ML20079S091
| ML20079S091 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 09/29/1994 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20079S088 | List: |
| References | |
| NUDOCS 9410170156 | |
| Download: ML20079S091 (66) | |
Text
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. TSCR 93-16 o .
PECO ENERGY Peach Bottom Atomic Power Station APPLICATION OF SCREENING CRITERIA O
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= 12;a =s -,s-O APPLICATION OF SELECTION CRITERIA TO THE PEACH BOTTOM ATOMIC POWER STATION UNITS 2&3 TECHNICAL SPECIFICATIONS O
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- 1. INTRODUCTION , . . . . . . . . . . . . . . . . . . . . . . . . . . I
- 2. SELECTION CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . 2 ,
- 3. PROBABILISTIC RISK ASSESSMENT INSIGHTS . . . . . . . . . . . . . . 6
- 4. RESULTS OF APPLICATION OF SELECTION CRITERIA . . . . . . . . . . . 9
- 5. REFERENCES . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 f ATTACHMENT ,
SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 l APPENDICES A. JUSTIFICATION FOR SPECIFICATION RELOCATION B. PBAPS SPECIFIC RISK SIGNIFICANT EVALUATION O
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- 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 Peach Bottom Atomic Power Station (PBAPS) Units 2 & 3.
PECO Energy Company has reviewed the application of the . selection criteria to each of the Technical Specifications utilized in BWROG report NED0-31466,
" Technical Specification Screening Criteria Application and Risk Assessment,"
(Reference 1) including Supplement 1 (Reference 1), NUREG 1433, Standard Technical. Specifications, General Electric Plants BWR/4," and applied the criteria to each of the current PBAPS Units 2 & 3 Technical Specifications.
Additionally, in accordance with the NRC guidance, this confirmation of the '
application of selection criteria to PBAPS Unit 2 & 3 includes confirming the risk insights from Probabilistic Risk Assessment (PRA) evaluations, provided in the Reference 1, as applicable to PBAPS Unit 2 & 3. !
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- 2. SELECTION CRITERIA PECO Energy Company (PECO Energy) used the selection criteria provided in the NRC Final Policy Statement on Technical Specification Improvements of July 22, 1993 to develop the results contained in the attached matrix. Probabilistic Risk Assessment (PRA) insights as used in the BWROG submittal were used, confirmed by PECO Energy, and are discussed in the next section of this report. The selection criteria and discussion provided in the NRC Final Policy statement 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 o 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 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 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 N18.2) (or equivalent) that either assume the failure of or present a i challenge to the integrity of a fission product barrier. ,
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- 2. (continued)
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 I not, however, limited to only those directly monitored and i controlled from the control room. These could also include other i features or characteristics that are specifically assumed in '
Design Basis Accident or Transient analyses if they cannot be directly observed in the control room (e.g., moderator temperature I coefficient and hot channel factors). I I
The purpose of this criterion is to capture those process l variables that have initial values assumed in the Design Basis l Accident and Transient analyses, and which are monitored and l 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) 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 O 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 FSAR (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 O
acceptance criteria.
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- 2. (continued)
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's 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:
. Reactor Core Isolation Cooling / Isolation Condenser,
. 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 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 the 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 O :
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- 2. (continued) accident sequences that are commonly found to dominate risk. I Similarly, the NRC staff will also employ risk insights and PSAs in evaluating Technical Specifications related submittals.
Further, as a part of the Commissions 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. ,
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- 3. PROBABILISTIC RISK ASSESSMENT INSIGHTS Introduction and Objectives The Final Policy Statement includes a statement that NRC expects licensees to utilize the available literature on risk insights 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 for relocation to other plant controlled documents will'be maintained under the 10 CFR 50.59, safety evaluation review program. These specifications have been compared to a variety of Probabilistic Risk Assessment (PRA) material with two purposes: 1) to identify if a component or variable is addressed by PRA, and 2) to judge if the component or variable is risk-important. In addition, in some cases risk was judged independent of any specific PRA material. The intent of the review was to provide a supplemental 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 PECO Energy for those Specifications to be relocated. Where Reference 1 did not review a Technical Specification against the criteria of Reference 3, PECO Energy performed a review similar (but not identical) to that described below for Reference 1. The results of these reviews are presented in Appendix B.
O Assumotions 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 I analysis steps were performed for each LCO proposed for relocation:
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- 3. (continued)
O a. List the function (s) affected by removal of the LC0 item.
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- b. Determine the effect of loss of the LC0 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 LC0 item is removed from Technical Specifications and controlled by other procedures or i programs, Use information from current PRAs and related analyses to establish the relative frequency.
- 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 sig7ificance. .
- f. Apply risk category criteria to establish the potential risk significance or non-significance of the LC0 item. Risk categories f were defined as follows:
RISK CRITERIA i Consequence i Frecuency Hiah Medium Low ,
High S S NS Medium S S NS l Low NS NS NS S = Potential Significant Risk Contributor NS - Risk Non-Significant
- g. 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.
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TABLE 3-1 BWR PRAs USED IN NED0-31466 (and Supplement 1)
RISK ASSESSMENT
. BWR/6 Standard Plant, GESSAR II, 238 Nuclear Island, BWR/6 Standard Plant Probabilistic Risk Assessment, Docket No. STN 50-447, March 1982.
. La Salle County Station, NED0-31085, Probabilistic Safety Analysis, February 1988.
. Grand Gulf Nuclear Station, IDCOR, Technical Report 86.2GG, Verification of IPE for Grand Gulf, March 1987.
. Limerick, Docket Nos. 50-352, 50-353, 1981, "Probabilistic Risk Assessment, Limerick Generating Station," Philadelphia Electric Company.
. Shoreham, Probabilistic Risk Assessment Shoreham Nuclear Power Station, Long Island Lighting Company, SAI-372-83-PA-01, June 24, 1983.
. Peach Bottom 2, NUREG-75/0104, " Reactor Safety Study," WASH-1400, October 1975.
. Millstone Point 1, NUREG/CR-3085, " Interim Reliability Evaluation l Program: Analysis of the Millstone Point Unit 1 Nuclear Power l Plant," January 1983.
. Grand Gulf, NUREG/CR-1659, " 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. l l
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- 4. RESULTS OF APPLICATION OF SELECTION CRITERIA The selection criteria from Section 2 were applied to the PBAPS Units 2 and 3 ,
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. PECO Energy will relocate those Specifications identified as not satisfying the criteria to licensee controlled documents whose changes are governed by 10 CFR 50.59.
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- 5. REFERENCES
- 1. NED0-31466 (and Supplement I), " Technical Specification Screening Criteria Application and Risk Assessment," November 1987.
- 2. NUREG 1433, " Standard Technical Specifications, General Electric Plants BWR/4," September 1992.
- 3. Final Policy Statement on Technical Specifications Improvements' July 22, 1993, (58FR39132)
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ATTACHMENT j
SUMMARY
DISPOSITION MATRIX !
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SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERlON UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUS!ON/ EXCLUSIONI *'
1.0 Definitions 1.1 1.0 1.1 Yes Definitions for selected terms used in the Technical Specifications are provided to improve understanding snd ensure consistent application. Application of the Technical Specification selection criteria to these definitions is not appropriate. However, definitions for those terms that remain in the Technical Specifications following the application of the selection criteria will be retained .
SAFETY LIMITS 1/2.1. A Safety Limits: 2.1 2.1.1 2.1 Yes Application of Technical Specification selection criteria to 1.1.8 Fuel Cladding integrity 3.3.1.1 2.1.2 3.3.1.1 Safety Limits and Limiting Safety System Settings (LSSS)is 1/2.1.C 3.3.S.1 2.1.4 3.3.5.1 not appropriate. The fuel cladding integrity LSSS (with the 2.1.D 3.3.5.2 2.2.1 3.3.5.2 exception of APRM Rod Blocks) are retained by their 2.1.E 3.3.6.1 3.3.2 3.3.6.1 incorporation into the RPS and ECCS instrumentation 2.1.F 3.3.3 Specificanons because the associated Functions either 2.1.G 3.3.S actuate to mitigate consequences of Design Basis Accidents 2.1.H (DBAs) and transients or are retained as directed by the NRC.
2.1.1 2.1.J 2.1.K 2.1.B Safety Limit: Relocated None None No See Appendix A, page 5.
Fuel Cladding Integrity - APRM Rod Block Trip Setting 1/2.2 Safety Limit: 2.1.2 2.1.3 2.1.2 Yes Application of Technical Specification selection criteria to Reactor Coolant System integrity 3.3.1.1 2.2 3.3.1.1 Safety Limits and Limiting Safety System Settings (LSSS)is 3.4.3 3.4.2.1 3.4.3 not appropriate. The Reactor Coolant System integnty LSSS 3.3.6.1 3.3.6.1 are retained by incorporation into the RPS and safety relief valve and safety valve Specifications because the associated components function to mitigate the consequences of events that would result in overpressurization of the RCS.
(a) The applicable safety analyses are discussed in the Bases for the individual Technical Specifications.
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SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUSION / EXCLUSION'*I 3.0.A Limiting Conditions For Operation: 3.0.1 3.0.1 3.0.1 Yes These Specifications provide generic guidance applicable to 3.0.B Applicability 3.0.2 3.0.2 3.0.2 one or more Specifications to facilitate understr.nding of 3.0.C 3.0.3 3.0.3 3.0.3 LCOs. Direct application of the Technica! Specification 3.0.D 3.8.1 3/4.8.1.1 3.8.1 selection criteria is not appropriate. The general requirements of 3.0 are retained in the Technical Specifications consistent with NUREG-1433.
RE ACTOR PROTECTION SYSTEM 3 /4.1. A Reactor Protection System: 3.3.1.1 3/4.3.1 3.3.1.1 Yes - 3, 4 All Functions retained because the various Functions: 1) 3.1.B Instrumentation that trutiste a Reactor actuate to mitigate consequences of DBAs and/or transients; 4.1.C Scram (Instruments in Table 3.1.1 and or, 21 are considered risk significant and retained in associated SRs in Table 4.1.1 and 4.1.2) accordance with the NRC Final Policy Statement on Technic W Specification Improvements; or,3) are part of the RPS/ Reactor Scram Function; or,4) provide an anticipatory scram to ensure the scram discharge volume and thus RPS remains operable.
3/4.1.D Reactor Protection System Power Supply 3.3.8.2 3/4.8.4.4 3.3.8.2 Yes - 3 Provides protection for the RPS bus powered instrumentation against unacceptable voltage and frequency conditions that could degrade instrumentation so that it would not perform the intended safety function.
PROTECTIVE INSTRUMENTATION 3/4.2.A Primary Containment Isolation: 3.3.6.1 3/4.3.2 3.3.6.1 Yes - 3, 4 All Functions retained because the Functions actuate to Instrumentation that initiates primary mitigate the consequences of a DBA LOCA or are considered containment isolation. (Instruments in risk significant and are retained in accordance with the NRC Tab!e 3.2.A and associated SRs in Table Final Policy Statement on Technical Specification 4.2.A) Improvements.
(a) The applicable safety analyses are discussed in the Bases for the individual Technical Specifications.
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SUMMARY
DISPOSITI0ft MATRIX FOR. PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS HbMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUStON!EXCLUSIONM PROTECTIVE INSTRUMENTATION (cont'd) 3/4.2.B Core and Containment Cooling Systems - 3.3.5.1 3/4.3.3 3.3.5.1 Yes - 3, 4 Functions retained (with exceptions listed below) because the Initiation & Control; Instrumentation that 3.3.5.2 3/4.3.5 3.3.5.2 variot.s Functions actuate to rnitigate the consequences of a initiates or controls the core and 3.3.6.1 3 /4.8.1.1 3.3.6.1 DBA LOCA or are considered risk significant and are retained containment cooling systems (LPCI. CS 3.3.8.1 3.3.S.1 in accordance with the NRC Final Policy Statement on ADS, HPCI and RCIC). (Instruments in Technical Specification improvements.
Table 3.2.B and associated SRs in Table 4.2.B) (Exceptions listed below.)
Table Reactor Low Pressure (50 s P s 75 psig) Relocated None None No See Primary Containment isolation Instrumentation technical 3/4.2.B change discussion (Rt for ITS 3.3.6.1).
Table Trip System Bus Power Monitors: Relocated None None No See Appendix A. Page 1 3/4.2.B RHR (LPCI), CS, ADS.HPCI and RCIC Trip Systems Table Core Spray Sparger to Reactor Pressure Relocated None None No See Appendix A. Page 3.
3/4.2.B Vessel d/p Table LPCI Cross-Connect Position Re!ocated hone None No See Appendix A. Page 4.
3/4.2.B 3/4.2.C Control Rod Block Actuation: 3.3.2.1 3/4.3.6 3.3.2.1 Yes Control Rod Block Actuation Instrumen'ation functions to Instrumentation that initiates Control Rod prevent violation of the MCPR Safety Linst and cladding Blocks. (Instruments in Table 3.2.C and plastic strain design limit during a single control rod associated SRs in Table 4.2.C) withdrawal error event, ensures the initial conditions of the (Exceptions listed below.) control rod drop accident analysis are not violated, and prevents inadvertent criticality when the reactor is shutdown (thereby preserving the safety analysis assunmtions).
Table APRM (Upscale Flow Biased. Upscale Relocated 3/4.3.6 None No See Appendix A, Page 5.
3 /4.2.C Startup Mode, Downscale)
Table IRM (Downscale. Detector not in Startup Relocated 3/4.3.6 None No See Appendix A, Page 7.
3/4.2.C Position, and Upscale)
Table SRM (Detector not in Startup Position. Relocated 3/4.3.6 None No See Appendix A, Page 8.
3/4.2.C Upscale)
(a) The applicable safety analyses are discussed in the Bases for the individual Technical Specifications.
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SUMMARY
DISPOSITI0f1 MATRIX FOR PBAPS UNITS 2 AtlD 3 i
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RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUSION / EXCLUSION N PROTECTIVE INSTRUMENTATION (cont'd)
Table Scram Discharge Instrument Volume High Relocated 3/4.3.6 None No See Appendix A. Page 9.
3/4.2.C Level 3/4.2.D Radiation Monitoring Systems - Isolation 3.3.6.1 3/4.3.2 3.3.6.1 Yes - 3 Functions actuate to rrutigate the consequences of a DBA and initiation Functions: 3.3.6.2 3/4.3.7 3.3.6.2 LOCA and Fuel Handling Accident. The isolation signals instrumentation that initiates Reactor 3.3.7.1 3.3.7.1 generated by the reactor building isolation instrumentation Building isolation, Standby Gas Treatment are implicitly assumed in the safety analyses to initiate System and Main Control Room closure of valves to limit offsite doses.
Emergency Venthtion. (Instruments in Table 3.2.0 and essociated SRs in Table 4.2.D.)
3/4.2.E Drywell Leak Detection: 3.4.S 3/4.4.3.1 3.4.6 Yes - 1 Leak detection instrumentation is used to indicate an instrumentation that morntors drywell abnormal condition of the reactor coolant pressure boundary.
leakage. (Instruments in Section 3.6.C and Table 4.2.E) 3/4.2.F Surveillance information Readouts (Post 3.3.3.1 3/4.3.7.S 3.3.3.1 Yes - 3 Regulatory Guide 1.97 Type A and Category 1 instruments Accident Monitoring Instruments): reta;ned. See Appendix A, Page 11 for full discussion of att Instrumentation that provides surveillance instruments in Table 3.2.F.
information. (Instruments in Table 3.2.F and associated SRs in Table 4.2.F.)
3/4.2.G Recirculation Pump Trip: Instrumentation 3.3.4.1 3/4.3.4.1 3.3.4.2 Yes - 4 ATWS-RPT is being retained in accordance with the NRC that trips the reactor recirculation pump to Final Pohey Statement on Technical Specification limit the consequences of a failure to improvements due to risk significance.
scram ( ATWS-RPT). (Instruments in Table 3.2.G and associated SRs in Table 4.2.GJ (Exceotions listed below) 3/4.2.G Alternate Rod insertion: Relocated None None No See Appendix A, Page 13.
Instrumentation that initiates an Alternate Rod insertion Scram to limit the consequences of a failure to scram.
(a) The apphcable safety analyses are discussed in the Bases for the individual Technical Specifications.
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SUMMARY
DISPOSITI0ff MATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER HUMBER INCLUSION BASIS FOR INCLUSION!EXCLUSIONW REACTIVITY CONTROL 3!4.3.A.1 Reactivity Margin - Core loading 3.1.1 3 /4.1.1 3.1.1 Yes - 2 Shutdown Margin (SDM) is assumed as an initial condition for the control rod removal error during a refueling event and the fuel assembly insertion error during a refueling event.
3/4.3.A.2. Reactivity Margin: 3.1.3 3 /4.1.3.1 3.i.3 Yes - 3 Control rods are part of the primary success path for inoperable Control Rods 3.1.5 3/4.1.3.5 3.1.5 mitigating the consequences of DBAs and transents.
4.3. A.2.b Reactivity Margin: 3.1.8 3 /4.1.3.1 3.1.8 Yes - 3 The capability to insert the control rods ensures the 4.3. A.2.c (Scram Discharge Volume) assumptions used for the scram reactivity in the DBA and transient analyses are maintained. The Scram Discharge Volume (SDV) vent and drain valves contribute to the operability of the control rod scram function.
3/4.3.B.1 Control Rod Drive Coupling 3.1.3 3/4.1.3.6 3.1.3 Yes - 3 Control rods are part of the primary success path for mitigating the consequences of DBAs and transients. Control rod drive coupling is a requirement that forms a part of the demonstration of control rod operability.
3/4.3.B.2 CRD Housing Support Relocated 3/4.1.3.8 None No See Control Rod Operability technical change discussion (R2 for ITS 3.1.3).
. 3/4.3.B.3 Rod Worth M nimizer (RWM) 3.3.2.1 3/4.1.4.1 3.3.2.1 Yes - 3 The RWM enforces the Banked Position Withdrawal Sequence (BPWS) to ensure that the initial conditions of the CRDA analysis are not violated.
3/4.3.B.4 Menimum SRM Count Rate for Control Rod 3.3.1.2 3/4.3.7.6 3.3.1.2 Yes Does not satisfy selection criteria. however is being retained
- Withdrawal because it is considered necessary for flux monitoring during shutdown. startup and refueling operations.
3/4.3.B.5 Operation with a Limiting Control Rod 3.3.2.1 3/4.1.4.3 3.3.2.1 Yes - 3 Prevents deviations from a banked position withdrawal Pattern sequence that if violated could allow high rod worth conditions that would challenge the MCPR Safety Limit and the cladding plastic strain design limit during a rod withdrawat error event.
(a) The applicable safety analyses are discussed in the Bases for the individual Technical Specifications.
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SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUSION /EXCLUSiONM REACTIVITY CONTROL (cont *d) 3/4.3.C Scram insertion Times 3.1.3 3/4.1.3.2 3.1.3 Yes - 3 Control rods are part of the primary success path for 3.1.4 3/4.1.3.3 3.1.4 mitigating the consequences of DBAs and transients. The '
3/4.1.3.4 DBA and transient analyses assume that control rods scram at a specified insertion rate.
3/4.3.D Reactivity Anomaties 3.1.2 3/4.1.2 3.1.2 Yes - 2 Not a measured process variable, but is important parameter that is used to confirm the acceptabihty of the accident analysis.
3/4.4 Standby Liquid Control System 3.1.7 3/4.1.5 3.1.7 Yes - 4 The Standby Liquid Control (SLC) is a backup system to the control rod scram function. This system is being retained per the NRC Final Policy Statement on Technical Specification improvements due to the risk significance.
CORE AND CONTAINMENT COOLING SYSTEMS 3/4.5.A Core Spray and LPCI Subsystems 3.5.1 3/4.5.1 3.5.1 Yes - 3 Core Spray and Low Pressure Coolant injection subsystems are part of the ECCS and function to provide cooling water to the reactor core to mitigate large Loss of Coolant Accidents.
3/4.5.B Containment Cooling System 3.6.2.3 3/4.6.2.2 3.6.2.3 Yes - 3 This system provides a reliable source of cooling water and 3.6.2.4 3/4.6.2.3 3.6.2.4 functions to provide cooling to the pnmary containment under 3.7.1 3/4.7.1.1 3.7.1 post accident conditions.
3/4.5.C HPCI Subsystem 3.5<1 3/4.5.1 3.5.1 Yes - 3 The HPCI System is part of the ECCS and functions to mitigate small break Loss of Coolant Accidents.
3/4.5.D Reactor Core Isolation Cooling (RCICI 3.5.3 3/4.7.4 3.5.3 Yes - 4 System retained in accordance with the NRC Final Policy System Statement on Technical Specification improvements due to risk significance.
l l (a) The apphcable safety analyses are discussed in the Bases for the individual Technical Specifications.
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SUMMARY
DISPOSITION 11 ATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERlON UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUSION / EXCLUSION M CORE AND CONTAINMENT COOLIP',
SYSTEMS (cont'd) 3/4.5.E Automatic Depressurization ( ADS) System 3.5.1 3 /4.5.1 3.5.1 Yes - 3 The ADS is part of the ECCS and is designed to mitigate a small or medium break Loss of Coolant Accident. The ADS acts to rapidly reduce reactor vessel pressure in a LOCA situation in which the HPCI System fails to automatically maintain reactor vessel water level. This depressurization enables the low-pressure emergency core cooling systems to deliver cooling water to the reactor core.
3/4.5.F Minimum Low Pressure Cooling Availability 3.5.2 3/4.5.2 3.5.2 Yes - 3 The low pressure ECCS requ>ements ensure systems are available to mitigate the consequeness of a vessel draindown event.
3/4.5.G Maintenance of Filled Discharge Pipe 3.5.1 3/4.5.1 3.5.1 Yes - 3. 4 This Specification ensures the operability of the ECCS and 3.5.2 3/4.5.2 3.5.2 RCIC Systems which function to mitigate the consequences 3.5.3 3/4.7.4 3.5.3 of a LOCA (ECCS) or is required to be retained by the NRC Final Policy Statement on Technical Specification Improvements (RCIC).
3/4.5.H Engineered Safeguards Compartrrants Relocated None None No See ECCS - Operating and RCIC technical change Cooling and Ventitation discussions (R5 for ITS 3.5.1 and R4 for ITS 3.5.3).
3/4.5.1 Average Planar Linear Heat Generation 3.2.1 3/4.2.1 3.2.1 Yes - 2 The APLHGR limit is an initial condition in the safety Rate (APLHGR) analyses.
3/4.5.J Local LHGR 3.2.3 3/4.2.4 3.2.3 Yes - 2 The LHGR limit is an initial condition in the safety analyses.
3/4.5.K Minimum Critical Power Ratio (MCPR) 3.2.2 3/4.2.3 3.2.2 Yes - 2 The MCPR limit is an initial condition in the safety analyses.
PRIMARY SYSTEM BOUNDARY 3/4.6.A Thermal and Pressurization Limitations 3.4.9 3/4.4.6.1 3.4.10 Yes - 2 Establishes initial conditions such that operation is prohibited 3.10.1 3.10.1 in areas or at temperature rate changes that might cause undetected flaws to propagste in turn challenging the reactor coolant pressure boundary integrity.
(a) The apphcable safety analyses are discussed in the Bases for the individual Technical Specifications.
7 Revision O
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SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 a
RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUSION! EXCLUSIONI *'
PRIMARY SYSTEM BOUNDARY (cont'd) 3/4.6.B.1 Coolant Chemistry (Coolant Activity 3.4.6 3/4.4.5 3.4.7 Yes - 2 The specif ic activity in the reactor coolant is an irutial Limits) condificn for evaluation of the consequence of an accident due to a main steam line break (MSLB) outside containment.
3/4.6.B.2 Coolant Chemistry (Reactor Water Quality) Relocated 3/4.4.4 None No See Appendix A. Page 14.
3.6.B.3 314.6.C Coolant Leakage 3.4.4 3/4.4.3.1 3.4.4 Yes - 1. 2 Leakage beyond limits would indicate an abnormal condition (Leakage Limits) 3.4.5 3/4.4.3.2 3.4.6 of the reactor coolant pressure boundary. Operation in this (Leakage Detection Instruments) condition may result in reactor coolant pressure boundary failure. Leakage detection instruments are used to indicate an abnormal condition of the reactor coolant pressure boundary.
3/4.6.D Safety and Relief Valves 3.4.3 3/4.4.2.1 3.4.3 Yes - 3 The Safety and Relief Valves are assumed to operate to maintain the reactor pressure below design limits.
3/4.6.E Jet Pumps 3.4.1 3/4.4.1.2 3.4.1 Yes - 2 Jet Pump operability is explicitly assumed in the design basis 3.4.2 3.4.2 LOCA to assure adequate core reflood capabihty.
3/4.6.F Recirculation Pumps (includes 3.4.1 3/4.4.1.1 3.4.1 Yes - 2 Recirculation loop flow is an initial condition in the safety requirements fer single loop operation) 3/4.4.1.3 analysis.
3/4.6.G Structural Integrity None 3/4.4.8 None No See Appendix A, Page 15.
CONTAINMENT SYSTEMS 3 /4.7. A.1 Primary Conteinment 3.6.2.1 3/4.6.2.1 3.6.2.1 Yes - 2. 3 The suppression pool water volume and temperature are 3.7 A.7 (Suppression Chamber ) 3.6.2.2 3.6.2.2 initial conditions in the DBA LOCA containment response analysis and mitigate the consequences of a DBA.
3 / 4.7. A . 2 Primary Containment 3.6.1.1 3/4.6.1.1 3.6.1.1 Yes - 3 Primary containment functions to rnatigate the consequences 3.7.A.7 (Primary Contamment Integnty) 3.6.1.2 3/4.6.1.2 3.6.1.2 of a DBA. Primary containment feakage is an assumption (Integrated Leak Rate Testing) 3.6.1.3 3/4.6.1.3 3.6.1.3 utilized in the LOCA safety analysis to ensure primary 3/4.6.1.5 containment operability.
(a) The applicable safety analyses are discussed in the Bases for the individual Technical Specifications.
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SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION Basis FOR INCLUSION / EXCLUSION'*I CONTAINMENT SYSTEMS (cont'd) 3!4.7.A.3 Pressure Suppression Chamber--Reactor 3.6.1.5 3/4.6.4.2 3.6.1.7 Yes - 3 Pressure suppression chamber-reactor building vacuum 3.7.A.7 Building Vacuum Breakers breaker operation is relied on to limit a negative pressure differential, secondary to primary containment, that could challenge primary containment entegrity.
3/4.7.A.4 Dryw ell Pressure Suppression Chamber 3.6.1.6 3/4.6.4.1 3.6.1.8 Yes - 3 Drywell-pressure suppression chamber vacuum breaker 3.7.A.7 Vacunm Breakers operation is assumed in the LOCA analysis to limit drywell pressure thereby ensuring primary containment integnty.
3/4.7.A.5 Primary Containment 3.6.3.2 3/4.6.6.4 3.6.3.3 Yes - 2 Oxygen concentration is limited such that, when combined 3.7 A.7 IOxygen Concentration) with hydrogen (that is postulated to evolve following a LOCA), the total explosive gas concentration remains below explosive levels. Therefore, primary containment integrity is maintained.
3/4.7.A.6 Primary Containment 3.6.3.1 3/4.6.6.2 3.6.3.4 Yes - 3 System ensures oxygen concentration is maintained below (Containment Atmospheric Dilution) 3.3.3.1 3.3.3.1 the explosive level following a LOCA by inerting the drywell (Post Accident Monitoring instrumentation) with nitrogen. Therefore, pnmary containment integrity is maintained.
3/4.7.B Standby Gas Treatment (SGT1 System 3.6.4.3 3/4.6.5.3 3.6.4.3 Yes - 3 System functions following a DBA to limit offsite releases.
3/4.7.C Secondary Containment 3.6.4.1 3/4.6.5.1 3.6.4.1 Yes - 3 Secondary containment integrity is relied on the 16mit the 3.6.4.2 3/4.6.5.2 3.6.4.2 offsite dose during an accident by ensunng a release to containment is delayed and treated prior to release to the environment. Damper operation within time limits establishes secondary containment and limits offsite releases to acceptable values.
3/4.7.D Pnmary Containment Isolation Valves 3.6.1.3 3/4.6.3 3.6.1.3 Yes 3 Isolation valves function to limit DBA consequences.
3/4.7.E Large Primary Containment Purge / Vent 3.6.1.3 3/4.6.1.8 3.6.1.3 Yes - 3 Isolation velves function to limit DBA consequences.
Isolation Valves (a) The applicable safety analyses are d:scussed in the Bases for the individual Technical Specifications.
9 Revision O
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SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUSION;' EXCLUSIONt )
RADIOACTIVE MATERIALS 3/4.8.A General Relocated None None No See Appendix A, Page 16.
3/4.8.B Liquid Radwaste Effluents Relocated None None No See Appendix A, Page 17, 3/4.8.C.1 Gaseous Effluents Relocated None None No See Appendix A. Page 20.
3/4.8.C.2 3/4.8.C.3 3/4.8.C.4 3/4.8.C.5 3/4.8.C.6 Gaseous Effluents Relocated None None tJo See Appendix A, Page 24.
(Explosive Ges Mixture) 3 /4.8.C.7.a Gaseous Effluents 3.7.5 None 3.7.6 Yes - 2 Main condenser offgas activity is an irutial condition in the (Main Condenser Offgas) offgas system failure event.
3/4.8.C.7.b Gaseous Effluents Relocated None None No See Appendix A. Page 20.
(Steam Jet Air Ejector Radiation Monitor) 3.8.C.8 Gaseous Effluents Relocated None None No See Appendix A. Page 20.
(Primary Containment Purging) 3/4.8.D 40 CFR 190 Relocated None None No See Appendix A. Page 25.
3/4.8.E Radiological Environmental Monitoring Relocated None None No See Appendix A, Page 26.
3/4.8.F Solid Radioactive Weste Relocated None None No See Appendix A, Page 28.
3/4.8.G Mechanical Vacuum Pump Relocated None None No See Mechanical Vacuum Pump technical change discussion (Ri for CTS 3/4.8L (a) The applicable safety analyses are discussed in the Bases for the individust Technical Specifications.
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SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUSION / EXCLUSION N AUXfLIARY ELECTRICAL SYSTEMS 3/4.9.A.1 Aumliary Electrical Equipment 3.8.1 3/4.8.1.1 3.8.1 Yes - 3 The operability of the AC power sources are part of the 3.9.A.2 (Offsite and Onsite AC Sources) 3.8.3 3.8.3 primary success path of the accident analyses.
3/4.9.B (Diesel Fuel Oil)
Operation with inoperable Equipment 3.9 A.3 Auxiliary Electrical Equipment 3.8.7 3 /4.8.3.1 3.8.9 Yes - 3 The operability of the distribution system is part of the 3.9.B.7 (Distribution) pnmary success path of the accident analyses.
4.9.A.3 (Swing Buses) 3.5.1 None None 3.9.A.4 Auxiliary Electrical Equipment 3.8.4 3/4.8.2.1 3.8.4 Yes - 3 The operability of the DC subsystems is consistent with the 3.9.B.5 (Batteries) 3.8.6 3.8.6 initial assumptions of the accident analyses.
4.9.A.2 3/4.9.C Emergency Service Water System 3.7.2 3/4.7.1.2 3.7.2 Yes - 3 The ability of this system to provide adequate cooling to the identified safety equipment is an implicit assumption for the safety analyses.
CORE ALTERATIONS 3 /4.10. A.1 Refueling interlocks 3.9.1 3/4.9.1 3.9.1 Yes - 3 The refueling interlocks protect against prompt reactivity 3 /4.10. A.2 3.9.2 3.9.2 excursions during the Refuel Mode. The safety analyses for 3.10. A.3 3.9.3 3.9.3 the control rod removal error during refueling and the fuel 3.10 A.4 assembly insertion error during refueling assume the functioning of the refueling interlocks.
3.10. A.5 (Withdrawal of two nonadjacent Control 3.10.5 3/4.9.10.2 3.10.5 Yes This requirement is being retained to si!ow relaxation of Rods for maintenance ) certain Limiting Coretions for Operation (LCOs) under specific conditions to allow testing and maintenance. This
, requirement is directly related to seveial LCOs. Direct
- application of the Technical Specification selection enteria is not appropriate. However, this requirement, directly tied to i LCOs that remain in Technical Specifications, will also remain i in Technical Specifications.
l (a) The applicable safety analyses are discussed in the Bases for the individual Technical Specifications.
11 Revision O
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SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITERION UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUSION! EXCLUSION
CORE ALTER ATIONS (cont'd; 3.10. A.6 Refueling Interlocks (Multiple Control Rod 3.10.6 3/4.9.10.2 3.10.6 Yes This requirement is being retained to allow relaxation of Withdrawal - Refueling) , certain Limiting Conditions for Operation (LCOs) under specific conditions to a!!ow testing and maintenance. This requirement is derectly related to several LCOs. Direct application of the Technical Specification selection criteria is not appropriate. However, this requirement, directly tied to LCOs that remain in Technical Specifications, will also remain in Technical Specifications.
3/4.10.8 Core Monitoring (Core Alterations) 3.3.1.2 3/4.9.2 3.3.1.2 Yes Does not satisfy criteria for inctusion but is retained because it is considered necessary for flux monitoring during shutdown, startup, and refueling operations.
3 /4.10.C Spent Fue! Storage Pool Water Level 3.7.7 3/4.9.9 3.7.8 Yes - 2 A minimum amount of water is required to assure adequate scrubbing of fission products following a fuel handling acci-dent.
3.10.D Heavy Loads Over Spent Fuel Relocated 3/4.9.7 None No See Appendix A. Page 29.
ADDITIONAL SAFETY RELATED PLANT CAPABILITIES 3 /4.1 1. A .1 Main Control Room Emergency Ventilation 3.7.4 3/4.7.2 3.7.4 Yes - 3 Maintains habitability of the control room so that operators 3 /4.1 1. A . 2 System can remain in the control room following an accident. As 3 /4.1 1. A .3 such, it mitigates the consequences of an accident by 3.11. A.4 allowing the operators to continue accident mitigation 3 /4.1 1. A.6 activities from the contro! room.
3.1 1. A.7 3 /4.1 1. A .5 Main Control Room Emergency Ventilation 3.3.7.1 3/4.3.7.1 3.3.7.1 Yes - 3 The ability of the radiation monitors to isolate the main 4.1 1. A .4 System control room ensures the control room remains habitable 3.11. A.7 (Radiation Monitors) following an accident.
(a) The applicable safety analyses are discussed in the Bases for the individus! Technical Specifications.
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SUMMARY
DISPOSITION MATRIX FOR PBAPS UNITS 2 AND 3 RETAINED:
CURRENT NEW CRITER!ON UNIT 2/3 TS UNIT 2/3 STS REV. 4 NUREG 1433 FOR NUMBER TITLE TS NUMBER NUMBER NUMBER INCLUSION BASIS FOR INCLUSION / EXCLUSIONI 'l ADDITIONAL SAFETY RELATED PLANT CAPABILITIES (cont'd) 3/4.11.B Emergency Heat Sink Facility 3.7.3 3 /4.7.1.3 3.7.2 Yes - 3 The seismic Class i emergency heat sink facility provides adequate cooling to support the shutdown of both PBAPS Units 2 and 3 in the event of a loss of the non-seismic normal heat sink as presented in the UFSAR.
3/4.11.C Emer0ency Shutdown Control Panel 3.3.3.2 3/4.3.7.4 3.3.3.2 Yes - 4 The Remote Shutdown System is considered an important (Remote Shutdown System) contnbutor to reducing risk. As such,it has been retained in accordance with the NRC Final Policy Statement on Technical Specification incrovements due to risk significance.
3/4.11.0 Shock Suppressors (Snubbers) on Safety Relocated 3/4.7.5 None No See Shock Suppressor technical change discussion (R for i Related Systems CTS 3/4.11.D).
3/4.12 River Water Level Relocated 3/4.7.3 None No See Appendix A. Page 30.
3/4.13 Miscellaneous Radioactive Materials Relocated 3/4.7.6 None No See Appendix A. Page '**.
Sources 3/4.14 Removed in Amendments 194/198 N/A N/A N/A N/A N/A, 3/4.15 Seismic Monitoring instrumentation Relocated 3/4.3.7.2 None No See Appendix A. Page 33.
5.0 Major Design Features 4.0 5.0 4.0 Yes Application of Technical Specification selection criteria is not appropriate. However, Design Features will be included in Technical Specifiestions as required by 10 CFR 50.36.
6.0 Administrative Controls 5.0 6.0 5.0 Yes Application of Technical Specification selection criteria is not appropriate. However. Administrative Controls will be included in Technical Specifications as required by 10 CFR 50.36.
(a) The applicable safety analyses are discussed in the Bases for the individual Technical Specifications.
13 Revision 0
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TABLE 3/4.2.B TRIP SYSTEM BUS POWER MONITORS FOR THE RHR (LPCI), CORE SPRAY, ADS, HPCI AND RCIC TRIP SYSTEMS LC0 Statement:
The limiting conditions for operation for the instrumentation that initiates or controls the core and containment cooling systems are given in Table 3.2.B.
Table 3.2.8 Instrumentation that Initiates or Controls the Core and Containment Coolina Systems 3/4.2.B.18 RHR (LPCI) Trip System Bus Power Monitor 3/4.2.B.19 Core Spray Trip System Bus Power Monitor 3/4.2.B.20 ADS Trip System Bus Power Monitor 3/4.2.B.21 HPCI Trip System Bus Power Monitor 3/4.2.B.22 RCIC Trip System Bus Power Monitor Discussion:
The Trip System Bus Power Monitors for the RHR (LPCI), Core Spray, ADS, HPCI and RCIC trip systems alarm if a fault is detected in the power system to the appropriate systems logic. No design basis accident (DBA) or transient analyses takes credit for the Trip System Bus Power Monitors. This instrumentation provides a monitoring / alarm function only.
q Comparison to Screenino Criteria:
b 1. The Trip System Bus Power Monitors are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The Trip System Bus Power Monitors are not process variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. The Trip System Bus Power Monitors are not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. ,
As discussed in Sections 3.5 and 6 of NED0-31466 and summarized in Table 4-1 (item 106) of NED0-31466, Supplement 1, and verified by PECO Energy, the loss of the RHR (LPCI), Core Spray, ADS, HPCI and RCIC Trip System Bus Power Monitors was found to be a non-significant risk contributor to core damage frequency and offsite releases.
l O l PBAPS UNITS 2 & 3 1 of 33 Revision 0 l
TABLE 3/4.2.8 TRIP SYSTEM BUS POWER MONITORS FOR THE RHR (LPCI), CORE (cont'd.) SPRAY, ADS, HPCI AND RCIC TRIP SYSTEMS.
Conclusion:
Since the screening criteria have not been satisfied, the RHR (LPCI), Core Spray, ADS, HPCI and RCIC Trip System Bus Power Monitors LC0 and Surveillances may be relocated to a licensee controlled document.
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PBAPS UNITS 2 & 3 2 of 33 Revision 0 l
TABLE 3/4.2.8 CORE SPRAY SPARGER TO REACTOR PRESSURE VESSEL d/p O LCO Statement:
The limiting conditions for operation for the instrumentation that initiates or controls the core and containment cooling systems are given in Table 3.2.B.
Table 3.2.8 Instrumentation that Initiates or Controls the Core and Containment Coolina Systems 3/4.2.B.23 Core Spray Sparger to Reactor Pressure Vessel d/p Discussion:
This instrumentation measures the differential pressure between the core spray sparger and the reactor pressure vessel above the core plate and alarms if a break is detected. This function does not actuate any equipment; it provides an alarm function only. This function monitors the integrity of the core spray system piping in the reactor annulus region which would not otherwise be apparent to the operators. It is not credited in the accident analysis.
Comparison to Screenina Criteria:
- 1. This instrumentation is not the primary method for detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
O 2. This instrumentation is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. This instrumentation is not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As discussed in Appendix B, Page 1, PECO Energy found the loss of the Core Spray Sparger to Reactor Pressure Vessel d/p Instrumentation to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the Core Spray Sparger to Reactor Pressure Vessel d/p Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.
O PBAPS UNITS 2 & 3 3 of 33 Revision 0 ,
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TABLE 3/4.2.B LPCI CROSS CONNECT POSITION G
V LCO Statement:
The limiting conditions for operation for the instrumentation that initiates or controls the core and containment cooling systems are given in Table 3.2.B.
Table 3.2.8 Instrumentation that Initiates or Controls the Core and Containment Coolina Systems 3/4.2.B.38 LPCI Cross-Connect Position Discussion:
This instrument initiates annunciation when the LPCI cross-connect valve is not closed. During normal operation, the LPCI cross tie valve is required to be closed. In addition, proposed SR 3.5.1.4 will require verification that the LPCI cross connect valve is closed and power is removed once per 31 days.
Thus, this instrument is not the primary method to ensure the valve remains closed, nor is it credited in any accident analysis.
Comparison to Screenino Criteria:
- 1. LPCI cross connect valve position instrumentation is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
' 2. LPCI cross connect valve position instrumentation is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. LPCI cross connect valve position instrumentation is not part of the primary success path that functions or actuates to mitigate a DBA or transient thot either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As discussed in Sections 3.5 and 6 of NED0-31466 and summarized in Table 4-1 (item 332) of NED0-31466, Supplement 1, and verified by PECO Energy, the loss of the LPCI Cross Connect Valve Position instrumentation was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the LPCI Cross Connect Valve Position Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.
PBAPS UNITS 2 & 3 4 of 33 Revision 0
l 2.1.8 APRM R0D BLOCK TRIP SETTING
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TABLE 3/4.2.C CONTROL R0D BLOCKS - APRM UPSCALE (FLOW BIASED, STARTUP MODE), APRM DOWNSCALE LSSS Statement:
The limiting safety system settings shall be as specified below:
B. APRMt R .kBlock Trio Settina Sag s (0.66W + 59% - 0.66aW)
(Clamp @ 108%)
LCO Statement:
The limiting conditions of operation for the instrumentation that initiates control rod blocks are given in Table 3.2.C.
Table 3.2.C Instrumentation that Initiates or Controls the Core and Containment Coolina Systems 3/4.2.C.1 APRM Upscale (Flow Biased) 3/4.2.C.2 APRM Upscale (Startup Mode) 3/4.2.C.3 APRM Downscale Discussion:
(' ) The Average Power Range Monitor (APRM) control rod blocks function to prevent a control rod withdrawal error during power range operations using LPRM signals to create the APRM rod block signal. APRMs provide information about the average core power and APRM rod blocks are not assumed to mitigate a DBA or transient.
Comparison to Screening Criteria:
- 1. The APRM control rod blocks are 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 instrumentatinn is not used to monitor a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. The APRM control rod blocks are not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
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PBAPS UNITS 2 & 3 5 of 33 Revision 0
2.1.8 APRM ROD BLOCK TRIP SETTING O TABLE 3/4.2.C (cont'd.)
CONTROL R0D BLOCKS - APRM UPSCALE (FLOW BIASED, STARTUP MODE), APRM DOWNSCALE As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 135) of NED0 31466, and verified by PECO Energy, the loss of the APRM control rod block functions was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the APRM Control Rod Block Instrumentation LC0 and Surveillances may be relocated to a licensee controlled document. j l
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TABLE 3/4.2.C CONTROL R0D BLOCKS - IRM DCWNSCALE, IRM DETECTOR NOT IN STARTUP POSITION, IRM UPSCALE LC0 Statement:
The limiting conditions of operation for the instrumentation that initiate control rod blocks are given in Table 3.2.C.
Table 3.2.C Instrumentation that Initiates or Controls the Core and Containment Coolina Systems 3/4.2.C 6 IRM Downscale 3/4.2.C.7 IRM Detector Not In Startup Position 3/4.2.C.8 IRM Upscale Discussion:
The Intermediate Range Monitor (IRM) control rod blocks function to prevent a control rod withdrawal error during reactor startup using IRM signals to create the rod block signal. IRMs are provided to monitor the neutron flux levels during refueling and startup conditions. No design basis accident or transient analysis takes credit for rod block signals initiated by IRMs.
Comparison to Screenina Criteria:
~ 1. The IRM control rod blocks are 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 instrumer.tation is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. The IRM control rod blocks are not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As discussed in Sections 3.5 and 6 and summarized in Table 4-I (item 138) of NED0-31466, and verified by PECO Energy, the loss of the IRM control rod block functions was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the IRM Control Rod Block Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.
O PBAPS UNITS 2 & 3 7 of 33 Revision 0
TABLE 3/4.2.C CONTROL R0D BLOCKS - SRM DETECTOR NOT IN STARTUP POSITION, O SRM UPSCALE V
LCO Statement:
The limiting conditions of operation for the instrumentation that initiates control rod blocks are given in Table 3.2.C.
Table 3.2.C Instrumentation that Initiates or Controls the Core '
and Containment Coolina Systems 3/4.2.C SRM Detector Not In Startup Position 3/4.2.C SRM Upscale .
Discussion:
The Source Range Monitor (SRM) control rod blocks function to prevent a control rod withdrawal error during reactor startup using SRM signals to create the rod block signal. SRM signals are used to monitor the neutron flux levels during refueling, shutdown, and startup conditions. No design basis accident or transient analysis takes credit for rod block signals initiated by the SRMs.
Comnarison to Screenino Criteria:
- 1. The SRM control rod blocks are not used for, nor capable of, detecting a O significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. The SRM control rod block instrumentation is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. ,
- 3. The SRM control rod blocks are is not part of the primary success path that functions or actuates to mitigate a DBA or transient that either i assumes the failure of or presents a challenge to the integrity of a ;
fission product barrier.
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 137) of NED0-31466, and verified by PECO Energy, the loss of the SRM control rod block l functions was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the SRM Control Rod Block Instrumentation LC0 and Surveillances may be relocated to a licensee controlled document.
O PBAPS UNITS 2 & 3 8 of 33 Revision 0 ;
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l TABLE 3.2.C CONTROL R0D BLOCKS - SCRAM DISCHARGE INSTRUMENT VOLUME HIGH j O LEVEL G
LC0 Statement:
The limiting conditions for operation for the instrumentation that initiates control rod blocks are given in Table 3.2.C.
Table 3.2.C Instrumentation that Initiate Control Rod Blocks 3/4.2.C.ll Scram Discharge Instrument Volume High Level Discussion:
The Scram Discharge Volume (SDV) control rod block functions to prevent control rod withdrawals during power range operations, utilizing SDV high level signals to create the rod block signal, if water is accumulating in the SDV. The purpose of monitoring the SDV water level is to ensure that there is sufficient volume remaining to contain the water discharged by the control rod drive 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 control rod withdrawal s . With continued water accumulation, a reactor protection system iritiated 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 analysis takes credit for b rod block signals initiated by the SDV high level instrumentation.
Comparison to Screenino Criteria:
- 1. The SDV control rod block 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 a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. The SDV control rod block is not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 139) of NEDO-31466, and verified by PECO Energy, the loss of the Scram Discharge Volume High Level Control Rod Block Instrumentation was found to be a non-significant risk contributor to core damage frequency and offsite releases.
O -
V PBAPS UNITS 2 & 3 9 of 33 Revision 0
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TABLE 3.2.C CONTROL R0D BLOCKS - SCRAM DISCHARGE INSTRUMENT VOLUME HIGH (cont'd.) LEVEL
Conclusion:
Since the screening criteria have not been satisfied, the Scram Discharge l Instrument Volume High Level Control Rod Block Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.
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PBAPS UNITS 2 & 3 10 of 33 Revision 0
3/4.2.F SURVEILLANCE INSTRUMENTATION O LC0 Statement:
The limiting conditions for the instrumentation that provides surveillance information readouts are given in Table 3.2.F Table 3.2.F Surveillance Instrumentation 3/4.2.F.1 Reactor Water Level (narrow range) 3/4.2.F.2 Reactor Water Level (wide range) 3/4.2.F.3 Reactor Water Level (fuel zone) 3/4.2.F.4 Reactor Pressure 3/4.2.F.5 Drywell Pressure 3/4.2.F.6 Drywell Pressure (wide range) 3/4.2.F.7 Drywell Pressure (subatmospheric range) 3/4.2.F.8 Drywell Temperature 3/4.2.F.9 Suppression Chamber Water Temperature 1 3/4.2.F.10 Suppression Chamber Water Level (narrow range) 3/4.2.F.ll Suppression Chamber Water Level (wide range) 3/4.2.F.12 Control Rod Position 3/4.2.F.13 Neutron Monitoring 3/4.2.F.14 Safety-Relief Valve Position Indication 3/4.2.F.15 Drywell High Range Radiation Monitors 3/4.2.F.16 Main Stack High Range Radiation Monitor 3/4.2.F.17 Reactor Building Roof Vent High Range Radiation j Monitor !
3/4.2.F.18 Drywell Hydrogen Concentration Analyzer and Monitor Discussion:
Each individual accident monitoring parameter has a specific purpose, however, the general purpose for all accident monitoring instrumentation is to provide sufficient information to confirm an accident is proceeding per prediction, i.e. automatic safety systems are performing properly, and deviations from expected accident course are minimal.
Comparison to Screenina Criteria:
The NRC position on application of the deterministic screening criteria to post-accident monitoring instrumentation is documented in letter dated May 7, 1988 from T. E. Murley (NRC) to R. F. Janecek (BWROG). 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 plants SER on Regulatory Guide 1.97, and all Regulatory Guide 1.97 Category 1 instruments. Accordingly, this position has been applied to the PBAPS Regulatory Guide 1.97 instruments. Those instruments not meeting this criteria have been relocated from the Technical Specifications to a licensee controlled document.
O PBAPS UNITS 2 & 3 11 of 33 Revision 0
3/3.2 F SURVEILLANCE INSTRUMENTATION (cont'd.)
The following summarizes the PBAPS position for those instruments currently in Technical Specifications.
From NRC SER dated 1/15/88,
Subject:
Conformance to RG 1.97 Tyne A Variables
- 1. Reactor Pressure
- 2. Reactor Water Level (wide range, fuel range)
- 3. Suppression Pool Water Temperature
- 4. Suppression Pool Water Level (wide range)
- 5. - Drywell Pressure (wide range, subatmospheric range)
- 6. Containment Oxygen Concentration Other Tyne, Category 1 Variables
- 1. Drywell High Range Radiation Monitor
- 2. Drywell Hydrogen Concentration Analyzer and Monitor For other post-accident monitoring instrumentation currently in Technical Specifications, their loss is not considered risk significant since the variable they monitored 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 O operator can perform necessary manual actions).
Conclusion:
Since the screening criteria have not been satisfied for non-Regulatory Guide
- 1.97 Type A or Category 1 variable Type A instruments, their associated LCO and Surveillances will be relocated to a licensee controlled document. The instruments to be relocated are as follows:
- 1. Reactor Water Level (Narrow Range)
- 2. Drywell Pressure
- 3. Drywell Temperature
- 4. Suppression Chamber Water Level (Narrow Range)
- 5. Control Rod Position
- 6. Neutron Monitoring
- 7. Safety-Relief Valve Position Indication
- 8. Main Stack High Range Radiation Monitor
- 9. Reactor Building Roof Vent High Range Radiation Monitor <
O PBAPS UNITS 2 & 3 12 of 33 Pevision 0
3/4.2.G ALTERNATE ROD INSERTION D
O L(0 Statement:
The limiting conditions for the instrumentation that initiates an Alternate Rod Insertion scram.. . are given in Table 3.2.G.
Table 3.2.G Instrumentation that Initiates Alternate Rod Insertion...
3/4.2.G.I Reactor High Pressure 3/4.2.G.2 Reactor Low-Low Water Level Discussion:
The Alternate Rod Insertion (ARI) Instrumentation functions as a backup to the Reactor Protection System. As such, ARI functions to limit the consequences of a Reactor Protection System failure to scram during an anticipated transient. However, this instrumentation is not credited in the accident or transient analyses.
Comparison to Screenina Criteria:
- 1. The ARI instrumentation is not used for, nor are they capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
O 2. The ARI instrumentation is not process variable that is a initial 1
l i'
condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. ;
- 3. The ARI instrumentation is not part of the primary success path that l functions or actuates to mitigate a DBA or transient that either assumes i the failure of or presents a challenge to the integrity of the fission product barrier.
As discussed in Appendix B, Page 2, PECO Energy found the loss of the j Alternate Rod Insertion Instrumentation to be a non-significant risk i contributor to core damage frequency and offsite releases. I
Conclusion:
Since the screening criteria have not been satisfied, the ARI Instrumentation LCO and Surveillances may be relocated to a licensee controlled document.
O PBAPS UNITS 2 & 3 13 of 33 Revision 0
3/4.6.B.2 PRIMARY SYSTEM BOUNDARY - COOLANT CHEMISTRY p
a 3.6.B.3 100 Statement:
The following limits shall be observed for reactor water quality prior to any startup and when operating at rated pressure:
a) Conductivity at 25'C - 5.0 pmho/cm b) Chloride concentration - 0.2 ppm '
Discussion:
Poor reactor coolant water chemistry contributes to the long-tera degradation of system materials and, thus, is not of immedicte importance to the plant operator. Reactor coolant water chemistry is maintained to reduce the possibility of failure in the reactor coolant system pressure boundary caused by corrosion. In summary, the chemistry monitoring activity is of a long term preventative purpose rather than mitigative.
Comparison to Screenina Criteria:
- 1. Reactor coolant water chemistry is not used for, nor capable of, detecting a significant abnormal degradatica of the reactor coolant pressure boundary prior to a DBA.
O 2. Reactor coolant water chemistry is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. Reactor coolant water chemistry is not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 211) of NED0-31466, and verified by PECO Energy, Coolant Chemistry requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the Coolant Chemistry (Conductivity and Chloride) LC0 and Surveillances may be relocated to a licensee controlled document.
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PBAPS UNITS 2 & 3 14 of 33 Revision 0 l
1 3/4.6 G PRIMARY SYSTEM BOUNDARY - STRUCTURAL INTEGRITY
~N (d L(0 Statement:
The structural integrity of the primary system boundary shall be maintained at the level required by the original acceptance standards throughout the life of the station.
Discussion:
The inspection programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity of those components will be maintained throughout the components life. Operability of the primary system boundary is ensured by separate Technical Specifications and therefore, the inspections are not required to be retained in the Technical Specifications. This Technical Specification is more directed toward prevention of component degradation and continued long term maintenance of acceptable structural conditions. However, it is not necessary to retain this Specification to ensure the operability of the primary system boundary.
Comparison to Screenina Criteria:
- 1. The inspections stipulated by this Specification are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
O The inspections stipulated by this Specification do not monitor process Q 2.
variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 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, however, only specifies inspection requirements for these components; and these inspections can only be performed when the plant is shutdown. Therefore, Criterion 3 is not satisfied.
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 216) of NE00-31466, and verified by PECO Energy, the lack of a Structural Integrity Specification was found to be a non-significant risk contributor to core damage frequency and offsite releases since the requirement is currently covered by 10 CFR 50.55a and the Inservice Inspection Program.
Conclusion:
Since the screening criteria have not been satisfied, the Structural Integrity LCO and Surveillance may be relocated to a licensee controlled document.
PBAPS UNITS 2 & 3 15 of 33 Revision 0
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3/4.8.A RADI0 ACTIVE MATERIALS - GENERAL A
V LCO Statement:
It is expected that releases of radioactive material in effluents will be kept at small fractions of the limits specified in Section 20.106 of 10 CFR Part 20 or as further specified in these Technical Specifications. )
i Discussion:
This requirement states releases of radioactive material in effluents will be kept within the limits of 10 CFR 20.106 and as further specified in the Technical Specifications. This requirement is duplicative of the general requirements in 10 CFR 20.106 and need not be included in Technical l Specifications. This general statement is not a condition of a DBA or a transient analysis that is based upon the integrity of the fission product barrier. j l
Comparison to Screenino Criteria:
l l
- 1. Radioactive materials requirements are not used for, nor capable of, '
detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. Radioactive materials requirements are not process variables that are :
initial conditions of a DBA or transient analysis that either assumes O
5 the failure of or presents a challenge to the integrity of a fission product barrier.
l
]
- 3. Radioactive materials requirements are not used in any part of a primary success path in the mitigation of a DBA or transient. 1 i
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (items 296, !
297, 298, 300, 301, 302, 303, 304, and 305) of NED0-31466, and verified by 1 PECO Energy, Radioactive Materials - General requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.
i
Conclusion:
l Since the screening criteria have not been satisfied, the Radioactive l Materials - General LC0 and Surveillances may be relocated to a licensee ,
controlled document. I l
O PBAPS UNITS 2 & 3 16 of 33 Revision 0 )
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3/4.8.B RADI0 ACTIVE MATERIALS - LIQUID RADWASTE EFFLUENTS n
V LCO Statement:
- 1. The concentration of radioactive material released to areas at and beyond the SITE BOUNDARY (See Figure 3.8.1) shall be limited to the Table II, Column 2 for concentration radionuclides otherspecified in 10gases than noble CFRand 20 Appendix 2X10' p,Ci/nl total activity of all dissolved or entrained noble gases.
- 2. The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluent releases from the two reactors at the site to areas at or beyond the SITE B0UNDARY (See Figure 3.8.1) shall be limited to:
- a. During any calendar quarter to s 3.0 mrem to the total body and to s 10.0 mrem to any organ, and,
- b. During any calendar year to s 6.0 mrem to the total body and to s 20.0 mrem to any organ.
- 3. During release of radioactive wastes, the following conditions shall be mat-
- a. The minimum dilution water required to satisfy 3.8.B.1 shall be met.
- b. The gross activity monitor and flow monitor on the waste effluent line shall be operable except as specified in 3.8.B.3.d and 3.8.B.3.e, below.
- c. The effluent control monitor shall be set in accordance with the methodology and parameters in the ODCM to alarm and automatically close the waste discharge valve prior to exceeding the limits specified in 3.8.B.1 above.
- 4. All liquid effluent releases at and beyond the SITE BOUNDARY shall be processed through one of the radwaste subsystems or combinations of these subsystems listed below, prior to release:
(i) The Waste Collection Filter and Demineralizer (ii) The Floor Drain Filter (iii) The Fuel Pool Filter Demineralizer (iv) The Chemical /0ily Waste Cleanup Subsystem O
PBAPS UNITS 2 & 3 17 of 33 Revision 0
3/4.8.B RADI0 ACTIVE MATERIALS - LIQUID RADWASTE EFFLUENTS (cont'd.)
Discussion:
10 CFR Part 20, BII(2) refers to liquid release to an unrestricted area of radioactive material in concentrations that exceed the specified limits. No screening criteria apply because the process variable of the LC0 (concentration of radioactive material in liquid effluents) is not an initial condition of a design basis accident (DBA) or transient analysis. Effluent control is for protection against radiation hazards from licensed activities, not accidents.
Limitation of the quarterly and annual projected doses to MEMBERS OF THE PUBLIC which result from cumulative liquid effluent discharge during normal operation over e;, tended periods is intended to assure compliance with the dose objectives of 10 CFR Part 50, Appendix I. These limits are not related to protection of the public from the consequences of any DBA or transient.
Radioactive liquid effluent instrumentation and associated requirements for effluent releases are used to assure conformance to the discharge limits of 10 CFR Part 20. They are not installed to detect excessive reactor coolant leakage. The radioactive liquid effluent monitors are used routinely to provide a continuous check on the release of radioactive liquid effluent from the normal plant effluent flow paths. These requirements ensure the various liquid effluent monitors are maintained operable with setpoints established in
~
accordance with the Offsite Dose Calculation Manual (ODCM). Plant DBA and
/ transient analyses do not assume any action, either automatic or manual, resulting from radioactive liquid effluent monitors.
The requirement for a liquid waste treatment system pertains to controlling the release of site liquid effluents during normal operational occurrences.
No loss of primary coolant is involved, neither is an accident condition assumed or implied. The limits for release in 10 CFR Part 50, Appendix I for liquids are design objectives for operation. In addition, the liquid radwaste subsystems are not credited in the safety sequence analysis and are not part ,
of the primary coolant pressure boundary. I 1
These requirements were also the subject of Generic letter 89-01 which allowed '
the removal of Radiological Effluent Technical Specifications (RETS) from the Technical Specifications.
Comparison to Screenina Criteria:
- 1. Liquid radwaste effluent requirements are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
O PBAPS UNITS 2 & 3 18 of 33 Revision 0
h 3/4.8.B RADI0 ACTIVE MATERIALS - LIQUID RADWASTE EFFLUENTS (cont'd.)
- 2. Liquid radwaste effluent requirements are not process variables that are an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. Liquid radwaste effluent requirements are not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the '
integrity of a fission product barrier.
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (items 188, 296, 297, and 298) of NED0-31466, and verified by PECO Energy, the Liquid Radwaste Effluent. requirements not being met was found to be a non-significant '
risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the Liquid Radwaste Effluent LC0 and Surveillances may be relocated to a licensee controlled ,
document.
O PBAPS UNITS 2 & 3 19 of 33 Revision 0
. _ = _ . . . . . - _ - -- . . - -.
3/4.8.C RADI0 ACTIVE MATERIALS - GASEOUS EFFLUENTS O (except 3/4.8.C.6 (EXPLOSIVE GAS MIXTURE) and 3/4.8.C.7.a (MAIN CONDENSER OFFGAS))
l LCO Statement:
- 1. The dose rate in areas at and beyond the SITE BOUNDARY (see Figure 3.8.1) due to radioactive materials in gaseous effluents released from the two reactors at the site shall be limited to the following:
- a. The dose rate for noble gases shall be limited to s 500 mrem /yr to i the total body and s 3000 mrem /yr to the skin.
- b. The dose rate for iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half lives greater than 8 ;
days shall be s 1500 mrem /yr to any organ.
- 2. The air dose in areas at and beyond the SITE BOUNDARY (see Figure 3.8.1) due to noble gases in gaseous effluents released from the two reactors at the site shall be limited to the following:
- a. During any calendar quarter for gamma radiation: s 10 mrad. L During any calendar quarter for beta radiation: s 20 mrad, i
- b. During any calendar year for gamma radiation: 5 20 mrad. During any calendar year for beta radiation: s 40 mrad.
O 3.
The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium and from all radionuclides in particulate form with half-lives greater i than 8 days in gaseous effluents released from the two reactors at the ;
site to areas at and beyond the SITE BOUNDARY (see Figure 3.8.1) shall be limited to the following:
- a. During any calendar quarter: s 15 mrem.
- b. During any calendar year: s 30 mrem. ,
- 4. During release of gaseous wastes the following conditions shall be met -
to avoid exceeding the limits specified in 3.8.C.1:
i
- a. The main off-gas stack minimum dilution flow of 10,000 cfm shall i be maintained. <
- b. One reactor building exhaust vent monitor and one main stack noble i gas monitor shall be operable and set to alarm in accordance with the methodology and parameters in the ODCM.
e PBAPS UNITS 2 & 3 20 of 33 Revision 0 t
3/4.8.C RADI0 ACTIVE MATERIALS - GASE0US EFFLUENTS O (cont'd.) (except 3/4.8.C.6 (EXPLOSIVE GAS MIXTURE) and 3/4.8.C.7.a (MAIN CONDENSER OFFGAS))
- c. One reactor building exhaust vent iodine filter and one main stack iodine filter and one reactor building exhaust vent particulate filter and one main stack particulate filter with their respective flow rate monitors shall be operable.
- d. One reactor building exhaust vent flow rate monitor and one main stack flow rate monitor shall be operable and set to alarm in accordance with the methodology and parameters in the ODCM.
- 5. Gaseous effluents shall be processed through the appropriate gaseous waste treatment system as described below prior to discharge:
- a. Gases from the Steam Jet Air Ejector Discharge shall be processed through the recombiner, holdup pipe, off-gas filter, and off-gas stack.
- b. Gases from the Mechanical Vacuum Pump and Gland Steam Exhauster discharge shall be processed through the off-gas stack.
- c. Reactor, turbine, radwaste, and recombiner building atmospheres shall be processed through permanently or temporarily installed equipment in the appropriate building ventilation system and the Reactor Building Ventilation Exhaust Stack, with the exception of the following unmanitored exhausts:
- 1. Recirculation M-G Set and Reactor Building Cooling Water equipment rooms.
- 2. Control room utility and toilet rooms.
- 3. Cable spread room.
- 4. Emergency switchgear rooms.
- 5. 125/250 VDC Battery rooms and the 250 VDC Battery rooms.
- 6. Administration Building maintenance decontamination areas 7b. One Steam Jet Air Ejector radiation monitor shall be operable during operation of a main condenser Steam Jet Air Ejector.
8a. Purging of primary containment shall be through the Standby Gas Treatment System whenever primary containment integrity is required as specified in 3.7.A.2.
8b. Primary containment purging via the Reactor Building Ventilation Exhaust System may be performed whenever primary containment integrity is not required as specified in 3.7.A.2.
O PBAPS UNITS 2 & 3 21 of 33 Revision 0
3/4.8.C RADI0 ACTIVE MKfERIALS - GASE0US EFFLUENTS O
V-(cont'd.) (except 3/4.3.C.6 (EXPLOSIVE GAS MIXTURE) and 3/4.8.C.7.a (MAIN CONDEilSER OFFGAS))
i Discussion:
Limitations are provided for the dose rate due to gaseous effluent in unrestricted areas r.t any time to a value less than the yearly dose limit of 10 CFR Part 20. This provides reasonable assurance that no MEMBER OF THE PUBLIC is exposed to annual average concentrations that exceed the limits of 10 CFR Part 20 Appendix B, Table II. These are limits which apply to normal operation of the plant. They are not assumed as an initial condition of any design basis accident (DBA) or transient analysis and are not relied upon to limit the consequences of such events.
Limitation of the quarterly and annual air doses from noble gases in plant gaseous effluents during normal operation over extended periods is intended to assure compliance with the dose objectives of 10 CFR Part 50, Appendix 1.
These limits are not related to protection of the public from the consequences of any DBA or transient.
Limitation of the quarterly and annual projected doses to MEMBERS OF THE PUBLIC from radionuclides other than noble gases during normal operation over extended periods is intended to assure compliance with the dose objectives of 10 CFR Part 50, Appendix 1. These limits are not related to protection of the public from consequences of any DBA or transient.
Radioactive gases effluent monitoring instrumentation and associated requirements for gaseous effluent releases are used to assure conformance to the discharge limits of 10 CFR Part 20. They are not installed to detect excessive reactor coolant leakage. The radioactive gaseous effluent monitors are used routinely to provide a continuous check on the release of radioactive gaseous effluents from the normal plant gaseous effluent flow paths. These requirements ensure the various effluent monitors are maintained operable with setpoints established in accordance with the Offsite Dose Calculation Manual (0DCM). Plant DBA and transient analyses do not assume any action, either automatic or manual, resulting from radioactive gaseous effluent monitors.
The requirement for use of the gaseous radwaste treatment system provides assurance the requirements of 10 CFR Part 50.36a, General Design Criteria 60 of Appendix A to 10 CFR Part 50, and the design objectives of 10 CFR Part 50, .
Appendix I, will be met. In addition, the operability of the gaseous radwaste treatment system is not assumed in the analysis or any DBA or transient.
O f PBAFS UNITS 2 & 3 22 of 33 Revision 0
3/4.8.C RADI0 ACTIVE MATERIALS - GASEOUS EFFLUENTS
(
(cont'd.) (except 3/4.8.C.6 (EXPLOSIVE GAS MIXTURE) i and 3/4.8.C.7.a (MAIN CONDENSER OFFGAS))
The requirement to treat the primary containment atmosphere prior to release is intended to provide reasonable assurance that releases of radioactive materials during normal operational occurrences are "as low as reasonably achievable" and to help assure compliance with the annual dose limits of 10 CFR Part 20 for areas at or beyond the SITE BOUNDARY. These limits are not related to protection of the public from any DBA or transient.
These requirements were also the subject of Generic Letter 89-01 which allowed the removal of Radiological Effluent Technical Specifications (RETS) from the Technical Specifications. t Comparison to Screenina Criteria:
- 1. Gaseous effluent requirements are not used for, nor capable of, i
detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA. +
- 2. Gaseous effluent requirements are not process variables that are initial i conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. Gaseous effluent requirements are not part of the primary success path !
that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a !
fission product barrier. !
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (items 189, ;
300, 301, 302, 303, and 305) of NED0-31466, and verified by PECO Energy, Gaseous Effluent requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the Gaseous Effluent LC0 and Surveillances may be relocated to a licensee controlled document.
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1 O i PBAPS UNITS 2 & 3 23 of 33 Revision 0
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3/4.8.C.6 RADI0 ACTIVE MATERIALS - GASEOUS EFFLUENTS, EXPLOSIVE GAS ,
A MIXTURE !
k_,)
LCO Statement:
The concentration of hydrogen downstream of the recombiners shall be limited to less than or equal to 4% by volume.
Discussion:
The explosive gas mixture Specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limit of hydrogen. l However, the waste gas holdup system is designed to contain detonations and will not affect the function of any safety related equipment. The,. ,
concentration of hydrogen in the offgas stream is not an initial assumption of any design basis accident (DBA) or transient analysis.
4 Comnarison to Screenina Criteria:
- 1. The explosive gas mixture requirements are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA. ,
- 2. The explosive gas mixture requirements are not process variables _ that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a
()
fission product barrier.
- 3. The explosive gas mixture requirements are not part of the primary success path that functions or actuates to mitigate a DBA or transient that eitner assumes the failure of or presents a challenge to the ,
integrity of a fission product barrier.
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 306) of NED0-31466, and verified by PECO Energy, an explosive gas mixture in the waste gas holdup system was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the Explosive Gas Mixture LCO and Surveillances may be relocated to a licensee controlled document.
O PBAPS UNITS 2 & 3 24 of 33 Revision 0
3/4.8.D RADI0 ACTIVE MATERIALS - 40 CFR 190 LC0 Statement: ,
The dose or dose commitment to a MEMBER OF THE PUBLIC from all uranium fuel cycle sources within 8 kilometers is limited to s 25 mrem to the total body or any organ (except the thyroid which is limited to s 75 mrem) over the calendar year.
Discussion:
This LCO limits the annual doses to individual members of the public from all plant sources. This is intended to assure that normal operation of the plant is in compliance with the provisions of 40 CFR Part 190. These limits are not related to protection of the public from any design basis accident (DBA) or transient. This requirement was also the subject of Generic Letter 89-01 which allowed the removal of Radiological Effluent Technical Specifications (RETS) from the Technical Specifications. '
Comparison to Screenina Criteria: ;
- 1. 40 CFR 190 requirements are not used for, nor capable of, detecting a 3 significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. 40 CFR 190 requirements are not process variables that are initial O conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. 40 CFR 190 requirements are not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 304) of NED0-31466, and verified by PECO Energy, 40 CFR 190 requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the 40 CFR 190 LC0 and Surveillances may be relocated to a licensee controlled document.
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PBAPS UNITS 2 & 3 25 of 33 Revision 0 t
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3/4.8.E RADI0 ACTIVE MATERIALS - RADIOLOGICAL ENVIRONMENTAL MONITORING LCO Statement:
- 1. All deviations from the sampling schedule for the radiological environmental monitoring program, as required by 4.8.E.1, shall be documented in the annual report.
- 2. A land use census shall be conducted and shall identify the location of the nearest milk animal in each of the 16 meteorological sectors within a distance of five miles.
- 3. Analyses shall be performed on radioactive materials supplied as part of the EPA Environmental Radioactivity Intercomparison Studies Program, or another Interlaboratory Comparison Program that has been approved by the Commission.
Discussion:
The radiological environmental monitoring requirements of this Specification provide measurements of radiation and radioactive materials in the exposure pathways and for those radionuclides which lend to the highest potential radiation exposure for members of the public resulting from station operations. This program monitors the long term impact of normal plant A operations. These requirements were also the subject of Generic Letter 89-01 V which allowed the removal of Radiological Effluent Technical Specifications (RETS) from the Technical Specifications.
Comnarison to Screenina Criteria:
- 1. Radiological environmental monitoring requirements are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. Radiological enviroamental monitoring requirements are not process variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. Radiological environmental monitoring requirements are not part of the primary success path that functions or actuates to mitigate a DBA or i transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. )
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 309) of NEDO-31466, and verified by PECO Energy, Radiological Environmental Monitoring ;
requirements not being met was found to be a non-significant risk contributor l to core damage frequency and offsite releases. !
i PBAPS UNITS 2 & 3 26 of 33 Revision 0 l
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3/4.8.E RADI0 ACTIVE MATERIALS - RADIOLOGICAL ENVIRONMENTAL (cont'd.) MONITORING r
Conclusion:
Since the screening criteria have not been satisfied, the Radiological Environmental Monitoring LC0 and Surveillances may be relocated to a licensee controlled document.
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O PBAPS UNITS 2 & 3 27 of 33 Revision 0 .
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3/4.8.F RADI0 ACTIVE MATERIALS - SOLID RADI0 ACTIVE WASTE LCO Statement:
The solid radwaste system shall be used in accordance with a Process Control Program (PCP) to process wet radioactive wastes to meet shipping and burial ground requirements.
Discussion:
The solid radioactive waste system is a logical continuation of the liquid radwaste system. It operates on the same requirement for effluent control, identified as controlling the release and handling of radioactive solid wastes. The system serves to control operational release of solid waste, not accidental release. This requirement was also the subject of Generic Letter 89-01 which allowed removal of Radiological Effluent Technical Specifications (RETS) from the Technical Specifications.
Comparison to Screenina Criteria:
- 1. Solid radioactive waste requirements are not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. Solid radioactive waste requirements are not process variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. Solid radioactive waste requirements are not part of the primary success path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 308) of NED0-31466, and verified by PECO Energy, the Solid Radioactive Waste requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the Solid Radioactive Waste LC0 and Surveillances may be relocated to a licensee controlled document. !
I O i PBAPS UNITS 2 & 3 28 of 33 Revision 0
3/4.10.D HEAVY LOADS OVER SPENT FUEL LCO Statement:
Loads in excess of 1000 lbs. (excluding the rigging and transport vehicle) shall be prohibited from travel over fuel assemblies in the spent fuel storage pool.
Discussion:
The restriction on movement of loads in excess of the approximate weight of a fuel assembly over spent fuel assemblies in the storage pool ensures that in the event the load is dropped, the activity release will be limited to that contained in a single fuel assembly and any possible distortion of the fuel in the storage racks will not result in a critical array. Monitoring of loads moving over the spent fuel storage racks is governed by strict administrative controls.
Although this Technical Specification supports the maximum refueling accident assumption in the DBA, the fuel handling crane limits are not monitored and controlled during operation; they are checked on a periodic basis to ensure the requirements are satisfied. The deterministic criteria for Technical Specification retention are therefore, not satisfied.
1 Comparison to Screenino Criteria: '
O 1. The fuel handling crano 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 maximum severity assumed for the fuel handling DBA is limited by the -
limits placed on the crane. These crane limits are not, however, process variables monitored and controlled by the operator. Therefore, )
Criterion 2 is not satisfied. l i
- 3. The fuel handling crane limits are not a structure, system, or component l that is part of the primary success path and which functions or actuates j to mitigate a DBA. i Traditional PRAs do not review risks associated with the spent fuel storage l pool. Design basis analyses indicate that the release associated with fuel l assembly damage in the spent fuel storage pool due to crane accidents is '
significantly lower than releases typically evaluated by PP.As.
Conclusion:
Since the screening criteria have not been satisfied, the Heavy Loads Over Spent Fuel LCO may be relocated to other licensee controlled documents.
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PBAPS UNITS 2 & 3 29 of 33 Revision 0 l I
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3/4.12 RIVER LEVEL LCO Statement: ,
A.1 If river level reaches a level of 113.0 ft (C.D.) at Peach Bottom and the predicted flow rate is greater than 840,000 cfs, the' reactors will be shut down to the cold condition using normal operating procedures.
A.2 If the river level exceeds 114.0 ft. (C.D.) at Peach Bottom the reactors will be manually scrammed and placed in the cold condition according to the applicable Special Event Procedure.
B.1 In the event of an unscheduled drop in river level to 104.0 ft (C.D.) at Peach Bottom, the reactor shall be shut down to the cold condition using normal operating procedures.
B.2 In the event of an unscheduled drop in water level to 98.5 ft (C.D.) at Peach Bottom, the reactors will be manually scrammed and placed in the cold condition according to the applicable Special Event Procedure.
C.1 Two of the three river water level indicators in the control room shall be continuously operable.
Discussion:
This Technical Specification has provisions for high and low river water levels and level instrumentation. A high river water level is a preliminary
(- indication of flood conditions. Low river water level is caused by an uncontrolled release at the conowingo Dam which leads to a lower level in the normal heat sink and potential loss of the normal heat sink. Neither the case of the flood or an uncontrolled release is a design basis accident or transient, thus river water level is not credited in any safety analysis. The river water level Technical Specification requirements were put in place to ensure the emergency heat sink was placed in service in a timely manner in advance of a total loss of the normal heat sink. This requirement is adequately controlled in plant emergency procedures and the Technical Specification operability requirements for the normal heat sink.
Comoarison to Screenina Criteria:
- 1. River level requirements not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure bounoary prior to a DBA.
- 2. River level requirements are not process variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
O PBAPS UNITS 2 & 3 30 of 33 Revision 0
3/4.12 RIVER LEVEL (cont'd.)-
- 3. River level-requirements are not part of the primary success path that function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As discussed in Appendix B, Page 3, PECO Energy found River Level requirements not being met to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the River Level LCO and Surveillances may be relocated to a licensee controlled document.
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O PBAPS UNITS 2 & 3 31 of 33 Revision 0
3/4.13 MISCELLANE0US RADI0 ACTIVE MATERIALS SOURCES m
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LCO Statement:
The leakage test shall be capable of detecting presence of 0.005 microcurie of radioactive material on the test sample.
A complete inventory of radioactive byproduct materials in sealed sources in possession shall be maintained current at all times.
Discussion:
The limitations on sealed source contamination are intended to ensure that the I total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation. This is done by imposing a limitation on the maximum amount of removable contamination on each sealed source. This requirement and the associated surveillance requirements bear no relation to the conditions or limitations which are necessary to ensure safe reactor l operation. I 1
Comparison to Screenina Criteria- 1 i
- 1. Miscellaneous radioactive materials sources requirements are not used l for, nor capable of, detecting a significant abnormal degradation of the l reactor coolant pressure boundary prior to a DBA. l l
- 2. Miscellaneous radioactive materials sources requirements are not process i variables that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
- 3. Miscellaneous radioactive materials sources requirements are not part of the primary success path that function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 267) of NED0 31466, and verified by PECO Energy, the Miscellaneous Radioactive Materials Sources requirements not being met was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the Miscellaneous Radioactive fiaterials Sources LCO and Surveillances may be relocated to a licensee controlled document. .
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PBAPS UNITS 2 & 3 32 of 33 Revision 0 i
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3/4.15 SEISMIC MONITORING INSTRUMENTATION V
LC0 Statement: }
The seismic monitoring instrumentation shown in Table 3.15 shall be operable. l Discussion:
In the event of an earthquake, seismic monitoring instrumentation is required to determine the magnitude of the seismic event. These instruments do not perform any automatic action. They are used to measure the magnitude of the seismic event for comparison to the design basis of the plant to ensure the '
design margins for plant equipment and structures have not been violated.
Since the determination of the magnitude of the seismic event is performed after the event has occurred, this instrumentation has no bearing on the mitigation of any design basis accident (DBA) or transient.
Comnarison to Screenino Criteria:
- 1. Seismic monitoring instrumentation is not used for, nor capable of, detecting a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA.
- 2. Seismic monitoring instrumentation is not a process variable that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product O barrier.
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- 3. Seismic monitoring instrumentation is not part of the primary success l path that functions or actuates to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
l As discussed in Sections 3.5 and 6 and summarized in Table 4-1 (item 151) of i NED0-31466, and verified by PECO Energy, the loss of the Seismic Monitoring !
Instrumentation was found to be a non-significant risk contributor to core damage frequency and offsite releases.
Conclusion:
Since the screening criteria have not been satisfied, the Seismic Monitoring Instrumentation LCO and Surveillances may be relocated to a licensee l controlled document. :
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APPENDIX B l
@ PBAPS SPECIFIC i
1 RISK SIGNIFICANT EVALUATIONS l
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TABLE 3/4.2.8 CORE SPRAY SPARGER TO REACTOR PRESSURE VESSEL d/p G
V LC0 Statement:
The limiting conditions for operation for the instrumentation that initiates or controls the core and containment cooling systems are given in Table 3.2.B.
Table 3.2.B Instrumentation that Initiates or Controls the Core and Containment Coolina Systems 3/4.2.B.23 Core Spray Sparger to Reactor Pressure Vessel d/p Descrintion of Reauirement:
This instrumentation measures the differential pressure between the core spray sparger and the reactor pressure vessel above the core plate and alarms if a break is detected. This instrumentation does not actuate any equipment.
Risk Justification:
The function of the instrumentation is to identify a break in the core spray sparger. The probability of a pipe break (EPRI TR-100380) is extremely low, therefore the relative probability as defined in NED0-31466 is low. LOCAs represent a small contribution to the Peach Bottom core damage frequency (CDF). A break in the sparger in the reactor pressure vessel would, without a q LOCA, provide injection to the core. Given the success of injection with a break in non-LOCA accidents and the small contribution of LOCA to the CDF, the V relative significance from an offsite radiological dose perspective would be low. The risk category would therefore be considered non-significant (NS).
Relative Probability Relative Sianificance Risk Cateaory Low Low NS l
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3/4.2.G ALTERNATE ROD INSERTION LCO Statement:
The limiting conditions for the instrumentation that initiates an Alternate Rod Insertion scram. . . are given in Table 3.2.G.
Table 3.2.G Instrumentation that Initiates Alternate Rod Insertion..
3/4.2.G.1 Reactor High Pressure 3/4.2.G.2 Reactor Low-Low Water Level Descriotion of Reauirement:
The Alternate Rod Insertion (ARI) instrumentation functions as a backup to the Reactor Protection System (RPS). As such, ARI functions to limit the consequences of an RPS failure to scram during an anticipated transient. '
However, this instrumentation is not credited in the accident or transient analyses.
Risk Justification:
The Peach Bottom Atomic Power Station probabilistic safety assessment specifically addresses the failure to scram subsequent to an initiating transient event. The failure to scram is analyzed by considering mechanical O faults (hydraulic) and electrical faults that prevent successful shutdown of the reactor. A mechanical failure will preclude the successful insertion of control rods by the electrical portion of RFS and ARI. Although the probability of failure of the electrical portion of RPS is much greater than the failure of the mechanical portion of RPS, the risk significance (importance) of the failure of ARI is small compared to the significance of the mechanical failure of RPS. This is due to the probability of the failure combination of the electrical portions of ARI and RPS that must occur for an ATWS to progress. The relative probability associated with the loss of the ARI function is considered to be medium when assessing the event frequency per Table 3-3 of NED0-31466. The relative significance in terms of offsite radiological dose (Table 3-4 of NED0-31466) of the functional loss of ARI is low due to the probability of additional failures that need to occur to affect the dose to the public. The risk category established in Table 3-5 of NED0-31466 would indicate that the combination of relative probability and relative significance would result in a risk non-significant categorization.
O PBAPS UNITS 2 & 3 2 of 4 Revision 0
3/4.12 RIVER LEVEL LC0 Statement:
3/4.12.A.1 If river level reaches a level of 113.0 ft (C.D.) at Peach Bottom and the predicted flow rate is greater than 840,000 cfs, the reactors will be shut down to the cold condition using normal operating procedures.
3/4.12.A.2 If the river level exceeds 114.0 ft. (C.D.) at Peach Bottom the reactors will be manually scrammed and placed in the cold condition according to the applicable Special Event Procedure.
3/4.12.B.1 In the event of an unscheduled drop in river level to 104.0 ft (C.D.) at Peach Bottom, the reactors shall be shut down to the cold condition using normal operating procedures.
3/4.12.B.2 In the event of an unscheduled drop in water level to 98.5 ft (C.D.) at Peach Bottom, the reactors will be manually scrammed and placed in the cold condition according to the applicable Special Event Procedure.
3/4.12.C.1 Two of the three river water level indicators in the control room shall be continuously operable.
Descriotion of Reauirements; This Technical Specification has provisions for high and low river water levels and level instrumentation. A high river water level is a preliminary indication of flood conditions. Low river water level is caused by an uncontrolled release at the Conowingo Dam which leads to a lower level in the normal heat sink and potential loss of the normal heat sink. Neither the case of the flood or an uncontrolled release is a design basis accident or 1 transient, thus river water level is not credited in any safety analysis. The '
river water level Technical Specification requirements were put in place to :
ensure the emergency heat sink was placed in service in a timely manner in advance of a total loss of the normal heat sink.
Risk Justification:
An analysis of the risk of external plant flooding was performed in NUREG/CR-4550 Vol . 4, Part 3 for Peach Bottom. Historical flood data was collected and analyzed to determine the frequency and magnitude of floods at Peach Bottom. All critical equipment essential to the safe shutdown of the plant are flood protected to an elevation well above that required in the 1
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3/4.12 RIVER LEVEL (cont'd) current LCO. Given the plant design features and the conservative analysis of flooding in NUREG/CR-4550, the contribution of flooding to overall plant risk (probability of occurrence and radiological consequence) is considered negligible.
Relative Probability Relative Sianificance Risk Category Low Low NS An analysis of the risk of low river water level was performed in NUREG/CR-4550 Vol .4, Rev.1, Part 3 for Peach Bottom. This event or condition was excluded from detailed consideration and thus resulted in a minimal risk impact for the following reasons: 1) The river flow and level are regulated by dams upstream and downstream of the plant, 2) low river level is of equal or lesser damage potential than the events for which the plant was designed, and 3) the event is slow in developing and there is sufficient time to eliminate the condition or to provide an adequate response such as using the emergency cooling tower (emergency heat sink).
Relative Probability Relative Sianificance Risk Category Low Low NS
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