ML051960212

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Improved Technical Specifications, Volume 1, Revision 0, Application of Selection Criteria to the Monticello Nuclear Generating Plant.
ML051960212
Person / Time
Site: Monticello Xcel Energy icon.png
Issue date: 06/29/2005
From:
Nuclear Management Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML051960212 (30)
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Text

IMPROVED TECHNICAL SPECIFICATIONS

.WeT I MONTICELLO NUCLEAR GENERATING PLANT

- VOLUME 1 Application of Selection Criteria to the MNGP Tech Specs I

Committed to ANud Excellecen>

Attachment 1, Volume 1, Rev. 0, Page 1 of 29 ATTACHMENT 1 VOLUME 1 MONTICELLO IMPROVED TECHNICAL SPECIFICATIONS CONVERSION APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS Revision 0 Attachment 1, Volume 1, Rev. 0, Page 1 of 29

Attachment 1, Volume 1, Rev. 0, Page 2 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS CONTENTS Page

1. INTRODUCTION........................................................................................................1
2. SELECTION CRITERIA ............................................... 2
3. PROBABILISTIC RISK ASSESSMENT INSIGHTS ............................................... 5
4. RESULTS OF APPLICATION OF SELECTION CRITERIA ........................................ 8
5. REFERENCES .............................................. 9 ATTACHMENT
1.

SUMMARY

DISPOSITION MATRIX FOR MONTICELLO APPENDIX A. JUSTIFICATION FOR SPECIFICATION RELOCATION Attachment 1, Volume 1, Rev. 0, Page 2 of 29

Attachment 1, Volume 1, Rev. 0, Page 3 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS

1. INTRODUCTION The purpose of this document is to confirm the results of the BWR Owners Group application of the Technical Specification selection criteria on a plant specific basis for Monticello Nuclear Power Station.

Nuclear Management Company, LLC has reviewed the application of the selection criteria to each of the Technical Specifications utilized in BWROG report NEDO-31466, 'Technical Specification Screening Criteria Application and Risk Assessment," including Supplement I (Reference 1), NUREG-1433, Standard Technical Specifications, General Electric Plants, BWR/4," (Reference 2) and applied the criteria to each of the current Monticello Technical Specifications. Additionally, in accordance with the NRC guidance, this confirmation of the application of selection criteria to Monticello includes confirming the risk insights from Probabilistic Risk Assessment (PRA) evaluations, provided in Reference 1, as applicable to Monticello.

Page 1 of 9 Attachment 1, Volume 1, Rev. 0, Page 3 of 29

Attachment 1, Volume 1, Rev. 0, Page 4 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS

2. SELECTION CRITERIA Nuclear Management Company, LLC has utilized the selection criteria provided in the NRC Final Policy Statement on Technical Specification Improvements of July 22, 1993 (Reference 3) to develop the results contained in the attached matrix. Probabilistic Risk Assessment (PRA) insights as used in the BWROG submittal were utilized, confirmed by Nuclear Management Company, LLC and are discussed in the next section of this report. The selection criteria and discussion provided in Reference 3 are as follows:

Criterion 1: Installed instrumentation that is used to detect, and indicate 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, III, or IV events (ANSI N1 8.2) (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).

Page 2 of 9 Attachment 1, Volume 1, Rev. 0, Page 4 of 29

Attachment 1, Volume 1, Rev. 0, Page 5 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO 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 su6cess 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 9 Attachment 1, Volume 1, Rev. 0, Page 5 of 29

Attachment 1, Volume 1, Rev. 0, Page 6 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS

2. SELECTION CRITERIA (continued)
  • Reactor Core Isolation Cooling/Isolation Condenser;

-

  • 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 this criterion 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.

Page 4 of 9 Attachment 1, Volume 1, Rev. 0, Page 6 of 29

Attachment 1, Volume 1, Rev. 0, Page 7 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS

3. PRA INSIGHTS Introduction and Obiectives Reference 3 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 a variety of 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. Those Technical Specifications proposed to remain part of the Improved Technical Specifications were not reviewed. This review was accomplished in Reference 1 except where discussed in Appendix A, "Justification For Specification Relocation," and has been confirmed by Nuclear Management Company, LLC for those Specifications to be relocated.

The Monticello plant-specific Probabilistic Risk Assessment (PRA) was reviewed during this process.

Assumptions and Approach Briefly, the approach used in Reference 1 was the following:

The risk assessment analysis evaluated the loss of function of the system or component whose LCO was being considered for relocation and qualitatively assessed the associated effect on core damage frequency and offsite releases. The assessment was based on available literature on plant risk insights and PRAs. Table 3-1 lists the PRAs used for making the assessments and is provided at the end of this section. A detailed quantitative calculation of the core damage and offsite release effects was not performed. However, the analysis did provide an indication of the relative significance of those LCOs proposed for relocation on the likelihood or severity of the accident sequences that are commonly found to dominate plant safety risks. The following analysis steps were performed for each LCO proposed for relocation:

a. List the function(s) affected by removal of the LCO item.
b. Determine the effect of loss of the LCO item on the function(s).
c. Identify compensating provisions, redundancy, and backups related to the loss of the LCO item.
d. Determine the relative frequency (high, medium, and low) of the loss of the function(s) assuming the LCO item is removed from Technical Specifications and controlled by other procedures or programs. Use information from current PRAs and related analyses to establish the relative frequency.

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Attachment 1, Volume 1, Rev. 0, Page 8 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS

3. PRA INSIGHTS (continued)
e. Determine the relative significance (high, medium, and low) of the loss of the function(s).

Use information from current PRAs and related analyses to establish the relative significance.

f. Apply risk category criteria to establish the potential risk significance or non-significance of the LCO item; Risk categories were defined as follows:

RISK CRITERIA Consequence Freauencv High Medium Low High S NS Medium S S NS Low NS NS NS S = Potential Significant Risk Contributor NS = Risk Non-Significant

9. List any comments or caveats that apply to the above assessment. The output from the above evaluation was a list of LCOs proposed for relocation that could have potential plant safety risk significance if not properly controlled by other procedures or programs.

As a result these Specifications will be relocated to other plant controlled documents outside the Technical Specifications.

Page 6 of 9 Attachment 1, Volume 1, Rev. 0, Page 8 of 29

Attachment 1, Volume 1, Rev. 0, Page 9 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS TABLE 3-1 BWR PRAs USED IN NEDO-31466 (and Supplement 1)

RISK ASSESSMENT

  • La Salle County Station, NEDO-31085, Probabilistic Safety Analysis, February 1988.
  • Grand Gulf Nuclear Station, IDCOR, Technical Report 86.2GG, Verification of IPE for Grand Gulf, March 1987.
  • Peach Bottom 2, NUREG-75/0104, "Reactor Safety Study," WASH-1 400, October 1975.
  • Millstone Point 1, NUREG/CR-3085, "Interim Reliability Evaluation Program: Analysis of the Millstone Point Unit I Nuclear Power Plant," January 1983.
  • Grand Gulf, NUREGICR-1 659, "Reactor Safety Study Methodology Applications Program:

Grand Gulf #1 BWR Power Plant," October 1981.

  • NEDC-30936P, "BWR Owners' Group Technical Specification Improvement Methodology (with Demonstration for BWR ECCS Actuation Instrumentation) Part 2," June 1987.

Page 7 of 9 Attachment 1, Volume 1, Rev. 0, Page 9 of 29

Attachment 1, Volume 1, Rev. 0, Page 10 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS

4. RESULTS OF APPLICATION OF SELECTION CRITERIA The selection criteria from Section 2 were applied to the Monticello Technical Specifications. The attachment is a summary of that application indicating which Specifications are being retained or relocated. Discussions that document the rationale for the relocation of each Specification which failed to meet the selection criteria are provided in Appendix A. No Significant Hazards Considerations (10 CFR 50.92) evaluations for those Specifications relocated are provided with the Discussion of Changes for the specific Technical Specifications. Nuclear Management Company, LLC will relocate those Specifications identified as not satisfying the criteria to licensee controlled documents whose changes are governed by 10 CFR 50.59.

Page 8 of 9 Attachment 1, Volume 1, Rev. 0, Page 10 of 29

Attachment 1, Volume 1, Rev. 0, Page 11 of 29 APPLICATION OF SELECTION CRITERIA TO THE MONTICELLO TECHNICAL SPECIFICATIONS

5. REFERENCES
1. NEDO-31466 (and Supplement 1), "Technical Specification Screening Criteria Application and Risk Assessment," November 1987 and July 1989.
2. NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4,"

Revision 3, June 2004.

3. Final Policy Statement on Technical Specifications Improvements, July 22,1993 (58 FR 39132).

Page 9 of 9 Attachment 1, Volume 1, Rev. 0, Page 11 of 29

, Volume 1, Rev. 0, Page 12 of 29 ATTACHMENT 1

SUMMARY

DISPOSITION MATRIX FOR MONTICELLO , Volume 1, Rev. 0, Page 12 of 29

C

SUMMARY

DISPOSITION MATRIX FOR MONTICELLO NUCLEAR GENERATING PLANT C'

CURRENT TS (CTS) CURRENT TITLE NEW TS RETAINED/ NOTES(a)

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, Su will remain as definitions in this section of Technical Specifications. Cs 3 2.0 SAFETY LIMITS AND LIMITING 2.0 ID a

SAFETY SYSTEM SETTINGS 2

CD 2.1 Safety Limits 2.1 2.1.A Reactor Core Safety Limits 2.1.1 YES Application of Technical Specification selection criteria is 0 not appropriate. However, Safety Limits will be included In CD 0

oF C3 2.1.B Reactor Coolant System Pressure 2.1.2 YES Same as above. 2

-D Safety Limit 0 2.2 Safety Limit Violations 2.2 YES Same as above.

a)

ED 314.0 SURVEILLANCE REQUIREMENTS - 3.0 X APPLICABILITY

-o a) 4.0.A Meeting Surveillance Requirements SR 3.0.1 YES This Specification provides generic guidance applicable to o ID ID and Time of Performance one or more Specifications. The information is provided to C;) facilitate understanding of Surveillance Requirements. As -0L 0o such, direct application of the Technical Specification selection criteria is not appropriate. However, the general (0 requirements of 4.0 will be retained in.Technical Specifications, as modified consistent with NUREG-1 433, Revision 3.

4.0.B Time Interval Extensions SR 3.0.2 YES Same as above.

4.0.C Noncompliance and Time of SR 3.0.1, YES Same as above.

Performance SR 3.0.4 4.0.D Missed Surveillances SR 3.0.1 YES Same as above.

4.0.E Delay Time for Missed Surveillances SR 3.0.3 YES Same as above.

(a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. I

C

SUMMARY

DISPOSITION MATRIX FOR MOTICELLO NUCLEAR GENERATING PLANT c,

CURRENT TS (CTS) CURRENT TITLE NEW TS RETAINED/ NOTES(a)

NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.1 REACTOR PROTECTION SYSTEM 3l4.1.A and B Reactor Protection System 1.1, YES-3 Instrumentation 3.3.1.1, 3.3.6.1, 3.3.6.2 3/4.1 .A and B and Turbine Condenser Low Vacuum Relocated NO See Appendix A, page 1.

Table 3.1.1 Trip Function 9, Table 4.1.1 Instrument Channel 5, s0)

C) and Table 4.1.2 ED Instrument Channel 7 3 0 3 0

3 314.1.C RPS Power Monitoring System 3.3.8.2 YES-3 314.2 PROTECTIVE INSTRUMENTATION 3.3

o M 0 3/4.2.A Primary Containment Isolation 3.3.6.1 YES-3, 4
0) Functions 3/4.2.B Emergency Core Cooling Subsystems 3.3.5.1, YES-3 7, Actuation 3.3.8.1
a 3/4.2.C Control Rod Block Actuation -U 0

AD 3/4.2.C.1 SRM, IRM, APRM and Scram Relocated NO See Appendix A, pages 2 through 5.

Discharge Volume Rod Blocks 3/4.2.C.2 Rod Block Monitor 3.3.2.1 YES-3 ID 3/4.2.D Other Instrumentation 3.3.5.1, YES -3, 4 to 3.3.5.2 3 3/4.2.E Reactor Building Ventilation Isolation 3.3.6.2 YES-3 to and Standby Gas Treatment System 0 Initiation 3/4.2.F Recirculation Pump Trip and Alternate 3.3.4.1 YES-4 CD Rod Injection Initiation 314.2.G Safeguards Bus Voltage Protection 3.3.8.1 YES-3 314.2.H Instrumentation for Safety/Relief Valve 3.3.6.3, YES-3 Low-Low Set Logic 3.6.1.5 3/4.2.1 Instrumentation for Control Room 3.3.7.1 YES-3 Habitability Protection (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. 2

C C

SUMMARY

DISPOSITION MATRIX FOR MONTICELLO NUCLEAR GENERATING PLANT C,

CURRENT TS (CTS) CURRENT TITLE NEW TS RETAINED/ NOTES(a)

NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 314.3 CONTROL ROD SYSTEM 3.1 3/4.3.A Reactivity Limitations 3/4.3.A.1 Reactivity margin - core loading 1.1, , YES-2 3.1.1 3/4.3.A.2 Reactivity margin - stuck control rods 3.1.3 YES-3 3/4.3.B Control Rod Withdrawal 3/4.3.B.1 Coupling 3.1.3, YES-3 i) 03 3.10.5 0 314.3.B1.2 Control Rod Drive Housing Support Deleted NO Deleted, see CRD Housing Support technical change C,

System discussion Inthe Discussion of Changes for CTS: ED a

3/4.3.B.2 M 3/4.3.B.3.(a) Control Rod Withdrawal Sequences 3.1.6, YES-3 3.3.2.1 0

3/4.3.B.3.(b) Rod Worth Minimizer 3.3.2.1 YES-3 0 CD 3/4.3.B1.4 Source Range Monitors for startup and 3.3.1.2 YES refueling

-A 3/4.3.C Scram Insertion Times 3.1.4 YES-3 -U 3/4.3.D Control Rod Accumulators 3.1.5, 3.9.5 YES-3 o

3/4.3.E Reactivity Anomalies 3.1.2 YES-2 3I4.3.F Scram Discharge Volume 3.1.8 YES-3 11) 3/4.3.G Required Action 3.1.1, YES This requirement provides the appropriate actions to take if to D 3.1.3, CTS 3.3.A through D are not met. As such, direct

-4 3.1.4, application of the Technical Specification selection criteria i to o 3.1.5, is not appropriate for actions. Therefore, changes to this 3.1.6, action are discussed in the technical change discussion in tD 3.3.1.2, the Discussion of Changes for ITS 3.1.1, 3.1.3, 3.1.4, 3.9.5 3.1.5, 3.1.6, 3.3.1.2, and 3.9.5.

3/4.4 STANDBY LIQUID CONTROL SYSTEM 3I4.4.A Standby Liquid Control System 3.1.7 YES-4 314.4.B Boron Solution Requirements 3.1.7 YES-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. 3

c CE

SUMMARY

DISPOSITION MATRIX FOR MONTICELLO NUCLEAR GENERATING PLANT C

CURRENT TS (CTS) CURRENT TITLE NEW TS RETAINED/ NOTES(a)

NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.4.C Required Action 3.1.7 YES This requirement provides the appropriate action to take if CTS 3.4.A or B is not met. As such, direct application of the Technical Specification selection criteria Is not appropriate for actions. Therefore, changes to this action are discussed Inthe technical change discussion Inthe Discussion of Changes for ITS 3.1.7.

314.5 CORE AND CONTAINMENT 3.5 IN)0 SPRAYICOOLING SYSTEMS 314.5.A ECCS Systems 3.5.1, YES-3 0 3 3.10.1 4.5.A.4 ADS Inhibit Switch Relocated NO See Appendix A, page 6. 0 0

3/4.5.B RHR Intertie Return Line Isolation 3.5.1 YES-3 Valves

00. 0 itl 3/4.5.C Containment Spray/Cooling System 3.6.2.3, YES-3 F 0 3.6.1.8, 03

-o 3.7.1 El Co 3/4.5.D RCIC 3.3.5.2, YES-4 01 Co 3.5.3, 0) 3.10.1 0 3/4.5.E Cold Shutdown and Refueling 3.5.2 YES-3 Requirements to 3/4.5.F Recirculation System 3.4.1 YES-2 (D 3/4.6 PRIMARY SYSTEM BOUNDARY 3.4 3/4.6.A Reactor Coolant Heatup and Cooldown 3.4.9 YES-2 3/4.6.B Reactor Vessel Temperature and 3.4.9 YES-2 Pressure 3/4.6.C Coolant Chemistry 3/4.6.C.1 Radiolodine concentration in the 3.4.6, YES-2 reactor coolant 3.10.1 3/4.6.C.2 and 3 Reactor Coolant Water Chemistry Relocated NO See Appendix A, page 7.

(a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. 4

  • C C S~UMMARY DI1SPOSITION MATRIX FOR MONTICELLO NUCLEAR GENERATING PLANT

(

CURRENT TS (CTS) CURRENT TITLE NEW TS RETAINED/ NOTES(a)

NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 314.6.C.4 Required Action 3.4.6 YES This requirement provides the appropriate actions to take if CTS 3.6.C.1 through 3 are not met. As such, direct application of the Technical Specification selection criteria Is not appropriate for actions. Therefore, changes to this action are discussed in the technical change discussion in the Discussion of Changes for ITS 3.4.6 and CTS 3/4.6.C.2 and 3.

3/4.6.D Reactor Coolant System (RCS)

A) a) 0 3/4.6.D.1 Operational Leakage 3.4.4 YES-2 0

314.6.D.2 RCS Leakage Detection 3.4.5 YES-1 0

Instrumentation 3/4.6.E Safety/Relief Valves 3.4.3, YES-3 4.,

0-3.6.1.5 0 3/4.6.F Deleted by Amendent 42 0 3/4.6.G Jet Pumps 3.4.2 YES-2 0, D 314.6.H Snubbers Deleted NO Deleted, see Snubbers technical change discussion in the Discussion of Changes for CTS: 314.6.H. 0 M

U 3/4.7 CONTAINMENT SYSTEMS 3.6 0

3/4.7.A Primary Containment CD 3/4.7.A.1 Suppression Pool Volume and 3.5.2, YES-2, 3 a) Temperature 3.6.1.1, to 0

3.6.2.1, 10 3;6.2.2 314.7.A.2 Primary Containment Integrity 0

f, 3/4.7.A.2.a Primary Containment Integrity 3.6.1.1, YES-3 (0 to 3.6.1.3, 3.10.1 3/4.7.A.2.b Deleted by Amendment 132 3/4.7.A.2.c Primary Containment Air Lock 3.6.1.1, YES-3 3.6.1.2 3I4.7.A.3 Pressure Suppression Chamber - 3.6.1.6 YES-3 Reactor Building Vacuum Breakers 3/4.7.A.4 Pressure Suppression Chamber - 3.6.1.1, YES-3 Drywell Vacuum Breakers 3.6.1.7 (a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. 5

c C

SUMMARY

DISPOSITION MATRIX FOR MONTICELLO NUCLEAR GENERATING PLANT C

CURRENT TS (CTS) CURRENT TITLE NEW TS RETAINED/ NOTES(a)

NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.7.A.5 Primary Containment Oxygen 3.6.3.1 YES-2 Concentration 3/4.7.3 Standby Gas Treatment System 3.6.4.3, YES-3 5.5.6 3/4.7.C Secondary Containment 3.6.4.1, YES-3 3.6.4.2 3/4.7.D Primary Containment Isolation Valves 3.6.1.3, YES-3 5.5.11 a) 3/4.7.E Deleted by Amendment 138 0 314.8 Main Condenser Offgas 3I4.8.A Main Condenser Offgas Activity 3.7.6 YES-2 0 C) 314.9 Auxiliary Electrical Systems 3.8 0 4."

3.9.A Operational Requirements for Startup 3.8.1, YES-3 3

3.8.4, 0

3.8.6, 3.8.7 ED 0

4.9.A Substation Switchyard Battery Deleted NO Deleted, see technical change discussion in the 0 Discussion of Changes for ITS 3.8.1. 0-E, 3.9.B Operational Requirements for 3.8.1, YES-3 ;U Co 0

Continued Operation 3.8.3, 3.8.4, 0:1 3.8.6, 0

-h; Co N,

3.8.7 4.9.B.3 Standby Diesel Generator 3.8.1, YES-3 CD 0)

(0 3.8.3, co 5.5.8 4.9.B.4 Station Battery Systems 3.8.4, YES-3 3.8.6, CD 3.8.7 4.9.B.5 24V Battery System Deleted NO Deleted, see technical change discussion in the Discussion of Changes for ITS 3.8.4.

3/4.10 REFUELING 3.9 3I4.10.A Refueling Interlocks 3.9.1, YES-3 3.9.2 3/4/10.B Core Monitoring 3.3.1.2 YES 3l4/10.C Fuel Storage Pool Water Level 3.7.8 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. 6

C

SUMMARY

DISPOSITION MATRIX FOR MONTICELLO NUCLEAR GENERATING PLANT C

CURRENT TS (CTS) CURRENT TITLE NEW TS RETAINED/ NOTES(a)

NUMBER (ITS) CRITERION NUMBER FOR INCLUSION 3/4.10.D Decay Time Deleted NO Deleted, see technical change discussion in the Discussion of Changes for CTS 3.10.D.

3/4.10.E Extended Core and Control Rod Drive 3.10.2, YES-3 Maintenance 3.10.6 314.11 REACTOR FUEL ASSEMBLIES 3.2 314.1.A Average Planar Linear Heat 3.2.1 YES-2 314.11.B Linear Heat Generation Rate 3.2.3 YES-2 W

D} 3/4.11 .C Minimum Critical Power Ratio (MCPR) 3.2.2 YES-2 Di 0

3 C)

ED 3/4.13 ALTERNATE SHUTDOWN SYSTEM 3/4.13.A Alternate Shutdown System 3.3.3.2 YES-4 0

a M 3/4.14 ACCIDENT MONITORING 3.3.3.1, YES-3 See Appendix A, pages 8 and 9. Instrumentation that does INSTRUMENTATION 3.3.6.3 not monitor Regulatory Guide 1.97 Type A or Category 1 0 a4 Z variables has been relocated in accordance with the guidance provided in NUREG-1433, Revision 3. CD 0

3/4.17 CONTROL ROOM HABITABILITY 3/4.17.A Control Room Ventilation System 3.7.5 YES-3 CD 3I4.17.B Control Room Emergency Filtration 3.7.4, YES-3 CD System 5.5.6 0)

UM ED 5.0 DESIGN FEATURES 4.0 YES Application of Technical Specification selection criteria is o not appropriate. However, specific portions of Design to 0 Features will be included in Technical Specifications as Ah co to required by 10 CFR 50.36. 0 ED 6.0 ADMINISTRATIVE CONTROLS 5.0 YES Application of Technical Specification selection criteria is not appropriate. However, specific portions of to Administrative Controls will be Included in Technical Specifications as required by 10 CFR 50.36.

(a) The Applicable Safety Analyses section of the Bases for the individual Technical Specifications describes the reason specific Technical Specification selection criteria are met. .7

, Volume 1, Rev. 0, Page 20 of 29 APPENDIX A JUSTIFICATION FOR SPECIFICATION RELOCATION , Volume 1, Rev. 0, Page 20 of 29

Attachment 1, Volume 1, Rev. 0, Page 21 of 29 3/4.1 A REACTOR PROTECTION SYSTEM LCO STATEMENT:

The setpoints, minimum number of trip systems, and minimum number of instrument channels that must be operable for each position of the reactor mode switch shall be given in Table 3.1.1.

The time from initiation of any channel trip to the de-energization of the scram pilot valve solenoids shall not exceed 50 milliseconds.

3/4.1.A and B, and Table 3.1.1 Trip Function 9, Table 4.1.1 Instrument Channel 5, and Table 4.1.2 Instrument Channel 7 (Turbine Condenser Low Vacuum).

DISCUSSION:

The turbine condenser low vacuum scram is provided to protect the main condenser from overpressurization in the event that vacuum is lost. A loss of condenser vacuum would cause the turbine stop valves to close, resulting in a turbine trip transient. The low condenser vacuum trip anticipates this transient and scrams the reactor. No design basis accidents or transients take credit for this scram signal.

COMPARISON TO SCREENING CRITERIA:

1. The turbine condenser low vacuum scram instrumentation is not an instrument 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 turbine condenser low vacuum scram instrumentation is not used for, nor capable of, monitoring a process variable that is an initial condition of a DBA or transient analysis.
3. The turbine condenser low vacuum scram instrumentation is not used as part of a primary success path in the mitigation of a DBA or transient.
4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 337) of NEDO-31466, Supplement 1, the loss of the turbine condenser low vacuum scram instrumentation was found to be a non-significant risk contributor to core damage frequency and offsite releases. Nuclear Management Company, LLC has reviewed this evaluation, considers it applicable to Monticello, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the portions of the LCO and Surveillances applicable to the Turbine Condenser Low Vacuum scram instrumentation may be relocated to other plant controlled documents outside the Technical Specifications.

Page 1 of 9 Attachment 1,Volume 1, Rev. 0, Page 21 of 29

Attachment 1, Volume 1, Rev. 0, Page 22 of 29 3/4.2.C.1 CONTROL ROD BLOCK ACTUATION LCO STATEMENT:

The limiting conditions for operation for the instrumentation that actuates control rod block are given in Table 3.2.3.

Table 3.2.3 Function 1, SRM

a. Upscale
b. Detector not fully inserted DISCUSSION:

SRM signals are used to monitor neutron flux during refueling, shutdown, and startup conditions. When IRMs are not above Range 2, the SRM control rod block functions to prevent a control rod withdrawal if the count rate exceeds a preset value or falls below a preset limit. No design basis accident (DBA) or transient analysis takes credit for rod block signals initiated by the SRMs.

COMPARISON TO SCREENING CRITERIA:

1. The SRM control rod block instrumentation is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
2. The SRM control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analysis.
3. The SRM control rod block instrumentation is not a part of a primary success path in the

- mitigation of a DBA or transient.

4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 137) of NEDO-31466, the loss of the SRM control rod block function was found to be a nonsignificant risk contributor to core damage frequency and offsite releases. Nuclear Management Company, LLC has reviewed this evaluation, considers it applicable to Monticello, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Control Rod Block Actuation LCO and Surveillances applicable to SRM instrumentation may be relocated to other plant controlled documents outside the Technical Specifications.

Page 2 of 9 Attachment 1, Volume 1, Rev. 0, Page 22 of 29

Attachment 1, Volume 1, Rev. 0, Page 23 of 29 3/4.2.C.1 CONTROL ROD BLOCK ACTUATION LCO STATEMENT:

The limiting conditions for operation for the instrumentation that actuates control rod block are given in Table 3.2.3.

Table 3.2.3 Function 2, IRM

a. Downscale
b. Upscale DISCUSSION:

IRMs are provided to monitor the neutron flux levels during refueling, shutdown, and startup conditions. The IRM control rod block functions to prevent a control rod withdrawal if the IRM reading exceeds a preset value, or if the IRM is inoperable. No design basis accident (DBA) or transient analysis takes credit for rod block signals initiated by IRMs.

COMPARISON TO SCREENING CRITERIA:

1. The IRM control rod block instrumentation is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
2. The IRM control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analysis.
3. The IRM control rod block instrumentation is not a part of a primary success path in the mitigation of a DBA or transient.
4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 138) of NEDO-31466, the loss of the IRM control rod block function was found to be a non-significant risk contributor to core damage frequency and offsite releases. Nuclear Management Company, LLC has reviewed this evaluation, considers it applicable to Monticello, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Control Rod Block Actuation LCO and Surveillances applicable to IRM instrumentation may be relocated to other plant controlled documents outside the Technical Specifications.

Page 3 of 9 Attachment 1, Volume 1, Rev. 0, Page 23 of 29

Attachment 1, Volume 1, Rev. 0, Page 24 of 29 -

3/4.2.C.1 CONTROL ROD BLOCK ACTUATION LCO STATEMENT:

The limiting conditions for operation for the instrumentation that actuates control rod block are given in Table 3.2.3.

Table 3.2.3 Function 3, APRM

a. Upscale (1) TLO Flow Biased (2) SLO Flow Biased (3) High Flow Clamp
b. Downscale DISCUSSION:

The APRM control rod block functions to prevent conditions that would require RPS action if allowed to proceed, such as during a "control rod withdrawal error at power." The APRMs utilize LPRM signals to create the APRM rod block signal and provide information about the average core power. However, the rod block function is not used to mitigate a design basis accident (DBA) or transient.

COMPARISON TO SCREENING CRITERIA:

1. The APRM control rod block instrumentation is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
2. The APRM control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analysis.
3. The APRM control rod block instrumentation is not a part of a primary success path in the mitigation of a DBA or transient.
4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 135) of NEDO-31466, the loss of the APRM control rod block function was found to be a non-significant risk contributor to core damage frequency and offsite releases. Nuclear Management Company, LLC has reviewed this evaluation, considers it applicable to Monticello, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Control Rod Block Actuation LCO and Surveillances applicable to APRM instrumentation may be relocated to other plant controlled documents outside the Technical Specifications.

Page 4 of 9 Attachment 1, Volume 1, Rev. 0, Page 24 of 29

Attachment 1, Volume 1, Rev. 0, Page 25 of 29 3/4.2.C.1 CONTROL ROD BLOCK ACTUATION LCO STATEMENT:

The limiting conditions for operation for the instrumentation that actuates control rod block are given in Table 3.2.3.

Table 3.2.3 Function 5, Scram Discharge Volume Water Level High

a. East
b. West DISCUSSION:

The Scram Discharge Volume (SDV) control rod block functions to prevent control rod withdrawals, utilizing SDV signals to create the rod block signal if water is accumulating in the SDV. The purpose of measuring the SDV water level is to ensure that there is sufficient volume remaining to contain the water discharged by the control rod drives during a scram, thus ensuring that the control rods will be able to insert fully. This rod block signal provides an indication to the operator that water is accumulating in the SDV and prevents further rod withdrawals. With continued water accumulation, a reactor protection system initiated scram signal will occur. Thus, the SDV water level rod block signal provides an opportunity for the operator to take action to avoid a subsequent scram. No design basis accident (DBA) or transient takes credit for rod block signals initiated by the SDV instrumentation.

COMPARISON TO SCREENING CRITERIA:

1. The SDV control rod block instrumentation is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
2. The SDV control rod block instrumentation is not used to monitor a process variable that is an initial condition of a DBA or transient analysis.
3. The SDV control rod block instrumentation is not a part of a primary success path in the mitigation of a DBA or transient.
4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 139) of NEDO-31466, the loss of the SDV control rod block function was found to be a nonsignificant risk contributor to core damage frequency and offsite releases. Nuclear Management Company, LLC has reviewed this evaluation, considers it applicable to Monticello, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the Control Rod Block Actuation LCO and Surveillances applicable to SDV instrumentation may be relocated to other plant controlled documents outside the Technical Specifications.

Page 5 of 9 Attachment 1, Volume 1, Rev. 0, Page 25 of 29

Attachment 1, Volume 1, Rev. 0, Page 26 of 29 4.5.A.4 ADS INHIBIT SWITCH SR STATEMENT:

ADS Inhibit Switch Operability Each Operating Cycle DISCUSSION CTS 4.5.A.4 requires the performance of an ADS Inhibit Switch Operability test. The ADS Inhibit Switch allows the operator to defeat ADS actuation as directed by the emergency operating procedures under conditions for which ADS would not be desirable. For example, during an ATWS event low pressure ECCS system activation would dilute sodium pentaborate injected by the Standby Liquid Control (SLC) System thereby reducing the effectiveness of the SLC System ability to shutdown the reactor. While 10 CFR 50.36(c)(2) criteria are not normally used for an individual Surveillance requirement, they are used in this case since the previous BWR Standard Technical Specifications included the ADS Manual Inhibit Switch as a separate Specification and the NRC evaluated it as such as documented in the NRC Staff Review of NSSS Vendor Owners Groups Application of the Commissions Interim Policy Criteria to Standard Technical Specifications, letter dated May 9, 1988. This SR does not meet the criteria for retention in the ITS; therefore, it will be retained in the Technical Requirements Manual (TRM).

COMPARISON TO THE SCREENING CRITERIA:

1. The ADS Inhibit Switch is not an instrument 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 ADS Inhibit Switch is not used for, nor capable of, monitoring a process variable that is an initial condition of a DBA or transient analysis.
3. The ADS Inhibit Switch is not used as part of a primary success path in the mitigation of a DBA or transient. The inhibit feature was added to allow defeating the automatic ADS function when such action is required by the Emergency Operating Procedures.

However, such manual operator action is not credited in a design basis accident or transient analysis.

4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 112B) of NEDO-31466, the loss of the ADS Inhibit switch was found to be a non-significant risk contributor to core damage frequency and offsite releases. Nuclear Management Company, LLC has reviewed this evaluation, considers it applicable to Monticello, and concurs with the assessment.

CONCLUSION:

Since the screening criteria have not been satisfied, the portions of the LCO and Surveillances applicable to the ADS Manual Inhibit switch may be relocated to other plant controlled documents outside the Technical Specifications.

Page 6 of 9 Attachment 1, Volume 1, Rev. 0, Page 26 of 29

Attachment 1, Volume 1, Rev. 0, Page 27 of 29 3/4.6.C.2 and 3/4.6.C.3 REACTOR COOLANT WATER CHEMISTRY LCO STATEMENT:

3/4.6.C.2. (a) The reactor coolant water shall not exceed the following limits with steaming rates less than 100,000 pounds per hour except as specified in 3.6.C.2.b.

Conductivity 5 pmho/cm Chloride ion 0.1 ppm 3/4.6.C.2. (b) For reactor startups the maximum value for conductivity shall not exceed p.mho/cm and the maximum value for chloride ion concentration shall not exceed

0. 1 ppm for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after placing the reactor in the power operating condition.

3/4.6.C.3.) Except as specified in 3. 6.C. 2.b above, the reactor coolant water shall not exceed the following limits with steaming rates greater than or equal to 100, 000 lbs. per hour.

Conductivity 5 gmho/cm Chloride ion 0.5 ppm DISCUSSION:

Poor coolant water chemistry contributes to the long term degradation of system materials of construction, and thus is not of immediate importance to the unit operator. Reactor coolant water chemistry is monitored for a variety of reasons. One reason is to reduce the possibility of failures in the Reactor Coolant System pressure boundary caused by corrosion. However, the chemistry monitoring activity is of a long term preventative purpose rather than mitigative.

COMPARISON TO SCREENING CRITERIA:

1. Reactor coolant water chemistry 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. Reactor coolant water chemistry is not a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient.
3. Reactor coolant water chemistry is not part of a primary success path in the mitigation of a DBA or transient.
4. As discussed in Sections 3.5 and 6, and summarized In Table 4-1 (item 211) of NEDO-31466, the reactor coolant water chemistry was found to be a non-significant risk contributor to core damage frequency and offsite releases. Nuclear Management Company, LLC has reviewed this evaluation, considers it applicable to Monticello, and concurs with the assessment.

CONCLUSION:

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

Page 7 of 9 Attachment 1, Volume 1, Rev. 0, Page 27 of 29

Attachment 1, Volume 1, Rev. 0, Page 28 of 29 3/4.14.F ACCIDENT MONITORING LCO STATEMENT:

Whenever irradiated fuel is in the reactor vessel and reactor coolant water temperature is greater than 2120 F, the limiting conditions for operation for accident monitoring instrumentation given in Table 3.14.1 shall be satisfied.

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.

COMPARISON TO SCREENING CRITERIA:

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 Monticello 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.

The following summarizes the Nuclear Management Company, LLC position for those instruments currently in Monticello Technical Specifications.

Tvye A Variables

1. Reactor Vessel Fuel Zone Water Level
2. Suppression Pool Temperature Other Type. Categorv I Variables
1. Drywell Wide Range Pressure
2. Suppression Pool Wide Range Level
3. Drywell High Range Radiation For other post-accident monitoring instrumentation currently in Technical Specifications, their loss is not risk-significant since the variables they monitor did not qualify as a Type A or Category 1 variable (one that is important to safety and needed by the operator, so that the operator can perform necessary normal actions).

Page8of9 Attachment 1, Volume 1, Rev. 0, Page 28 of 29

Attachment 1, Volume 1, Rev. 0, Page 29 of 29 CONCLUSION:

Since the screening criteria have not satisfied for non-Regulatory Guide 1.97 Type A or Category I variable instruments, their associated LCO and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications. The instruments to be relocated are as follows:

1. Safety/Relief Valve Position
2. Offgas Stack Wide Range Radiation
3. Reactor Bldg Vent Wide Range Radiation Page 9 of 9 Attachment 1, Volume 1, Rev. 0, Page 29 of 29
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