ML082270661

Improved Technical Specification Conversion License Amendment Request, Volume 1, Revision 1, Application of Selection Criteria to the Dbnps Technical Specifications

ML082270661
Person / Time
Site:	Davis Besse
Issue date:	08/07/2008
From:	FirstEnergy Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation
References
L-08-240, TAC MD6398
Download: ML082270661 (41)

Text

DAVIS-BESSE NUCLEAR POWER STATION UNIT 1 IMPROVED TECHNICAL SPECIFICATION CONVERSION LICENSE AMENDMENT REQUEST VOLUME 1 (Rev. 1)APPLICATION OF SELECTION CRITERIA TO THE DBNPS TECHNICAL SPECIFICATIONS

--7 -, -, , , 7, , :_ 7, 7?,L7 " 71ý -, ý ý,, 7 '77, 7 7! ý .T 77 Attachment 1, Volume 1, Rev. 1, Page i of i Summary of Changes Split Report Change Description Affected Pages Typographical error corrected.

New ITS number for Page 13 CTS 4.0.5 has been changed from "5.5.6" to "5.5.7." New ITS number for CTS 3/4.3.2.3 has been Pages 16 and 31 changed from "Relocated" to "3.3.16" and the Appendix A, Page 7 discussion has been deleted to be consistent with the Davis-Besse response to question 200805211307.

Typographical error corrected.

Criterion for Inclusion Page 20 for CTS 3/4.7.1.7 has been changed from "3" to "4." Changes have been made to be consistent with the Pages 20 and 21 Davis-Besse response to question 200801161532.

a) New ITS Number for CTS 3/4.7.6.1 has been changed from "3.3.16" to "3.3.15";

b) New ITS Number for CTS 3/4.9.3 has been changed from"Relocated" to "3.9.3," Retained/Criterion For Inclusion has been changed from "NO" to "YES-2," and the Note discussion has been deleted; and c)New ITS Number for CTS 3/4.9.4 has been changed from "3.3.15, 3.9.3" to "Deleted," Retained/Criterion For Inclusion has been changed from "YES-3" to"NO," and a new Note discussion has been provided.Changes have been made to be consistent with the Pages 16 and 32 Davis-Besse response to question 200801231234.

New ITS number for CTS 3/4.3.3.5.2 has been changed from "Relocated" to "3.3.18," Retained/Criterion For Inclusion has been changed from "No" to "YES," the Note discussion has been deleted, and the Appendix A, Page 8 discussion has been deleted.0 0 Page 1 of 1 Attachment 1, Volume 1, Rev. 1, Page i of i Attachment 1, Volume 1, Rev. 1, Page 1 of 39 0 ATTACHMENT 1 VOLUME 1 DAVIS-BESSE IMPROVED TECHNICAL SPECIFICATIONS CONVERSION APPLICATION OF SELECTION CRITERIA TO THE DAVIS-BESSE NUCLEAR POWER STATION TECHNICAL SPECIFICATIONS Revision 1 0 0 Attachment 1, Volume 1, Rev. 1, Page 1 of 39 Attachment 1, Volume 1, Rev. 1, Page 2 of 39 APPLICATION OF SELECTION CRITERIA TO THE DAVIS-BESSE NUCLEAR POWER STATION TECHNICAL SPECIFICATIONS 0 CONTENTS Page 1. IN T R O D U C T IO N ....................................................................................................

1 2. SELECTIO N C R ITER IA ...........................................................................................

2 3. PROBABILISTIC RISK ASSESSMENT (PRA) INSIGHTS ......................................

5 4. RESULTS OF APPLICATION OF SELECTION CRITERIA .......................

7 5. R E FE R E N C E S .8..........................................................................................................

8 0 ATTACHMENT

1.

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION APPENDIX A. JUSTIFICATION FOR SPECIFICATION RELOCATION Attachment 1, Volume 1, Rev. 1, Page 2 of 39 Attachment 1, Volume 1, Rev. 1, Page 3 of 39 APPLICATION OF SELECTION CRITERIA TO THE DAVIS-BESSE NUCLEAR POWER STATION TECHNICAL SPECIFICATIONS

1. INTRODUCTION The purpose of this document is to confirm the results of the Babcock & Wilcox (B&W) Owners Group application of the Technical Specification selection criteria on a plant specific basis for the Davis-Besse Nuclear Power Station. Davis-Besse has reviewed the application and confirmed the. applicability of the selection criteria to each of the Technical Specifications utilized in BAW-1923, Volume 1, "Justification and Background for Technical Specification Improvements," submitted by letter dated February 16, 1987 (Reference 1); B&W Owners Group Technical Report 47-1170689-00, "Application of Selection Criteria to the B&W Standard Technical Specifications," submitted by letter dated October 15, 1987 (Reference 2); NRC Staff Review of Nuclear Steam Supply System (NSSS) Vendor Owners Groups'Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications (Wilgus/Murley letter), dated May 9, 1988 (Reference

3) and as revised in NUREG-1430, "Standard Technical Specifications, Babcock & Wilcox Plants," Revision 3.1 (Reference

4) and applied the criteria to each of the current DBNPS Unit 1 Technical Specifications.

Additionally, in accordance with the NRC Final Policy Statement on Technical Specification Improvements (Reference 5), this confirmation of the application of selection criteria includes confirming the risk insights from site-specific Probabilistic Risk Assessment (PRA) evaluations for Davis-Besse Nuclear Power Station.0 Page 1 of 7 Attachment 1, Volume 1, Rev. 1, Page 3 of 39 Attachment 1, Volume 1, Rev. 1, Page 4 of 39 APPLICATION OF SELECTION CRITERIA TO THE DAVIS-BESSE NUCLEAR POWER STATION* TECHNICAL SPECIFICATIONS

2. SELECTION CRITERIA Davis-Besse has utilized the selection criteria provided in the NRC Final Policy Statement on Technical Specification Improvements of July 22, 1993 (Reference

5) to develop the results contained in the attached matrix. Site-specific PRA insights were utilized, and are discussed in the next section of this report. The selection criteria of 10 CFR 50.36 and discussion provided in Reference 5 are as quoted below: Criterion 1: Installed instrumentation that is used to detect, andindicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary: Discussion of Criterion 1: A basic concept in the adequate protection of the public health and safety is the prevention of accidents.

Instrumentation is installed to detect significant abnormal degradation of the reactor coolant pressure boundary so as to allow operator actions to either correct the condition or to shut down the plant safely, thus reducing the likelihood of a loss-of-coolant accident.This criterion is intended to ensure that Technical Specifications control those instruments specifically installed to detect excessive reactor coolant system leakage. This criterion should not, however, be interpreted to include instrumentation to detect precursors to reactor coolant pressure boundary leakage or instrumentation to identify the source of actual leakage (e.g., loose parts monitor, seismic instrumentation, valve position indicators).

Criterion 2: A process variable, design feature, or operating restriction that is an initial condition of a design basis accident (DBA) or transient analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier: Discussion of Criterion 2: Another basic concept in the adequate protection of the public health and safety is that the, plant shall be operated within the bounds of the initial conditions assumed in the existing design basis accident and transient analyses and that the plant will be operated to preclude unanalyzed transients and accidents.

These analyses consist of postulated events, analyzed in the Final Safety Analysis Report (FSAR), for which a structure, system, or component must meet specified functional goals. These analyses are contained in Chapters 6 and 15 of the FSAR (or equivalent chapters) and are identified as Condition II, Ill, or IV events (ANSI N18.2) (or equivalent) that either assume the failure of or present a challenge to the integrity of a fission product barrier.As used in Criterion 2, process variables are only those parameters for which specific values or ranges of values have been chosen as reference bounds in the design basis accident or transient analyses and which are monitored and controlled during power operation such that process values remain within the analysis bounds. Process variables captured by Criterion 2 are not, however, limited to only those directly monitored and controlled from the control room.These could also include other features or characteristics that are specifically assumed in design basis accident and transient analyses even if they cannot be directly observed in the control room (e.g, moderator temperature coefficient and hot channel factors).0 Page 2 of 7 Attachment 1, Volume 1, Rev. 1, Page 4 of 39 Attachment 1, Volume 1, Rev. 1, Page 5 of 39 I APPLICATION OF SELECTION CRITERIA TO THE DAVIS-BESSE NUCLEAR POWER STATION TECHNICAL SPECIFICATIONS

2. SELECTION CRITERIA (continued)

The purpose of this criterion is to capture those process variables that have initial values assumed in the design basis accident and transient analyses, and which are monitored and controlled during power operation.

As long as these variables are maintained within the established values, risk to the public safety is presumed to be acceptably low. This criterion also includes active design features (e.g.,'high pressure/low pressure system valves and interlocks) and operating restrictions (pressure/temperature limits) needed to preclude unanalyzed accidents and transients.

Criterion 3: A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier: Discussion of Criterion 3: A third concept in the adequate protection of the public health and safety is that in the event that a postulated design basis accident or transient should occur, structures, systems, and components are available to function or to actuate in order to mitigate the consequences of the design basis accident or transient.

Safety sequence analyses or their equivalent have been performed in recent years and provide a method of presenting the plant response to an accident.

These can be used to define the primary success paths.A safety sequence analysis is a systematic examination of the actions required to mitigate the consequences of events considered in the plant's design basis accident and transient analyses, as presented in Chapters 6 and 15 of the plant's Final Safety Analysis Report (or equivalent chapters).

Such a safety sequence analysis considers all applicable events, whether explicitly or implicitly presented.

The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criteria), so that the plant response to design basis accidents and transients limits the consequences of these events to within the appropriate acceptance criteria.It is the intent of this criterion to capture into Technical Specifications only those structures, systems, and components that are part of the primary success path of a safety sequence analysis.

Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function.

The primary success path for a particular mode of operation does not include backup and diverse equipment (e.g., rod withdrawal block which is a backup to the average power range monitor high flux trip in the startup mode, safety valves which are backup to low temperature overpressure relief valves during cold shutdown).

Criterion 4: A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety: Discussion of Criterion 4: It is the Commission policy that licensees retain in their Technical Specifications LCOs, action statements and Surveillance Requirements for the following systems (as applicable), which operating experience and PSA have generally shown to be significant to public health and safety and any other structures, systems, or components that meet this criterion:

Page 3 of 7 Attachment 1, Volume 1, Rev. 1, Page 5 of 39 Attachment 1, Volume 1, Rev. 1, Page 6 of 39 APPLICATION OF SELECTION CRITERIA TO THE DAVIS-BESSE NUCLEAR POWER STATION TECHNICAL SPECIFICATIONS

2. SELECTION CRITERIA (continued)

• Reactor Core Isolation Cooling/Isolation Condenser,* Residual Heat Removal,* Standby Liquid Control, and* Recirculation Pump Trip.The Commission recognizes that other structures, systems, or components may meet this criterion.

Plant and design-specific PSA's have yielded valuable insight to unique plant vulnerabilities not fully recognized in the safety analysis report design basis accident or transient analyses.

It is the intent of thiscriterion that those requirements that PSA or operating

.experience exposes as significant to public health and safety, -consistent with the Commission's Safety Goal and Severe Accident Policies, be retained or included in Technical Specifications.

The Commission expects that licensees, in preparing their Technical Specification related submittals, will utilize any plant specific PSA or risk survey and any available literature on risk insights and PSAs. This material should be employed to strengthen the technical bases for those requirements that remain in Technical Specifications, when applicable, and to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that are commonly found to dominate risk.Similarly, the NRC staff will also employ risk insights and PSAs in evaluating Technical Specifications related submittals.

Further, as a part of the Commission's ongoing program of improving Technical Specifications, it will continue to consider methods to make better use of risk and reliability information for defining future generic Technical Specification requirements.

0 Page 4 of 7 Attachment 1, Volume 1, Rev. 1, Page 6 of 39 Attachment 1, Volume 1, Rev. 1, Page 7 of 39 APPLICATION OF SELECTION CRITERIA TO THE DAVIS-BESSE NUCLEAR POWER STATION* TECHNICAL SPECIFICATIONS

3. PRA INSIGHTS Introduction and Objectives Reference 5 includes a statement that NRC expects licensees to utilize any plant-specific PSA or risk survey and any available literature on risk insights and PSAs to strengthen the technical bases for these requirements that remain in Technical Specifications, and to verify that none of the requirements to be relocated contain constraints of prime importance in limiting the likelihood or severity of the accident sequences that. are commonly found to dominate risk.Those Technical Specifications proposed as being relocated to other plant controlled documents will be maintained under programs subject to the 10 CFR 50.59 review process. These Relocated Specifications have been compared to site-specific PRA material with two purposes:

1) to identify if a Specification component or topic is addressed by PRA; and 2) if addressed, to judge if the Relocated Specification component or topic is risk-important.

The intent of the PRA review was to provide an additional screen to the deterministic criteria.

This review was accomplished in the generic B&W Owners review (Reference 2). The results of this generic review have been confirmed by Davis-Besse for the Davis-Besse Nuclear Power Station Technical Specifications to be relocated.

Where Reference 2 did not review a Davis-Besse Nuclear Power Station Technical Specifications against the criteria of Reference 5, Davis-Besse performed a review similar (but not identical) to that described below for Reference 2.Assumptions and Approach O Any relocated system or component specifically addressed by PRA material is assumed to participate in core melt or plant risk. The first step in the screening process was to identify those systems and components.

The risk significance of the contribution of an identified system or component was then assessed.

PRA data, initiating events, sequence frequencies, fault trees, and event trees were examined to aid in the judgment of the risk significance.

No specific screening criteria were used, and analyst's judgment and experience were relied upon to make the decision for risk significance.

In some cases the judgments were clearly supported by the PRA material used. In other cases the judgments were subjective.

When making the decisions based on PRA, the general approach used was to assume a loss or degradation of the function for those systems or components.

In one sense this provides a crude sensitivity analysis to permit judgments on the importance of the subject of the specification under review. This approach is conservative since the systems or components will be managed by the utility to prevent significant degradation of system performance.

0 Page 5 of 7 Attachment 1, Volume 1, Rev. 1, Page 7 of 39 Attachment 1, Volume 1, Rev. 1, Page 8 of 39 APPLICATION OF SELECTION CRITERIA TO THE DAVIS-BESSE NUCLEAR POWER STATION TECHNICAL SPECIFICATIONS

4. RESULTS OF APPLICATION OF SELECTION CRITERIA The selection criteria from Section 2 were applied to the Davis-Besse Nuclear Power Station Technical Specifications.

The following Summary Disposition Matrix is a summary of that application indicating which Specifications are being retained or relocated, the criteria for inclusion, if applicable, the NRC results of the criteria application as expressed in the NRC Staff Review of NSSS Vendor Owners Groups' Application of The Commission's Interim Policy Statement Criteria To Standard Technical' Specifications (Wilgus/Murley letter), dated May 9, 1988 (Reference 3), and any necessary explanatory notes. Discussions that document the rationale for the relocation of each Specification which failed to meet the selection criteria are provided in Appendix A, except as noted in the Summary Disposition Matrix. In addition, Appendix A includes a discussion of the PRA evaluations performed for those Davis-Besse Nuclear Power Station specific Technical Specifications being relocated.

0 Page 6 of 7 Attachment 1, Volume 1, Rev. 1, Page 8 of 39 Attachment 1, Volume 1, Rev. 1, Page 9 of 39 APPLICATION OF SELECTION CRITERIA TO THE DAVIS-BESSE NUCLEAR POWER STATION TECHNICAL SPECIFICATIONS

5. REFERENCES

1. BAW-1923, Volume 1, "Justification and Background for Technical Specification Improvements," submitted by letter dated February 16, 1987.2. B&W Owners Group Technical Report 47-1170689-00, "Application of Selection Criteria to the B&W Standard Technical Specifications," submitted by letter from R. Gill (BWOG)to T. Murley (NRR), dated October 15, 1987.3. "NRC Staff Review of Nuclear Steam Supply Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications," submitted by letter from T. Murley (NRR) to W. Wilgus (BWOG), dated May 9, 1988.4. NUREG-1430, "Standard Technical Specifications, Babcock & Wilcox Plants," Revision 3.1, December, 2005.5. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).Page 7 of 7 Attachment 1, Volume 1, Rev. 1, Page 9 of 39 Attachment 1, Volume 1, Rev. 1, Page 10 of 39 ATTACHMENT I

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION 0 Attachment 1, Volume 1, Rev. 1, Page 10 of 39

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION CID 0 CD CD CID_0 (a3 (0 CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 1.0 DEFINITIONS 1.1 YES This section provides definitions for several defined terms used throughout the remainder of Technical Specifications.

They are provided to improve the meaning of certain terms. As such, direct application of the Technical Specification selection criteria is not appropriate.

However, only.those definitions for defined terms that remain as a result of application of the selection criteria, will remain as definitions in this section of Technical Specifications.

2.0 SAFETY

LIMITS AND LIMITING 2.0 SAFETY SYSTEM SETTINGS 2.1 Safety Limits 2.1 2.1.1 Reactor Core -Combination of the 2.1.1.2, YES Application of Technical Specification selection criteria is not reactor coolant core outlet pressure and 2.2.2 appropriate.

However, Safety Limits will be included in Technical outlet temperature Specifications as required by 10 CFR 50.36.2A1.2 Reactor Core -Combination of reactor 2.1.1.1, YES Same as above THERMAL POWER and AXIAL 2.2.1 POWER IMBALANCE 2.1.3 Reactor Coolant System Pressure 2.1.2, YES Application of Technical Specification selection criteria is not 2.2.3, appropriate.

However, Safety Limits will be included in Technical 2.2.4 Specifications as required by 10 CFR 50.36.2.2 Limiting Safety System Settings 2.2.1 ReactorProtection System Setpoints 3.3.1, YES-3 The RPS LSSS have been included as part of the RPS 3.3.2 instrumentation Specification, which has been retained since the Functions -either actuate to mitigate consequences of design basis accidents and transients or are retained as directed by the NRC as the Functions are part of the RPS.(a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 1 of 13 0)CD 0 CD CD CD

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION CD CD0 CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.0 LIMITING CONDITIONS FOR 3.0 OPERATION AND SURVEILLANCE REQUIREMENTS

-APPLICABILITY

3.0.1 Operational

Modes LCO 3.0.1 YES This Specification provides generic guidance applicable to one or more Specifications.

The information is provided to facilitate understanding of Limiting Conditions for Operation and Surveillance Requirements.

As such, direct application of the Technical Specification-selection criteria is not appropriate.

However, the general requirements of 3.0/4.0 will be retained in Technical Specifications, as modified consistent with NUREG-1430, Revision 3.3.0.2 Noncompliance LCO 3.0.2 YES Same as above.3.0.3 Generic Actions LCO 3.0.3 YES Same as above.3.0.4 Entry into Operational Modes LCO 3.0.4 YES Same as above.3.0.5 Operability Exception 3.8.1 YES The application of Technical Specification selection criteria is not appropriate.

However, this exception to the definition of OPERABILITY has been included as part of the Required Actions in new LCO 3.8.1.3.0.6 Actions Exceptions LCO 3.0.5 YES This Specification provides generic guidance applicable to one or more Specifications.

The information is provided to facilitate understanding of Limiting Conditions for Operation and Surveillance Requirements.

As such, direct application of the Technical Specification selection criteria is not appropriate.

However, the general requirements of 3.0/4.0 will be retained in Technical Specifications, as modified consistent with NUREG-1430, Revision 3.CD 0 CD CD CD-o 0)C.0 (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 2 of 13

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION 3)CD 0 C 3D CD CD CD 0A Q0 CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 4.0.1 Operational Modes SR 3.0.1 YES Same as above.4.0.2 Time of Performance SR 3.0.2 YES Same as above.4.0.3 Noncompliance SR 3.0.1, YES Same as above.SR 3.0.3 4.0.4 Entry into Operational Modes SR 3.0.4 YES Same as above.4.0.5 ASME Code Class 1, 2, and 3 5.5.7 YES This Specification is actually a Surveillance Requirement which has Components been retained in the Administrative Controls programs for Inservice Testing.3/4.1 REACTIVITY CONTROL SYSTEMS 3.1 3/4.1.1 Boration Control 3/4.1.1.1 SHUTDOWN MARGIN 3.1.1, YES-2 3.1.2, 3.2.1 3/4.1.1.2 Boron Dilution Deleted NO Deleted, see Boron Dilution technical change discussion in the Discussion of Changes for CTS 3/4.1.1.2.

3/4.1.1.3 Moderator Temperature Coefficient 3.1.3 YES-2 3/4.1.1.4 Minimum Temperature for Criticality 3.4.2. YES-2 3/4.1.2 Boration Systems 3/4.1.2.1 Flow Paths -Shutdown Relocated NO See Appendix A, Page 1.3/4.1.2.2 Flow Paths -Operating Relocated NO See Appendix A, Page 3.3/4.1.2.3 Makeup Pump -Shutdown Relocated NO See Appendix A, Page 1.(a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 3 of 13 0)CD 0 CD CD CD co (0 CD

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION 0 0)=)0 E CD CD CD CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.1.2.4 Makeup Pumps -Operating Relocated NO See Appendix A, Page 3.3/4.1.2.5 Decay Heat Removal Pump -Shutdown Relocated NO See Appendix A, Page 5.3/4.1.2.6 Boric Acid Pump -Shutdown Relocated NO See Appendix A, Page 1.3/4.1.2.7 Boric Acid Pumps -Operating Relocated NO See Appendix A, Page 3.3/4.1.3 Movable Control Assemblies 3/4.1.3.1 Group Height -Safety and Regulating 3.1.4 1YES-2 Rod Groups 3/4.1.3.2 Group Height- Axial Power Shaping 3.1.6 YES-2 Group 3/4.1.3.3 Position Indicator Channels 3.1.7 YES-2.3/4.1.3.4 Rod Drop Time 3.1.4 YES-2 This Specification has been incorporated as a Surveillance Requirement (SR 3.1.4.3) in ITS 3.1.4.3/4.1.3.5 Safety Rod Insertion Limit 3.1.5 YES-2 3/4.1.3.6 Regulating Rod Insertion Limits 3.2.1 YES-2 3/4.1.3.7 Rod Program Relocated NO See Appendix A, Page 6.3/4.1.3.8 Xenon Reactivity

-Deleted NO Deleted, see Xenon Reactivity technical change discussion in the Discussion of Changes for CTS 3/4.1.3.8.

3/4.1.3.9 Axial Power Shaping Rod Insertion 3.2.2 YES-2 Limits (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 4 of 13 0)CD 0 2 CD CD CD 0 CA)Q0

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION 0)CD 0 CD CD CD, 0-h WA (0 CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.2 POWER DISTRIBUTION LIMITS 3.2 3/4.2.1 Axial Power Imbalance 3.2.3 YES-2 3/4.2.2 Nuclear Heat Flux Hot Channel Factor -3.2.5 YES-2 FQ 3/4.2.3 Nuclear Enthalpy Rise Hot Channel 3.2.5 YES-2 Factor -F NAH 3/4.2.4 Quadrant Power Tilt 3.2.4 YES-2 3/4.2.5 DNB Parameters 3.4.1 YES-2 3/4.3 INSTRUMENTATION 3.3 3/4.3.1 Reactor Protection System 3.3.1, YES-3 Instrumentation 3.3.2, 3.3.3, 3.3.4,-3.3.9, 3.3.10 3/4.3.2 Safety System Instrumentation 3/4.3.2.1 Safety Features Actuation System 3.3.5, YES-3 Instrumentation 3.3.6, 3.3.7, 3.3.8, 3.4.14, 3.8.1 3/4.3.2.2 Steam and Feedwater Rupture Control 3.3.11, YES-3 System Instrumentation 3.3.12, 3.3.13 (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 5 of 13 0 CD 0 C: CID CID CD C,, 0 Q0

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION 0)C)CD 0 C: CD CD CE)0 C.0 CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.3.2.3 Anticipatory Reactor Trip System 3.3.16 YES Instrumentation 3/4.3.3 Monitoring Instrumentation 3/4.3.3.1 Radiation Monitoring Instrumentation Instrument 1 Area Monitors Instrument 1.a -Fuel Storage Pool Area Emergency 3.3.14 YES-3 Ventilation System Actuation Instrument 2 Process Monitors Instrument 2.a.i Containment Gaseous Activity RCS 3.4.15 YES-1 Leakage Detection Instrument 2.a.ii Containment Particulate Activity RCS 3.4.15 YES-1 Leakage Detection 3/4.3.3.2 Relocated by Amendment 234 NA NA 3/4.3.3.3, Relocated by Amendment 201 NA NA 3/4.3.3.4 3/4.3.3.5.1 Remote Shutdown Instrumentation 3.3.18 YES-4 3/4.3.3.5.2 Appendix R Remote Shutdown 3.3.18 YES Requirements 3/4.3.3.6 Post-Accident Instrumentation 3.3.17 YES-3 See Appendix A, Page 9. Instrumentation that does not monitor Regulatory Guide 1.97 Type A or Category 1 variables has been relocated in accordance with the guidance provided in NUREG-1430, Revision 3.(a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 6 of 13 0 CID 0 CD CD Co CD 0')"0

SUMMARY

DISPOSITION MATRIX FORAVIS-BESSE NUCLEAR POWER STATION CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.4 REACTOR COOLANT SYSTEM 3.4 3/4.4.1 Reactor Coolant Loops and Coolant Circulation 0 3/4.4.1.1 Startup and Power Operation 3.4.4 YES-2 _Y CD W4.4.1.2 Shutdown and Hot Standby 3.4.5, YES-3 3.4.6, 3.4.7, 3.4.8 o 3/4.4.2 Safety Valves -Shutdown 3.4.12 YES-3 3 CD 3/4.4.3 Safety Valves and Pilot Operated Relief 3.4.10, YES-3 Valve -Operating 3.4.11 X CD 3/4.4.4 Pressurizer 3.4.9 YES-2 3/4.4.5 Steam Generators 3.4.17 YES-2 3/4.4.6 Reactor Coolant System Leakage CD 3/4.4.6.1 Leakage Detection Systems 3.4.15 YES-1 -4 0 3/4.4.6.2 Operational Leakage 3.4.13, YES-2 h 3.4.14 QO 3/4.4.7 Relocated by Amendment 234 NA NA 3/4.4.8 Specific Activity 3.4.16 YES-2 3/4.4.9 Pressure/Temperature Limits 3/4.4.9.1 Reactor Coolant System 3.4.3 YES-2 (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 7 of 13

SUMMARY

DISPOSITION MATRIX FOR AVIS-BESSE NUCLEAR POWER STATION 0, CE)=3 0 C 3D CD CD 0-h C.4 C.0 CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.4.10 Structural Integrity 3/4.4.10.1 ASME Code Class 1, 2 and 3 Relocated NO See Appendix A, Page 11. The Reactor Coolant Pump Flywheel Components and the Reactor Vessel Internals Vent Valves Surveillances are being retained as Programs in ITS 5.5.3/4.5 EMERGENCY CORE COOLING 3.5 SYSTEMS 3/4.5.1 Core Flooding Tanks 3.5.1 YES-3 3/4.5.2 ECCS Subsystems -Tavg > 280°F 3.5.2, -YES-3 3.4.14, YES-2 3.6.7 YES-3 3/4.5.3 ECCS Subsystems

-Tavg < 2801F 3.5.3, YES-3 3.6.7 3/4.5.4 Borated Water Storage Tank 3.5.4 YES-3 3/4.6 CONTAINMENT SYSTEMS 3.6 3/4.6.1 Primary Containment 3/4.6.1.1 Containment Integrity 3.6.1 YES-3 3/4.6.1.2 Containment Leakage 3.6.1, YES-3 Containment leakage is being retained as a Surveillance

3.6.3 Requirement

(SR 3.6.1.1) in ITS 3.6.1 and the containment purge and exhaust isolation valves with resilient seals leakage is being retained as a Surveillance Requirement (SR 3.6.3.5) in ITS 3.6.3.3/4.6.1.3 Containment Air Locks 3.6.2 YES-3 3/4.6.1.4 Internal Pressure 3.6.4 YES-2 (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 8 of 13 0, C)CD 0 CD (0 0 C-0

SUMMARY

DISPOSITION MATRIX FOR AVIS-BESSE NUCLEAR POWER STATION CE)0 E CD CD CD 0 C.0 CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.6.1.5 Air Temperature 3.6.5 YES-2 3/4.6.1.6 Deleted in Amendment 205 NA NA 3/4.6.1.7 Containment Ventilation System 3.6.3 YES-3 Containment purge valves are being retained as a Surveillance Requirement (SR 3.6.3.1) in ITS 3.6.3.3/4.6.2 Depressurization and Cooling Systems 3/4.6.2.1 Containment Spray System 3.6.6 YES-3 3/4.6.2.2 Containment Cooling System 3.6.6 YES-3 3/4.6.3.1 Containment Isolation Valves 3.6.3 YES-3 3/4.6.4 Deleted by Amendment 265 NA NA 3/4.6.5 Shield Building 3/4.6.5.1 Emergency Ventilation System 3.7.12, YES-3 3.7.13 3/4.6.5.2 Shield Building Integrity 3.7.12 YES-3 3/4.7 PLANT SYSTEMS 3.7 3/4.7.1 Turbine Cycle 3/4.7.1.1 Safety Valves 3.7.1 YES-3 3/4.7.1.2 Auxiliary Feedwater System 3.7.5 YES-3 3/4.7.1.3 Condensate Storage System 3.7.6 YES-2, 3 (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 9 of 13 C)CID 0 C: CID CID CD (0 0-h w0

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 0 CD 0 CD CD CD CD 0-h wA (D 3/4.7.1.4 3/4.7.1.5 3/4.7.1.6 3/4.7.1.7 3/4.7.1.8 3/4.7.1.9 3/4.7.2 3/4.7.3.1 3/4.7.4.1 3/4.7.5.1 3/4.7.6.1 3/4.7.7 Activity Main Steam Line Isolation Valves Deleted by Amendment 47 Motor Driven Feedwater Pump System Main Feedwater Control Valves and Startup Feedwater Control Valves Turbine Stop Valves Steam Generator Pressure/Temperature Limitation Component Cooling Water System Service Water System Ultimate Heat Sink Control Room Emergency Ventilation System Snubbers Sealed Source Contamination Steam Generator Level 3.7.17 3.7.2 NA 3.7.5 3.7.3 3.7.4 Relocated 3.7.7 3.7.8 3.7.9 3.3.15, 3.7.10 3.7.11 LCO 3.0.8 Relocated 3.7.18 YES-2 YES-3 NA YES-4 YES-3 YES-3 NO YES-3 YES-3 YES-2 YES-3 See Appendix A, Page 13.0 CD 0 CD CD CD 0 (0-h Co YES NO YES-2 The Actions are retained as LCO 3.0.8. The LCO and Surveillances are deleted. See Snubbers technical change discussion in the Discussion of Changes for CTS 3/4.7.7.See Appendix A, Page 14.3/4.7.8 3/4.7.9 (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific technical Specification selection criteria are met.Page 10 of 13

SUMMARY

DISPOSITION MATRIX FOR DAVIS-BESSE NUCLEAR POWER STATION 0 CD CD CD 0)Q0 CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.8 ELECTRICAL POWER SYSTEM 3.8 3/4.8.1 A.C. Sources 3/4.8.1.1 Operating 3.8.1, YES-3 3.8.3 3/4.8.1.2 Shutdown 3.8.2, YES-3 3.8.3 3/4.8.2 Onsite Power Distribution Systems 3/4.8.2.1 A.C. Distribution

-Operating 3.8.9 YES-3 3/4.8.2.2 A.C. Distribution

-Shutdown 3.8.10 YES-3 3/4.8.2.3 D.C. Distribution

-Operating 3.8.4, YES-3 3.8.6, 3.8.9 3/4.8.2.4 D.C. Distribution

-Shutdown 3.8.5, YES-3 3.8.6, 3.8.10 3/4.9 REFUELING OPERATIONS 3.9 3/4.9.1 Boron Concentration 3.9.1 YES-2 3/4.9.2 Instrumentation 3.9.2 YES-3 3/4.9.3 Decay Time 3.9.3 YES-2 3/4.9.4 Containment Penetrations Deleted NO Deleted, see Containment Penetrations technical change discussion in the Discussion of Changes for CTS 3/4.9.4.(a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 11 of 13 0)C., CD CD CD 0)CD

SUMMARY

DISPOSITION MATRIX FOR VIS-BESSE NUCLEAR POWER STATION CD.0 CD CD CD 0 CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a).(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.9.6 Fuel Handling Bridge OPERABILITY Relocated NO See Appendix A, Page 15.3/4.9.8 Decay Heat Removal and Coolant Circulation 3/4.9.8.1 All Water Levels 3.9.4, YES-4 3.9.5 3/4.9.8.2 Low Water Level 3.9.5 YES-4 3/4.9.9 Deleted by Amendment 186 NA NA 3/4.9.10 Water Level -Reactor Vessel 3.9.6 YES-2 3/4.9.11 Storage Pool Water Level 3.7.14 YES-2, 3 3/4.9.12 Storage Pool Ventilation System 3.7.13 YES-3 3/4.9.13 Spent Fuel Assembly Storage 3.7.16 YES-3 3/4.10 SPECIAL TEST EXCEPTIONS NA 3/4.10.1 Group Height, Insertion and Power 3.1.8 YES This Specification is provided to allow relaxation of certain LCOs Distribution Limits under certain specific conditions to allow testing. Direct application of the Technical Specification selection criteria is not appropriate.

However, since this special test exception is directly tied to LCOs that remain in Technical Specifications and the testing described in the special test exception is still required, this Specification will remain in the Technical Specifications.

3/4.10.2 Physics Tests 3.1.9 YES Same as above.3/4.10.3 Reactor Coolant Loops Deleted NO Deleted, see Reactor Coolant Loops technical change discussion in the Discussion of Changes for CTS 3/4.10.3.(a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 12 of 13 CD 0 C: 3D CD CD CD 0

SUMMARY

DISPOSITION MATRIX FORIVIS-BESSE NUCLEAR POWER STATION 0.CD 0 E CD CD (.0 CD 0 CIO CURRENT TS CURRENT TITLE NEW TS RETAINED/

NOTES(a)(CTS) NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.10.4 Shutdown Margin Deleted NO Deleted, see Shutdown Margin technical change discussion in the Discussion of Changes for CTS 3/4.10.4.314.11 RADIOACTIVE EFFLUENTS NA 3/4.11.1 Liquid Holdup Tanks 5.5 YES Although this Specification does not meet any Technical Specification selection criteria, it has been retained in accordance with the NRC letter from W. T. Russell to the industry ITS Chairpersons, dated October 25, 1993.3/4.11.2 Explosive Gas Mixture 5.5 YES Same as above.5.0 DESIGN FEATURES 4.0 YES Application of Technical Specification selection criteria is not appropriate.

However, specific portions of Design Features will be included in Technical Specifications as required by 10 CFR 50.36.6.0 ADMINISTRATIVE CONTROLS 5.0 YES Application of Technical Specification selection criteria is not appropriate.

However, specific portions of Administrative Controls will be included in Technical Specifications as required by 10 CFR 50.36.-0-CD 0 C CD CID CD 0-h C-0 (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met.Page 13 of 13 Attachment 1, Volume 1, Rev. 1, Page 24 of 39 APPENDIX A JUSTIFICATION FOR SPECIFICATION RELOCATION 0 Attachment 1, Volume 1, Rev. 1, Page 24 of 39 Attachment 1, Volume 1, Rev. 1, Page 25 of 39 Appendix A -Justification For Specification Relocation O 3/4.1.2.1:

FLOW PATHS -SHUTDOWN 3/4.1.2.3:

MAKEUP PUMP -SHUTDOWN 3/4.1.2.6:

BORIC ACID PUMP -SHUTDOWN LCO STATEMENT:

3/4.1.2.1 At least one of the following boron injection flow paths shall be OPERABLE.a. A flow path from the concentrated boric acid storage system via a boric acid pump and a makeup or decay heat removal (DHR) pump to the Reactor Coolant System, if only the boric acid storage system is OPERABLE, or b. A flow path from the borated water storage tank via a makeup or DHR pump to the Reactor Coolant System if only the borated water storage tank is OPERABLE.3/4.1.2.3 At least one makeup pump in the boron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE essential bus.3/4.1.2.6 At least one boric acid pump shall be OPERABLE and capable of being powered from an OPERABLE essential bus if only the flow path through the boric acid pump in Specification 3.11.2.1 a is OPERABLE.DISCUSSION:

Theboration subsystems of the Makeup and Purification System and Chemical Addition System provide the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the SHUTDOWN MARGIN.COMPARISON TO SCREENING CRITERIA: 1. The boration subsystems are not used for, nor are capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.2. The boration subsystems are not used to indicate status of, or monitor a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient.

3. The boration subsystems are not part of a primary success path in the mitigation of a DBA or transient.

4. As discussed in Reference 2 (Appendix A pages A-3 through A-6), the loss of the boration subsystems was found to be non-significant risk contributor to core damage frequency and offsite releases.

Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

Page 1 of 15 Attachment 1, Volume 1, Rev. 1, Page 25 of 39 Attachment 1, Volume 1, Rev. 1, Page 26 of 39 Appendix A -Justification For Specification Relocation CONCLUSION:

Since the screening criteria have not been satisfied, the Flow Paths -Shutdown LCO and Surveillances, Makeup Pump -Shutdown LCO and Surveillances, and the Boric Acid Pump -Shutdown LCO and Surveillances may be relocated to other plant controlled documents outside Technical Specifications.

0 Page 2 of 15 Attachment 1, Volume 1, Rev. 1, Page 26 of 39 Attachment 1, Volume 1, Rev. 1, Page 27 of 39 Appendix A -Justification For Specification Relocation 3/4.1.2.2:

FLOW PATHS -OPERATING 3/4.1.2.4:

MAKEUP PUMPS -OPERATING 3/4.1.2.7:

BORIC ACID PUMPS -OPERATING LCO STATEMENT:

3/4.1.2.2 Each of the following boron injection flow paths shall be OPERABLE: a. A flow path from the concentrated boric acid storage system via a boric acid pump and makeup or decay heat removal (DHR) pump to the Reactor Coolant System, and b. A flow path from the borated water storage tank via a makeup or DHR pump to the Reactor Coolant System.3/4.1.2.4 Two makeup pumps shall be OPERABLE.3/4.1.2.7 At least one boric acid pump in the boron injection flow path required by Specification 3.1.2.2a shall be OPERABLE and capable of being powered from an OPERABLE essential bus.DISCUSSION:

The boration subsystems of the Makeup and Purification System and Chemical Addition System provide the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the SHUTDOWN MARGIN.COMPARISON TO SCREENING CRITERIA: 1. The boration subsystems are not used for, nor are capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.2. The boration subsystems are not used to indicate status of, or monitor a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient.

3. The boration subsystems are not part of a primary success path in the mitigation of a DBA or transient.

4. As discussed in Reference 2 (Appendix A pages A-3 through A-6), the loss of the boration subsystems was found to be non-significant risk contributor to core damage frequency and offsite releases.

Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

Page 3 of 15 Attachment 1, Volume 1, Rev. 1, Page 27 of 39 Attachment 1, Volume 1, Rev. 1, Page 28 of 39 Appendix A -Justification For Specification Relocation CONCLUSION:

Since the screening criteria have not been satisfied, the Flow Paths -Operating LCO and Surveillances, Makeup Pumps -Operating LCO and Surveillances, and Boric Acid Pumps -Operating LCO and Surveillances may be relocated to other plant controlled documents outside Technical Specifications.

0 Page 4 of 15 Attachment 1, Volume 1, Rev. 1, Page 28 of 39 Attachment 1, Volume 1, Rev. 1, Page 29 of 39 Appendix A -Justification For Specification Relocation 3/4.1.2.5:

DECAY HEAT REMOVAL PUMP -SHUTDOWN LCO STATEMENT:

At least one decay heat removal (DHR) pump in the boron injection flow path required by Specifications 3.1.2.1 or 3.1.2.2 shall be OPERABLE and capable of being powered from an OPERABLE essential bus.DISCUSSION:

The boration subsystems of the Makeup and Purification System and Chemical Addition System provide the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the SHUTDOWN MARGIN.COMPARISON TO SCREENING CRITERIA: 1. The boration subsystems are not used for, nor are capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.2. The boration subsystems are not used to indicate status of, or monitor a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient.

3. The boration subsystems are not part of a primary success path in the mitigation of a DBA or transient.

4. As discussed in Reference 2 (Appendix A pages A-3 and A-4), the loss of the boration subsystems was found to be non-significant risk contributor to core damage frequency and offsite releases.

Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Decay Heat Removal Pump -Shutdown LCO and Surveillances may be relocated to other plant controlled documents outside Technical Specifications.

Page 5 of 15 Attachment 1, Volume 1, Rev. 1, Page 29 of 39 Attachment 1, Volume 1, Rev. 1, Page 30 of 39 Appendix A -Justification For Specification Relocation 3/4.1.3.7:

ROD PROGRAM LCO STATEMENT:

Each control rod assembly (safety, regulating and APSR) shall be programmed to operate in the core location and rod group specified in the CORE OPERATING LIMITS REPORT.DISCUSSION:

The location of control rod assemblies (safety, regulating and APSR) is stipulated in the reload report for each fuel cycle, and are reflected as core location and rod group assignments in the CORE OPERATING LIMITS REPORT. These constraints on control rod assembly core locations and rod group assignments function to optimize core burnup and minimize local power peaking during operation.

Programming (or "patching")

of control rod assemblies is also determined by the reload report for each fuel cycle to ensure that adequate shutdown margin can be achieved when the control rods are tripped.Incorrect programming of control rod assemblies in regulating groups would be revealed during measurement of group rod worths performed during startup testing, and verification that control rod assemblies in safety groups are fully withdrawn is performed using the control rod position indication system. Unlatched control rod assemblies would be detected via core power tilt measurements during power escalation.

When test, reprogramming, or maintenance of the control rod drive patch panel and associated cables and instrumentation is performed, control rod control "programming" is also validated.

O If rod assemblies are not programmed correctly at some point the applicable insertion, overlap, and alignment limit may not be met. The Technical Specifications still include appropriate compensatory actions for insertion, overlap, and alignment limits not met. This will ensure the safety analysis is met or the plant will be required to be shut down within the .specified time frame.COMPARISON TO SCREENING CRITERIA: 1. Rod Program is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.2. Rod Program is not a process variable that is an initial condition in a DBA or transient analyses.3. Rod Program does not act as a part of a primary success path in the mitigation of a DBA or transient.

4. As discussed in Reference 2 (Appendix A pages A-1 3 and A-1 4), the loss of this Specification was found to be non-significant risk contributor to core damage frequency and offsite releases.Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Rod Program LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.

0 Page 6 of 15 Attachment 1, Volume 1, Rev. 1, Page 30 of 39 Attachment 1, Volume 1, Rev. 1, Page 31 of 39 Appendix A -Justification For Specification Relocation 0 Not Used 0 0 Page 7 of 15 Attachment 1, Volume 1, Rev. 1, Page 31 of 39 Attachment 1, Volume 1, Rev. 1, Page 32 of 39 Appendix A -Justification For Specification Relocation Not Used Page 8 of 15 Attachment 1, Volume 1, Rev. 1, Page 32 of 39 Attachment 1, Volume 1, Rev. 1, Page 33 of 39 Appendix A -Justification For Specification Relocation 3/4.3.3.6:

POST-ACCIDENT MONITORING INSTRUMENTATION LCO STATEMENT:

The post-accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.DISCUSSION:

Each individual accident monitoring parameter has a specific purpose; however, the general purpose for all accident monitoring instrumentation is to ensure sufficient information is available following an accident to allow an operator to verify the response of automatic safety systems, and to take preplanned manual actions to accomplish a safe shutdown of the plant.The NRC position on application of the deterministic screening criteria to post-accident monitoring instrumentation is documented in letter dated May 9, 1988 from T.E. Murley (NRC) to W.S. Wilgus (NRC Split Report to Owners Groups). The position taken was that the post-accident monitoring instrumentation table list should contain, on a plant specific basis, all Regulatory Guide 1.97 Type A instruments specified in the plant's Safety Evaluation Report (SER) on Regulatory Guide 1.97, and all Regulatory Guide 1.97 Category 1 instruments.

Accordingly, this position has been applied to the DBNPS Unit 1 Regulatory Guide 1.97 instruments.

Those instruments meeting these criteria have remained in Technical Specifications.

The instruments not meeting this criteria will be relocated from the Technical Specifications to plant controlled documents.

A review of the Davis-Besse Nuclear Power Station UFSAR and the NRC Regulatory Guide 1.97 Safety Evaluation for Davis-Besse Nuclear Power Station shows that the following CTS Tables 3.3-10 and 4.3-10 Instruments do not meet Category 1 or Type A requirements.

Instrument 10 RC System Subcooling Margin Monitor Instrument 11 PORV Position Indicator Instrument 12 PORV Block Valve Position Indicator Instrument 13 Pressurizer Safety Valve Position Indicator Instrument 15 Containment Normal Sump Level COMPARISON TO SCREENING CRITERIA: 1.. These instruments are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).2. The monitored parameters are not process variables, design features, or operating restrictions that are initial conditions of a DBA or transient.

3. These instruments are not part of a primary success path in the mitigation of a DBA or transient.

4. As discussed in Reference 2 (Appendix A pages A-37 through A-43), the loss of the (above listed) instruments were found to be non-significant risk contributor to core damage frequency and offsite releases.

Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

Page 9 of 15 Attachment 1, Volume 1, Rev. 1, Page 33 of 39 Attachment 1, Volume 1, Rev. 1, Page 34 of 39 Appendix A -Justification For Specification Relocation CONCLUSION:

Since the screening criteria have not been satisfied for instruments which do not meet Regulatory Guide 1.97 Type A variable requirements or Category 1 variable requirements, their associated LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.

Page 10 of 15 Attachment 1, Volume 1, Rev. 1, Page 34 of 39 Attachment 1, Volume 1, Rev. 1, Page 35 of 39 Appendix A -Justification For Specification Relocation

• 3/4.4.10.1:

STRUCTURAL INTEGRITY

-ASME CODE CLASS.1,2 AND 3 COMPONENTS LCO STATEMENT:

The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.10.1.DISCUSSION:

The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of these components will be maintained throughout the life of the components.

ASME Code Class 1, 2, and 3 components are monitored so that the possibility of component structural failure does not degrade the safety function of the system. The monitoring activity is of a preventive nature rather than a mitigative action. Other Technical Specifications require important systems to be OPERABLE (for example, Emergency Core Cooling Systems) and in a ready state for mitigative action. This Technical Specification is more directed toward prevention of component degradation and continued long term maintenance of acceptable structural conditions.

Hence, it is not necessary to retain this Specification to ensure immediate OPERABILITY of safety systems.Further, this Technical Specification prescribes inspection requirements that are performed during plant shutdown.

It is, therefore, not directly important for responding to design basis accidents.

COMPARISON TO SCREENING CRITERIA: 1. The programmatic inspections stipulated by this Specification are not installed instrumentation used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary during operations prior to a design basis accident (DBA).2. The programmatic inspections stipulated by this Specification are not a process variable, design feature, or operating restriction that is an initial assumption in a DBA or transient.

3. The ASME Code Class 1, 2, and 3 components inspected per this Specification are assumed to function to mitigate a DBA. Their capability to perform this function is addressed by other Technical Specifications.

This Technical Specification only specifies programmatic inspection requirements for these components, and these inspections can only be performed when the plant is shutdown.

Therefore, Criterion 3 is not satisfied.

4. As discussed in Reference 2 (Appendix A pages A-63 and A-64), the assurance of operability of the entire system as verified in the system operability Specification dominates the risk contribution of the system. The lack of a long term assurance of structural integrity as stipulated by this Specification was found to be non-significant risk contributor to core damage frequency and offsite releases.

Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Structural Integrity

-ASME Code Class 1, 2, and 3 Components LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.

In addition, Surveillances, except for the reactor coolant pump (RCP) flywheel inspection and the internal vent valve requirements, are already required by regulations in 10 CFR 50.55a to be performed in accordance with Section XI of the ASME Boiler and Pressure O Vessel Code and applicable addenda. The RCP flywheel inspection requirement and the internal vent valve requirements are not covered by other regulatory requirements and are needed for safe Page 11 of 15 Attachment 1, Volume 1, Rev. 1, Page 35 of 39 Attachment 1, Volume 1, Rev. 1, Page 36 of 39 Appendix A -Justification For Specification Relocation operation of the plant; therefore, these requirements will be maintained in the Davis-Besse Improved Technical Specifications.

Chapter 5.0 of the Davis-Besse Improved Technical Specifications will contain a section which provides a programmatic approach to the requirements relating to the structural integrity of ASME Code Class 1, 2, and 3 components.

0 Page 12 of 15 Attachment 1, Volume 1, Rev. 1, Page 36 of 39 Attachment 1, Volume 1, Rev. 1, Page 37 of 39 Appendix A -Justification For Specification Relocation 3/4.7.2: STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION LCO STATEMENT:

The temperatures of the secondary coolant in the steam generators shall be > 1 10°F when the pressure of the secondary coolant in the steam generator is > 237 psig.DISCUSSION:

The limitation on steam generator pressures and temperatures ensures that pressure-induced stresses on the steam generators do not exceed the maximum allowable fracture toughness limits. These pressure and temperature limits are based on maintaining a steam generator RTNDT sufficient to prevent brittle fracture.

As such, the Technical Specification places limits on variables consistent with structural analysis results. However, these limits are not initial condition assumptions of a DBA or transient.

These limits represent operating restrictions and Criterion 2 includes operating restrictions.

However, it should be noted that in the Final Policy Statement the Criterion 2 discussion specified only those operating restrictions required to preclude unanalyzed accidents and transients be included in Technical Specifications.

COMPARISON TO SCREENING CRITERIA: 1. The steam generator pressure and temperature limits are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).2. The steam generator pressure and temperature limits are not process variables, design features, or operating restrictions that are an initial condition of a DBA or transient.

3. The steam generator pressure and temperature limits are not part of a primary success path in the mitigation of a DBA or transient.

4. As discussed in Reference 2 (Appendix A pages A-73 and A-74), the steam generator pressure and temperature limits were found to be non-significant risk contributor to core damage frequency and offsite releases.

Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Steam Generator steam generator pressure and temperature Limitation LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.

0 Page 13 of 15 Attachment 1, Volume 1, Rev. 1, Page 37 of 39 Attachment 1, Volume 1, Rev. 1, Page 38 of 39 Appendix A -Justification For Specification Relocation 3/4.7.8: SEALED SOURCE CONTAMINATION LCO STATEMENT:

Each sealed source containing radioactive material either in excess of 100 microcuries of beta and/or gamma emitting material or 5 microcuries of alpha emitting material, shall be free of >0.005 microcuries of removable contamination.

DISCUSSION:

The limitations on sealed source contamination are intended to ensure that the total body and individual organ irradiation doses do not exceed allowable limits in the event of ingestion or inhalation.

This is done by imposing a maximum limitation of < 0.005 microcuries of removable contamination on each sealed source. This requirement and the associated surveillance requirements bear no relation to the conditions or limitations that are necessary to ensure safe reactor operation.

COMPARISON TO SCREENING CRITERIA: 1. Sealed source contamination is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).2. Sealed source contamination is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient.

3. Sealed source contamination is not part of a primary success path in the mitigation of a DBA or transient.

4.As discussed in Reference 2 (Appendix A pages A-77 and A-78), sealed source contamination was found to be non-significant risk contributor to core damage frequency and offsite releases.Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Sealed Source Contamination LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.

Page 14 of 15 Attachment 1, Volume 1, Rev. 1, Page 38 of 39 Attachment 1, Volume 1, Rev. 1, Page 39 of 39 Appendix A -Justification For Specification Relocation S 3/4.9.6: FUEL HANDLING BRIDGE OPERABILITY LCO STATEMENT:

The control rod hoist and fuel assembly hoist of the fuel handling bridge shall be used for movement of control rods or fuel assemblies and shall be OPERABLE with: a. The control rod hoist having: 1. A minimum capacity of 3000 pounds, and 2. An overload cutoff limit < 2650 pounds.b. The fuel assembly hoist having: 1. A minimum capacity of 3000 pounds, and 2. An overload cutoff limit < 2700 pounds.DISCUSSION:

Operability of the fuel handling bridge hoists ensures that the equipment used to handle fuel within the reactor pressure vessel functions as designed and that the equipment has sufficient load capacity for handling fuel assemblies and/or control rod assemblies.

Although the interlocks designed to provide the above capabilities can prevent damage to the refueling equipment and fuel assemblies, they are not assumed to function to mitigate the consequences of a design basis accident.COMPARISON TO SCREENING CRITERIA: 1. Fuel Handling Bridge OPERABILITY is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a design basis accident (DBA).2. Fuel Handling Bridge OPERABILITY is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient.

3. Fuel Handling Bridge OPERABILITY is not part of a primary success path in the mitigation of a DBA or transient.

4. As discussed in Reference 2 (Appendix A pages A-89 and A-90), Fuel Handling Bridge OPERABILITY was found to be non-significant risk contributor to core damage frequency and offsite releases.

Davis-Besse has reviewed this evaluation, considers it applicable to Davis-Besse Nuclear Power Station, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Fuel Handling Bridge OPERABILITY LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications.

0 Page 15 of 15 Attachment 1, Volume 1, Rev. 1, Page 39 of 39