and forward flow tested at the same interval.
76 of 105
FENOC FirstErnergy Nuclear Operating Comaeny 10.0 REFUELING JUSTIFICATIONS Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Refueling Justification RJ-1 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Component Cooling Water CC 183, CC283, CC383 & CC483, Component Cooling Water Inlet Check Valves to the Reactor Coolant Pump Seals CC183, CC283, CC383, and CC483 check valves must close to prevent reverse flow upon failure of the RCP seal cooling heat exchanger tubes to prevent high pressure reactor coolant from flowing into the low pressure Component Cooling Water System.
Verify forward flow and reverse flow closure quarterly.
The CC183, CC283, CC383, and CC483 reverse flow closure tests cannot be performed quarterly during normal operation since the system is in operation supplying cooling water to the RCP Seals and cannot be isolated.
The valves are located in a high radiation area inside Containment approximately 25 feet above the floor. The setup and preparation for performance and restoration from these tests involve a significant amount of scaffolding and personnel radiation exposure and are impractical to be performed during a cold shutdown.
CC 183, CC283, CC383, and CC483 will be reverse flow closure tested each refueling outage and forward flow tested at the same interval.
77 of 105
FENOC FirstEnergy Nuclear Operating COMp8nY System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Refueling Justification RJ-2 Core Flood CF28, CF29, CF30 & CF31, Core Flood Tank Discharge Check Valves to the RCS These valves must open to provide a flow path from the Core Flood Tanks to the Reactor Coolant System to prevent core damage during events that depressurize the Reactor Coolant System.
Verify forward flow and reverse flow closure quarterly.
The Core Flood Tanks are isolated from the RCS by these normally closed check valves. Each Core Flood Tank is charged with a nitrogen blanket of approximately 600 PSIG. This pressure is insufficient during operation to inject into the RCS.
Exercising these valves requires the reactor head to be off and the refueling canal partially filled. This is required because sufficient space is necessary to except the large volume of water dumped from the CFTs.
This plant condition only exists during refueling outages.
CF28, CF29, CF30 & CF31 will be forward flow tested during each refueling outage and reverse flow closure tested at the same interval.
78 of 105
FENOC FirstErergy Nuclear Operating Company System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Refueling Justification RJ-3 Decay Heat DH9A & DH9B, Decay Heat/LPI Containment Emergency Sump Isolation Valves DH9A and DH9B receive a close command on an SFAS Level 2 actuation for containment isolation.
These valves must open upon an SFAS Level 5 (low BWST level) actuation.
Exercise, time open and closed quarterly.
DH9A and DH9B are closed and de-powered during normal operation to prevent possible repositioning to address 10 CFR 50 Appendix R fire protection concerns. Operator action is required to close the 480VAC supply breakers before any valve movement can occur. DH9A and DH9B are interlocked to prevent opening unless DH7A and DH7B are fully closed, or an SFAS Level 5 signal is present. The control circuitry for these valves does not allow partial stroking.
Opening DH9A and DH9B during normal operation would require closing DH7A and DH7B to prevent draining the BWST to the Containment Emergency Sump. Closing DH7A and DH7B during normal operation isolates the normal suction supply for the Decay Heat/LPI Pumps, High Pressure Injection Pumps, Containment Spray Pumps and the alternate suction for the Makeup Pumps. This necessitates the opening of the power supplies to this equipment, prohibiting damage due to pump start. This results in a complete loss of an Emergency Core Cooling Train, placing the plant in an unacceptable risk category.
To preclude draining water from the ECCS Pump suctions into the Containment Emergency Sump, blank flanges are installed in the Containment Emergency Sump suction lines. Installation of the blank flanges during power operation requires Containment entry.
To install the flanges, the emergency sump debris screens, and the anti vortex flange attachments must be removed. After stroking, the water trapped between the blank flanges and the valves must be drained, the blank flanges must be removed, and the anti-vortex flange attachments, debris screens must be reinstalled and the ECCS systems filled and vented. The burden of performing these activities makes it impractical to performed this testing during a cold shutdown.
DH9A and DH9B will be exercised and timed open and closed each refueling outage.
79 of 105
FENOC FirstEnerqy Nuclear Operating Company Refueling Justification RJ-4 Decay Heat DH49, RCS Thermal Expansion Check Valve Safety Function:
Code Testing:
Justification:
Alternate Testing:
This valve must open to relieve thermally induced pressure that could accumulate between normally closed valves DH 11 and DH12, RCS to DH System Isolation Valves. Once opened, this valve must close to form a pressure boundary between the Reactor Coolant System and the Decay Heat System.
Verify forward flow and reverse flow closure quarterly.
This valve is located inside Containment within the Decay Heat Valve Pit. The Decay Heat Valve Pit is a sealed enclosure that must be watertight during normal plant operation.
Surveillance requirements for the Decay Heat Valve Pit are addressed in Technical Specification 3.5.2f. Testing DH49 would require Containment entry, opening the Decay Heat Valve Pit, test equipment setup and test performance, removal of test equipment, sealing of the Decay Heat Pit access plates, and performance of a leak test on the Decay Heat Pit once it has been sealed. The burden of opening, resealing, and testing the Decay Heat Pit integrity makes it impractical to perform this testing during cold shutdown.
DH49 will be forward flow tested and reverse flow closure tested each refueling outage.
80 of 105 Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 System:
Valve(s):
FENOC FirstEnaery Nuclear Operating Company System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Refueling Justification RJ-5 High Pressure Injection HP10, HP22, HP 1I & HP23 High Pressure Injection Pump 1 and 2 Suction and Discharge Check Valves These valves must open to allow a forward flow rate of 826 GPM into the RCS (with zero RCS pressure) providing makeup during a LOCA Verify forward flow and reverse flow closure quarterly.
During normal operation, forward flow through these valves is limited to approximately 380-430 GPM by a pump test recirculation line to the BWST equipped with an orifice. The design accident flow rate for these valves is 826 GPM. HPI Pump flow must be injected directly into the RCS to obtain this flow rate. The HPI pumps have insufficient discharge pressure to overcome normal RCS operating pressure.
Due to the potential for low temperature over-pressurization of the RCS, the HPI system is Safety tagged out when the RCS is below 150 pounds.
Therefore this testing is performed with the Reactor head off during the fill or draining of the refueling canal. Verification of full forward flow can only be performed with the RCS depressurized and vented and is impractical to be performed during cold shutdown.
HP 10, -P22, HP 1I and HP23 will be forward flow tested during each refueling outage and reverse flow closure tested at the same interval.
81 of 105
FENOC FinsfEnergy NuClear Opera77nQ o System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Refueling Justification RJ-6 High Pressure Injection HP48, HP49, HP50, HP51, HP56, HP57, HP58 & HP59, HPI Lines to RCS Loop 1 and 2 Check Valves These valves must open to provide the required HPI flow to the RCS.
Verify forward flow and reverse flow closure quarterly.
Forward flow testing must be performed while injecting water at a flow rate of 413 GPM into the Reactor Coolant System using the HPI Pump.
The HPI pumps have insufficient discharge pressure to overcome normal RCS operating pressure.
Due to the potential for low temperature over-pressurization of the RCS, the HPI system is Safety tagged out when the RCS is below 150 pounds.
Therefore this testing is performed with the reactor head off during the fill or draining of the refueling canal. Verification of full forward flow can only be performed with the RCS depressurized and vented and is impractical to be performed during cold shutdown.
HP48, HP49, HP50, HP51, HP56, HP57, HP58 and HP59 will be forward flow tested during each refueling outage and reverse flow closure tested at the same interval.
82 of 105
FENOC FirstEnergy Nuclear Operating Corpeny System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Refueling Justification RJ-7 Nitrogen NN1000, NN1002, NN1004, NN1006, NN1008, NN1010, NN1012, NN1014, NN1016, NN1018, NN1020, NN1022, NN1024, NN1026, NN1028, NN1030, NN1032, NN1034, NN1036, NN1038, NN1040, NN1042, NN1044, NN1046, NN1050, NN1052, NN1054, NN1056, NN1058, NN1060, NN1062, NN1064, NN1066, NN1068, NN1070, NN1072, NN1074, NN1076, NN1078, NN1080, NN1082, NN1084, NN1086, NN1088, NN1090, NN1092, NN1094 & NN1096, Nitrogen Supply to Containment Electrical Penetrations Check Valves These valves must allow forward flow to supply nitrogen to the electrical penetrations and prevent reverse flow to ensure nitrogen pressure will remain in each penetration assembly if the non-safety-grade nitrogen supply system is lost.
Verify forward flow and reverse flow closure quarterly.
Testing these check valves during normal operation requires personnel to enter the Shield Building Annulus. During normal operation the Shield Building Annulus has significant gamma and neutron fields, and is designated as a Locked High Radiation Area.
Testing these check valves during cold shutdown requires the building of a significant amount of scaffolding for access to the Electrical Penetration Assemblies. The amount of work required to access these penetrations is very extensive and presents a significant burden to the station and is impractical to be performed during cold shutdown.
NN1000, NN1002, NN1004, NN1006, NN1008, NN1010, NN1012, NN1014, NN1016, NN1018, NN1020, NN1022, NN1024, NN1026, NN1028, NN1030, NN1032, NN1034, NN1036, NN1038, NN1040, NN1042, NN1044, NN1046, NN1050, NN1052, NN1054, NN1056, NN1058, NN1060, NN1062, NN1064, NN1066, NN1068, NN1070, NN1072, NN1074, NN1076, NN1078, NNI080, NN1082, NN1084, NN1086, NN1088, NN1090, NN1092, NN1094 and NN1096, will be forward flow and reverse flow closure tested during each refueling outage.
83 of 105
FENOC FirstEnergy Nuclear Operating Company Refueling Justification RJ-8 Reactor Coolant RC5 1, Pressurizer Auxiliary Spray Line Check Valve Safety Function:
Code Testing:
Justification:
Alternate Testing:
This valve must open to provide a flow path through the auxiliary spray line to the RCS to provide dilution water to preclude boron precipitation following a LOCA.
Verify forward flow and reverse flow closure quarterly.
Forward flow operability is verified by injecting flow into the Pressurizer at a rate of 250 GPM. The RCS must be partially drained to provide adequate space to accommodate the inventory of water during testing and the Decay Heat Pump 2 must be lined up to the suction of the HPI Pump 2 suction. This is and abnormal line up placing the plant in an unacceptable risk category. The control circuitry for these valves does not allow partial stroking.
The RCS is not normally drained during cold shutdown.
RC51 will be forward flow tested during each refueling outage and reverse flow closure test at the same interval.
84 of 105 Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 System:
Valve(s):
FENOC FirStEnargy Nuclear operating C7778n y 11.0 SAMPLE DISASSEMBLY JUSTIFICATIONS Sample Disassembly Justification SDJ-1 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Decay Heat DH42 & DH43, Decay Heat Pump 1 & 2 Discharge Check Valves These valves must open allow a forward flow rate of 3000 gpm to meet low-pressure injection and decay heat removal requirements.
Verify forward flow and reverse flow closure quarterly.
These valves are welded directly to the downstream gate valve bodies with no intervening piping or test connections. To perform reverse flow testing during normal operation requires taking out of service a Train of Decay Heat/LPI. This results in increased system Maintenance Rule unavailability and places the plant in an elevated risk condition. There is no safety function in the reverse direction, making quarterly reverse flow closure testing impractical with no compensating increase is plant reliability or safety.
DH42 & DH43 have the same manufacturer, design, service, size, and material of construction and orientation. Additionally Kalsi Engineering, Inc. using the SOER 86-3 Check Valve Application and Prioritization (CVAP) program modeled these valves. This program takes into account valve orientation and piping geometry when calculating the Wear/Fatigue Index. The results of the model indicated a Wear/Fatigue Index of 1 (Very Low). This index equates to a Wear/Fatigue limit of 14 plant cycles or 21 years.
One valve will be disassembled, inspected and manually full-stroke exercised each refueling outage, alternating between valves. Verification the valve will pass forward flow will occur after it has been reassembled following inspection.
85 of 105 Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0
FENOC FirstEnergy Nucear Operating Company Sample Disassembly Justification SDJ-2 System:
Valve(s):
Safety Function:
Code Testing:
Justification:
Alternate Testing:
Main Steam MS726 & MS727, Main Steam to Auxiliary Feedwater Pump Turbine Line Check Valves These valves must pass forward flow to allow the Auxiliary Feed Pumps to achieve their design flow rate and differential pressure. Additionally, these valves must prevent reverse flow to ensure steam is not diverted from an AFPT when it is being supplied from the opposite side Steam Generator Verify forward flow and reverse flow closure quarterly.
These valves are welded directly to the upstream gate valve bodies with no intervening piping or test connections. The gate valves can not be isolated from the Main Steam line. Reverse flow closure testing it is impractical because there are no system design provisions to allow testing.
MS726 & MS727 have the same manufacturer, design, service, size, material of construction and orientation. Additionally Kalsi Engineering, Inc. using the SOER 86-3 Check Valve Application and Prioritization (CVAP) program modeled these valves. This resulted in a Wear/Fatigue index of 1 (Very Low). This index equates to a Wear/Fatigue limit of 14 plant cycles or 21 years.
One valve will be disassembled, inspected and manually full-stroke exercised each refueling outage, alternating between valves. Forward flow testing will occur after the valve has been reassembled following inspection.
86 of 105 Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0
FENOC FinrvEerqy Nuclear Operating Company 12.0 IST PUMP TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 PUMP INFORMATION PARAMETER r
C1, V
0 Pump Description E
q Auxiliary Feedwater P14-1 3
B C{P 006D D-7 RP-1 Q/2Y Q/2Y 2Y Q/2Y Pumps P14-2 3
B CHP 006D G-5 RP-1 Q/2Y Q/2Y 2Y Q/2Y P38-1 NC A CVL NA NA RP-1 Q/2Y Q/2Y Q/2Y NV Boric Acid Pumps P38-2 NC A CVL NA NA RP-1 Q/2Y Q/2Y Q/2Y NV P43-1 3
A CBP 036A D-4 RP-1,2 Q/2Y Q/2Y Q/2Y NV Component Cooling P43-2 3
A CET 036A J-4 RP-1,2 Q/2Y Q/2Y Q/2Y NV Water Pumps P43-3 3
A CBIP 036A G-4 RP-1,2 Q/2Y Q/2Y Q/2Y NV Containment Spray P56-1 2
B CBIP 034 D-10 RP-1 Q/2Y Q/2Y 2Y NV Pumps P56-2 2
B CBIP 034 B-10 RP-1 Q/2Y Q/2Y 2Y NV Decay Heat P42-1 2
A CHP 033B G-10 RP-1 Q/2Y Q/2Y Q/2Y NV Removal Pumps P42-2 2
A CHP 033C F-8 RP-1 Q/2Y Q/2Y Q/2Y NV EDG Fuel Oil P195-1 3
A C-P 017A C-4 RP-3 NM 2Y NM NV Transfer Pumps P195-2 3
A CHP 017A C-7 RP-3 NM 2Y NM NV High Pressure P58-I 2
B CHP 033A H-7 RP-1 Q/2Y Q/2Y 2Y NV Injection Pumps P58-2 2
B CRP 033A E-7 RP-1 Q/2Y Q/2Y 2Y NV P37-1 NC A CHP 031C D-9 RP-1 Q/2Y Q/2Y Q/2Y NV Makeup Pumps P37-2 NC A CHP 031C G-9 RP-1 Q/2Y Q/2Y Q/2Y NV P3-1 3
A CVL 041A G-2 RP-1,4 Q/2Y Q/2Y Q/2Y NV Service Water Serc WP3-2 3
A CVL 041A G-5 RP-1,4 Q/2Y Q/2Y Q/2Y NV Pumps P3-3 3
A CVL 041A G-9 RP-1,4 Q/2Y Q/2Y Q/2Y NV 87 of 105
FENOC FirstEaergy Nutclear Opera ing Com.pany, 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 Si4Z Vveie
- r.
=0 Alternate Tests Number D
Q Code Required Testing 0
N u m b e r
'W
( n 6.
03 P e r f o r m e d T
to A
I CST to AFPl2Suction AF2 L CST ValSt 50 3
C At 006D D4 8
CK SA C
N C
FF-Q RF-Q CS-1 FF-SI RF-C Line Check Valve I__
CStAFPP12inSuctlon AF2 Line Check Valve 50 3
C At 006D G2 8
CK SA C
N C
FF-Q RF-Q CS-I FF-SI RF-C AFP iMinimum Flow AF165 Le k
I alow 50 3
C At 006D B9 2
CK SA C
N 0
FF-Q RF-Q CS-2 FF-SI RF-C
_____Line Check Valve AF16 AFP 2Minimum Flow 50 3
C At 006D ES 2
CK SA C
N 0
____Line Check Valve CSI IS RF-CI AF9 AFP1DischargeLine 50 3 C
At 006D D10 6
CK SA C
N O FF-Q RF-Q AF19 Check Valve AF20 AFP 2 Discharge Line 50 3
C At 006D G10 6
CK SA C
N O FF-Q RF-Q Check Valve AFW to OTSG 1 AF39 Injection Line Check 50 2
C At 007B Bil 6
CK SA C
N O/C FF-Q RF-Q CS-3 FF-C RF-SI Valve AFW to OTSG 2 AF43 Injection Line Check 50 2
C At 007B D5 6
CK SA C
N O/C FF-Q RF-Q CS-3 FF-C RF-SI Valve MDFP to AFWLine 2 AF49 Discharge Check Valve 50 3
C At 006D B6 6
CK SA C
N C
FF-Q RF-Q CS-4 FFSI RF-C MDEP to AEW Line 1 AF52 Discharge Check Valve 50 3
C At 006D B6 6
CK SA C
N C
FF-Q RF-Q CS-4 FF-SI RF-C AFP I Cooling Water AF63 Return to AFP 1 50 3
C At 006D E7 1
CK SA C
N O/C FF-Q RF-Q Suction.
AEP 2 Cooling Water AF68 Return toAFP2 50 3
C At 006D H4 1
CK SA C
N O/C FF-Q RF-Q Suction.
C At 007B C9 6
CK SA C
N O/C FF-Q RF-Q CS-5 FF-C RF-C Supply Check Valve AF73 AFW to OTSG2 50 3
C At 007B B8 6
CK SA C
N 2
FF-Q RF-Q CS-5 FF-C RF-C Supply Check Valve AF7-4 AFW toOTSG 1 50 3
C At 007B B7 6
CK SA C
N 0
FF-Q RF-Q CS-6 FF-C RF-C Supply Check Valve I
3 AF75 AFW to OTSGV2 50 3
C At 007B C6 6
CK SA C
N O/C FF-Q RF-Q CS-6 FF-C RF-C Supply Check Valve 5
3 C
AF599 AFWtoOTSG2 Line 50 2
B At 007B D3 6
GT MO LO AI C
TC-Q PV-CS-7 TC-C Stop Valve 2Y AF608 AFWtoOTSG I Line 50 2
B At 007B B12 6
GT MO LO Al C
TC-Q PV-CS-7 TC-C Stop Valve 5
2Y AF3869 AFP ItoOTSG2Stop 50 3
B At 007B B8 6
GT MO C AI O/C TO-Q TC-Q PV Valve 2Y AF3870 AFPItoOTSG1Stop 50 3
B At 007B B9 6
GT MO 0 Al O/C TO-Q TC-Q PVT Valve 2Y AEP~oOTGIS~p
-PV AF3871 AFP2toOTSG I Stop 50 3
B At 007B B7 6
GT MO C Al O/C TO-Q TC-Q 2Y I
AF3872 AFP2toOTSG2Stop 50 3
B At 007B B6 6
GT MO 0 AI O/C TO-Q TC-Q PV Valve 2Y AEPI1Cooler Line SR AF4979 50 3
C At 006D E7 lxl RL SA C
N 0
Relief Valve 10Y AF4980 AFP2CoolerLine 50 3
C At 006D H5 IxI RL SA C
N 0
SR Relief Valve 1
0 AF6451 AFP2DischargeFlow 50 3
B At 006D Gil 4
GL SO 0
0 0
TO-Q FO-Q Control SOV.
AF6452 AFP1Discharge Flow 50 3
B At 006D DlI 4
GL so 0 0
0 TO-Q FO-Q Control SOV.
II (Continued) 88 of 105
FENOC FirslEnergy Nuclear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve a
.CAlternate Test Numbe Valve Description Z
Z Code Required Testing Performed Number r-* *
+/-
a
>j >
z 0j Auxiliary Steam to AFP AS274 turbines Line Check 50 3
C At 003C D3 6
CK SA C
N C
FF-Q RF-Q CS-8 FF-SI RF-C Valve BW26 HPI to BWSTReturn 49 2
C At 033A A13 3
CK SA E
N C
FF-Q RF-Q Line Check Valve CCW Pump I CC17 Discharge Line Check 16 3
C At 036A D5 16 CK SA E
N O/C FF-Q RF-Q CS-9 FF-C RF-SI Valve CCW Pump 3 CC18 Discharge Line Check 16 3
C At 036A G5 16 CK SA E
N O/C FF-Q RF-Q CS-9 FF-C RF-SI Valve CCW Pump 2 CC19 Discharge Line Check 16 3
C At 036A J5 16 CK SA E
N O/C FF-Q RF-Q CS-9 FF-C RF-SI Valve MU Pump 1 CCW Inlet CC127 Check Valve to Lube 16 3
C At 036B G3 1.5 SC SA 0 N O/C FF-Q RF-Q Oil Cooler.
MU Pump 2 CCW Inlet CC128 Check Valve to Lube 16 3
C At 036B G4 1.5 SC SA 0
N O/C FF-Q RF-Q Oil Cooler.
MU Pump I CCW CC129 Outlet Valve from Lube 16 3
C At 036B F2 1.5 SC SA T
N 0
FF-Q RF-Q Oil Cooler.
MU Pump 2 CCW CC130 Outlet Valve from Lube 16 3
C At 036B F4 1.5 SC SA T
N 0
FF-Q RF-Q Oil Cooler.
CCW to RCP 1-1 CC183 Thermal barrier Line 16 3
C At 040D J3 1.5 CK SA 0
N C
FF-Q RF-Q RJ-I FF-S, RF-R Check Valve CC256 CCW Cooling from 16 3
C At 036B G3 1.5 SC SA C
N 0
FF-Q RF-Q Loop I for MU Pump 1.
3 CC263 CCW Cooling from 16 3
C At 036B G4 1.5 SC SA C
N 0
FF-Q RF-Q Loop 2 for MU Pump 2..
CCW to RCS Pump 1-2 CC283 Thermal Barrier Line 16 3
C At 040D J3 1.5 CK SA 0
N C
FF-Q RF-Q Ri-i FF-SI RF-R Check Valve CCW to RCS Pump 2-1 CC383 Thermal Barrier Line 16 3
C At 040D J3 1.5 CK SA 0 N
C FF-Q RF-Q RJ-1 FF-SI RF-R Check Valve CCW to RCS Pump 2-2 CC483 Thermal Barrier Line 16 3
C At 040D J3 1.5 CK SA 0
N C
FF-Q RF-Q Ri-i FF-SI RF-R Check Valve CCW Line 2 to Non CC532 essential Header 16 3
C At 036A F9 20 CK SA E
N C
PF-Q RF-Q Isolation Check Valve CCW Line 1 to Non CC533 essential Header 16 3
C At 036A H9 20 CK SA E
N C
FF-Q RF-Q Isolation Check Valve CCW Inlet to CRDC PV CC1328 Booster Pump I Block 16 3
B At 036C H9 3
GT MO 0 Al C
TC-Q 2Y Valve CCW Inlet to CRDC PV CC1338 Booster Pump 2 Block 16 3
B At 036C H7 3
GT MO 0 Al C
TC-Q 2Y Valve CC1407A CCWReturn from 16 2
A At 036C CS 12 BF MO 0 Al C
TC-Q U-B CS-10 TC-C CTMT CIV 2Y CCW Return from PV CC1407B C CW t 16 2
A At 036C C9 12 BF MO 0 AlI C TC-Q U
(Continued) 89 of 105
FENOC Ficr*Energy Nuclar Operating CoMpany 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve
~Alternate Tests Nuber Valve Description Z
Z m
Code Required Testing Perfored Number Performed CCW Penetration #4 CC1407C Check Valve 16 2
N O/C FF-Q RF-Q IU-B CS-l1 FF-C RF-C CC1411A CCWtoCTMTCIV 16 2
A At 036C G3 12 BF MO 0 Al C
TC-Q PV U-13-B CS-12 TC-C 2Y CC1411B CCWtoCTMTCIV 16 2
A At 036C H3 12 BF MO 0 Al C
TC-Q PV U-B CS-12 TC-C 2Y CC141 IC CCW Penetration #3 16 2
N O/C FF-Q RF-Q LJ-B CS-13 FF-C RF-C lCheck ValveI I
II II III CC1460 CCWtoMUPump 16 3
B At 036A FlI 1.5 GL AO 0 C
C TC-Q PVY FC-Q Header Inlet Valve 2Y DH Cooler I CCW PV CC1467 Outlet Isolation Valve 16 3
B At 036B B8 18 BF AO C
0 0
TO-Q 2Y FO-Q DH Cooler 2 CCW PV CC1469 Outlet Isolation Valve 16 3
0 0
TO-Q 2Y FO-Q CCW to Aux Bldg CC1495 Non-essentials Isolation 16 3
B At 036A HI1 16 BF AO 0 C
C TC-Q PVY FC-Q Valve C16ACCW Inlet to CRD 2
CC567A Cooling CIV 16 2
A At 036C E9 3
GT MO 0 Al C
TC-Q PV U
L-B CS-14 TC-C CCW Inlet to CRD IJ-BC-4 T
CC1567B CoolInlet 16 2
A At 036C EIO 3
GT MO 0 Al C
TC-Q PV U-B CS-14 TC-C Cooling CIV 2Y CC1568 CCW Penetration #12 16 2
N O/C FF-Q RF-Q U-B CS-15 FF-C RF-C Check ValveI I
III CC1643 CCW Surge Tank 16 3
C At 036A A8 3x4 RL SA C
N 0
SR Relief Valve 10Y CCW Return Line from PV CC2645 Aux Bldg non-essential. 16 3
B At 036B C3 16 GT MO 0 Al C
TC-Q 2Y Isolation Valve CCW Return Line from PV CC2649 Aux Bldg Non-essential 16 3
B At 036B C3 16 GT MO C Al C
TC-Q 2Y Isolation Valve CC3602 CCW Surge Tank 16 3
C At 036A A9 3x4 RL SA C
N 0
SR Relief Valve lOY CC4100 RCPIW-Seat Cooler 16 3
B At 040D K3 1.5 GL MO 0 Al C
TC-Q 2Y CS-16 TC-C CCW Return Valve 2
CC4200 RCP 1-2 Seal Cooler 16 3
B At 040D K3 1.5 GL MO 0 Al C
TC-Q PV-CS-16 TC-C CCW Return Valve I
2Y CC4300 RCP 2-1Seal Cooler 16 3
B At 040D K3 1.5 GL MO 0 A]
C TC-Q PV-CS-16 TC-C CCW Return Valve I
I
_I 2Y CC4400 RCP 2-2 Seal Cooler 16 3
B At 040D K3 1.5 GL MO 0 Al C
TC-Q PV-CS-16 TC-C CCW Return Valve 2Y CCW Line 1 Discharge PV CC5095 Header Cross-tie Line 16 3
B At 036A G9 20 GT MO 0 Al C
TC-Q 2Y Block Valve CCW Line 2 Discharge PV CC5096 Header Cross-tie Line 16 3
B At 036A H9 20 GT MO C AI C
TC-Q 2Y Block Valve CC5097 CCW LinelReturn 16 3
B At 036B C5 12 GT MO 0 Al C
TC-Q PV Block Valve 121
__2Y CCW ine2 ReurnPV CC5098 CCW Line2Return 16 3
B At 036B C5 12 GT MO C Al C
TC-Q 2Y Block Valve 2
Letdown Cooler CCW SR CC11203 Let Reler Vav 16 3
C At 036C E3 6X10 RL SA C
N 0
SRY Inlet Relief Valve
_oy cFrT2to RCS Isolation PV CFIA Valv 51 2
B P
034 J3 14 GT MOO A
0 2Y _
_____Valve 2Y (Continued) 90 of 105
FENOC Firs(Enargy Nuclear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve Z
- '--~
Alternate Tests N V sa Code Required Testing er Valve Description Performed Nube a
CFlB CF I toRCS Isolation 51 2
B P
034 J8 14 GT MO 0 Al 0
PV0 Valve 2Y LV-B CF2A CFT2BIeedLineCIV 51 2
A P
034 H5 I
GL MO LC Al C
2Y
_J-B PV CF2B CFTIBleedLineCIV 51 2
A P
034 H6 1
GL MO LC Al C
2Y LJ-B CF2C CFPenetration#47A 51 2 AC At 034 J6 3/8 CK SA C
N O/C FF-Q RF-Q LI-B CS-17 FF-C RF-C Check ValveIIII I
PV CF5A CFr2VentLineCIV 51 2
A P
034 G4 1
GL MO LC Al C
2Y LJ-B PV-II CF5B CFT 1VentLineCIV 51 2
A P
034 G6 i
GL MO LC Al C
2Y LI-B CF7A CFT12SafetyRelief 51 2
C At 034 G3 lx2 RL SA C
N 0
SR Valve 10Y CF7B CFT 1 Safety Relief 51 2
C At 034 G8 1x2 RL SA C
N 0
SR Valve 10Y CFT 2 Fill and CF15 Pressurizing Stop 51 2
AC At 034 G3 1
SC SA C
N C
FF-Q RF-Q LJ-B CS-18 FF-C RF-SI Check CIV CFr I Fill and CFI6 Pressurizing Stop 51 2
AC At 034 G9 1
SC SA C
N C
FF-Q RF-Q LJ-B CS-18 FF-C RF-SI Check CIV CF28 CFT2toRXCheck 51 1
C At 033B B4 14 CK SA C
N 0
FF-Q RF-Q RJ-2 FF-R RF-SI Valve CF29 CFT I toRXCheck 51 1
C At 033B B2 14 CK SA C
N 0
FF-Q RF-Q RJ-2 FF-R RF-SI Valve CF30 CFT 2/LPI Injection to 51 1 AC At 033B B3 14 CK SA C
N O/C FF-Q RF-Q LK-RJ-2 FF-R RF-SI Reactor Check Valve 2Y II_2 CF31 CFT1 /LPI Injection to 51 1 AC At 033B B2 14 CK SA C
N O/C FF-Q RF-Q LK-RJ-2 FF-R RF-SI Reactor Check Valve I
2Y II_2 CF1541 51 2
A At 034 F10 I
GL AO C
C C
FC-Q TC-QPV LJ-B
_F_541 Pressurization CIV 2Y PV-I.
CF1542 CFT Vent to Waste Gas 51 2
A At 034 ElO I
GL AO C
C C
FC-Q TC-Q 2Y LI-B CIV CF1544 CFTI Filland 51 2
A At 034 GIO I
GL AO C
C C
FC-Q TC-Q PVY L-B Pressurization CIV 2Y CF1545 CFrBleedLineCIV 51 2
A At 034 FlO 1
GL AO C
C C
FC-Q TC-Q Y
L-B C59 S
Pmp Dishare
~LK CS9 CSLinePUmP2 Discharge 61 2 AC At 034 B9 8
CK SA C
N O/C FF-Q RF-Q 2Y CS10 CS Pump I Discharge 61 2 AC At 034 D9 8
CK SA C
N O/C FF-Q RF-Q LK Line Check Valve I
III2Y CS17 CS Pump 2 Test Line 61 2
A P
034 A6 8
GL MA C N
C LJ-B CIV CS18 CS Pump l Test Line 61 2
A P
034 C6 8
GL MA C N
C IJ-B CIV CS33 CS Pump 2TestLine 61 2
A P
034 A6 8
OT MALC N C
LJ-B CIV CS36 CS Pump l Test Line 61 2
A P
034 C6 8
GT MA LC N C
LI-B CIV CS1530 CS Pump I Discharge 61 2
A At 034 D7 8
GL MO C Al O/C TO-Q TC-Q PV-IY-B Line CIV 2Y C1531CS Pup 2 Discharge 61 2
A At 034 B7 8
GL MO C AI 0/C TO-Q TC-Q PV-LJ-B Line CIV 2Y CV124 CTMT Gas Analyzer 60 2
AC At 029B G9 I
CK SA 0
N 0/C F-Q RF-Q LJ-B CS-19 FF-SI RF-C Return CIV (Continued) 91 of 105
FENOC FirstEnaery Nutclea* OperahIg
- 170aQ,
13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve 8
A Valve Description Z
Code Required Testing l
e Number Performed CV>
>2 CTISa naye CV125 CRMT Gas Analyzer 60 2 AC At 029B H9 1
CK SA 0
N O/C FF-Q RF-Q LJ-B CS-19 FF-SI RF-C Return CIV H2 Dilution Blower 1 CV186 Diseharge Line Check 60 3
C At 029D E12 2
CK SA C
N 0
FF-Q RF-Q Valve H2 Dilution Blower 2 CV187 Discharge Line Check 60 3
C At 029D GIl 2
CK SA C
N 0
FF-Q RF-Q Valve C29H2 Dilution BlowerI CV209 Discharge Line CIV 60 2 AC At 029D G9 4
CK SA C
N O/C FF-Q RF-Q U-B CS-20 FF-SI RF-C H2DishreLuion Boer 2I CV210 H2 Dilution Blower 2 60 2 AC At 029D G9 4
CK SA C
N O/C FF-Q RF-Q U-B CS-20 FF-SI RF-C Discharge Line CIV I_
CTMT leak Test Line CV343 CIV 60 2
A P 029B G4 8
GT MA LC N C
LJ-B CIV CTMT/nnulu D/PPV CV624B CTMT/Annulus D/P 60 2
B P 029B EIO 0.75 GT MO 0 Al 0
Y Sensing Line CIV 2Y CV4BCTMT/Annulus D/P 2V CV645B Line CIV 60 2
B P 029B D4 0.75 GT MO O AI O Sensing Line CIV 2Y CTMT Pressure PV CV2000B Sensing Line CIV for 60 2
B P 029B D10 1
GT MO 0 Al 0
2Y SFAS & RPS CTMT Pressure PV CV2001B Sensing Line CIV for 60 2
B P 029B D10 1
GT MO 0 Al 0
2Y SFAS & RPS CTMT Pressure PV CV2002B Sensing Line CIV for 60 2
B P 029B F4 1
GT MO 0 Al 0
2Y SFAS & RPS CTMT Pressure PV CV2003B Sensing Line CIV for 60 2
B P 029B F4 1
GT MO 0 Al 0
2Y SFAS & RPS H2 Dilution Blower 2 SR CV3876 P s Reli ef Vlve 60 3
C At 029D G0l lxl.5 RL SA C
N 0
Pressure Relief Valve 10Y H2 Dilution Blower 1 SR CV3877 Pressure Relief Valve 60 3
C At 029D Ell lxl.5 RL SA C
N 0O CV05CIV ugeSpl 60 2
A At 029E Ell 48 BF AO C
C C
FC-Q TC-Q PV-2 LJ-B CS-21 FC-C TC-C CV5006 CTMT Purge Supply 60 2
A At 029E El0 48 BF AO C
C C
FC-Q TC-Q PV U J-B CS-21 FC-C TC-C CIV 2Y CV5007 CTMT Purge Exhaust 60 2
A At 029E G5 48 BF AO C
C C
FC-Q TC-Q PV-LJ-B CS-21 FC-C TC-C CIV 2Y CV08CTMT Purge Exhaust LJP-B S2 CCT CV5008 60 2
A At 029E G4 48 BF AO C
C C
FC-Q TC-Q PV U-B CS-21 C-C TC-C CV5010A CTMT H2 Analyzer
!60 2
A At 029B H5 I
BL, MOO0 Al O/C TO-Q TC-Q PV-I.-B ISample Line CIV 2Y CTMT Purg ExhaustzPV CV5010B Saml LinelCIe 60 2
A At 029B J4 4
BL MAO CA O/C TO-Q TC-Q PV2y LJ-B CIVSml ieCV2 CTMT H2 AnalyzerPV CV5010C SaMp H2iAnalCIV 60 2
A At 029B
-9 I
BL MO 0 Al O/C TO-Q TC-Q PV2y LJ-B Sample Line CIV 2Y I
CTMT H2 Analyzer PV CV5010D 60 2
A At 029B GI1 1
BL MO 0 Al O/C TO-Q TC-Q IJ-B ISample Line CIV I__ 2YI CV5010E CTMT H2 Analyzer 60 2
A At 029B GIl 1.5 DA MOO0 Al O/C TO-Q TC-Q 2Y L
CIV 2
CTMT H2 Analyzer CV501C1A 60 2
A At 029B H3 I
BL MO Al 0/C TO-Q TC-Q PV2y U-B Sample Line CIV I
II 2 PV-I.
CV5011B CTMT H2 Analyzer 60 2
A At 029B J5 1
BL MO 0 Al 0/C TO-Q TC-Q U-B Sample Line CIV I
I_
2Y (Continued) 92 of 105
FENOC FirsrEnerqy Nuclear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve
- -L p on Nubr Valve Description 910 Code Required Testing Aertest Number Performed C51ICCTMT H2 Analyzer 602AA02BIl IIIM 0Al/CTQ V
U1 IV01 Sample Line CIV 6I I
A t02BFl1 LMO AI O/
OQT-V2Y PV CV501 I CTMT H2 Analyzer 60 2
A At 029B GF 1 BL MO 0 Al O/C TO-Q TC-Q U-13-B ISample Line CIV II2Y CV5011I CTMT H2 Analyzer 60 2
A At 029B G9 4
BF MO 0 Al O/C TO-Q TC-Q PV U
I-B Samp L2Y CV503C H2 utnayzer 60 2
A At 029D GH1 4
BF MO C Al O/C TO-Q TC-Q PV U-B C
I 2Y CV5038 H2 Purge CIV 60 2
A At 029D G4 4
BF MO C Al 0/C TO-Q TC-Q PV U
U-B CV5065 H2 Pilutiorge akCiV 60 2
A At 029D 04 4
BF MO C Al
/C TO-Q TC-Q PV
-B 2Y H Diltion System2 PV CV5073 60 2
A At 029B C5 8
BF MO C Al C
TC-Q 1U-B Breaker CIV 2Y CV5071 CTMT Vacuum 60 2
A At 029B C5 8
BF MO 0 Al C
TC-Q PV U
LJ-B Breaker CIV 2Y CV5072 CTMT Vacuum 60 2
A At 029B C5 8
BF MO 0 Al C
TC-Q PV-LJ-B Breaker CIV 2
CV5073 CTMT Vacuum 60 2
A At 029B C5 8
BF MO 0 Al C
TC-Q PV2 U-B Breaker CIV 2Y CV07CTMT Vacuum IV-B CV507BeakCV 60 2
A At 029B C5 8
BF MO 0 Al C
TC-Q PV U-B Breaker CIV 2Y CV5075 CTMT Vacuum 60 2
A At 029B C5 8
BF MO 0 Al C
TC-Q Y
U Breaker CIV 2Y CV5076 CTMT Vacuum 60 2
A At 029B CS 8
BF MO 0 Al C
TC-Q V
U-B Breaker CIV 2Y CV5079 CTMT Vacuum 60 2
A At 029B CS 8
BF MO 0 Al C
TC-Q PV U-B Breaker CIV 2Y CV5078 CTMT Vacuum 60 2
A At 029B C5 8
BF MO 0 Al C
TC-Q WT-U-B3 Breaker CIV 2Y CTTVacuueV CV5079m 60 2
A At 029B CS 8
BF MO 0 Al C
TCQ U-B Breaker CIV I2Y CV5080 CTMT Vacuum Relief 60 2 AC At 029B C5 8
VR SA C
N O/C SO-R SC-R WT-R BY Valve CIV 2Y CV5081 Valf 60 2
VR SA C
N 0/C SO-R SC-R W-R 2LK-RV-1 LJ-B CTVacuum Reie I2 CV5082 CTMT Vacuum Relief 60 2 AC At 029B C5 8
VR SA C
N O/C SO-R SC-R WT-R 2Y RV-1 U-B Valve R
IK CV5083 CTMT Vacuum Relief 60 2 AC At 029B C5 8
VR SA C
N O/C SO-R SC-R WT-R 2Y RV-1 U-B Valve Rf CV5084 CTMT Vacuum Relief 60 2 AC At 029B C5 8
VR SA C
N O/C SO-R SC-R WT-R LK-RV-1 U-B CV5085 CTMT Vacuum Relief 60 2
AC At 029B C5 8
VR SA C
N 0/C SO-R 5C-R WT-R LK-RV-1 U-B Valve 2Y CV5085 CTMT Vacuum Relief 60 2 AC At 029B C5 8
VR SA C
N O/C SO-R SC-R WT-R 2Y RV-1 U-B Valve CV5086 CTMT Vacuum Relief 60 2 AC At 029B C5 8
VR SA C
N O/C SO-R SC-R WT-R LK-RV-1 U
Valve 2Y CV5087 CTMT Vacuum Relief 60 2 AC At 029B C5 8
VR SA C
N O/C SO-R SC-R WT-R 2Y RV-1 U-B Valve CTMT Vacuum Relief LK CV5088 9 Valv 60 2 AC At 029B C5 8
VR SA C
N O/C SO-R SC-R WT-R 2YRV-U-B CV5089 CTMT Vacuum Relief 60 2 AC At 029B C5 8
VR SA C
N 0/C SO-R SC-R WTR LK2Y Valve 2Y H2 Diltion ystem I- -B CV5090 H2 Dilution System 1 60 2
A At 029D FIO 4
BF MO C Al O/C TO-Q TC-Q PV2 U-B CIV 2Y (Continued) 93 of 105
FENOC FirstEnerOy Nuclear Operati
- 7.
C 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve Alternate Tests r Valve Description C
Code Required Testing Number~~
CI 1.Prfre EDG Air Start DA24 compressor to Air DA4 cmpressor to ine 24 3
C At 017B C3 0.75 CK SA E
N C
FF-Q RF-Q Receiver Tk 1-1 Line Check Valve EDG Air Start DA25 Compressor to Air 24 3
C At 017B G3 0.75 CK SA E
N C
FF-Q RF-Q Receiver Tk 2-1 Line Check Valve EDG Air Start DA38 Compressor to Air 24 3
C At 017B D3 0.75 CK SA E
N C
FF-Q RF-Q Receiver Tk1-2 Line Check Valve EDG Air Start DA39 Compressor to Air 24 3
C At 017B J3 0.75 CK SA E
N C
FF-Q RF-Q ReceiverTk 2-2 Line Check Valve DA1135 Air Start Receiver Tank 3
C At 017B B4 RL SA C
N 0
S 1-1 Relief Valve 1.5 SR DAII38 Air Start Receiver Tank 24 3
C At 017B D4 1.0X SR 1-2 Relief Valve.
I SA 1.5 toy O
DAI141 Air Start Receiver Tank 10x RL SA C
N 0
SR 2-1 Relief Valve 2
C 17BF4 1.5 SR DAl 144 Air Start Receiver Tank 24 3
N 0
SR 2-2 Relief Valve._R 1.5 SAtCN o 0Y DH7A BWST to ECCS Train 2 49 2
B At 033A DlI 14 GT MO LO AI O/C TO-Q TC-Q 2Y CS-22 TO-C TC-C Isolation Valve II2Y BWST to ECCS Train IP DH7B BIsolation Valve 49 2
B At 033A D10 14 GT MO LO Al O/C TO-Q TC-Q 2Y CS-22 TO-C TC-C DH Pump 2 Suction PV DH9A from CTMT 49 2
B At 033C H3 14 GT MO LC AI O/C TO-Q TC-Q 2Y RJ-3 TO-R TC-R Emergency Sump.
DH Pump 1 Suction PV DH9B from CTMT 49 2
B At 033B K6 14 GT MO LC Al O/C TO-Q TC-Q 2Y RJ-3 TO-R TC-R Emergency Sump.
I II_
DHll RCS to DH System 49 1
B At 033B H3 12 GT MO C AI O/C TO-Q TC-Q PV-CS-23 TO-C TC-C Isolation Valve 2Y DH12 RCS to DH System 49 1
B At 033B H2 12 GT MO C Al O/C TO-Q TC-Q PV-CS-23 TO-C TC-C Isolation Valve 2Y DH13A DH Cooler 2 Bypass 49 2
B At 033C Flo 6
BF AO C C
C TC-Q FC-Q PV Flow Control Valve II_2Y PV DH13B DH Cooler 1 Bypass 49 2
B At 033B G12 6
BF AO C C
C TC-Q FC-Q 2Y Flow Control Valve DH Cooler 2 Outlet PV DH14A Flow Control Valve 49 2
TO-Q FO-Q 2Y DHo Contole 1VOalvet I2 DH14B DH Cooler 1 Outlet 49 2
TO-Q FO-Q PV Flow Control Valve II2Y DH42 DH Pump 2 Discharge 49 2
C At 033C F9 10 CK SA C
N 0
FF-Q RF-Q SDJ-l PF-RE SD-RE Check Valve DH43 DH Pump 1 Discharge 49 2
C At 033B G10 10 CK SA C
N 0
FF-Q RF-Q SDJ-l PF-RO RO 4
Check Valve RO RCS Thermal DH49 Expansion Check 49 1
C At 033B J3 1.5 CK SA C
N O/C FF-Q RF-Q RJ-4 FF-R RF-R Valve DH Pump 2 Discharge PV DH63 to HPI Pump 2 Suction 49 2
B At 033B Bll 4
GT MO C Al 0/C TO-Q TC-Q 2Y Isolation Valve (Continued) 94 of 105
FENOC Firstffnergy Nuclear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued) a o
~a CISa Valve A
Tests Number Valve Description Code Required TestingAlternateTests Number Q W I
Ce ud t
V.
Performed DH Pump 1 Discharge T
DH64 to HPIPump I Suction 49 2
B At 033B C9 4
GT MO C Al 0/C TO-Q TCQ PV Isolation Valve 2Y DH Pump 2 Discharge LK DH76 to RCS Stop Check 49 1 AC At 033B B4 10 SC SA C
N 0/C FF-Q !RF-Q 2Y CS-24 FF-C RE-C Valve DH Pump 1 Discharge LK DH77 to RCS Stop Check 49 1 AC At 033B D3 10 SC SA C
N 0/C FF-Q RF-Q 2YCS-24 FE-C RF-C Valve 2
DH Pump 1 Suction DH81 from BWST Check 49 2
C At 033A H10 14 CK SA C
N 0/C FF-Q RF-Q CS-25 FF-C RF-C Valve DH Pump 2 Suction DH82 from BWST Check 49 2
C At 033A FlI1 14 CK SA C
N 0/C FF-Q RF-Q CS-25 FE-C RE-C Valve DH87 Penetration P49 CIV 49 2
A P 033B E6 8
GT MA LC N C
U.-B DH88 Penetration P49 CIV 49 2
A P 033B E5 8
GT MA LC N C
UJ-B DH Cooler Cross DH125 connect Bypass Stop 49 2
C At 033C DI0 8
SC SA C
N C
FF-Q RF-Q CS-26 FE-SI RF-C Check Valve DH Cooler Cross DH126 connect Bypass Check 49 2
C At 033C DI0 8
CK SA C
N C
_____Valve DH Cooler Cross DH127 connect Bypass Stop 49 2
C At 033B FI2 8
SC SA C
N C
FF-Q RF-Q CS-26 FF-SI RE-C Check Valve DH Cooler Cross DH128 connect Bypass Check 49 2
C At 033B FlI1 8
CK SA C
N C
FE-Q RF-Q CS-26 FE-SI RF-C Valve DH Coler Coss-PV DH830 DHcoolrctCross-49 2
B At 033C DI0 8
GT MO LC AI C
TC-Q 2
DH Coler Coss-PV DH831 DHcoolrctCross-49 2
B At 033B Fll1 8
GT MO LC Al C
TC-Q 2
DHPumpl1Emergency 49 2 C
At 033B 7
.75x1 RL SA C N O SR DH58Sump Line Relief Valve 10Y DHpump2Emergency 92CAt03 H4.75xl RL SA C
N 0
SR DH 5 9 Sump Line Relief Valve 10Y_
DH57DH Pump I Suction 49 2
B At 033B G6 12 GT MO C Al 0/C TO-Q TC-Q 2V DH57from RCS.
2 DH58DHPump2Suction 492 B At 033C F3 12 GT MO C
AI 0/C TO-Q TC-Q 2V from RCS.
DH1529 DH Pump I Injection 49 2
C At 033B D8 1.5x2 RL SA C
N 0
SR Line Relief Valve 10 IY DHpump2lnjection 492C At 033B AI0 1.5x2 RL SA C
N 0
R DH55 49 2
=
I' Y DH50Line Relief Valve 10Y DH Pump I Suction PV DH2733 Valve from BWST or 49 2
B At 033B J8 18 GT MO LO Al O/C TC-Q TO-Q 2Y Emergency Sump.
DH Pump 2 Suction PV DH2734 Valve from BWST or 49 2
B At 033C H6 18 GT MO LO AN O/C TC-Q TO-Q 2Y Emergency Sump.
DH2735 DH Auxiliary Spray 49 1
A At 033B A6 1.5 GT MO LC AN O/C TO-Q TC-Q UJ-B PV aLine Stop CIV 2Y DH AuxpIlir Sprayon DH81 from S
r 49 2
A At 033B B6 1.5 4
L MO LC AN O/C TO-Q TC-Q PV-LJB DH76throttle CIV 2
(Continued) 95 of 105
FENOC FrstEneray Nuclear Operati,19nlPn y
13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve Valve Description Z
Z z
Code Required Testing "0
Alternate Tests Number.
a etn Performed Nube J
(-4
-=
1.0 Z
m=
DH2761 DH to Spent Fuel Pool 49 3
C At 033A G8 1.5x2 RL SA C
N 0
SR Relief Valve I1I0I_
__Y DH4633 DHTrainlSample 49 2
B P 033B Hll I
C C
PV DH4633 Isolation Valve II2Y DH4636 DHTrain2Sample 49 2
B P 033C FIO I
C C
PV2 Isolation Valve I_2Y DH4849 DHCool downLine 49 2
C At 033B H5 4x6 RL SA C
N 0
SR DH89Relief Valve I10lY Pntaon13Rle75xSR-UJ-B DR2012 Penetration #13 Relief 20 2
C At 046 C7 75 x RL SA C
N O/C 10Y Valve N Sum I
PV DR2012A CTMTNormalSump 20 2
A At 046 C6 4
GT MO O Al C
TC-Q U-B Inside CIV 2Y CTM Nrml um
-PV-L DR2012B CTMTNormalSump 20 2
A At 046 C7 4
GT MOO Al C
TC-Q U-B Outside CIV 2Y DW Makeup to CCW PV DW2643 Surge Tank Line 37 3
B At 036A C9 1
GL AO C
C C
FC-Q TO-Q TC-Q 2Y Isolation Valve DW6831ADWSupplyLineCIV 37 2
A At 010C P1" 4
GL AO 0
C C
FC-Q TC-Q PV-LJ-B 2Y DW6831B DW Supply Line CIV 37 2
A At 010C F7 4
GL AO 0
C C
FC-Q TC-Q PV-U-B I
I--
2Y FW601 OTSG2MainFWStop 45 2
B At 007B E3 18 GT MO 0 AI C
TC-Q PV-CS-27 TC-C Valve 2Y FW612 OTSG1MainFWStop 45 2
B At 007B E10 18 GT MO 0 Al C
TC-Q PV-CS-27 TC-C Valve I
I 2Y HP2A HPI to RCS Injection 52 2
B At 033A E3 2.5 GL MO C Al O/C TO-Q TC-Q PV Line 2-1 CIV 52 2Y HP2B HPI to RCS Injection 52 2
B At 033A F3 2.5 GL MO C Al O/C TO-Q TC-Q PV Line 2-2 CIV 2Y HP2C HPI to RCS Injection 52 2
B At 033A H3 2.5 GL MO C
Al O/C TO-Q TC-Q PV Line 1-1 CIV 2Y HP2 HPtoRCSInjection 2
2 B
At 033A J3 2.5 GL MO C Al O/C TO-Q TC-Q PV HPD Line 1-2 CIV
___2Y HPIO HPI Pump 1 Suction 52 2
C At 033A H9 6
CK SA C
N O/C FF-Q RF-Q RJ-5 FF-R RF-SI Line Check Valve H
C At 033A Ell 6
CK SA C
N O/C FF-Q RF-Q RJ-5 FF-R RF-SI Line Check Valve HP22 HPI Pump I Discharge 52 2
C At 033A H6 4
CK SA C
N 0
FF-Q RF-Q RJ-5 FF-R RF-SI Line Check Valve HP23 HPI Pump 2 Discharge 52 2
C At 033A E6 4
CK SA C
N 0
FF-Q RF-Q RJ-5 FF-R RF-SI Line Check Valve HP31 HPIPump2Reculation 52 2 BC At 033A D6 1.5 SC MO C N O/C FF-Q RF-Q PV2 TC-Q Stop Check Valve 2Y HP32 HPI Pump 1 Reculation 52 2
BC At 033A K6 1.5 SC MO C N O/C PF-Q RF-Q PV-TC-Q CS-28 FF-SI RF-C Stop Check Valve I
2Y HPI Pumps Reculation HP33 Line to BWST Check 52 2
C At 033A B8 3
CK SA C
N 0
FF-Q RF-Q CS-29 FF-SI RF-C Valve HPI Alternate MU to HP48 RCS Injection Line 1-1 52 1
C At 033A H2 2.5 SC SA C
N 0
FF-Q RF-Q RJ-6 FF-R RFSI Stop Check Valve I
HPI to RCS Injection HP49 Line 1-2 Stop Check 52 1
C At 033A J2 2.5 SC SA C
N 0
FF-Q RF-Q RJ-6 FF-R RF-SI Valve I
III
_ I (Continued) 96 of 105
FENOC Frs&Enerqy Nuclear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve Vav Description Alternate Tests Number Valve Description 4.)
0 0
Code Required Testing Pernated Performed 7i na 0
HPI Alternate MU to HP50 RCS Injection Line 1-1 52 1
C At 033A H2 2.5 CK SA C
N 0
FF-Q RF-Q RJ-6 FF-R RF-S1 Check Valve HPI to RCS Injection 52 1
C At 033A J2 2.5 CK SA C
N 0
FF-Q RF-Q RJ-6 FF-R RF-SI HP51 Line 1-2 Check Valve HPI/MU to RCS HP56 Injection Line 2-2 Stop 52 1
C At 033A F2 2.5 SC SA 0
N 0
FF-Q RF-Q RJ-6 FF-R RF-SI Check Valve HPI to RCS Injection HP57 Line 2-1 Stop Check 52 1
C At 033A E2 2.5 SC SA C
N 0
FF-Q RF-Q RJ-6 FF-R RF-SI Valve HPI/MU to RCS HP58 Injection Line 2-2 52 1
C At 033A F2 2.5 CK SA 0 N
0 FF-Q RF-Q RJ-6 FF-R RF-SI Check Valve HPI pump to RCS HP59 Injection Line 2-52 1
C At 033A E2 2.5 CK SA C
N 0
FF-Q RF-Q RJ-6 FF-R RF-SI 1Check Valve HP1510 HPI Pump 1 Suction 52 2
C At 033A H8 1.5 x RL SA C
N SR Line Relief Valve 1.5 oy HP1511 HPIPump2Suction 52 2
C At 033A E8 1
RL SA C
N O Line Relief Valve 1.5 10Y IA5o IA to CTMT Check 18 2 AC At 015A E7 1
CK SA 0
N C
FF-Q RF-Q LJ-B CS-30 FF-SI RF-C
_ Valve IA2011 IAtoCTMTCIV 18 2
A At 015A E6 I
GL AO 0
C C
FC-Q TC-Q PV-LJ-B 2Y SA502 SAtoCTMTCheck 18 2
N C
FF-Q RF-Q U-B CS-31 FF-SI RF-C Valve CIV SA532 UpperCS HeaderAir 18 2
A P
034 B4 2
GL MA C N
C U-B Test CIV Lower CS Header Air SA533 Toest CS edeVi 18 2 1A P
034 C5 2
GL MA C
N C
LJ-B SA535 Lower CS Header Air 18 NC A P
034 C5 2
GT MA LC N C
LJ-B Test CIV SA536 Upper CS Header Air 18 NC A P
034 A5 2
GT MA LC N C
LJ-B Test CIV SA2010 SAtoCTMTCIV 18 2
A At 015D F6 1.5 GL AO C
C C
FC-Q TC-Q PV2 U-B MS-Lin eY 2Oatmospheri ICSI1A MS Line 2 atmospheric 83 2
B At 007A D7 8x36 AN AO C
C O/C TO-Q TC-Q PV-Y C-Q CS-32 TO-C TC-C PC-C Vent Valve 2Y ICS11B MS Line I atmospheric 83 2
B At 007A D8 8x36 AN AO C
C O/C TO-Q TC-Q PV Vent Valve 2
C C
T T
PC-C M~lO V-CS-33 TC-C P MSie2SVPV V)alsValeII I2 MS100-1L 83 2
B At 003A F4 2
GL AO C
C C
FC-Q TC-Q PVY Bypass Valve 2YPV MS101 MS Line I Isolation 83 2
B At 003A C4 36 BS AO 0 C
C FC-Q TC-Q 2Y CS-33 TC-C PC-C Valve 2Y MSI01-1 MS L
ve 83 2
B At 003A C4 2
GL AO C
C C
FC-Q TC-Q PV2y Bypass Valve I
II2 MSLne 1toAMSIVV MS106 MSI.ineltoAFPT1 83 2
B At 003C B3 6
GT MO C Al O/C TO-Q TC-Q PV Isolation Valve 2Y MS Line 2 to AFPT 1 PV MS106A cross-tie Isolation 83 2
B At 003C C5 6
GT MO 0 AI O/C TO-Q TC-Q 2Y Valve MS107 MSLine2toAFPT2 83 2
B At 003C B7 6
GT MO C AI O/C TO-Q TC-Q 2Y Isolation Valve
__________2Y (Continued) 97 of 105
FENOC FrsfEnergy Nuclear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve q
0 C
- d 9e't Alternate Tests Valve Description 0
Code Required Testing er e
Number-Performed MS Line I to AFPT 2 PV MS107A cross-tie Isolation 83 2
B At 003C C6 6
GT MO 0 Al 0/C TO-Q TC-Q 2Y Valve AFPT MS Minimum MS 145 Flow Line Check 83 3
C At 003C E3 1.5 CK SA 0
N C
FF-Q RF-Q CS-34 FF-SI RF-C Valve AFPT MS Minimum MS 146 Flow Line Check 83 3
C At 003C E7 1.5 CK SA 0
N C
FF-Q RF-Q CS-34 FF-SI RF-C Valve MS375 MS Line 2 Warm-up 83 2
B At 003A G7 1.5 GL AO 0
C C
FC-Q TC-Q 2Y Drain CIV PV MS394 MS Line 1 Warm-up 83 2
B At 003A D7 1.5 GL AO 0
C C
FC-Q TC-Q 2Y Drain CIV MS603 S 2 Blowdown Line 83 2
B At 007B GI 4
GT MO C
AI C
TC-Q 2Y Isolation Valve MS611 SG 1 Blowdown Line 83 2
B At 007B H12 4
GT MO C Al C
TC-Q 2Y Isolation Valve MS Line I to AFW SD MS726 Pump Turbine 1 Supply 83 3
C At 003C B3 6
CK SA C
N O/C FF-Q RF-Q SDJ-2 PF-RO RO Line Check Valve MS Line 2 to AFW MS727 Pump Turbine 2 Supply 83 3
C At 003C B7 6
CK SA C
N 0/C FF-Q RF-Q SDJ-2 PF-RE SD-RE Line Check Valve MS Line 2 to AFW MS734 Pump Turbine I Cross-83 3
C At 003C C3 6
CK SA 0 N 0/C FF-Q RF-Q tie Check Valve MS Line 1 to AFW MS735 Pump Turbine 2 Cross-83 3
C At 003C C7 6
CK SA 0
N O/C FF-Q RF-Q tie Check Valve MS5889A AFW Pump Turbine 1 83 3
B At 003C E2 4
GL AO C
0 0
TO-Q PV-FO-Q steam Admission Valve 8 3
B A C
2Y MS5889B AFW Pump Turbine 2 83 3
B At 003C E8 4
GL AO C
0 0
TO-Q PV-FO-Q steam Admission Valve 8 3
B A
C 2Y SPI7A1 MS Line 2 Code Safety 83 2
C At 007A A7 6x10 RL SA C
N 0
SR-5Y Relief Valve SP17A2 MS Line 2 Code Safety 83 2
C At 007A B7 6x10 RL SA C
N 0
SR-5Y Relief Valve SPI7A3 MS Line 2 Code Safety 83 2
C At 007A B5 6x10 RL SA C
N 0
SR-5Y Relief Valve SP17A4 MS Line 2 Code Safety 83 2
C At 007A A6 6x10 RL SA C
N 0
SR-5Y Relief Valve SP17A5 MS Line 2 Code Safety 83 2
C At 007A B6 6x10 RL SA C
N 0
SR-5Y Relief Valve SP17A6 MS Line 2 Code Safety 83 2
C At 007A A6 6x8 RL SA C
N 0
SR-5Y Relief Valve SP17A7 MS Line 2 Code Safety 83 2
C At 007A B6 6x8 RL SA C
N 0
SR-5Y Relief Valve I
SPI7A8 MS Line 2 Code Safety 83 2
C At 007A B5 6x10 RL SA C
N 0 SR-5Y Relief Valve 8
2R SP17A9 MS Line 2 Code Safety 83 2
C At 007A A5 6x10 RL SA C
N 0
SR-5Y Relief Valve SP17BI MS Line1CodeSafety 83 2
C At 007A A9 6xl0 RL SA C
N 0
SR-5Y Relief Valve SP17B32 IMS Line I Code Safety 83 2
C At 007A 6x10 RL SA C
N 0
SR-5Y Relief Valve 8
2 C
A 0A 8
R MS Line 1 Code Safely
-0 SP17B3 fety 83 2
C At 007A B10 6x10 RL SA C
N 0
SR-5Y Relief Valve I
I I
I I
I I
(Continued) 98 of 105
FENOC FirsrEnergy Nuclear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve Ce Q
Alternate Tests Number Valve Description 5
0 U
2 a
Code Requir Testing Numbe Performed SP17B4 MS Line 1 Code Safety 83 2
C At 007A A10 6x10 RL SA C
N 0
SR-5Y Relief Valve SP17B5 MSLine I Code Safety 83 2
C At 007A B9 6x10 RL SA C
N 0
SR-5Y
_ Relief Valve P7B6MS Line Code Safety 83 2
C At 007A A9 6x8 RL SA C
N 0
SR-5Y SP17B6_ iRelief Valve SP17B7 MS Line l Code Safety 83 2
C At 007A B9 6x8 RL SA C
N 0
SR-5Y Relief ValveIII SP17B8 MS Line I Code Safety 83 2
C At 007A BlI 6x10 RL SA C
N 0
SR-5Y Relief Valve SP17B9 MS Line I Code Safety 83 2
C At 007A A1O 6x10 RL SA C
N 0
SR-5Y Relief Valve RC Letdown Cooler!
PV-MU1A RnLetdown Vale 65 1
B At 031A E5 2.5 GT MO 0 Al C
TC-Q PV2 CS-35 TC-C Inlet Isolation Valve 2Y MUIB RC Letdown Cooler 2 65 1
B At 031A F5 2.5 GT MO 0 Al C
TC-Q PV CS-35 TC-C Inlet Isolation Valve
__2Y Letdwn oole OuletPV MU2A Letdown Cooler Outlet 65 2
A At 031A E8 2.5 GT MO 0 Al C
TC-Q Y
-B CS-35 TC-C CIV 2Y MU2B RC letdown Iolation 65 1
B At 031A E3 2.5 GT MO 0 Al C
TC-Q PV-CS-35 TC-C Valve I2Y PV-L-
S3 CCT MU3 Letdown CIV 65 2
A At 031A E9 2.5 GT AO 0 C
C FC-Q TC-Q 2
U-CS-35 PC-C TC-C I
2Y PV-L-
S3 CCT MU38 RCP Seal Return CIV 65 2
A At 031B B7 I
GL AO 0 C
C FC-Q TC-Q U-B CS-36 FC-C TC-C 2Y M5ARCP 2-I Seal Return LJBCSV T
MU59A 65 2
A At 031B B2 I
GL MO 0 Al C
2YI MU59B RCP 2-2 Seal Return 65 2
A At 031B C2 I
L MO 0 Al C
TC-Q PV-U-B CS-36 TC-C CIV 2Y MU59C RCP 1-1 Seal Return 65 2
A At 031B D3 GL MO 0 Al C
TCQ PV LJ-B CS-36 TC-C CIV 2Y MU59D RCP 1-2 Seal Return 65 2
A At 031B E3 1
GL MO 0 A]
C TCQ PV-U-B CS-36 TC-C CIV.
2Y MU66A RCP 2-1 Seal Injection 65 2
A At 031B K6 1.5 GL AO 0 C
C FC-Q TC-Q PV-L-B CS-36 FC-C TC-C CIV 6
2 A
2Y PV-U-B CS-36 PC-C TC-C MU66B RCP 2-2 Seal Injection 65 2
A At 031B J5 1.5 GL AO 0 C
C FC-Q TC-Q PV CIV 2Y MU66C RCP I -I Seal Injection 65 2
A At 031B H5 1.5 GL AO 0 C
C FC-Q TC-Q PV-U-B CS-36 FC-C TC-C CIV t I3-B 2Y T
MU66D RCP 1-2 Seal Injection 65 2
A At 031B F6 1.5 GL AO 0 C
C FC-Q TC-Q PV2 U-B CS-36 FC-C TC-C CIV 2Y Normal MU to RCS MU169 Injection Line Check 65 2
C At 031C G2 2
CK SA 0
N O/C FF-Q RF-Q CS-37 FF-SI RF-C
_ Valve RCP 2-1 Seal Injection MU242 Stp65 2
N C
FF-Q RF-Q U-B CS-38 FF-SI RF-C
_____Stop Check CIV MU243 RCP2-2 Seal Injection 65 2
N C
FF-Q RF-Q U-B CS-38 FF-SI RF-C Stop Check CIV MU244 RCP I-1 Seal Injection 65 2
N C
FF-Q RF-Q U-B CS-38 FF-SI RF-C Stop Check CIV MU245 RtpCPh-ea nec tio 6I 5
2 JACI At 031B P5 1.5 SC SA 0
N C
PP-Q RP-Q U-B CS-38 FE-SI RE-C Alternate MU to RCS MU800 Injection Line Check 65 2
C At 031C D2 2.5 CK SA C
N O/C FF-Q RF-Q CS-39 FF-C RF-SI Valve 3-way Valve to Align PV MU3971 MU Pump Suction to 65 2
B At 031C F11 4
TW MO 0 Al C
TC-Q 2Y CS-40 TC-C I BWST or MU Tank.
I IIIIIII
_I (Continued) 99 of 105
FENOC FirstEnergy Nuclear Operaring Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve W
AlternateTests Valve Description a
0, Z
""Number Code Required Testing Performed V,.ve i
Number Perf"*
med 3-way Valve to align PV MU6405 MU Pump Suction to 65 2
B At 031C D1O 4
TW MO 0 Al C
TC-Q 2Y CS-40 TC-C BWST or MU Tank.
MU6421 Alternate MU to RCS 65 2
A At 031C D3 2.5 GT MO C Al C
TC-Q PV2y I-B MU6422 NormalMUtoRCS 65 2
A At 031C G3 2.5 GT MO 0 Al C
TC-Q PV U
L-B CS-41 TC-C CIV 2Y N2 Supply to NN58 Pressurizer Quench 74 2
AC At 019 E9 1
CK SA E
N C
FF-Q RF-Q U-B CS-42 FF-SI RF-C Tank CIV N2 Supply to PV NN236 Pressurizer Quench 74 2
A At 019 ElO 1
GL AO 0 C
C FC-Q TC-Q UY IJ-B Tank CIV N2 Supply Line Check LK NN1OOO Valve to Electrical 74 2
AC At 019 B4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PAPIB.
2Y N2 Supply Line Check LK NN1002 Valve to Electrical 74 2
AC At 019 C4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PBP1C.
N2 Supply Line Check LK NN1004 Valve to Electrical 74 2 AC At 019 C4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PBP1 D.
N2 Supply Line Check LK NN1006 Valve to Electrical 74 2 AC At 019 C4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PBL1E.
N2 Supply Line Check LK NN1008 Valve to Electrical 74 2 AC At 019 D4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration P4L1G.
N2 Supply Line Check LK NN1010 Valve to Electrical 74 2 AC At 019 D4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PBC2D.
N2 Supply Line Check LK NNI012 Valve to Electrical 74 2 AC At 019 D4
.375 CK SA C N O/C FF-Q RF-Q 2Y RJ-7 FF-R RFR Penetration PCL2E.
N2 Supply Line Check LK NN1014 Valve to Electrical 74 2 AC At 019 D4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCL2F.
N2 Supply Line Check LK NN1016 Valve to Electrical 74 2 AC At 019 E4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCL2G.
N2 Supply Line Check LK NN1018 Valve to Electrical 74 2
AC At 019 E4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RFR
,Penetration P1P3B.
I_
N2 Supply Line Check LK NN1020 Valve to Electrical 74 2
AC At 019 S4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PAL3D.
N2 Supply Line Check LK NN1022 Valve to Electrical 74 2
AC At 019 E4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PAC3E.
II N2 Supply Line Check LK NN1024 Valve to Electrical 74 2 AC At 019 F4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PAP3F.
N2 Supply Line Check LK NN1026 Valve to Electrical 74 2 AC At 019 F4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RFR Penetration PBP4A.
I N2 Supply Line Check LK NN1028 Valve to Electrical 74 2 AC At 019 F4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PAP4B.
(Continued) 100 of 105
FENOC FirstEnergy Nuclear Operatig Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve U
Code Required Testing Alternate Tests Valv Decito U
Number Valve Description Q
C Performed z
09 N2 Supply Line Check LK NN1030 Valve to Electrical 74 2
AC At 019 F4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration P3P4C.
N2 Supply Line Check LK NN1032 Valve to Electrical 74 2
AC At 019 G4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PBC4D.
N2 Supply Line Check LK NN1034 Valve to Electrical 74 2
AC At 019 G4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PBL4E.
2Y N2 Supply Line Check LK NN1036 Valve to Electrical 74 2
AC At 019 G4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration P2L4G.
N2 Supply Line Check LK NN1038 Valve to Electrical 74 2
AC At 019 G4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PBP5A.
N2 Supply Line Check LK NN1040 Valve to Electrical 74 2
AC At 019 G4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PAP5B.
I___
N2 Supply Line Check LK NN1042 Valve to Electrical 74 2
AC At 019 H4
.375 CK SA C
N O/C FF-Q RF-Q RJ-7 FF-R RF-R Penetration PBP5D.
I N2 Supply Line Check LK NN1044 Valve to Electrical 74 2
AC At 019 H4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R
,Penetration P2P5F.
N2 Supply Line Check LK NN1046 Valve to Electrical 74 2
AC At 019 H4
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration P2C5G.
N2 Supply Line Check LK NN1050 Valve to Electrical 74 2 AC At 019 C7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PILI L.
L N2 Supply Line Check LK NN1052 Valve to Electrical 74 2 AC At 019 C7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PAC IN.
N2 Supply Line Check LK NN1054 Valve to Electrical 74 2 AC At 019 C7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PAP IP.
N2 Supply Line Check LK NN1056 Valve to Electrical 74 2 AC At 019 C7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PBP1R.
I__
N2 Supply Line Check LK NN1058 Valve to Electrical 74 2 AC At 019 C7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R 2Y Penetration PI C2L.
N2 Supply Line Check LK NN1060 Valve to Electrical 74 2 AC At 019 D7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PI P2M.
I I
I N2 Supply Line Check LK NN1062 Valve to Electrical 74 2 AC At 019 D7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RFR Penetration PAL2N.
I N2 Supply Line Check LK NN1064 Valve to Electrical 74 2 AC At 019 D7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PAP2P.
N2 Supply Line Check LK NN1066 Valve to Electrical 74 2 AC At 019 D7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PBC3P.
I_
N2 Supply Line Check LK NN1068 Valve to Electrical 74 2 AC At 019 E7
.375 CK SA C
N O/C FF-Q RF-Q RJ-7 FF-R RF-R 2Y Penetration PBL3Q.
(Continued) 101 of 105
FENOC FirstEnargy Nucjear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
"Val*ve"
- o Alternate Tests Valve Valve Description Code Required Testing Number s
Performed 0~
a.>
N2 Supply Line Check LK NN1070 Valve to Electrical 74 2
AC At 019 E7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCP4N.
N2 Supply Line Check LK NN1072 Valve to Electrical 74 2
AC At 019 E7
.375 CK SA C
N O/C FF-Q RF-Q LK RJ-7 FF-R RF-R Penetration PCP4P.
2Y N2 Supply Line Check LK NN1074 Valve to Electrical 74 2 AC At 019 E7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCP4Q.
_I I
N2 Supply Line Check LK NN1076 Valve to Electrical 74 2 AC At 019 E7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration P3L4S.
III N2 Supply Line Check LK NN1078 Valve to Electrical 74 2 AC At 019 F7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCC4TX N2 Supply Line Check LK NNI080 Valve to Electrical 74 2 AC At 019 F7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCC4UX N2 Supply Line Check LK NN1082 Valve to Electrical 74 2
AC At 019 F7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCC4V.
N2 Supply Line Check LK NN1084 Valve to Electrical 74 2
AC At 019 F7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCL4W.
I__
N2 Supply Line Check LK NN1086 Valve to Electrical 74 2 AC At 019 G7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCP5N.
N2 Supply Line Check LK NN1088 Valve to Electrical 74 2 AC At 019 G7
.375 CK SA C
N O/C FF-Q RF-Q 2RJ-7 FF-R RF-R Penetration PCP5P.
N2 Supply Line Check LK NN1090 Valve to Electrical 74 2 AC At 019 G7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCP5Q.
I N2 Supply Line Check LK NN1092 Valve to Electrical 74 2
AC At 019 H7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R Penetration PCC5T.
I__
N2 Supply Line Check LK NN1094 Valve to Electrical 74 2
AC At 019 H7
.375 CK SA C
N O/C FF-Q RF-Q 2Y RJ-7 FF-R RF-R 2Penetration PCC5U.
N2 Supply Line Check LK NN1096 Valve to Electrical 74 2
AC At 019 H7
.375 CK SA C
N O/C FF-Q RF-Q RJ-7 FF-R RF-R Penetration PCC5V.
I__
RC1O Pressurizer Spray Line 64 1
B At 030A B8 2.5 GT MO 0 Al C
TC-Q PV-CS-43 TC-C Isolation Valve 2Y PORV Line Block PV RC O
I Valve 64 1
B At 030A B4 2.5 GT MO 0 AIO/C TO-Q TC-Q 2Y RC13A Pressurizer Code Safety 64 1
C At 030A B6 4x6 RL SA C
N 0
SR-5Y Relief Valve RC13B Pressurizer Code Safety 64 1
C At 030A A5 4x6 RL SA C
N 0
SR-5Y Relief Valve I
I RC51 Pressurizer Auxiliary 64 1
C At 030A B8 1.5 CK SA C
N 0
FF-Q RF-Q RI-8 FF-R RF-SI Spray Check Valve I
IIII RClI13 Quench Tank Inlet Line 64 2 AC At 040A E3 2
CK SA E
N C
FF-Q RF-Q U-B CS-44 FF-SI RF-C CIV Pressurizer Vent Line 64 1
B P 030A B3 1
GT MO C A C
Stop Valve M
C PV2Y (Continued) 102 of 105
FENOC FirstEnergy Nuclear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve
.45odR
- ~t Alternate Tests Number Valve Description C
R equired Testing M
Perf S
RC229A Quench Tank Outlet 64 2
A At 040A G4 3
GL AO 0 C
C FC-Q TC-Q PVT LJ-B CIV 2Y RC229B Quench Tank Outlet 64 2
A At 040A G2 3
GL AO 0 C
C FC-Q TC-Q PV-IU-B CIV 2Y RC229C RC Penetration#48 64 2 AC At 040A G2 3/8 CK SA C
N O/C FF-Q RF-Q IU-B CS-45 FF-SI RF-C Check Valve RC232 Quench Tank Inlet CIV 64 2
A At 040A E4 2
GL AO 0
C C
U-B 2Y Pressurizer Vapor PV RC239A Space Sample Isolation 64 1
B P 030A B3 1
GT MO C AI C
2Y
_____Valve RC240A Pressurizer Sample 64 1 A At 030A B2 I
GT MO C
Al C
TC-Q PV-U-B Line CIV I
2Y RC240B Pressurizer Sample 64 2
A At 030A B1 1
GT MO C Al C
TC-Q PV-U-B Line CIV I
2Y RC262 Pressurizer Spray 64 1
B P 030A B10 2.5 GL MA C N
C PV Bypass 2Y RC1719A CTMT Vent Header 64 2
A At 040A H7 3
C FC-Q TC-Q U-B CV I-2Y RC1719B CTMT Vent Header 64 2
A At 040A H8 3
C C
FC-Q TC-Q PV-U-B CIV 2Y RC1773A CTMT Drain Header 64 2
A At 040A B9 3
C C
FC-Q TC-Q PV2
-B CIV 2Y RC1773B CTMT Drain Header 64 2
A At 040A C9 3
C C
FC-Q TC-Q PV-U-B CIV 2Y RC4608A RCSLoop I High Point 64 1
B At 030A DlI I
GL SO C
C O/C TO-Q TC-Q 2Y FC-Q CS-46 TO-C TC-C EC-C Vent Valve RC4608B RCS Loop 1 High Point 64 1
B At 030A DI0 1
GL SO C C O/C TO-Q TC-Q PV-FC-Q CS-46 TO-C TC-C FC-C Vent Valve III 2YI I
VentVale
___V2 FC-Q CS-46 TFO-C TC-C FC-C RC4610A RCS Loop 2 High Point 64 1
B At 030A F2 I
GL SO C
C O/C TO-Q TC-Q 2Y FC-Q CS-46 TO-C TC-C FC-C Vent Valve H
P RC4610B RCS Loop2 High Point 64 1
B At 030A G2 I
GL SO C
C O/C TO-Q TC-Q PV-FC-Q CS-46 TO-C TC-C FC-C Vent Valve I
2Y SS235A Quench Tank Vapor 38 2
A At 040A D4 1
GL AO C C
C FC-Q TC-Q PV-U-B Sample CIV 2Y SS235B Quench Tank Vapor 38 2
A At 040A D3 I
GL AO C C
C FC-Q TC-Q PV U-B SS235B Sample CIV 2Y SS598 OTSG2 SampleLine 38 2
B At 007A H4
.75 GL AO 0 C
C FC-Q TC-Q PV2 S 5 8
-PV SS607 TI a
382 B
At 007A F8
.75 GL AO G
C C
FC-Q TC-Q 2Y SS607__
CIV SW17 SW Pump 1 Discharge 11 3
C At 041A G3 20 CK SA E
N O/C FF-Q RF-Q CS-47 FF-C RF-SI Check Valve 1
C sW18 SWPump2 Discharge 11 3
C At 041A G10 20 CK SA E
N O/C FF-Q RF-Q CS-47 FF-C RF-SI Check Valve SW19 SW Pump 3 Discharge 11 3
C At 041A G7 20 CK SA E
N O/C FF-Q RF-Q CS-47 FF-C RF-SI Check Valve Return from TPCW HX SW57 to SW Line Check 11 3
C At 041A B6 20 CK SA 0
N C
FF-Q RF-Q CS-48 FF-SI RF-C Valve SW Train 2 Emergency EX SW232 MUSplytoCCWTRN 11 3
B At 036A K2 1
GL MA C
N O/C EX-Q RV-2 2Y 2 Isolation Valve SW Train I Emergency EX SW233 MUSplytoCCWTRN 11 3
B At 036B B2 I
GL MA C N 0/C EX-Q RV-2 EY 2 Isolation Valve
[
M (Continued) 103 of 105
FENOC Frst'eErnergy Nuclear Operating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued)
Valve Decitin' Alternate Tests
"" cd Code Required Testing AlentTss Number Valve Description Code*equired eting Per or"e Nme S
7i.
Performed SW Train 1 Emergency EX SW234 MUSplytoCCWTRN 11 3
B At 041B J6 1
GL MA C
N O/C EX-Q RV-2 EX 2 Isolation Valve 2Y SW Train 2 Emergency EX SW236 MUSplytoCCWTRN 11 3
B At 041B A8 1
GL MA C N O/C EX-Q RV-2 2Y 2 Isolation Valve SW1356 CAC1SWReturnLine 11 2
B At 041C C1 8
BL AO 0 0 O/C TO-Q TC-Q PV2 FO-Q Isolation Valve 2Y SW1357 CAC 2 SW Return Line 11 2
B At 041C C9 8
BL AO 0
0 O/C TO-Q TC-Q PV-FO-Q Isolation Valve 1
2 2Y PV-F SW1358 CAC3SWReturnLine 11 2
B At 041C C5 8
BL AO C
0 O/C TO-Q TC-Q 2Y FO-Q SW 1366 SW Supply to CAC 1slto av 11 2
B At 041 C H5 8
B L MO 0O A I O/C TO-Q TC-Q PVT2
_____Isolation Valve I2 PV SW1367 SWSupplytoCAC2 11 2
B At 041C H12 8
BL MO 0 AI O/C TO-Q TC-Q 2Y Isolation Valve PV SW1368 SWSupplytoCAC3 11 2
B At 041C H8 8
BL MO 0 Al O/C TO-Q TC-Q 2Y Isolation Valve SW Pump I Strainer SW1379 Blowdown Line Block 11 3
B At 041A H4 4
GT MO C
Al O/C TO-Q TC-Q Valve SW Pump 2 Strainer SW1380 Blowdown Line Block 11 3
B At 041A Hit 4
GT MO C Al O/C TO-Q TC-Q
,Valve SW Pump 3 Strainer SW1381 Blowdown Line Block 11 3
B At 041A H7 4
GT MO C Al O/C TO-Q TC-Q Valve SW to AFW Pump I PV SW1382 Suction Line Block 11 3
B At 041C J4 6
BF MO C Al 0
TO-Q 2Y Valve SW to AFW Pump 2 PV SW1383 Suction Line Block 11 3
B At 041C K9 6
BF MO C Al 0
TO-Q 2Y Valve SW Supply to TPCW PV SW1395 HXLine Isolation 11 3
B At 041A CIO 20 BF MO 0 Al C
TC-Q 2Y Valve SW Supply to TPCW PV SW1399 HXLinelsolation 11 3
B At 041A D8 20 BF MO C AI C
TC-Q 2Y Valve CCW HX I SW Outlet PV SW1424 Line Temperature 11 3
B At 041B C7 12 BL AO C 0 0
TO-Q 2Y FO-Q Control Valve CCW HX 3 SW Outlet PV SW1429 Line Temperature 11 3
B At 041B C9 12 BL AO C
0 0
TO-Q 2Y FO-Q Control Valve CCW HX 2 SW Outlet PV SW1434 Line Temperature 11 3
B At 041B Cit 12 BL AO C
0 0
TO-Q 2Y FO-Q Control Valve C.R. Emergency PV SW2927 Condenser 1 SWSply 11 3
B At 041B F7 1.5 OT MO C AI 0
TO-Q 2Y Line Isolation Valve C.R. Emergency PV SW2928 Condenser2SWSply 11 3
B At 041B Fll 1.5 GT MO C
AI 0
TO-Q 2Y Line Isolation Valve SW Discharge to Intake PV SW2929 Structure Isolation 11 3
B At 041C A4 20 BF MO C
AI O/C TO-Q TC-Q 2Y Valve SW2930 SW Discharge to Intake PV Forebay Isolation Valve I 3 B At 041C A5 30 BF MO C AI 0/C TO-Q TC-Q 2Y (Continued) 104 of 105
FENOC FirstEnorqy Nuclear Gpirating Company 13.0 VALVE TEST TABLE Davis-Besse Nuclear Power Station Unit 1 Third Ten Year Inservice Testing Program Revision 0 (Continued) 105 of 105 Valve Valve Description 4
Code Required Testing Alternate Tests Numbr
'0 Performed SW Discharge to PV SW2931 Cooling Tower MU 11 3
B At 041C A6 30 BF MO 0 Al C
TC-Q 2Y Isolation Valve SW Discharge to PV SW2932 Collection Box 11 3
B At 041C A8 30 BF MO C AI C
TC-Q 2Y Isolation Valve SW3962 SW Discharge Header 11 3
C At 041A E5 6x8 RL SA C
N 0
SR Relief Valve I0Y I___
SW3963 SW Discharge Header 11 3
C At 041A ElO 6x8 RL SA C
N 0
SR Relief Valve 10Y SW to H2 Dilution PV SW5067 Blower I Line Isolation 11 3
B At 041C J8 1
GT MO C Al 0
TO-Q 2Y Valve SW to H2 Dilution PV SW5068 Blower2 Line Isolation 11 3
B At 041B J4 2
GT MO C AI 0 TO-Q 2Y Valve
Docket Number 50-346 License Number NPF-3 Serial Number 2751 Page 1 of 1 COMMITMENT LIST The following list identifies those actions committed to by the Davis-Besse Nuclear Power Station (DBNPS) in this document. Any other actions discussed in the submittal represent intended or planned actions by the DBNPS. They are described only for information and are not regulatory commitments. Please notify the Manager - Regulatory Affairs (419-321-8450) at the DBNPS of any questions regarding this document or associated regulatory commitments.
COMMITMENTS DUE DATE None
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D-1
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DWG. NO. ISID2-003AJ REV. 5 "MAIN STEAM AND REHEAT SYSTEM SHEET 1" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-003A) REV. 5 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-2
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D-3
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DWG. NO. ISID2-006D. REV. 5 "AUXILIARY FEEDWATER SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-006D, REV. 5 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-4
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DWG. NO. ISID2-007A, REV. 5 "STEAM GENERATOR SECONDARY SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-007A, REV. 5 NOTE: Because of this page's large file size, it may be more convenient to copy the fe to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-5
THIS PAGE IS AN OVERSIZED DRAWING OR FIGURE, THAT CAN BE VIEWED AT THE RECORD TITLED:
DWG. NO. ISID2-007B) REV. 4 "STEAM GENERATOR SECONDARY SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-007B, REV. 4 NOTE: Because of this page's large file size, it may be more convenient to copy the fe to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-6
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DWG. NO. ISID2-010C, REV. 3 "MAKE-UP WATER TREATMENT SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-010C, REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the'Programs/Accessories menu.
D-7
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DWG. NO. ISID2-015A, REV. 3 "INSTRUMENT AIR SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISD2-015A) REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-8
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DWG. NO. ISID2-015D, REV. 3 "STATION AIR SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-015D, REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-9
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DWG. NO. ISID2-017A, REV. 3 "DIESEL GENERATORS" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-017A) REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-10
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DWG. NO. ISID2-017B, REV. 8 "DIESEL GENERATORS AIR START" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-017B, REV. 8 NOTE: Because of this page's large fe size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-11
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DWG. NO. ISID2-019) REV. 5 "NITROGEN SUPPLY SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-019) REV. 5 NOTE: Because of this page's large Mfle size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-12
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DWG. NO. ISID2-023, REV. 3 "CONTAINMENT LEAK-RATE TEST DIAGRAM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-023, REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-13
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DWG. NO. ISID2-029B, REV. 8 "AUX. BLDG. RADWASTE, FUEL HANDLING AND ACCESS CONTROL AREAS,SH. 2" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-029B) REV. 8 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-14
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DWG. NO. ISID2-029C, REV. 3 "CONTAINMENT AND PENETRATION ROOMS SHEET 2" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENTREPORT NUMBER ISID2-029C, REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-15
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DWG. NO. ISID2-029D, REV. 3 "CONTAINMENT AND PENETRATION ROOMS SHEET 3" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-029D, REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-16
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DWG. NO. ISID2-029E, REV. 6 "CONTAINMENT AND PENETRATION ROOMS SHEET 4" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-029E, REV. 6 NOTE: Because of this page's large file size, it may be more convenient to copy the fMle to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-17
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DWG. NO. ISID2-030A, REV. 9 "REACTOR COOLANT SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-030A, REV. 9 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-18
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DWG. NO. ISID2-031A, REV. 3 "MAKE-UP & PURIFICATION SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-031A, REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-19
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DWG. NO. ISID2-031B, REV. 3 "MAKE-UP & PURIFICATION SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-031B, REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-20
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DWG. NO. ISID2-031C, REV. 6 "MAKE-UP & PURIFICATION SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-031C, REV. 6 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-21
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DWG. NO. ISID2-033A, REV. 8 "HIGH PRESSURE INJECTION" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-033A, REV. 8 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-22
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DWG. NO. ISID2-033B, REV. 9 "DECAY HEAT TRAIN 1" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-033B, REV. 9 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-23
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DWG. NO. ISID2-033C, REV. 7 "DECAY HEAT TRAIN 2" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-033C, REV. 7 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-24
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DWG. NO. ISID2-034, REV. 7 "EMERG. CORE COOLING SYSTEM CTMT. SPRAY & CORE FLOODING SYSTEMS" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-034, REV. 7 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-25
OVERSIZED DRAWING OR FIGURE, THAT CAN BE VIEWED AT THE RECORD TITLED:
DWG. NO. ISID2-035, REV. 3 "SPENT FUEL POOL COOLING SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-035, REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-26 THIS PAGE IS AN
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DWG. NO. ISID2-036A, REV. 7 "COMPONENT COOLING WATER SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-036A, REV. 7 NOTE: Because of this page's large file size, it may be more convenient to copy the fie to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-27
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DWG. NO. ISID2-036B, REV. 8 "COMPONENT COOLING WATER SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-036B, REV. 8 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-28
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DWG. NO. ISID2-036C, REV. 10 "COMPONENT COOLING WATER SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-036C, REV. 10 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-29
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D-30
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D-31
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DWG. NO. ISID2-041A, REV. 3 "SERVICE WATER PUMPS AND SECONDARY SERVICE WATER SYSTEM" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-041A, REV. 3 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-32
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D-33
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DWG. NO. ISID2-041C, REV. 10 "SERVICE WATER SYSTEM FOR CONTAINMENT AIR COOLERS" WITHIN THIS PACKAGE...OR, BY SEARCHING USING THE DOCUMENT/REPORT NUMBER ISID2-041C, REV. 10 NOTE: Because of this page's large file size, it may be more convenient to copy the file to a local drive and use the Imaging (Wang) viewer, which can be accessed from the Programs/Accessories menu.
D-34
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D-35