ML20150C520
| ML20150C520 | |
| Person / Time | |
|---|---|
| Issue date: | 03/10/1988 |
| From: | Butcher E Office of Nuclear Reactor Regulation |
| To: | Rossi C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8803210041 | |
| Download: ML20150C520 (45) | |
Text
r March 10, 1988 MEMORANDUM FCR: Charles E. Rossi, Director Division of Operational Events Assessment, NRR FROM Edward J. Butcher, Chief Technical Specifications Branch Division of Operational Events Assessment, NRR
SUBJECT:
TECHNICAL SPECIFICATIONS IMPR0VEMEliT PROGRAM DATE & TIME:
Tuesday, March 29, 1988 9:00 a.m. - 11:00 a.m.
LOCATION:
One White Flint North Building Room 12B11 11555 Rockville Pike Rockville, Maryland 20852 PURPOSE:
For NRC Management and Utility Owners Groups Managerent Representatives to discuss the NRC Staff's Draft Technical SpecificationsSplitReport(Attacheo)
NRC INDUSTRY PARTICIPANTS:
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A. Thadani N. Rutherford (B&WOG)
C. E. Rossi L. Shao J. Richardson D. Grace (BWROG) fWOG))
J. Stohr R. Barrett R. Newton (CEOG J. Gasper T. Tipton (NUMARC)
Original Signed By:
Edward J. Butcher Edward J. Butcher, Chief Technical Specifications Branch Division of Ope.ational Events Assessment, NRR (MEETING NOTICE MEMO)
Distribution:
Murley/Sniezek BBHayes, 01 PDR TCox SAVarga VStello, E00 ADT CHBerlinger FMiraglia MClausen, OC TSB Members WDLanning TMartin SDEbneter, OSP
\\ TSB R/F WKenneday DMCrutchfield JLieberman, OE
\\ Central Files JGPartlow WGMcDonald, 0ARti TSB S/F - (MEETING NOTICE)
FCongel WBKerr, SDBU/CR Regional Administrators FGillespie ELJordan, AEOD JRoe PEBird OP SRConnelly, 01A HLThompson, i4 MSS g
WCParler, CGC NRC Participants SJChilk, SECY TSB: DOE RR TSB:00EA:NRR C
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NRC STAFF REVIEW 0F NUCLEAR STEAM SUPPLY SYSTEM VENDOR OWNERS GROUPS' APPLICATION OF
-THE C0 MISSION'S INTERIM POLICY STATEMENT CRITERIA TO STANDARD TECHNICAL SPECIFICATIONS
4 1.
INTRODUCTION On February 6,1987, the Commission issued its Interim Policy Statement on Technical Specification Improvements (52 FR 3788).
The Policy Statement encourages the industry to develop new Standard Technical Specifications (STS) to be used as guides for licensees in preparing improved Technical Specifications (TS) for their facilities.
The Interim Policy Statement contains criteria (including a dis:ussion of t'ach) for determining which regulatory requirements and operating restrictions should be retained in the new STS and ultimately in plant TS.
It also identifies four additional systems th.t are to be retained on the basis of operating experience and probabi1Mic risk assessments (PRA).
Finally, the Policy Statement indicates that risk evaluations are an appropriate tool for defining requirements that should be retained in the STS/TS where including such requirements is consistent with the purpose of TS (as stated in the Policy Statement).
Requirements that are not retained in the new STS would generally not be retained in individual plant TS.
Current TS requirements not retained in the STS will be relocated to other licensee-controlled documents.
One of the first steps in the program to implement the Commission's Interim Policy Statement is to determine which Limiting Conditions for Operation (LCOs) contained in the existing STS should be retained in the new STS.
An early decision on this issue will facilitate efforts to make the other improvements (described in the Policy Statement) to the text and Bases of those requirements that must be retained in the new STS.
Each Nuclear Steam Supply System (NSSS) vendor Owners Group has submitted a report to the NRC for review that identifies which STS LCOs the group believes should be retained in the new STS and which can be relocated to other licensee-controlled documents.
These four NSSS vendor submittals are as follows:
(1) A letter dated October 15, 1987, R. L. Gill, B&W Owners Group, to Dr. T. E. Murley, NRC,
Subject:
"B&W Owners Group Technical Specification Committee Application of Selection Criteria to the B&W Standard Technical Specifications."
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1 (2) A letter dated November 12, 1987, R. A. Newton, Westinghouse Owners Group, to NRC Document Control Desk,
Subject:
"Westinghouse Owners Group HERITS Program Phase II, Task 5, Criteria Application Topical Report."
(3) A letter dated December 11, 1987, J. K. Gasper, Combustion Engineering Owners Group, to Dr. T. E. Murley, NRC
Subject:
"CEN-355, CE Owners Group Restructured Standard Technical Specifications - Volume 1 (Criteria Application)."
(4) A letter dated November 12, 1987, R. F. Janecek, BWR Owners Group, to R. E. Starostecki, NRC,
Subject:
"BWR Owners Group Technical Specification screening Criteria Application and Risk Assessment."
These submittals provide the rationale for why each STS requirement (e.g.
Limiting Condition for Operation) should be retained in the new STS or why it can be relocated to a licensee-controlled document.
They also describe how each Owners Group used risk insights in determining the appropriate content of the new STS.
2.
STAFF REVIEW The staff focused its review on those requirements identified by the Owners Groups as candidates for relocation.
The staff evaluated each of these requirements to determine whether it agreed with the Owners Groups' conclusions.
During the NRC Staff's review, several issues were raised concerning the proper interpretation or application of the criteria in the Commission's Interim Policy The NRC Staff has considered these issues and concluded the following:
Statement.
. (1) Criterion 1 should be interpreted to include only instrumentati.on used to detect actual leaks and not more broadly to include instrumentation used to detect precursors to an actual breech of the reactor coolant pressure boundary or instrumentation to identify the source of actual leakage (e.g.,
loose parts monitor, seismic instrumentation, valve position indicators).
(2) The "initial conditions" captured under Criterion 2 should not be limited to only "process variables" assumed in safety analyses.
They should also include certain active design features (e.g., high pressure / low pressure system valves and interlocks) and operating restriction (e.g., pressure-temperature operating limit curve, needed to preclude unanalyzed accidents; In this context, "active design features" includes only design features under the control of operations personnel (i.e., licensed operators and personnel who perform control functions at the direction of licensed opera-tors).
This position is consistent with the conclusions reached by the Staff during the trial application of the criteria to the Wolf Creek and Limerick Technical Specifications.
(3) The "initial conditions" of design-basis accidents (DBA) and transients, as used in Criterion 2, should not be limited to only those directly "monitored and controlled" from the control room.
Initial conditions should also in-clude other features / characteristics that are specifically assumed in DBA and transient analyses even if they can not be directly observed (e.g.,
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moderator temperature coefficient, shutdown margin).
Initial conditions do not, however, include things that are purely design requirements.
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DRUT (4) The phrase "primary success path," used in Criterion 3, should be interpreted to include only the primary equipment (including redundant trains / components) to mitigate accidents and transients.
Primary success path does not include backup and diverse equipment or instrumentation used to prevent analyzed accidents or transients or to improve reliability of the mitigation function (e.g., rod withdrawal block which is backup to the average power range monitor high flux trip in the startup mode, safety valves which are backup to low temperature over pressure relief valves during cold shutdown).
(5) Post-Accident Monitoring Instrumentation that satisfies the definition of Type A variables in Regulatory Guide 1.97, "Instrumentation for Water,
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," meets Criterion 3 and should be retained in Technical Specifications.
Type A variables provide primary information (i.e., information that is essential for the direct accomplishment of the specified manual actions (including long-term recovery actions) for which no automatic control is provided and that is required for safety systems to accomplish their safety functions for DBAs or transients. Type A variables do not include those variables associated with contingency actions that may also be identified in written procedures to compensate for failures of primary equipment.
The STS should contain a narrative statement that indicates that individual plant Technical Specifications should contain a list of Post-Accident Instrumentation that includes only Type A variables.
The Type A variables at each operating power reactor were identified to the NRC in licensees' responses to Generic Letter 82-33.
The result of the staff's review of these responses will form the basis for the plant-specific list of Post-Accident Monitoring Instrumentation to be included in Technical Specifications. This plant specific list should be included in the Post-Accident Monitoring Instrumentation Specification regardless of whether or not the instrumentation is included in other Specifications.
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' (6) The NRC's design basis for licensing a plant is the plant's Fin.al Safety Analysis Report (FSAR) as supplemented by the analysis performed by the Staff in accordance with the acceptance criteria in the NRC's Standard Review Plan (NUREG-0800, SRP) and documented in the Staff's safety evaluation report (SER).
For example, the dose limits used in licensing the plant may be "some small fraction" of those specified in the Commission's regulations in Title 10 of the Code of Federal Regulations Part 100 (10 CFR 100).
These are the limits that should be used to define the equipment in the primary success path for mitigating accidents and transients. These types of conservatisms are required to compensate for uncertainties in analysis techniques and provide reasonable assurance that the absolute numerical limits of the regulations:will be satisfied.
Consequently, systems and equipment that are assumed by the NRC staff in its safety review as functioning to meet the review acceptance criteria in the SRP should be captured by Criterion 3 (e.g., radiation monitoring instrumentation that initiates an isolation function, penetration room exhaust air cleanup system).
(7) DBA and transients, as used in Criteria 2 and 3, should be interpreted to include any design-basis event described in the FSAR (i.e., not just those events described in Chapters 6 and 15 of the FSAR).
For example, there may be requirements for some plants which should be retained in Technical Speci-fications because the risks associated with some site-specific characteristic (e.g., although not normally required a Technical Specification on the chlorine detection system might be appropriate where a significant chlorine hazard exists in the site vicinity; similarly, a Technical Specification on flood protection might be appropriate where a plant is particularly vulner-able to flooding and is designed with special flood' protection features).
Criteria 2 and 3 should not be interpreted to include purely generic design requirements applicable to all plants (e.g., the requirements of General Design Criterion 19 in Appendix A to 10 CFR Part 50 for control room design).
The NRC staff has used the Commission's Interim Policy Statement and the conclusions described above to define the appropriate content of the new STS.
The staff plans to factor these conclusions into the final Policy Statement on Technical Specification Improvements that will be proposed to the Commis3 ion.
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DREI
- The staff reviewed the methodology used by each Owners Group to veri.fy that none of the requirements proposed for relocation contains constraints of prime importance in limiting the likelihood or severity of accident sequences that are commonly found to dominate risk.
The Staff found the approach of each Owners Groups to be reasonable.
The staff also reviewed each STS requirement being considered for relocation to ensure that no requirement of prime importance in limiting the likelihood or severity of an accident is relocated.
LCOs retained in Technical Specifications solely on the basis of risk are identified in Table 1 of Appendices A through D to this report ("Risk" is specified in the criteria column.)
As stated in the Commission's Interim Policy Statement, licensees should use plant-specific PRAs or risk surveys as they prepare license amendments to adopt the revised STS to their plant.
Where PRAs or surveys are available, licensees should use them to strengthen the Bases as well as to screen those Technical Specifications to be relocated.
Where such plant-specific risk surveys are not available, licensees should use the literature available on risk insights and PRAs.
Licensees need not complete a plant-specific PRA before they can adopt the new STS.
The NRC staff will also use risk insights and PRAs in evaluating the plant-specific submittals.
3.
RESULT 5 0F THF STAFF'S REVIEW Appendices A through D present the detailed results of the Staff's review of the Babcock and Wilcox, Westinghouso, Combustion Engineering, and General Electric application of the selection criteria to the existing STS.
Each Appendix consists of two tables.
Table 1 identifies those LCOs that must be retained in the new STS.
Table 2 lists those LCOs that may be relocated to
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licensee-controlled documents. Where the staff placed specific conditions on relocation of particular LCOs the staff has so noted in the Tables. As a part of the plant specific implementation of the new STS, the staff plans to review the location of, and controls over, relocated requirements.
In as much as practicable, the Owners Groups should propose standard locations for, and controls over, relocated requirements.
i DRAFT For each LCO listed in Table 1, the criterion (criteria) that required that the LC0 be retained in Technical Specifications is identified.
If an LC0 was retained in Technical Specification solely on the basis of risk, "Risk" appears in the criteria column. Where an Owners Group determined that an LC0 had to stay in Technical Specifications (because of either a particular criterion or risk) and the Staff agreed that the LC0 should be retained in Technical Specif-ications, the staff did not, in general, verify the Owners Group's basis for retention. However, in several instances the Owners Groups cited risk consider-ations alone as the basis for retaining Technical Specifications and the staff disagreed with the Owners Groups.
In these instances, the staff's basis for retention appears in the criteria column of Table 1.
Any 1.00 not specifically identified in Table 1 or Table 2 (e.g., an LC0 unique to an STS not addressed in the Owners Groups submittals such as the BWR5 STS) should be retained in Technical Specification until industry proposes and the staff makes a specific determination that it can be relocated to a licensee-controlled document.
Notwithstanding the results of this review, the staff will give further consideration for relocation of additional LCOs as the staff and industry proceed with the develcpment of the new STS.
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4.
CONCLUSION The results of the effort of the Owners Groups and of the NRC staff to apply the Policy Statement selection criteria to the existing STS are an important i
I step toward ensuring that the new STS contain only those requirements that are j
consistent with 10 CFR 50.36 and have a sound safety basis.
As shown in the following table, application of the criteria contained in the Commission's l
Interim Policy Statement resulted in a significant reduction in the number of l
LCOs to be included in the new STS.
The development of the new STS based on the staff's conclusions will result in more efficient use of NRC and industry resources. Safety improvements are expected through more operator-oriented Technical Specifications, improved Technical Specification Bases, a reduction i
in action statement-induced plant transients, and a reduction in testing at
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BABC0CK GENERAL COMBUSTION ELECTRIC LCOs WILCOX WESTINGHOUSE ENGINEERING BWR4/BWR6 Toial Number 137 165 159 124/144 Retained 75 92 88 81/86 Relocated 62 73 71 43/58 Percent Relocated 45%
44%
45%
35%/40%
We are confident that the staff's conclusions will provide an adequate basis for the Owners Groups to proceed with the development of complete new STS in accordance with the Commission's Interim Policy Statement.
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APPENDIX A a
RESULTS OF THE NRC STAFF REVIEW BABC0CK & WILCOX OWNERS GROUP'S SUBMITTAL RETENTION AND RELOCATION OF SPECIFIC TECHNICAL SPECIFICATIONS O
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DRMT APPENDIX A TABLE 1 LCOs TO BE RETAINED IN BABC0CK & WILCOX STANDARD TECHNICAL SPECIFICATIONS 3
LC0 CRITERIA 3.1 REACTIVITY CONTROL SYSTEM 3.1.1.1 Shutdown Margin (Note 1) 2 3.1.1.2 Moderator Temperature Coefficient 2
3.1.1.3 Minimum Temperature for Criticality 2
3.1.3.1 Group Height - Safety and Regulating Rod Groups 2
3.1.3.2
. Group Height - Axial Power Shaping Rod Group 2
3.1.3.6 Safety Rod Insertion Limit 2&3 3.1.3.7 Regulating Rod Insertion Limits 2
3.1.3.9 Xenon Reactivity 2
3.2 POWER DISTRIBUTION LIMITS 3.2.1 Axial Power Imb.ilance 2
3.2.2 Nuclear Heat Flux Hot Channel Factor 2
3.2.3 Nuclear Enthalpy Rise Hot Channel Factor 2
3.2.4 Quadrant Power Tilt 2
3.2.5 DNB Parameters 2
3.3 INSTRUMENTATION 3.3.1 Reactor Protection System Instrumentation (Note 2) 3 3.3.2 Engineered Safety Feature Actuation System Instrumentation (Note 2) 3 3.3.3.1 Radiation Monitoring Instrumentation (Notes 2 & 3) 3 3.3.3.5 Remote Shutdown Instrumentation (Note 2)
Risk 3.3.3.6 Accident Monitoring Instrumentation (Note 2) 3 3.4 REACTOR COOLANT SYSTEM l
3.4.1.1 Startup and Power Operation 3
3.4.1.2 Hot Standby 3
3.4.1.3 Hot Shutdown 3
3.4.1.4 Cold Shutdown Policy Statement (DHR) 3.4.3 Safety Valve - Operating 3
l 3.4.4 Pressurizer 2&3 l
3.4.5 Relief Valve 2
3.4.6 Steam Generators - Water Level 2
3.4.7.1 Leakage Detection System 1
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LCO CRITERIA 3.4.7.2 Operational Leakage 2
3.4.9 Specific Activity 2
3.4.10.1 Reactor Coolant System Pressure / Temperature Limits 2
3.4.10.3 Overpressure. Protection System 2
3.5 EMERGENCY CORE COOLING SYSTEM (ECCS) 3.5.1 Core Flooding Tanks 2&3 3.5.2 ECCS Subsystems - T
> (305) F 3
yg 3.5.3 ECCS Subsystems - Tu g <(305) F 3
3.5.4 Borated Water Storage Tank 2&3
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3.6 CONTAINMENT SYSTEMS 3.6.1.1 Containment Integrity 3
3.6.1.3 Containment Air Locks 3
- 3. 6.1. 5 Internal Pressure 2
3.6.1.6 Air Temperature 2
- 3. 6.1. 8 Containment Ventilation System 3
3.6.2.1 Containment Spray System 3
3.6.2.2 Spray Additive System 2&3 3.6.2.3 Containment Cooling System 3
3.6.3 Iodine Cleanup System 3
3.6.4 Containment Isolation Valves 3
3.6.5.1 Hydrogen Analyzers 3
3.6.5.2 Electric Hydrogen Recombiners 3
3.6.6 Penetration Room Exhaust Air Cleanup System 3
3.7 PLANT SYSTEMS 3.7.1.1 Safety Valves 3
3.7.1.2 Auxiliary Feedwater System 3
3.7.1.3 Condensate Storage Tank 2&3 3.7.1.4 Activity 2
3.7.1.5 Main Steam Line Isolation Valves 3
3.7.3 Component Cooling Water System 3
3.7.4 Service Water System 3
l 3.7.5 Ultimate Heat Sink 3
3.7.6 Flood Protection (optional) 3 l
3.7.7 Control Room Emergency Air Cleanup System 3
l 3.7.8 ECCS Pump Room Exhaust Air Cleanup System 3
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DRMI B&W-TABLE 1 (Continued)
LCO CRITERIA 3.8 ELECTRICAL POWER SYSTEMS 3.8.1.1 A.C. Sources - Operating 3
3.8.1.2 A.C. Sources - Shutdown Policy Statement (DHR) 3,8.2.1 A.C.. Distribution - Operating 3
3.8.2.2 A.C. Distribution - Shutdown Policy ftatement (DHR) 3.8.2.3 0.C. Distribution - Operating 3
3.8.2.4 D.C. Distribution - Shutdown Policy Statement (DHR) 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration 2
3.9.2 Instrumentation 2
3.9.3 Decay Time 2
3.9.4 Containment Building Penetration 3
3.9.8.1 Residual Heat Removal and Coolant Circulation -
All Water Levels Policy Statement (DHR) 3.9.8.2 Residual Heat Removal and Coolant Circulation -
Low Water Levels Policy Statement (DHR) 3.9.9 Containment Purge and Exhaust Isolation System 3
3.9.10 Water Level - Reactor Vessel 2
3.9.11 Water Level - Storage Pool 2
3.9.12 Storage Pool Air Cleanup System 2
Notes:
1.
Required for Modes 3 through 6.
May be relocated for Modes 1 and 2.
2.
The LC0 for this system should be retained in STS.
The Policy Statement criteria should not be used as the basis for relocating specific trip functions, channels, or instruments within these LCOs.
3.
These specifications may be included with RETS if RETS are relocated as a program requirement in Section 6.8 of the STS (See Note 3 of Table 2).
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3.1.3.3 Position Indication Channels - Shur 46wn (Note 2).s 3.1.3.4 O
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B&W-TABLE 2 (Continued) 3.7.11.1 Fire Suppression Water System 3.7.11.2 Spray and/or Sprinkler Systems 3.7.11.3 C0 System 3
3.7.11.4 HaTon System 3.7.11.5 Fire Hose Stations 3.7.11.6 Yard Fire Hydrants and Hydrant Hose Houses 3.7.12 Fire Barrier Penetrations 3.7.13 Area Terrperature Monitoring 3.9 REFUELING OPERATIONS 3.9.5 Communice.tions 3.9.6 Fuel Handling Bridge 3.9.7 Crane Travel - Spent Fuel Storage Pool Building 3.10 SPECIAL TEST EXCEPTIONS 3.10.1 Shutdown Margin (Note 5) 3.10.2 Group Height Insertion Limits and Power Dist.ribution Limits (Note 5) 3.10.3 Physics Tests (Note 5) 3.10.4 Reactor Coolant Loops (Note 5) 3.11 RADI0 ACTIVE EFFLUENTS (Note 3) 3.11.1.1 Concentration 3.11.1.2 Dose 3.11.1.3 Liquid Radwaste Treatment System 3.11.1.4 Liquid Holdup Tanks 3.11.2.1 Dose 3.11.2.2 Dose - Noble Gases
-3.11.2.3 Dose - Iodine - 131, Tritium and Radionuclides in Particulate Form 3.11.2.4 Gaseous Radwaste Treatment Systems 3.11.2.5 Explosive Gas Mixture l
3.11.2.6 Gas Storage Tanks 3.11.3 Solid Radioactive Waste 3.11.4 Total Dose l
l 3.12 RADIOACTIVE ENVIRONMENTAL MONITORING 3.12.1 Monitoring Program 3.12.2 Land Use Census 3.12.3 Interlaboratory Comparison Program l
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DRAFT B&W-TABLE 2 (Continued)
Notes
- 1. Specifications listed in this table may be relocated contingent upon NRC staff approval of the location of and controls over relocated requirements.
- 2. This LC0 may be removed from the STS.
However, the associated Surveillance Requirement (s) should be relocated to an LCO which meets one of the criteria if the Surveillance Requirement is required in order to meet the operability requirements for the retained LCO.
- 3. This LCO may be relocated.
However, it must be controlled as a program requirement of Section 6.6 of STS.
- 4. This LC0 may be relocated.
However, Pa, La, Ld, and Lt may be either retained in TS or in the Bases of the appropriate Containment LCO.
- 5. Special Test Exceptions may be included with corresponding LCOs.
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DRAFT Li APPENDIX B RESULTS OF THE NRC STAFF REVIEW WESTINGHOUSE OWNERS GROUP'S SUBMITTAL RETENTION AND RELOCATION OF SPECIFIC. TECHNICAL SPECIFICATIONS 4
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APPENDIX B TABLE 1 LCOs TO BE RETAINED IN WESTINGHOUSE STANDARD TECHNICAL SPECIFICATIONS LC0 CRITERIA 3.1' REACTIVITY CONTROL SYSTEMS 3.1.1.1 Shutdown Margin - Tave >200 deg. F (Note 1) 2 3.1.1.2 Shutdown Margin - Tave < 200 deg. F (Note 1) 2 3.1.1. 3 Moderator Temperature Coefficient 2
3.1.1.4 Minimum Temperature for Criticality 2
3.1.3.1 Moveable Control Assemblies - Group Height 3
3.1.3.5 Shutdown Rod Insertion Limit 2
3.1.3.6 Control Rod Insertion Limits 2
3.2 POWER DISTRIBUTION LIMITS 3.2.1 Axial Flux Difference 2
3.2.2 Heat Flux Hot Channel Factor 2
3.2.3 RCS Flow Rate and Nuclear Enthalpy Rise Hot Channel 2
Factor 3.2.4 Quadrant Power Tilt Ratio 2
3.2.5 DNB Parameters 2
3.3.
INSTRUMENTATION 3.3.1 Reactor Trip System Instrumentation (Note 2) 3
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3.3.2 Engineered Safety Feature Actuation System 3
Instrumentation (Note 2) l 3.3.3.1 Radiation Monitoring Instrumentation (Note 2) 1&3 3.3.3.5 Remote Shutdown Instrumentation (Note 2)
Risk 3.3.3.6 Accident Monitoring Instrumentation (Note 2) 3 3.4 REACTOR COOLANT SYSTEM 3.4.1.1 RCS Startup and Power Operation 3
3.4.1.2 RCS Hot Standby 3
3.4.1.3 RCS Hot Shutdown 3
l 3.4.1.4.1 RCS Cold Shutdown - Loops Filled 3
3.4.1.4.2 RCS Cold Shutdown - Loops Not Filled 3
3.4.1.5 RCS Isolated Loop (Optional) 2 3.4.1.6 RCS Isolated Loop Startup (Optional) 2 3.4.2.2 RCS Safety valves - Operation 3
3.4.3 Pressurizer 2&3 3.4.4 Relief Valves (minus PORV block valves) 3 l
3.4.6.1 Leakage Detection System 1
3.4.6.2 Operational Leakage 2
3.4.8 Specific Activity 2
L 3.4.9.1 Pressure / Temperature Limits - RCS 2
3.4.9.3 Overpressure Protection Systems 2
B-1 J
f..
,'.4%
.(
W-TABLE 1 (Continued)
CRITERIA LCO 3.5 EMERGENCY' CORE COOLING SYSTEMS 3.5.1.1 Cold Leg Injection Accumulators 2&3 3.5.1.2 Upper Head Injection Accumulators (STS REV-5) 2&3 3.'5.2 ECCS Subsystems, Tavg > 350 deg F 3
3.5.3 ECCS Subsystems, Tavg < 350 deg F 3
3.5.4.1 Boron Injection Tank 2&3 3.5.5 Refueling Water Storage Tank 2&3 t:
3.6 CONTAINMENT SYSTEMS 3.6.1.1 Containment Integrity 3
3.6.1.3 Containment Air Locks 3
3.6.1.4 Containment Isolation Valve and ChanneS Weld 3
Pressurization System (Optional) 3.6.1.5 Internal Pressure 2
2 3.6.1.6 Air Temperature 3
3.6.1.8 Containment Ventilation System 3.6.1.9' Shield Building Air Cleanup System (Ice Condenser) 3 3.6.2.1 Containment Quency Spray System (Sub-ATM Containment) 3 3.6.2.1 Containment Spray System 3
3.6.2.2 Containment Recirculation Spray System (Sub-ATM 3
Containment) 3.6.2.2 Spray Additive System (Optional) 2&3 3.6.2.3 Containment Cooling System (Optional) 3 3.6.3 Iodine Cleanup System (Optional) 3 3.6.4 Containment Isolation Valves (minus response time) 3 3.6.5.1 Hydrogen Monitors 3
3.6.5.2 Electric Hydrogen Recombiners 3
3.6.5.3-Hydrogen Control Distributed Ignition System (STS 3
REV-5, Ice Condenser) 3.6.5.4 Hydrogen Mixing System (Optional) 3
{
3.6.6 Penetration Room Exhaust Air Cleanup System (Optional) 3 3.6.7 Vacuum Relief Valves 3
3.6.7.1 Ice Bed (Ice Condenser) 2&3 3.6.7.3 Ice Condenser Doors (Ice Condenser) 2&3 3.6.7.5 Divider Barrier Personnel Access Doors and Equipment 2&3 Hatches (Ice Condenser) 3.6.7.6 Containment Air Recirculation Systems (Ice Condenser) 2&3 3.6.7.7 Floor Drains (Ice Condenser) 2&3 3.6.7.8 Refueling Canal Drains (Ice Condenser) 3 3.6'.7.9 Divider Barrier Seal (Ice Condenser) 2&3 3.6.8.1 Shield Building Air Cleanup System (Dual) 3 3.6.8.2 Shield Building Integrity (Oual) 3 B-2 i
DRAFT W-TABLE 1 (Continued)
LCO CRITERIA 3.7 PLANT SYSTEMS 3.7.1.1 Turbine Cycle Safety Valves 3
3.7.1.2 Auxiliary Feedwater System 2&3 3.7.1.3 Condensate Storage Tank 2&3 3.7.1.4 Activity 2
3.7.1.5 Main Steam Line Isolation Valves 3
3.7.3 Component Cooling Water System 3
3.7.4 Service Water System 3
3.7.S Ultimate Heat Sink (Optional) 3 3.7.7 Control Room Emergency Air Cleanup System 3
3.7.8 ECCS Pump-Room Emergency Air Cleanup System 3
3.8 ELECTRICAL POWER SYSTEMS 3.8.1.1 A.C. Sources - Operating 3
3.8.1.2 A.C. Sources - Shutdown 3
3.8.2.1 0.C. Sources - Operating 3
3.8.2.2 D.C. Sources - Shutdown 3
3.8.3.1 Onsite Power Distribution - Operating 3
3.8.3.2 Onsite Power Distribution - Shutdown 3
3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration (Some plants) 2 3.9.2 Instrumentation (Some plants) 3 3.9.3 Decay Time 2
3.9.4 Containment Building Penetrations 3
3.9.8.1 Residual Heat Removal and Coolant Circulation - High Water Level Policy Statement (RHR) 3.9.8.2 Residual Heat Removal and Coolant Circulation - Low Water Level Policy Statement (RHR) 3.9.9 Containment Purge and Exhaust Isolation System 3
3.9.10 Water Level - Reactor Vessel 2
3.9.11 Water Level - Storage Pool 2
j 3.9.12 Storage Pool Air Cleanup System 3
Notes:
- 1. Required for Modes 3 through 6.
May be relocated for Modes 1 and 2,
- 2. The LC0 for this system should be retained in STS.
The Policy Statement criteria should not be used as the basis for relocating specific trip functions, channels, or instruments within these LCOs.
I
requirement of Section 6.8 of STS (See Note 3 of Table 2).
l B-3 i
.,.__,___..7,
DRAFT TABLE 2 (Note 1)
WESTINGHOUSE STANDARD TECHNICAL SPECIFICATIONS LCOs WHICH MAY BE RELOCATED LC0 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2.1 Flow Paths - Shutdown 3.1.2.2 Flow Paths - Operating 3.1.2.3 Charging Pumps - Shutdown 3.1.2.4 Charging pumps - Operating 3.1.2.5 Borated Water Sources - Shutdown 3.1.2.6 Borated Water Sources - Operating 3.1.3.2 Position Indication System - Operating (Note 2) 3.1.3.3 Position Indication System - Shutdown (Note 2) 3.1.3.4 Rod Drop Time (Note 2) 3.3 INSTRUMENTATION 3.3.3.2 Movable Incore Detectors 3.3.3.3 Seismic Instrumentation 3.3.3.4 Meteorological Instrumentation 3.3.3.7 Chlorine Detection Systems 3.3.3.8 Fire Detection Instrumentation 3.3.3.9 Loose-Part Detection Instrumentation 3.3.3.10 Radioactive Liquid Effluent Monitoring Instrumentation (Note 3) 3.3.3.11 Radioactive Gaseous Effluent Monitoring Instrumentation (STS REV - 5) (Note 3) 3.3.4 Turbine Overspeed Protection 3.4 REACTOR COOLANT SYSTEM 3.4.2.1 RCS Safety Valves - Shutdown 3.4.4 Relief Valves (PORV block valves) 3.4.5 Steam Generators (Note 2) 3.4.7 Chemistry 3.4.9.2 Pressure / Temperature Limits - Pressurizer 3.4.10 RCS Structural Intgerity (Note 2) 3.4.11 Reactor Coolant System Vents (STS REV-5) 3.5 EMERGENCY CORE COOLING SYSTEMS 3.5.4.2 Heat Tracing B-4
s W-TABLE 2 (Continued)
LCO 3.6 CONTAINMENT SYSTEMS 3.6.1.2 Containment Leakage (Note 4) 3.6.1.7 Containment Structural Integrity (Note 2) 3.6.1.8 Shield Building Structural Integrity (Ice Condenser) (Note 2) 3.6.4 Containment Isolation Valves (response times) (Note 2) 3.6.5.1 Steam Jet Air Ejector (Sub-ATM Containment) 3.6.5.2 Mechanical Vacuum Pumps (SUB-ATM. Containment) 3.6.5.3 Hydroden Purge Cleanup System 3.6.7.2 Ice Bed Temperature Monitoring System (Ice Condenser) 3.6.7.4 Inlet Door Position Monitoring System (Ice Condenser) 3.6.8.3 Shield Building Structural Integrity (Dual) 3.7 PLANT SYSTEMS 3.7.2 Steam Generator Pressure / Temperature Limitation 3.7.6 Flood Protection (Optional) 3.7.9 Snubbers 3.7.10 Sealed Source Contamination 3.7.11.1 Fire Suppression Water System 3.7.11.2 Spray and/or Sprinkler Systems 3.7.11.3 CO2 Systems 3.7.11.4 Halon Systems 3.7.11.5 Fire Hose Stations 3.7.11.6 Yard Fire Hydrants and Hydrant Hose Houses 3.7.12 Fire Rated Assemblies 3.7.13 Area Temperature Monitoring 3.8 ELECTRICAL POWER SYSTEMS 3.8.4.1 A.C. Circuits Inside Primary Containment (STS REV-5) 3.3.4.2 Containment Penetration Conductor Overcurrent Protective Devices 3.8.4.3 Motor-0perated Valves Thermal Overload Protection and Bypass Devices 3.9 REFUELING OPERATIONS 3.9.5 Communications 3.9.6 Manipulator Crane 3.9.7 Crane Travel - Spent Fuel Storage Pool 3.10 SPECIAL TEST EXCEPTIONS (Note 5)
B-5
W-TABLE 2 (Continued)
LCO 3.11 RADIOACTIVE EFFLUENTS 3.11.1.1 Liquid Effluents Concentration (STS REV-5) (Note 3) 3.11.1.2 Dose (STS REV-5) (Note 3) 3.11.1.3 Liquid Radwaste Treatment System (STS REV-5) (Note 3) 3.11.1.4 Liquid Holdup Tanks (STS REV-5) (Note 3) 3.11.2.1 Dose Rate (STS REV-5) (Note 3) 3.11.2.P.
Dose - Noble Gases (STS REV-5) (Note 3) 3.11.2.3 Dose I-131, I-133, Tritium and Radioactive Material In Particulate Form (Note 3) 3.11.2.4 Gaseous Radwaste Treatment (STS REV-5) (Note 3) 3.11.2.5 Explosive Gas Mixture (STS REV-5) (Note 3) 3.11.2.6 Gas Storage Tanks (Note 3) 3.11.3 Solid Radioactive Waste (STS REV-5) (Note 3) 3.11.4 Total Dose (STS REV-5) (Note 3) 3.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3.12.1 Monitcring Program (STS REV-5) (Note 3) 3.12.2 Land Use Census (STS REV-5) (Note 3) 3.12.3 Interlaboratory Comparison Program (STS REV-5) (Note 3)
Notes:
- 1. LCOs listed in thi. table may be relocated contingent upon NRC staff approval of the location of and controls over relocated requirements.
- 2. This LC0 may be removed from the STS.
Howevt.r, the associated Surveillance Requirement (s) should be relocated to an LCO which meets one of the criteria if the Surveillance Requirement is required in order to meet the operability requirements for the retained LCO.
- 3. This LCO may be relocated.
However, it must be controlled as a Program requirement of Section 6.8 of the STS.
- 4. This LC0 may be relocated.
However, Pa, La, Ld and Lt may be either retained in TS or in the Bases of the appropriate containment LCO.
- 5. Special Test exceptions 3.10.1 through 3.10.4 may be included with corresponding LCOs which are remaining in Technical Specifications.
Special Test Exception 3.10.5 may be relocated outside of Technical Specifications along with LCO 3.1.3.3.
l l
l
-m
DRAFT l
APPENDIX C RESULTS OF THE NRC STAFF REVIEW COMBUSTION ENGINEERING OWNERS GROUP'S SUBMITTAL RETENTION AND RELOCATION OF SPECIFIC TECHNICAL SPECIFICATIONS 4
DRAFT APPENDIX C TABLE 1 LCOs TO BE RETAINED IN COMBUSTION ENGINEERING STANDARD TECHNICAL SPECIFICATIONS LCO CRITERIA 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1.1 Shutdown Margin --Tcold. > 210F (Note 1) 2 3.1.1.2 Shutdown Margin - Tcold. < 210F (Note 1) 2 3.1.1.3 Moderator Temperature Coefficient 2
3.1.1.4 Minimum Temperature for Criticality 2
3.1.3.1 CEA Position 2&3 3.1.3.5 Shutdown CEA Insertion Limit 2
3.1.3.6 Regulating CEA Insertion Limits 2
3.1.3.7 Part Length CEA Insertion Limits 2
3.2 POWER DISTRIBUTION LIMITS 3.2.1 Linear Heat Rate 2
3.2.2 Planar Radial Peaking Factors--Fxy 2
3.2.3 Azimuthal Power Tilt -- Tq 2
3.2.4 DNBR Margin 2
3.2.5 RCS Flow Rate 2
3.2.6 Reactor Coolant Colo Leg Temperature 2
3.2.7 Axial Shape Index 2
3.2.8 Pressurizr:r Pres ce 2
3.3 INSTRUMENTATION 3.3.1 Reactor Protective Instrumentation (Note 2) 3 l
3.3.2 ESFAS Instrumentation (Note 2) 3 3.3.3.1 Radiation Monitoring Instrumentation (Notes 2 & 3) 3 I
3.3.3.5 Remote Shutdown System (Note 2)
Risk 3.3.3.6 Post-Accident Monitoring Instrumentation (Note 2) 3 3.4 REACTOR COOLANT SYSTEM 3.4.1.1 Startup and Power Operation 2&3 1
3.4.1.2 Hot Standby 2
3 1
3.4.1.3 Hot Shutdown 2( 3 l
3.4.1.4.1 Cold Shutdown - Loops filled 2&3 l
3.4.1.4.2.
Cold Shutdown - Loops not filled 2&3 C-1
DRAFI CE-TABLE 1 (Continued)
LCO CRITERIA 3.4.2.2 Safety Valves - Operating 3
3.4.3 Pressurizer 3
3.4.3.1 Pressurizer 2&3 3.4.6.1 Leakage Detection Systems 3
3.4.6.2 Operational Leakage 3
3.4.8 Specific Activity 2
3.4.9.1 Reactor Coolant System 2
3.4.9.3 Overpressure Protection Systems-LTOP 2
3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Safety Injection Tanks 3
3.5.2 ECCS Subsystems -- Tcold. > 350F 3
3.5.3 ECCS Subsystems -- Tcold. < 350F 3
3.5.4 Refueling Water Tank 3
3.6 CONTAINMENT SYSTEMS 3.6.1.1 Containment Integrity 3
3.6.1.3 Containment Air Locks 3
3.6.1.5 Internal Pressure 2
3.6.1.6 Air Temperature 2
3.6.1.8 Containment Ventilation System 3
3.6.2.1 Containment Spray System 3
3.6.2.2 Spray Additive System 3
3,6.2.3 Containment Cooling System 3
3.6.3 Iodine Cleanup System 3
3.6.4 Containment Isolation Valves 3
3.6.5.1 Hydrogen Monitors 3
~
3.6.5.2 Electric Hydrogen Combiners 3
3.6.5.3 Hydrogen Purge Cleanup System 3
3.6.5.4 Hydrogen Mixing System 3
3.6.5.5 Hydrogen Monitors 3
3.6.6 Penetration Room Exhaust Air Cleanup System 3
3.6.7 Vacuum Relief Valves 3
3.6.8.1 Shield Building Air Cleanup System 3
~
3.7 PLANT SYSTEMS i
3.7.1.1 Safety Valves 3
3.7.1.2 Auxiliary Feedwater System 3
3.7.1.3 Condensate Storage Tank 3
3.7.1.4 Activity 3
3.7.1.5 Main Steam Isolation Valves 3
C-2
i
-4 i
6 uL CE-TABLE 1 (Continued)
LCO CRITERIA 3.7.3 Component Cooling Water System 3
3.7.4 Service Water System 3
3.7.5 Ultimate Heat Sink 3
3.7.7 Essential Chilled Water System 3
3.7.9.
ECCS Pump Room Air Exhaust Cleanup System 3
3.8 ELECTRICAL POWER SYSTEMS 3.8.1.1 A.C. Sources - Operating 3
3.8.1.2 A.C. Sources - Shutdown 3
3.8.2.1 D.C. Sources - Operating 3
3.8.2.2 D.C. Sources - Shutdown 3
3.8.3 Onsite Power Distribution Systems 3.8.3.1 Operating 3
3.8.3.2 Shutdown 3
3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration 2
3.9.2 Instrumentation 3
3.9.3 Decay Time 2
3.9.4 Containment Building Penetrations 3
3.9.8.1 Shutdown Cooling and Coolant Circulation -
High Water Level 2
3.9.8.2 Shutdown Cooling and Coolant Circulation -
Low Water Level 2
3.9.9 Containment Purge Valve Isolation System 3
3.9.10 Water Level-Reactor Vessel 2
3.9.11 Water Level-Storage Pool 2
3.9.12 Fuel Building Air Cleanup System 3
Notes:
1.
Required for Modes 3 through 6.
May be relocated for Modes 1 and 2.
2.
LCOs for this system should be retained in STS.
The Policy Statement Criteria should not be used to relocate specific trip functions, channels, or instruments within these LCOs.
3.
These LCOs may be included with RETS if RETS are relocated as a program requirement of Section 6.8 of STS (See Note 3 of Table 2).
C-3
DRAFT
~
TABLE 2 (Note 1)
COMBUSTION ENGINEERING STANDARD TECHNICAL SPECIFICATION LCOs WHICH MAY BE RELOCATED LC0 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2.1 Flow Paths -- Shutdown 3.1.2.2 Flow Paths-Operating 3.1.2.3 Charging Pumps -- Shutdown 3.1.2.4 Charging Pumps-Operating 3.1.2.5 Boric Acid Makeup Pumps -- Shutdown 3.1.2.6 Boric Acid Makeup Pumps-Operating 3.1.2.7 Borated Water Source - Shutdown 3.1.2.8 Borated Water Sources - Operating 3.1.3.2 Position Indicator Channels-Operating (Note 2) 3.1.3.3 Position Indicator Channels-Shutdown (Note 2) 3.1.3.4 CEA Drop Time (Note 3) 3.3 INSTRUMENTATION 3.3.3.2 Incore Detectors 3.3.3.3 Seismic Instrumentation 3.3.3.4 Meteorological Instrumentation 3.3.3.7 Fire Detection Instrumentation 3.3.3.8 Chlorine Detection Systems 3.3.3.9 Loose Part Detection Instrumentation 3.3.3.10 Radioactive Liquid Effluent Monitor (Note 3) 3.3.3.11 Radioactive Gaseous Effuent Monitor (Note 3) 3.3.4 Turbine Overspeed Protection 3.4 REACTOR COOLANT SYSTEM 3.4.2.1 Safety Valves-Shutdown 3.4.4 P.:It:f Valves 3A=5 Steam Generators (Note 2) 3.4.7 Chemistry 3.4.9.2 Pressurizer Heatup/Cooldown Limits 3.4.10 Structural Integrity (Note 2) 3.4.11 Reactor Coolant System Vents
- 3. 6 CONTAINMENT SYSTEMS 3.6.1.2 Containment Leakage (Note 4) 3.6.1.4 Containment Isolation Valve and Channel Weld Pressure System 3.6.1.7 Containment Vessel Structural Integrity (Note 2) 3.6.8.2 Shield Building Integrity 3.6.8.3 Shield Building Structural Integrity (Note 2)
C-4
CE-TABLE 2 (Continued)
LCO 3.7 PLANT SYSTEMS 3.7.2 Steam Generator Pressure / Temperature Limitation 3.7.6 Flood Protection 3.7. 8 Control Room Emergency Air Cleanup System 3.7.10 Snubbers 3.7.11 Sealed Source Contamination 3.7.12 Fire Suppression Systems 3.7.12.1 Fire Suppression Water System 3.7.12.2 Spray and/or Sprinkler Systems 3.7.12.3 CO2 Systems 3.7.12.4 Halon Systems 3.7.12.5 Fire Hose Stations 3.7.12.6 Yard Fire Hydrants and Hose Houses 3.7.13 Fire-Rated Assemblies 3.8 ELECTRICAL POWER SYSTEMS 3.8.4.1 Containment Penetration Conductor Overcurrent Protection Device 3.8.4.2 Motor-Operated Valves-Thermal Overload Protection 3.9 REFUELING OPERATIONS 3.9.5 Communication 3.9.6 Manipulator Crane (Refueling Machine) 3.9.7 Crane Travel - Spent Fuel Pool Building 3.10 SPECIAL TEST EXCEPTIONS 3.10.1 Shutdown Margin (Note 5) 3.10.2 Group Height, Insertion, and Power Dist. (Note 5) 3.10.3 Reactor Coolant Loops (Note 5) 3.10.4 CEA Position, Reg CEA Ins, and Cold Leg Temp. (Note 5) 3.11 RADIOACTIVE EFFLUENTS 3.11.1.1 Liquid Waste Discharge to Evap. Ponds -
Concentration (Note 3) 3.11.1.2 Liquid Waste Discharge to Evap. Ponds (Note 3) -
Dose (Note 3) 3.11.1.3 Liquid Holdup Tanks (Note 3) 3.11.2.1 Gaseous Effluents - Dose Rate (Note 3) 3.11.2.2 Gaseous Effluents - Dose-Noble Gases (Note 3) 3.11.2.3 Gaseous Effluents - Dose--I-131, 133, Tritium & Radionuclides (Note 3) i 3.11.2.4 Gaseous Radwaste Treatment (Note 3) l 3.11.2.5 Explosive Gas Mixture (Note 3) l 3.11.2.6 Gas Storage Tanks (Note 3) l 3.11.3 Solid Radioactive Waste (Note 3) 3.11.4 Total Dose (Note 3)
C-5
CE-TABLE 2 (Continued)
LCO 3.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3.12.1 Monitoring Program 3.1? 2 Land Use Census 3.12.3 Interlaboratory Comparison Program 1
1 Notes:
- 1. Specifications listed in this table may be relocated contingent upon NRC staff approval of the location of and controls over relocated requirements.
- 2. This LCO may be removed from the STS.
However, the associated Surveillance Requirement (s) should be relocated to an LCO which meets one of the criteria if the Surveillance Requirement is required in order to meet the operability requriements for the retained LCO.
- 3. This LCO may be relocated.
However, it must be controlled by as a Program requirement of Section 6.8 of the STS.
- 4. This LCO may be relocated.
However, Pa, La, Ld, and Lt may be either retained in TS or in the Bases of the appropriate containment LCO.
5.
Special Test Exceptions may be included with the corresponding LCOs.
C-6
4 i
DRAFT
~
t a
~
APPENDIX 0 RESULTS OF THE NRC STAFF REVIEW BWR OWNERS GROUP'S SUBMITTAL RETENTION AND RELOCATION OF SPECIFIC TECHNICAL SPECIFICATIONS t
O 6
?
e 1
i t
I f
i t
l
APPENDIX D TABLE 1 LCOs TO BE RETAINED IN GENERAL ELECTRIC STANDARD TECHNICAL SPECIFICATIONS REPORT LCO ITEM PLANT
- CRITERIA 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 1
ShutdownMargin(Note 1)
H.GG 2
3.1.3 Control Rods 3
Control Rods Operability H GG 3
5 HaximumScramTimes(BWR/6)
GG 3
l 6
Average Scram Times H
3 7
Fastest 3-out-of-4 Scram H
3 l
Times 8
Scram Accumulators H GG 3
9 Control Rod Drive Coupling H.GG 3
10 Control Rod Position H.GG 3
Indication 11 Control Rod Drive Housing H.GG 3
Support 3.1.4 Control Rod Progran Controls 12 RodWorthMinimizer(BWR/2-5)
H 3
13 ControlRodWithdrawal(BWR/6)
GG 2
14 Rod Pattern Control System GG 3
(BWR/6) 15 Rod Sequence Control Systems H
3 i
16 Rod Block Honitor i!
3
~
3.1.5 17 Standby Liquid Control System H,GGPolicyStatement(SBLC) 3.1.6 18 Scram Discharge Volume Vent H
3 ano Drain Valves 3.2 POWER DISTRIBUTION LIMITS 3.2.1 19 Average Planar Linear Heat H,GG 2
Generation (APLHGR) 3.2.3 21 Hinimum Critical Power Ratio H GG 2
(HCPR) 3.2.4 22 Linoar Heat Generation Rate H,GG 2
(LHGR)
- H-Hatch Unit 2 GG-Grand Gulf D-1
BWR-TABLE 1(Continued)
REPORT LCO ITEM PLANT CRITERIA 3.3 INSTRUMENTATION 3.3.1 Reactor Protection System Instrumentation (Note 2) 23 Ave' rage Power Range Monitors H.GG 3
(APRM) 24 Intermediate Range Monitors H,GG 3
(IRM) 25 Vessel Pressure - High H,GG 3
26 Reactor Vessel Water H.GG 3
Level-Low (Level 3) 27 Reactor Yessel Water GG 3
Level - High (Level 8) 28 MSIV Closure H GG 3
29 MSL Radiation - High H,GG 3
(RPSInst.)
30 Drywell Pressure - High H,GG 3
31 SDV Water Level - High H.GG 3
34 Mcde Switch H,GG 3
35 Manual Scram H,GG 3
3.3.2 Isclation Actuation Instrumentation (Note 2)
Primary Containment Isolation 36 Reactor Vessel Water H
3 Level-Low (Level 5) 37 Reactor Vessel Water H,GG 3
Level-Low (Level 2) 38 Reactor Vessel Water H,GG 3
Level - Low (Level 1) 39 Drywell Pressure - High H,GG 3
40 Containment and Drywell GG 3
Ventilation Exhaust Radiation - High High Main Steam Line Isolation 41 Manual Initiation GG 3
(Primary Containment) 42 Reactor Vessel Water GG 3
Level - Low (Level 1) 43 Main Steam Line Radiation -
H,GG 3
High(MSLI) 44 Hain Steam Line Pressure -
H,GG 3
Low 45 Main Steani Line Flow - High H,GG I&3 D-2
BWR-TABLE 1(Ccntinued)
REPORT LCo T1T PLANT CRITERM 46 Condenser Vacuum - Low H.GG 3
47 Main Steam Line Tunnel H,GG 1&3 Temperature - High 48 Main Steam Line Tunnel GG 1&3 Differential Temperature -
High 49 Manual Initiaticn (MSLI)
GG 3
50 Turbine Building Area h
1&3 Temperature - High Secondary Containment Isolation 51 Reactor Building Exhaust H
3 Radiation - High 52 Reactor Vessel Water H,GG 3
Level - Low (Level 2) 53 Drywell Pressure - High H.GG 3
54 Refueling Floor Exhaust H
3 Radiation - High 55 Fuel Handling Area GG 3
Ventilation Exhaust Radiation - High High 56 Fuel Handling Area Fool GG 3
Sweep Exhaust Radiation -
High High Reactor Water Cleanup System Isolation I
57 Manual Initiation GG 3
(Secondary Containment) 58 Differential Flow - High H,GG 1&3 59 Differential Flow Timer GG 2
60 Equipment Area H,GG 1&3 Temperature - High 61 Equiprent Area Differential H,GG 1&3 Temperature - High 62 Reactor Vessel Water H,GG 3
Level - (Level 2) 63 Main Steam Line Tunnel GG 1&3 Temperature - High 64 Main Steam Line Tunnel GG 1&3 Differential Temperature -
High 65 SLCS Initiaticn H.GG Policy Stater:ent (SBLC) 0-3 l
$U
~
BWR-TABLE 1(Continued)
REPORT PLANT CRITERIA LCO ITEn High Pressure Coolant Injection System Isolation 66 HanualInitiation(RWCS)
GG 3
67 HPCI Steam Line Flow - High H
1&3 68 HPCI Steam Supply H
3 Pressure - Low 69 HPCI Turbine Exhaust Diaphragm Pressure - High H
3 70 HPCI Pipe Penetration Room H
I&3 Temperature - High 71 Suppression Pool Area H
1&3 Ambient Temperature -
High 72 Suppression Pool Area H
1&3 Differential Temperature -
High 73 Suppression Poc1 Area H
2&3 Temperature Timer Relays 74 Emergency Area Cooler H
1&3 Temperature - High 76 Logic Power Monitor H
3 Reactor Core Isolation Cooling System Isolation 77 RCIC Steam Line Flow - High H,GG 1&3 78 RCIC Steam Supply H.GG Policy Statement (RCIG)
Pressure - Low 79 RCIC Turbine Exhaust H.GG Policy Statement (RCIC)
Diaphragm Pressure - High 80 RCIC Equipment Area H,GG 1&3 Temperature - High 81 Suppression Pool Area H
1&3 Ambient Temperature - High 82 Suppression Pcol Area H
1&3 Differential Temperature -
High 23 Suppression Pool Area H
2&3 Temperature Timer Relays 85 Logic Power Monitor H
3 86 RCIC Equipment Room GG 1&3 Differential Temperature -
High 87 Main Steam Line Tunnel CG 1&3 Temperature - High 68 Main Steam Line Tunnel GG 1&3 Differential Temperature -
l High I
D-4
a BWR-TABLE 1 Continued)
REPORT LCO ITEM PLAf!T CRITERIA 89 Main Steam Line Tunnel GG 3
Temperature Timer 90 RHR Equipment Room GG 1&3 Temperature - High 91 RHR Equipment Room GG 1&3 Differential Temperature -
High 92 RHR/RCIC Steam Line GG 1&3 Flow - High RHR System Iso 10 tion 93 ManualInitiation(RCIC)
GG 3
94 RHR Equipment Area GG 1&3 Temperature - High 95 RHR Equipment Room GG 1&3 Differential Temperature -
High 96 Reactor Vessel Water H,GG 3
Level - Low (Level 3) 97 Reactor Vessel (RHR Cut-In H,GG Policy Statement (RHR)
Permissive) Pressure -
High 98 Drywell Pressure - High GG Policy Statement (RHR) 99 ManualInitiation(RHR)
C1 3.3.3 ECCSActuationInstrumentation(Note 2)
RHR(LPCI/LPCS/CoreSpray) 1C0 Reactor Vessel Water H,GG 3
Level-Low (Level 1) 101 Drywell Pressure - High H,GG 3
102 RHR Pump Time Delay H GG 3
103 Manual Initiation GG 3
RHR (LPCI/LPCS/ Core Spray) 104 Reactor Steam Dome H,GG 3
Pressure - Low 105 Reactor Vessel Shroud H
3 Level - Low 106 Logic Power Monitor H
3 Automatic Depressurization System 106A Control Power Monitor H
3 107 Reactor Vessel Water Level H,GG 3
Low (Level 1) 108 Drywell Pressure - High H,GG 3
109 ADS Initiation Timer H GG 3
110 Low Water Level Timer H
3 D-5 I
$c b
6 DRAR BWR-TABLE _1 Continued)
REPORT PLANT CRITERIA LCO ITEM 111 Reactor Vessel Water Level H GG 3
Low (Level 3) 112 LPCI/LPCS/ Core Spray H,GG 3
Discharge Pressure - High 112A ADS Bypass Timer GG 3
High Pressu*e Core S MsnualInhabit(ADS) pray GG 3
112B 113 HanualInitiation(ADS)
GG 3
114 Dryweil Pressure - High GG 3
115 Reactor Vesse? Water Level CG 3
Low (Level 2) 116 Reactor Yessel Water Level GG 2
I;igh (Level 8) 117 CST Level - Low GG 3
118 Supp. Pool Water GG 3
Level - High HPCI 119 HanualInitiation(HPCS)
GG 3
120 Drywell Pressure - High H
3 121 Reactor Vessel Water H
3 Level-Low (Level 2) 122 Reactor Vessel Water H
2 Level-High(Level 8) 123 Condensate Storage Tank H
3 Level - Low 124 Suppression Chamber Water H
3 Level - High 106 Logic Power Monitor H
3 ECCS Inst.
125 Loss of Power GG 3
126 Reactor Pressure - High H
3 (LowLowSetInterlock) 3.3.4 Recirculation Pump Trip Actuation Instrumentation 127 E0C-RPT H,GG S
128 ATWS-RPT H,GG Policy Statement (RPT) 3.3.5 RCIC Instrumentation 129 Reactor Vessel Water H.GC Policy Statement (RCIC)
Level - Low (Level 2) 130 Reactor Yessel Water GG Policy Statement (RCIC)
Level - High (Level 8)
-s DRAFI BWR-TABLE 1(Continued REPORT LC0 ITEM PLANT CRITERIA 131 CST Level - Low H GG PolicyStatement(RCIC) 132 Supp. Pool Water Level'- High H.GG 3
133 ManualInitiation(RCIC)
GG 2
3.3.6 Control Rod Withdrawal Block Instrun>entation 134 Rod Pattern Control System GG 3
136 RBM H
3 141 Reactor Mode Switch GG 3
Shutdown Position 3.3.7 Honitoring Instrumentation t
(Radiation Monttors) (Notes 2 & 3) 146 Control Room Vent H,GG 3
153 Remote Shutdown Instrumentation H.GG Risk Accident Monitoring Instrumentation (Note 2)
182 SRM 3.3.8 Plant Systems Actuation Instrumentation 190 Drywell Press (Cont. 3 pray)
GG 3
191 Cont. Press (Cont. Spray, GG 3
192 Water Level 1 (Cont. Spray)
GG 3
193 Timers (Cont. Spray)
GG 3
194 WaterLevel8(FW/TT)
GG 2
195 Drywell Pressure GG 3
(Supp. Pool Nakeup System-SPMS) 196 Level 1(SPMS)
GG 3
197 Level 2 (SPMS)
GG 3
198 Supp. Pool Level (SPMS)
GG 3
199 Supp.PoolMakeupTimer(SPMS)
GG 3
200 ManualInitiation(SPMS)
GG 3
3.3.10 201A Neutron Flux Monitoring GG 2
3.3.11 202 Degraded Voltage H
3 3.4 REACTOR COOLANT SYSTEM 3.4.1 203 Recirculation Loops H,GG 2
204 Jet Pumps H,GG 3
205 Idle Recirculetion Loop H,GG 2
Startup 206 Recirculation Loop Flow CG 2
D-7
t DRAFI BWR-TABLE 1 (Continued)
REPORT LC0 ITEM PLANT CRITERIA 3.4.2 207 Safety / Relief Valves H,GG 3
208 S/RV Low-Low Set H GG 3
3.4.3 209 Leak Detection Systems H GG 1
3.4.3 210 Operational Leakage Limits H,GG 1
3.4.5 212 Specific Activity H,GG 2
3.4.6 213 Pressure /Tenperature Limits 214 Reactor Steam Dome Pressure H,GG 2
3.4.9 217 RHR - Hot Shutdown GG Policy Statement (RHR) 218 RHR - Cold Shutdown GG PolicyStategent(RHR) 3.5 EMERGENCY CORE CCOLING SYSTEMS 3.5.1 219 hPCI H
3 3.5.2 220 ADS H
3 3.5.3 221 CSS H
3 222 LPCI H
3 3.5.4 223 Supp. Pool H,GG 3
3.6 CONTAINMENT SYSTEMS 3.6.1 Primary Containment 226 Cont. Integrity H,GG 3
226 Air Locks H GG 3
229 MSLIV-LCS H GG 3
231 Structural Integrity H,GG 3
232 Cont. Internal Pressure H.GG 2
233 Cont. Air Terap GG 2
234 Containment Purge Systera H,GG 3
3.6.2 Drywell 235 Drywell Integrity H.GG 3
236 Drywell Air Temperature H,GG 2
237 Drywell Bypass Leakage GG 2
238 Drywell Air Locks GG 3
239 Drywell Structural Integrity GG 3
240 Drywell Internal Pressure GG 2
241 Dryvell Vent and Purge GG 2
0-8
s
. s.
DRAFT BWR-TABLE 1 (Continued) i REPORT LCO ITEN PLANT CRITERIA
?.6.3 Depressurization Systems 242 Cont. Spray GG 3
243 SuppressionChamber(Pool)
H.GG 2&3 i
244 Suppression Pool Makeup GG 3
245 Suppression Pool Cooling H.GG 3
3.6.4 246 1 solation Valves H,GG 3
3.6.5 247 Supp. Chamber - Drywell VB H
3 249 Drywell Post LOCA VB GG 3
3.6.6 Secondary Containment 250 Secondary Containment H.GG 3
Integrity 251 Auto Isolation Dampers H,GG 3
3.6.7 Containment Atmosphere Control 252 SGTS H,GG 3
253 H Recombiner H,GG 3
254 H Mixing System H
3 255 0 Conc.
H 3
256 H Ignition System GG 3
2 3.7 PLANT SYSTEMS 3.7.1 256 RHR Service Water H
3 259 Standby Service Water GG 3
260 Plant Service Water H
3 261 HPCS Service Water GG 3
~
262 Ultimate Heat Sink GG 3
3.7.2 263 Control Room Environmental H
3 Control 264 Control Room Emergency Filter GG 3
3.7.3 265 RCIC H,GG Policy Statement (RCIC) j 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 274 Electrical Power Systems H,GG 3
j (AC/DCSources,On-Site Distribution)(6 Sections) 3.8.4 277 Power Monitoring of RPS H,GG 3
1 Protection y
,n.
3.9,l
[.;
e 4
i, SWR-TABLE 1(Continued)
REPORT LCO ITEM
. PM CRITERIA l
3.9 REFUELING OPERATIONS
1 i
j
\\%
4 h
ig DRAET CN!-TABLE 2 (Note 1)m 'i s
s
-, GENERAL ELECTRIC STANDARD TECHNICAL '5PECIFIC".T!0N LCOs'VH,1CH f Y BE RELOCAT(D_
l S
i
\\1EI'03T PLANT LC0 jdFM_
3.1 i
, REACTIVITY CONTROL TYSTEMS 2
' Reacti/ity Avon aly (Note 21 H,GG 3.1.2 Maximug 3cra Tires (7 Sech M
3.1.3 4
x) s 3
3.3
' INSIMEENT/UON J
3 IsolationActuationInstrumentOion 3.3.2 s
g4
/5
- ' Drywell Pressure - Hioh (yPCI)
H t
1 <:<
1 84 l Drywell Pressure;- High (kCIC)
H,GG
\\
3.3.6\\,,
'Copf"ol Rod yithdrawal Block Instrumentation i
N 1
i H,GG 135
- APR:t ' i S!
! S-4 /,'
H
\\
137
.i !M
' '1RM k ',, )
H,GG
' i 138
'i 139 SDV Water Level it H,GG 140 Reactor Ccolant System GG Recirculatiot Finw-Upsrale l
l
)
' Monitoring.InstruSentatic.i l
3.3.7 4
GG ComponenhCooling(Water 14?
)
Radition Monitor Noe 3 5 143 Standby Servjce Water (tiote 3) uG
,144 Carbon Bed Vault (Note 3)
?5 Offgas Post Treatment (Note 3)
H.
1 '7 Cont. & Drywell Vent Exhaust (N6te 3) aiG t.
\\
'FuelHandlingArea'holSweep(Note 3))GG fuel Handling arm htilation (Note 3 l
14 GG 14i.1 i-15U AreaMonitors-(Radiativ),(Note 3)
GG, H.G /,
i 151\\
Seismic Monitths r
l 157 ?
Meteorological Ir t.,
'183 L TIP 184 Main Control Rocm H
s
/
Environmental System i
(Chlorine and Amonia)
.I D?ttttion System 18f )
Fire Protes. tion GG l
s' 18P. '
Lorde Parts GG 188 radioactive Liquid Effluent (Note 3)
H,G?
Monitoring instruinentation 189 Radioae',iva Gasedus Effluent (Note 3)
H,GG
/ J, Monitoring Instrunintation l
1 D-11 l'
.J h
s.
y DRAR BWR-TABLE 2(Continued)
REPORT PLANT LCO ITEM 3.3.9 201 Turbine Overspeed Protection H,GG 3.4 REACTOR COOLANT SYSTEM 3.4.4 211 Chemistry H,GG
~
3.4.8 216 Structural Integrity (Note 2)
H,GG 3.6 CONTAINHENT SYSTEMS 3.6.1 227 ContainmentLeakage(Note 4)
H.GG 3.6.2 230 Feedwater Leakage Control GG 3.6.7 257 Combustible Gas Control GG Purge System 3.7 PLANT SYSTEMS 3.7.4 266 Snubbers H,GG 3.7.5 267 Sealed Source Contamination H,GG 3.7.6 268 Fire Suppression Systems GG (6 Sections) 3.7.7 269 Fire Rated Assemblies GG 3.7.8 270 Area Temp Monitoring GG 271 Settlement of Class 1 H
Structure 3.7.9 272 Spent Fuel Pool Temp GG 3.7.10 273 Flood Protection H,GG 3.6 ELECTRICAL POWER SYSTEMS 3.8.2 275 AC Circuits inside Containment H
3.8.3 276 Overcurrent Protection Devices H,GG 3.0 REFUELING OPERATIONS 3.9.6 286 Comunications H,GG 3.9.7 287 Refueling Equipment H,GG (3 Sections) 3.9.10 291 Control Rod Removal (2 Sections)
H GG 3.9.12 294 Horizontal fuel Transfer GG System 3.10 295 SPECIAL TEST EXCEPTIONS (Note 5)
H.GG D-12 I
i A
k.
DRM BWR-TABLE 2(Continued)
REPORT LCO ITEM PLANT 3.11 RADICACTIVE EFFLUENTS 3.11.1 296 LiquidEffluents(Note 3)
H GG 297 LiquidEffluentsDose(Note 3)
H,GG 298 LiquidWasteTreatment(Note 3)
H,GG 299 Liquid Holdup Tanks (Note 3)
H,GG 3.11.2 300 GaseousEffluentDoseRate(Note 3)H,GG 301 GaseousEffluentDose-(Note 3)
H,GG Noble Gases 302 Gaseous Effluent Dose - (Note 3)
H,GG Other than Noble Gas 303 GaseousRadwasteTreatment(Note 3)HGG 304 Total Dose H,GG 305 VentilationExhaust(Note 3)
GG Treatment System 305 ExplosiveGasMixture(Note 3)
H GG 3.11.3 308 SolidRadwasteSystem(Note 3)
H,GG 3.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l
309 Environmental Monitoring H,GG (3 Sections) l l
l Notes:
- 1. LCOs listed in this table may be relocated to other licensee-controlled document contingent upon NRC staff approval of the location of and controls l
over relocated requirements.
b However, the associated Surveillance
- 2. This LCO may(s)e removed from the STS.should be relocated to an LCO whic Requirement if the Surveillar.ce Pequirement is required in order to meet the operability l
l requirements for the retained LCO.
l
- 3. These TS may be relocated from LCOs. However, they must be controlled by including them in program requirements of 6.8 in Section 6 of STS.
- 4. This LCO may be relocated, however, Pa, La, Ld and Lt may be either retained in TS or in the Bases of the appropriate containment LCO.
- 5. Special Test Exceptions may be included with the corresponding LCOs.
1 0-13 l
1
- - - ~ -.
-