ML20248C520
| ML20248C520 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/29/1989 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 89-465, NUDOCS 8910030493 | |
| Download: ML20248C520 (116) | |
Text
_ _ _ _ - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ .
VIRGINIA ELECTRIC AN9 POWER COMPANY RICHMOND, VIRGINIA 23261 September 29, 1989 U. S. Nuclear Regulatory Commission Serial No.89-465 Attention: Document Control Desk N0/EJL:rmh Whshington, D. C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 NPF-7' Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2 MERITS PROGRAM CRITERIA APPLICATI(M REPORT North Anna Power Station is the lead plant for the Westinghouse Owners Group MERITS Program. As discussed with members of the NRC staff in a meeting held on June 15, 1989, Virginia Electric and Power Company will be making several document submittals to the NRC to support implementation ' of the North Anna MERITS Program. Attached for your review is our first submittal, " North Anna Units 1 & 2 MERITS Program Criteria Application Report".
The NRC's Interim Policy Statement on Technical Specification Improvement (52FR3788, February 6,1987) provided screening criteria to evaluate technical specifications and determine whether a specification must be retained, or may be relocated to another licensee controlled document. The attached report documents the application of the screening criteria to the North Anna Technical Specifications. The report identifies the current requirements that we propose to retain in the North Anna Technical Specifications and 'those .that will be relocated outside of the Technical Specificationsifor the North Anna MERITS Program implementation. The Technical Specification Screening Forms that were 1 used to document the evaluation of the Technical Specifications were reviewed and approved by the Station , Nuclear Safety and Operating Committee.
We considered our initial meeting with the NRC staff to be productive, and look forward to working with your staff to implement this program.
Very truly yours,
\
.L Stewart Senior Vice President - Power Attachment 0 0 m
8
$U P
$$$ 0 g
cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.
Suite 2900 Atlanta, Georgia 30323 l Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station Mr. J. A. Calvo - USNRC, NRR/0TSB U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Operational Events Assessment Technical Specifications Branch Mr. R. L. Emch )
U. S. Nuclear Regulatory Commission 1 Office of Nuclear Reactor Regulation i Division of Operational Events Assessment Technical Specifications Branch Mr. C. W. Moon U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Operational Events Assessment i Technical Specifications Branch i
i l
4
_ - - _ _ - _ - - _ _ - _ - - _ _ - _ _ _ - _ _ _ _ - _ _ _ - 1
r: ,-
1
~
NORTH ANNA UNITS 1 & 2 MERITS PROGRAM CRITERIA APPLICATION REPORT
~ __
INTRODUCTION Virginia Power and the Westinghouse Owners Group (WOG) have jointly agreed to identify the North Anna Power Station as the lead plant in the implementation of the HERITS Technical Specification Improvement Program.
One of the initial tasks in the implementation of this program is the application of the Technical Specification screening criteria contained in the NRC's Interim Policy Statement on Technical Specification Improvement (52FR3788), February 6,1987. . These criteria have been provided to evaluate Technical Specifications and determine whether a specification must be, retained, or may be relocated to another licensee controlled document.
This report documents the application of these screening criteria to the Technical Specifications of the North Anna Power Station, Units 1 and 2, and provides justification for the results. The Report also identifies the proposed new location, and controlling mechanisms for the relocated specifications.
APPLICATION OF THE NRC SELECTION CRITERIA An original application of these criteria by the WOG can be found in Reference 1, where the criteria were applied to the Standard Technical Specifications (STS) for Westinghouse Pressurized Water Reactors (NUREG 0452, Revision 4 and draft Revision 5).
In addition to the criteria application process, the NRC's Interim Policy Statement requires that certain specific systems, which operating experience and probabilistic risk assessments have generally shown to be important to public health and safety, be retained in the Technical Specifications. For '
Pressurized Water Reactors, the only system which is mentioned specifically is the Residual Heat Removal System.
1 b
x.
The Interim Policy Statement also stipulates that the owners groups will 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. Reference 1 was submitted to the NRC in November 1987. The NRC documented the results of their review in Reference 2.
Reference 2 identified which specifications in the Westinghouse Standard Technical Specification must be retained, and which specifications may be relocated. These determinations formed the basis for the application of the criteria to the North Anna Technical Specifications. Specifically, the North Anna Technical Specification. requirements were compared with the Westinghouse Standard Technical Specification requirements (Revision 5). .The North Anna-specifications with comparable requirements to the standard requirements were dispositioned in accordance with the provisions of Reference 2. For those requirements which were unique to the North Anna Technical Specifications, or were inconsistent with the requirements in the Standard Technical Specifications, an individual application of the criteria was performed. This involved the use of individual screening forms similar to those employed in Reference 1. In this way, a decision to retain or relocate each of the Technical Specifications for North Anna Units 1 and 2 was obtained.
RESULTS The recommended disposition of the LCOs in Chapter 3 of the Technical Specifications for the North Anna Units 1 and 2 is shown in Tables 1 and 2, respectively. These results are based on the application of the criteria and the NRC conclusions, documented in Reference 2. Those LCOs in the Technical Specifications for North Anna, Units I and 2, which were not addressed in Reference 2, and, hence were not reviewed by the NRC, are indicated with "Not Reviewed" in column four of Tables 1 and 2, respectively. . Supplementary information related to the basis of the information contained in the two tables is provided as referenced notes at the end of each-table.
Each separate LCO which has not been previously reviewed by the NRC has been subjected to the screening process. This process is documented on screenirig I forms, which are shown in Tables 3 and 4, respectively. The screering forms TJrnish the justification for the retent'on of the specification or the relocation outside of the technical specifications. Where applicable, the screening forms reference Appendix A, which documents the PRA evaluations performed to determine if the technical specification requirement is contained in a dominant risk sequence. If a requirement is not risk significant, and does not satisfy any of the NRC screening criteria for retention, then it has been concluded that it may be relocated to another licensee controlled document.
Appendix A provides a detailed account of the analytical approach and assumptions used in the PRA evaluation. This Appendix also includes a description of the application of PRA to the MERITS program and the available FRA studies for other three loop plants. A justification is provided for the use of this material to assess the sensitivity of North Anna Units 1 and 2 to plant incidents involving risk to the general public.
ADMINISTRATIVE CONTROL OF THE RELOCATED TECHNICAL SPECIFICATIONS The Technical Specifications that do not meet any of the Technical Specification screening criteria and are not significant risk contributors will be relocated outside of the North Anna Technical Specifications. Some of these Technical Specifications, for example, the Radiological Effluent Technical Specifications (RETS) and Fire Protection Technical Specifications, will be relocated and controlled in accordance with NRC staff guidance provided in Generic Letters. These relocations may precede the other MERITS program Technical Specification relocations.
A station Administrative Procedure will be used for the " relocated" Technical Specifications. The residence of these " relocated" Technical Specifications will be in the Updated Final Safety Analysis Report (UFSAR). It is currently envisioned that the " relocated" Technical Specifications to the UFSAR will be controlled as " Technical Requirements" in Section 16. A copy of the " Technical Requiren.<nts" will be maintained with the Technical Specifications but as a separate documen'.. Changes to the " Technical Requirements" and UFSAR will be controlled in accordance with the criteria of 10CFR50.59 and any other controlling regulation (e.g.,10CFR50.55a in the case ofISI/ISTPrograms). Additionally, proposed changes to this Administrative Procedure and the " Technical Requirements" (UFSAR) will be reviewed and approved by the Station Nuclear Safety and Operating Committee prior to implementation. The change control process will be described in the station administrative procedare. A copy of the station administrative procedure will be provided to the NRC to allow the staff to ensure the accountability of each of these " relocated" requirements.
REFERENCES
- 1. WCAP-11618, " Methodically Engineered, Restructured, and Improved Technical Specifications", HERITS Program - Phase II, Task 5, Criteria Application.
- 2. T. E. Nurley to W. S. Wilgus, "NRC Staff Review of Nuclear Steam Supply System Vunder Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specification", Nay,1988 (NRC Letter).
- 3. " Guidance for the Implementation of Programmatic Controls for RETS in the Administrative Controls Section of Technical Specifications and the Relocation of Procedural Details of Current RETS to the Offsite Dose Calculation Manual or Process Control Program", United States Nuclear Regulatory Commission Generic Letter 89-01, January 31, 1989.
- 4. 52FR3788, " Interim Policy Statement on Technical Specification improvements for Nuclear Power Reactors," U.S. Nuclear Regulatory Commission, February 6, 1987.
TABLE 1 NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REACTIVITY CONTROL SYSTEMS Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.1.1.1 3.1.1.1 Shutdown Margin, Retain Retain 1 T,yg> 200'F 3.1.1.2 3.1.1.2 Shutdown Nargin, Retain Retain 1 T,yg 5 200'r 3.1.1.3.1 None Boron Dilution, RCS Flow Not Reviewed Relocate 2 3.1.1.3.2 None Boron Dilution, Valve Position Not Reviewed Retain 2 3.1.1.4 3.1.1.3 Moderator Temperature Coefficient Retain Retain 3.1.1.5 3.1.1.4 Ninimum Temperature for Criticality Retain Retain 3.1.2.1 3.1.2.1 Flow Paths, Shutdown Relocate Relocate 3.1.2.2 3.1.2.2 Flow Paths, Operating Relocate Relocate 3.1.2.3 3.1.2.3 Charging Pump, Shutdown Relocate Relocate 3.1.2.4 3.1.2.4 Charging Pumps, Operating Relocate Relocate 3.1.2.5 None Boric Acid Transfer Pumps, Shutdown Not Reviewed Relocate 2 3.1.2.6 None Boric Acid Transfer Pumps, Operating Not Reviewed Relocate 2 3.1.2.7 3.1.2.5 Borated Water Sources, Shutdown Relocate Relocate 3.1.2.8 3.1.2.6 Borated Water Sources, Operating Relocate Relocate 3.1.3.1 3.1.3.1 Novable Control Assemblies, Group Height Retain Retain 3.1.3.2 3.1.3.2 Position Indicator Channels, Operating Relocate Retain 3 l
5-
T,ACLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REACTIVITYCONTROLSYSTEMS(Cont)
Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.1.3.3 3.1.3.3 Position Indicator Channels, Shutdown Relocate Relocate 4 3.1.3.4 3.1.3.4 Rod Drop T:.ae Relocate Relocate 4 3.1.3.5 3.1.3.5 Shutdown Rod Insertion Limit Retain Retain 3.1.3.6 3.1.3.6 Control Rod Insertion Limit Retain Retain l
l b
I 1 .
l .
l l
TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION POWER DISTRIBUTION LIMITS Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.2.1 3.2.1 Axial Flux Difference Retain Retain 3.2.2 3.2.2 Heat Flux Hot Channel Factor-Fg(Z) Retain Retain 3.2.3 3.2.3 Nuclear EnthalpyNHt Channel Factor-F AH Retain Retain 3.2.4 3.2.4 Quadrant Power Tilt Ratio Retain Retain l 3.2.5 3.2.5 DNB Parameters Retain Retain l
l 4
TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION INSTRUMENTATION Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.3.1.1 3.3.1 Reactor Trip System Instrumentation Retain Retain 3.3.2.1 3.3.2 Engineered Safety Featura Actuation System Instrumentation Retain Retain 3.3.3.1 3.3.3.1 Radiation Monitoring Instrumentation Retain Retain 5 3.3.3.2 3.3.3.2 Novable Incore Detectors Relocate Relocate 3.3.3.3 3.3.3.3 Seismic Instrumentation Relocate Relocate 3.3.3.4 3.3.3.4 Meteorological Instrumentation Relocate Relocate 3.3.3.5 3.3.3.5 Auxiliary Shutdown Panel Monitoring Instrumentation Retain Retain 3.3.3.6 3.3.3.6 Accident Monitoring Instrumentation Retain Retain 6 3.3.3.7 3.3.3.8 Fire Detection Instrumentation Relocate Relocate 3.3.3.9 3.3.3.9 Loose Parts Nonitoring Systems Relocate Relocate 3.3.3.10 3.3.3.10 Radioactive Liquid Effluent Monitoring Instrumentation Relocate Relocate 7 3.3.3.11 3.3.3.11 Radioactive Gaseous Effluent Monitoring Instrumentation Relocate Relocate 8
.g.
TABLE 1 (Continued) .l NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION i
Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.4.1.1 3.4.1.1 Reactor Coolant Lom:$ and Coolant Circulation Startup and Power Operation Retain Retain 3.4.1.2 3.4.1.2 Hot Standby Retain Retain 3.4.1.3 3.4.1.3 Shutdown Retain Retain 9 3.4.1.4.1 3.4.1.4.2 j 3.4.1.4 3.4.1.5 Isolated Loop Retain Retain 3.4.1.5 3.4.1.6 Isolated Loop Startup Retain Retain 3.4.2 3.4.2.1 Safety Valves, Shutdown Relocate Relocate 3.4.3 3.4.2.2 Safety Valves, Operating Retain Retain 3.4.3.2 3.4.4 Relief Valves Retain Retain 3.4.4 3.4.3 Pressurizer Retain Retain 3.4.5 3.4.5 Steam Generators Relocate Relocate 4
)
3.4.6.1 3.4.6.1 Leakage Detection Systems Retain Retain 3.4.6.2 3.4.6.2 Operational Leakage Retain Retain 3.4.6.3 None Primary to Secondary Leakage Not Reviewed Relocate 2 3.4.6.4 None Primary to Secondary Leakage Dection Systems Not Reviewed Relocate 2 3.4.7 3.4.7 Chemistry Relocate Relocate 3.4.8 3.4.8 Specific Activity Retain Retain ,
l l
]
_ 9
TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REACTORCOOLANTSYSTEM(Cont)
Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit i Number Number Title Results Results Notes 3.4.9.1 3.4.9.1 Pressure / Temperature Limits, Reactor Coolant Retain Retain System 3.4.9.2 3.4.9.2 Pressurizer, (Pressure / Temperature) Relocate Relocate 3.4.9.3 3.4.9.3 Overpressure Protection Systems Retain Retain 3.4.10.1 3.4.10.1 Structural Integrity, ASME Code Class 1, 2 & 3 Components Relocate Relocate 4 3.4.11.1 3.4.11 Reactor Vessel Head Vent Relocate Relocate 10 i
10
TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION EMERGENCY CORE COOLING SYSTEMS Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.5.1 3.5.1 Accumulators Retain Retain 3.5.2 3.5.2 ECCS Subsystem Tavg
> 350'F Retain Retain 3.5.3 3.5.3 ECCS Subsystem, T avg
< 350*F Retain Retain 3.5.4.1 3.5.4.1 Boron Injection Tank Retain Retain 3.5.4.2 3.5.4.2 Heat Tracing Relocate Relocate 3.5.5 3.5.5 Refueling Water Storage Tank Retain Retain e
f
k TABLE 1(Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION CONTAINMENT SYSTEMS
{
Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.6.1.1 3.6 1.1 Containment Integrity Retain Retain 3.6.1.2 3.6.1.2 Containment Leakage Relocate Relocate 14 3.6.1.3 3.6.1.3 Containment Airlock Retain Retain 3.6.1.4 3.6.1.5 Internal Pressure Retain Retain j
3.6.1.5 3.6.1.6 Air Temperature Retain Retain l
3.6.1.6 3.6.1.7 Containment Structural l Integrity Relocate Relocate 3.6.2.1 3.6.2.1 Containment Quench Spray System Retain Retain 3.6.2.2 3.6.2.2 Containment Recirculation Spray System Retain Retain 3.6.2.3 3.6.2.3 Chemical Addition System Retain Retain 3.6.3.1 3.6.3 Containment Isolation i Valves Retain Retain I i
3.6.4.1 3.6.5.1 Hydrogen Analyzers Retain Retain 3.6.4.2 3.6.5.2 Electric Hydrogen Recombiners Retain Retain 3.6.4.3 3.6.4.3 Waste Gas Charcoal Filter System Retain Retain l 3.6.5.1 3.6.5.1 Steam Jet Air Ejector Relocate Relocate l
TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION PLANT SYSTEMS Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 i Number Number Title Results Results Notes 3.7.1.1 3.7.1.1 Safety Valves Retain Retain 3.7.1.2 3.7.1.2 Auxiliary Feedwater System Retain Retain 3.7.1.3 3.7.1.3 . Emergency condensate Storage Tank Retain Retain 3.7.1.4 3.7.1.4 Activity Retain Retain 3.7.1.5 3.7.1.5 Main Steam Trip Valves Retain Retain 3.7.1.6 None Steam Turbine Assembly Not Reviewed Relocate 2 3.7.1.7 3.3.4 Turbine Overspeed Relocate Relocate 3.7.2.1 3.7.2 Steam Generator Pressure / ,
Temperature Limitation Relocate Relocate j 3.7.3.1 3.7.3 Component Cooling Water Subsystem Retain Retain 3.7.4.1 3.7.4 Service Water System Retain Retain j 3.7.5.1 3.7.5 Ultimate Heat Sink Retain Retain 3.7.6.1 3.7.6 Flood Protection Relocate Relocate j J
3.7.7.1 3.7.7 Control Room Emergency Habitability System Retain Retain 3.7.8.1 3.7.8 Safeguards Area j Ventilation System Retain Retain ;
3.7.9.1 None Residual Heat Removal l System, Operating Not Reviewed Retain 15 3.7.9.2 3.4.1.4.1 Residual Heat Removal 3.4.1.4.2 System, Shutdown Retain ketain 11 3.7.10 3.7.9 Snubbers Relocate Relocate i l
l i
_ - _ - _ - - _ . m ________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
- 1 l
TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION
, PLANTSYSTEMS(Cont)
Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.7.11.1 3.7.10 Sealed Source Contamination Relocate Relocate 3.7.12.1 None Settlement of Class 1 Structures Not Reviewed Relocate 2 3.7.13 None Groundwater Level, Service Water Reservoir Not Reviewed Relocate 2 3.7.14.1 3.7.11.1 Fire Suppression Systems Relocate Relocate 3.7.14.2 3.7.11.3 Low Pressure CO Systems Relocate Relocate 2
3.7.14.3 3.7.11.3 High Pressure CO Systems 2 Relocate Relocate 3.7.14.4 3.7.11.4 Halon Systems Relocate Relocate i 3.7.14.5 3.7.11.5 Fire Hose Station Relocate Relocate 3.7.14.6 3.7.11.2 Spray and/or Sprinkle Systems Relocate Relocate 3.7.15 3.7.12 Penetration Fire Barriers Relocate Relocate I
i i
l
_-----___A
, TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION ELECTRICAL POWER SYSTEMS 1 1 Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 j Number Number Title Results Results Notes
]
3.8.1.1 3.8.1.1 A.C. Sources, Operating Retain Retain 3.8.1.2 3.8.1.2 A'.C. Sources, Shutdown Retain Retain 3.8.2.1 3.8.3.1 Onsite Power Distribution Systems, A.C. Distribution ,
Operating Retain Retain 3.8.2.2 3.8.3.2 Onsite Power Distribution Systems, A.C. Distribution Shutdown Retain Retain 1
3.8.2.3 3.8.2.1 Onsite Power Distribution {
Systems, D.C. Distribution ;
Operating Retain Retain
]
3.8.2.4 3.8.2.2 Onsite Power Distribution Systems, D.C. Distribution Shutdown Retain Retain l
I l
I
TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REFUELING OPERATIONS Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.9.1 3.9.1 Boron Concentration Retain Retain 3.9.2 3.9.2 I'strumentation n Retain Retain 3.9.3 3.9.3 Decay Time Retain Retain 3.9.4 3.9.4 Containment Building Penetrations Retain Retain 3.9.5 3.9.5 Communications Relocate Relocate 3.9.6 3.9.6 Manipulator Crane Operability Relocate Relocate 3.9.7 3.9.7 Crane Travel, Spent Fuel Pit Relocate Relocate 3.9.8.1 3.9.8.1 Residual Heat Removal 3.9.8.2 and Coolant Circulation, All Water Levels Retain Retain 12 3.9.8.2 3.9.8.2 Residual Heat Removal and Coolant Circulation, Low Water Level Retain Retain 12 3.9.9 3.9.9 Containment Purge and Exhaust Isolation System Retain Retain 3.9.10 3.9.10 Water Level, Reactor Retain Retain Yassel 3.9.11 3.9.11 Spent Fuel Pit Water Level Retain Retain 3.9.12 3.9.12 Fuel Building Ventilation System Retain Retain
TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION SPECIAL TEST EXCEPTIONS Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.10.1 3.10.1 Shutdown Margin Relocate Relocate 13 3.10.2 3.10.2 IroupHeight, Insertion and Power Distribution Limits Retain Retain 3.10.3 3.10.3 Physics Tests Retain Retain 3.10.4 3.10.4 Reactor Coolant Loops Retain Retain 3.10.5 3.10.5 Position Indicator ,
Channels, Shutdown Relocate Relocate !
l I
i I
1
'I TABLE 1 (Continued)
NORTH' ANN' A UNIT 1 TECHNICAL SPECIFICATIONS j
1 CRITERIA APPLICATION l l
RADI0 ACTIVE EFFLUENTS Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 .
Number Number Title Results Results Notes !
3.11.1.1 3.11.1.1 Liquid Effluents Concentration Relocate Relocate 7 3.11.1.2 3.11.1.2 Dose Relocate Relocate 7 3.11.1.3 3.11.1.3 Liquid Radwaste Treatment Relocate Relocate 7 3.11.1.4 3.11.1.4 Liquid Holdup Tanks Relocate Relocate 3.11.2.1 3.11.2.1 Gaseous Effluents, Dose Rate Relocate Relocate 7 3.11.2.2 3.11.2.2 Dose, Noble Gases Relocate Relocate 7 3.11.2.3 3.11.2.3 Dose, Iodine 131, Tritium and Radionuclides in Particulate Form Relocate Relocate 7 3.11.2.4 3.11.2.4 Gaseous Radwaste Treatment ' Relocate Relocate 7 l 3.11.2.5 3.11.2.6 Explosive Gas Mixture Relocate Relocate 3.11.2.6 3.11.2.5 Gas Storage Tanks Relocate Relocate 3.11.3 3.11.3 Solid Radioactive Waste Relocate Relocate 7 3.11.4 3.11.4 Total Dose Relocate Relocate 7 l
1 j
l
TABLE 1 (Continued)
NORTH ANNA UNIT 1 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION RADIOLOGICAL ENVIRONMENTAL MONITORING Unit 1 STS-Rev 5 North Anna Unit 1 NRC Unit 1 Number Number Title Results Results Notes 3.12.1 3.12.1 Nonitoring Program Relocate Relocate ,
7 3.12.2 3.12.2 L'and Use Census Relocate Relocate 7 3.12.3 3.12.3 Interlaboratory Comparison Program Relocate Relocate 7 k
-)?
, TABLE 1 (Continued)
NOTES
- 1) Reference 2 allows relocation for Nodes 1 and 2, and specifies retention for Nodes 3, 4 and 5. The Modes 1 and 2 requirements will be relocated.
- 2)- This LCO was not evaluated in Reference 2.
- 3) Reference 2 allows relocation of this LCO, but specifies retention of the associated surveillance requirements. The entire specification will be retained.
- 4) This LCO will be relocated and the associated surveillance requirements will be retained in,the Technical Specifications per Reference 2.
- 5) The Radiation Effluent Monitoring Instrumentation will be dispositioned in accordance with Reference 3. The remaining instruments will be included in other appropriate specifications.
- 6) The Regulatory Guide 1.97 Type A variables are listed in Table 1A and will be retained. The Regulatory Guide 1.97 Category 1, non-Type A variables are identified and evaluated in Table 3.
- 7) This LCO will be dispositioned in accordance with the provisions of Reference 3.
- 8) The radioactive gaseous effluent monitoring instrumentation will be dispositioned in accordance with the provisions of Reference 3. The explosive gas monitoring instrumentation will be relocated in accordance with Reference 2.
- 9) Unit 1 Technical Specifications LCO 3.4.1.3 applies to Modes 4 and 5.
Standard Technical Specifications LCO 3.4.1.3 pertains only to Mode 4.
Standard Technical Specifications LCO 3.4.1.4.1 and LCO 3.4.1.4.2 address Mode 5 with, respectively, loops filled and loops not filled.
- 10) Unit 1 Technical Specifications LCO 3.4.11.1 applies to Modes 1, 2 and
- 3. Modes 4, 5 and 6 (with vessel head on) are covered in LCO 3.4.9.3..
Standard Technical Specifications LCO 3.4.11 covers Modes 1, 2, 3 and 4.
Modes 4, 5 and 6 (with vessel head on) are covered in LCO 3.4.9.3. Note, Mode 4 is covered in both sections of the Standard Technical Specifications.
- 11) North Anna Unit 1 LCO 3.7.9.2 is applicable in Modes 4 and 5. Standar'd Technical Specifications LCO 3.4.1.4.1 and LCO 3.4.1.4.2 are applicable in N#
- with reactor coolant loops filled and reactor coolant loops not filled, respectively.
- 12) Standard Technical Specifications LCO 3.9.8.1 addresses Residual Heat Removal and Coolant Circulation - High Water Level. Standard Technical Specifications LCO 3.9.8.2 addresses Residual Heat Removal and Coolant Circulation - Low Water Level.
, TABLE 1 { Continued)
NOTES
- 13) Reference 2 allows relocation of the Mode 2 requirements for LCO 3.1.1.1. The Mode 2 requirements for LCO 3.1.1.1 will be relocated per ,
Note 1 above. Therefore, LCO 3.10.1 will not be needed and will be j relocated.
- 14) This LCO will be relocated; however, Pa, La, Ld and Lt will be retained in the Technical Specifications per Reference 2.
- 15) This LCO was not evaluated in Reference 2. Reference 4 requires that the Residual Heat Removal System be retained in the Technical Specifications.
I i
21-
TABLE 1A l
NORTH ANNA UNIT 1 ACCIDENT MONITORING INSTRUMENTATION REGULATORY GUIDE 1.97 TYPE A VARIABLES l 1. Steam generator narrow range level
- 2. Steam generator pressure
- 3. Core exit temperature (incore thermocouple)
- 4. Reactor coolant system (RCS) cold leg temperature
- 5. RCS hot leg temperature
- 6. RCS wide range pressure
- 7. Emergency condensate storage tank. level
- 8. Refueling water storage tank level
- 9. Pressurizer power operated relief valve (PORV) position indication
- 10. Pressurizer level
- 11. Containment intermediate range pressure
- 12. Reactor coolant system subcooling margin monitor (degree of subcooling)
- 13. Containment sump wide range level
- 14. Containment area radiation
- 15. Auxiliary feedt:ater flow
. i TABLE 2 ,
, i NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS l
l CRITERIA APPLICATION REACTIVITY CONTROL SYSTEMS -
1 j
Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 !
Number Number Title Results Results Notes 3.1.1.1 3.1.1.1 Shutdown Margin, T"V9
> 200'F Retain Retain 1 ,
3.1.1.2 3.1.1.2 Shutdown Margin, T avg 1
< 200*F
~
Retain Retain 1 l 3.1.1.3 No.'e Boron Dilution, Valve Position Not Reviewed Retain 2 3.1.1.4 3.1.1.3 Moderator Temperature Coefficient Retain Retain 3.1.1.5 3.1.1.4 Minimum Temperature for Criticality Retain Retain 3.1.2.1 3.1.2.1 Flow Paths, Shutdown Relocate Relocate 3.1.2.2 3.1.2.2 Flow Paths, Operating Relocate Relocate 3.1.2.3 3.1.2.3 Charging Pump, Shutdown Relocate Relocate !
3.1.2.4 3.1.2.4 Charging Pumps, Operating Relocate Relocate 3.1.2.7 3.1.2.5 Borated Water Sources, Shutdown Relocate Relocate 3.1.2.8 3.1.2.6 Borated Water Sources, Operating Relocate Relocate 1
3.1.3.1 3.1.3.1 Movable Control l Assemblies, Group Height Retain Retain i 3.1.3.2 3.1.3.2 Position Indicator Channels, Operating Relocate Retain 3 3.1.3.3 3.1.3.3 Position Indicator .
Channels, Shutdown Relocate Relocate 4 3.1.3.4 3.1.3.4 Rod Drop Time Relocate Relocate 4 3.1.3.5 3.1.3.5 Shutdown Rod Insertion Limit Retain Retain 3.1.3.6 3.1.3.6 Control Rod Insertion Limit Retain Retain
TABLE 2 (Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION POWER DISTRIBUTION LIMITS Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes 3.2.1 3.2.1 Axial Flux Difference Retain Retain 3.2.2 3.2.2 Heat Flux Hot Channel Factor, F (Z) Retain Retain n
3.2.3 3.2.3 Nuclear Enthalpy Hot Channel Factor, F N 3H Retain Retain 3.2.4 3.2.4 Quadrant Power Tilt Ratio Retain Retain 3.2.5 3.2.5 DNB Parameters Retain Retain I
1
, 4 5
___------_----_i-___-------- - - . - _ - - - - - - - . - -]
A TABLE 2(Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION INSTRUMENTATION Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes 3.3.1.1 3.3.1 Reactor Trip System I, instrumentation Retain Retain 3.3.2.1 3.3.2 Engineered Safety Feature Actuation System Instrumentation Retain Retain 3.3.3.1 3.3.3.1 Radiation Monitoring Instrumentation Retain Retain 5 3.3.3.2 3.3.3.2 Movable Incore Detectors Relocate Relocate 3.3.3.5 3.3.3.5 Auxiliary Shutdown Panel Monitoring Instrumentation Retain Retain 3.3.3.6 3.3.3.6 Accident Monitoring Instrumentation Retain Retain 6 3.3.3.7 3.3.3.8 Fire Detection Instrumentation Relocate Relocate 3.3.3.9 3.3.3.10 Radioactive Liquid #
Effluent Monitoring Instrumentation Relocate Relocate 7 3.3.3.10 3.3.3.11 Radioactive Gaseous 1 Effluent Monitoring l Instrumentation Relocate Relocate 8 i i
I i
I l
_______-- - l
i i
TABLE 2 (Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION l REACTOR COOLANT SYSTEM Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Title Results Results Notes Number I 3.4.1.1 3.4.1.1 Reactor Coolant Loops and Coolant Circulation i
' Startup and Power Operation Retain Retain 3.4.1.2 3.4.1.2 Hot Standby Retain Retain 3.4.1.3 Retain Retain 9- j 3.4.1.3 Shutdown l
3.4.1.4.1 3.4.1.4.2 3.4.1.4 3.4.1.5 Isolated Loop Retain Retain 3.4.1.5 3.4.1.6 Isolated Loop, Startup Retain Retain 3.4.2 3.4.2.1 Safety Valves, Shutdown Relocate Relocate 3.4.3.1 3.4.2.2 Safety Valves, Operating Retain Retain 3.4.3.2 3.4.4 Relief Valves Retain Retain I
3.4.4 3.4.3 Pressurizer Retain Retain 3.4.5 3.4.5 Steam Generators Relocate Relocate 4 I
3.4.6.1 3.4.6.1 Leakage Detection Systems Retain Retain 3.4.6.2 3.4.6.2 Operational Leakage Retain Retain 3.4.6.3 None Primary to Secondary Leakage Not Reviewed Relocate 2 l 1
3.4.6.4 None Primary to Secondary _
1 '
Leakage Detection Systems Not Reviewed Relocate 2-3.4.7 3.4.7 Chemistry Relocate Relocate 3.4.8 3.4.8 Specific Activity Retain Retain I
)
i
z l
1 I
. TABLE 2 (Continued) l NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REACTOR COOLANT SYSTEM (Cont)
Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes 3.4.9.1 3.4.9.1 Pressure / Temperature Limits, Reactor Coolant System Retain Retain 3.4.3.2 3.4.9.2 Pressurizer,(Pressure /
Temperature) Relocelt Relocate 3.4.9.3 3.4.9.3 Overpressure Protection Systems Retain Retain 3.4.10.1 3.4.10.1 Str uctural Integrity, ASME Code Class 1, 2 & 3 Components Relocate Relocate 4 3.4.11.1 3.4.11 Reactor Vessel Head Vent Relocate Relocate 10 i
i 0
i i
I
~
27-
TABLE 2 (Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION l
EMERGENCY CORE COOLING SYSTEMS Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 <
Number Number Title Results Results Notes 3.5.1 3.5.1 Accumulators Retain Retain 3.5.2 3.5.2 E'CCS Subsystem, T""8 1 350*F Retain Retain 3.5.3 3.5.3 ECCS Subsystem, T avg Retain
< 350*F Retain 3.5.4.1 3.5.4.1 Boron Injection Tank Retain Retain 3.5.4.2 3.5.4.2 Heat Tracing- Relocate Relocate 3.5.5 3.5.5 Refueling Water Storage ,
Tank Retain Retain i
i 9
TABLE 2 (Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS ,
1 CRITERIA APPLICATION CONTAINMENT SYSTEMS Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes
! -3.6.1.1 3.6.1.1 Containment Integrity Retain Retain 3.6.1.2 3.6.1.2 C'ontainment Leakage -Relocate Relocate 14 3.6.1.3 3.6.1.3 Containment Airlocks Retain Retain 3.6.1.4 3.6.1.5 Internal Pressure Retain Retain 3.6.1.5 3.6.1.6 Air Temperature Retain Retain 3.6.1.6 3.6.1.7 Containment Structural Integrity Relocate Relocate 3.6.2.1 3.6.2.1 Containment Quench Spray System Retain Retsin 3.6.2.2 3.6.2.2 Containment Recirculation Spray System Retain Retain 3.6.2.3 3.6.2.3 Chemical Addition System Retain Retain 1 l
3.6.3.1 , 3.6.3 Containment Isolation Valves Retain Retain ,
I 3.6.4.1 3.6.4.1 Hydrogen Analyzers Retain Retain 3.6.4.2 3.6.4.2 Electric Hydrogen Recombiners Retain Retain 3.6.4.3 3.6.4.3 Waste Gas Charcoal 1 Filter System Retain Retain l 3.6.5.1 3.6.5.1 Steam Jet Air Ejector Relocate Relocate
< i
-_ _ _ _ . __---_____a
. TABLE 2 (Continued) ;
{
NDRTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS l l CRITERIA APPLICATION l l
PLANT SYSTEMS !
Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 )
Number Number Title Results Results Notes 1 3.7.1.1 3.7.1.1 Safety Valves Retain Retain 3.7.1.2 3.7.1.2 A'uxiliary Feedwater System Retain Retain 3.7.1.3 3.7.1.3 Emergency Condensate Storage Tank Retain Retain i 3.7.1.4 3.7.1.4 Activity Retain Retain 3.7.1.5 3.7.1.5 Main Steam Trip Valves Retain Retain 3.7.1.6 None Steam Turbine Assembly Not Reviewed Relocate 2 3.7.1.7 3.3.4 Turbine Overspeed Relocate Relocate 3.7.2.1 3.7.2 Steam Generator Pressure /
Temperature Limitation Relocate Relocate l
3.7.3.1 3.7.3 Component Cooling Water Subsystem Retain Retain 3.7.4.1 3.7.4 Service Water System Retain Retain 3.7.5.1 3.7.5 Ultimate Heat Sink Retain Retain 3.7.6.1 3.7.6 Flood Protection Relocate Relocate 3.7.7.1 3.7.7 Control Room Emergency Habitability System Retain Retain 3.7.8.1 3.7.8 Safeguards Area Ventilation System Retain Retain 3.7.9.1 None Residual Heat Removal System, Operating Not Reviewed Retain 15 3.7.9.2 3.4.1.4.1 Residual Heat Removal 3.4.1.4.2 System, Shutdown Retain Retain 11 3.7.10 3.7.9 Snubbers Relocate Relocate A _ _ - - _ _ . _ _ . _ _ . - _ - - - . _ _ _ - - _ - - - - _
1 TABLE 2 (Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION PLANT SYSTEMS (Cont)
Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes 3.7.11.1 3.7.10 Sealed Source Contamination Relocate Relocate 3.7.12.1 None Settlement of Class 1 Structures Not Reviewed Relocate 2 3.7.13 None Groundwater Level, Service Water Reservoir Not Reviewed Relocate 2 i 3.7.14.1 3.7.11.1 Fire Suppression Systems Relocate Relocate 3.7.14.2 3.7.11.3 Low Pressure CO Systems. Relocate Relocate 2
3.7.14.3 3.7.11.3 High Pressure CO2 i Systems Relocate Relocate 3.7.14.4 3.7.11.4 Halon Systems Relocate Relocate 3.7.14.5 3.7.11.5 Fire Hose Stations Relocate Relocate 3.7.14.6 3.7.11.2 Spray and/or Sprinkler Systems Relocate Relocate 3.7.15 3.7.12 Penetration Fire Barriers Relocate Relocate I
l l
I i
i
- __-__a
TABLE 2(Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION ELECTRICAL POWER SYSTEMS Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes 3.8.1.1 3.8.1.1 A.C. Sources, Operating Retain Retain 3.8.1.2 3.8.1.2 A'.C. Sources Shutdown Retain Retain 3.8.2.1 3.8.3.1 Onsite Power Distribution Systems, A.C.
Distribution Operating Retain Retain 3.8.2.2 3.8.3.2 Onsite Power Distribution Systems, A.C.
Distribution Shutdown Retain Retain 3.8.2.3 3.8.2.1 Onsite Power Distribution Systems, D.C.
Distribution Operating Retain Retain 3.8.2.4 3.8.2.2 Onsite Power Distribution )
Systems, D.C '
Distribution Shutdown Retain Retain 3.8.2.5 3.8.4.2 Containment Penetration )
Conductor Overcurrent i i
Protective Devices Relocate Relocate 3.8.2.6 3.8.4.3 Notor-Operateo Valves Thermal Overload i Protection and/or !
Bypass Devices Relocate Relocate 3.8.2.7 None Normally De-energized i Power Circuits Not Reviewed Relocate- 2 3
TABLE 2 (Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION REFUELING OPERATIONS I
Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes l 3.9.1 3.9.1 Boron Concentration Retain Retain 3.9.2 3.9.2 Instrumentation Retain Retain i
3.9.3 3.9.3 Decay Time Retain Retain j i
3.9.4 3.9.4 Containment Building Penetrations Retain Retain 3.9.5 3.9.5 Communications Relocate Relocate 3.9.6 3.9.6 Manipulator Crane Operability Relocate Relocate 3.9.7 3.9.7 Crane Travel, Spent Fuel Pit Relocate Relocate 3.9.8.1 3.9.8.1 Residual Heat Removal 3.9.8.2 and Coolant Circulation, All Water Levels Retain Retain 12 !
l 3.9.8.2 3.9.8.2 Residual Heat Removal I and Coolant Circulation, !
Low Water Level Retain Retain 12 !
3.9.9 3.9.9 Containment Purge and l Exhaust Isolation System Retain Retain !
l 3.9.10 3.9.10 Water Level, Reactor Retain Retain Vessel 3.9.11 3.9.11 Spent Fu21 Pit Water Level Retain Retain :
3.9.12 3.9.12 Fuel Building Ventilation j System Retain Retain 4
TABLE 2 (Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION SPECIAL TEST EXCEPTIONS Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes 3.10.1 3.10.1 Shutdown Margin Relocate Relocate 13 3.10.2 3.10.2 E'roup Height, Insertion and Power Distribution Limits Retain Retain 3.10.3 3.10.3 Physics Tests Retain Retain 3.10.4 3.10.4 Reactor Coolant Loops Retain Retain i 3.10.5 3.10.5 Position Indicator Channels, Shutdown Relocate Relocate 4
l l
l l
l
TABLE 2 (Continued)
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS CRITERIA APPLICATION RADI0 ACTIVE EFFLUENTS Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes 3.11.1.1 3.11.1.1 Liquid Effluents Concentratier. Relocate Relocate 7 q
3.11.1.2 3.11.1.2 Dose Relocate Relocate 7 3.11.1.3 3.11.1.3 Liquid Radwaste Treatment Relocate Relocate 7 i
l 3.11.1.4 3.11.1.4 Liquid Holdup Tanks Relocate Relocate 3.11.2.1 3.11.2.1 Gaseous Effluents, i Dose Rate Relocate Relocate 7 l
3.11.2.2 3.11.2.2 Dose, Noble Gases Relocate Relocate 7 3.11.2.3 3.11.2.3 Dese, Iodine 131. Tritium and Radionuclides in Particulate Form Relocate Relocate 7 3.11.2.4 3.11.2.4 Geseous Radwaste Treatment Relocate Relocate 7 I
3.11.2.5 3.11.2.6 Explosive Gas Mixture Relocate Relocate 3.11.2.6 3.11.2.5 Gas Storage Tanks Relocate Relocate 3.11.3 3.11.3 Solid Radioactive Weste Relocate Relocate 7 3.11.4 3.11.4 Total Dose Relocate Relocate 7
i TABLE 2(Continued) i i
NORTH ANNA UNIT 2 TECHNICAL SPECIFICATIONS
]
CRITERIA APPLICATION RADIOLOGICAL ENVIRONMENTAL NONITORING Unit 2 STS-Rev 5 North Anna Unit 2 NRC Unit 2 Number Number Title Results Results Notes 1
3.12.1 3.12.1 Nonitoring Program Relocate Relocate 7 l l
3.12.2 3.12.2 L'and Use Census Relocate Relocate 7 3.12.3 3.12.3 Interlaboratory Comparison Program Relocate Relocate 7 I
i 1
s I
_ _ - - - - - - - - - - - - 1
TABLE 2(Continued)
NOTES -
- 1) Reference 2 allows relocation for Modes 1 and 2, and specifies retention for Modes 3, 4 and 5. The Modes 1 and 2 requirements will be relocated.
2)- This LCO was not evaluated in Reference 2.
- 3) Reference 2 allows relocation of this LCO, but specifies retention of the associated surveillance requirements. The entire specification will be retained.
- 4) This LCO will be relocated and the associated surveillance requirements will be retained in,the Technical Specifications per Reference 2.
- 5) The Radiation Effluent Monitoring Instrumentation will be dispositioned in accordance with Reference 3. The remaining instruments will be included in other appropriate specifications.
- 6) The Regulatory Guide 1.97 Type A variables are listed in Table 2A and will be retained; The Regulatory Guide 1.97 Category 1, non-Type A variables are identified and evaluated in Table 4.
- 7) This LCO will be dispositioned in accordance with the provisions of Reference 3.
- 8) The radioactive gaseous effluent monitoring instrumentation will be dispositioned in accordance with the provisions of Reference 3. The explosive gas monitoring instrumentation will be relocated in accordance with Reference 2.
- 9) Unit 2 Technical Specifications LCO 3.4.1.3 applies to Modes 4 and 5.
Standard Te:hnical Specifications LCO 3.4.1.3 pertains only to Mode 4.
Standard Technical Specifications LCO 3.4.1.4.1 and LCO 3.4.1.4.2 address Mode 5 with, respectively, loops filled and loops not filled.
- 10) Unit 2 Technical Specifications LCO 3.4.11.1 applies to Modes 1, 2 and
- 3. Modes 4, 5 and 6 (with vessel head on) are covered in LCO 3.4.9.3.
Standard Technical Specifications LCO 3.4.11 covers Modes 1, 2, 3 and 4.
Modes 4, 5 and 6 (with vessel head on) are covered in LCO 3.4.9.3. Note, Wode 4 is covered in both sections of the Standard Technical Specifications.
- 11) North Anna Unit 2 LCO 3.7.9.2 is applicable in Modes 4 and 5. Standard Technical Specifications LCO 3.4.1.4.1 and LCO 3.4.1.4.2 are applicable in Mode 5 with reactor coolant loops filled and reactor coolant loops not filled, respectively.
- 12) Standard Technical Specifications LCO 3.9.8.1 addresses Residual Heat Removal and Coolant Circulation - High Water Level. Standard Technical Specifications LCO 3.9.8.2 addresses Residual Heat Removal and Coolant Circulation - Low Water Level.
TABLE 2 (Continued)
NOTES
- 13) Reference 2 allows relocation of the Mode 2 requirements for LCO 3.1.1.1. The Mode 2 requirements for LCO 3.1.1.1 will be relocated per Note 1 above. Therefore, LCO 3.10.1 will not be needed and will be relocated.
- 14) This LCO will be relocated; however, Pa, La, Ld and Lt will be retained in the Technical Specifications per Reference 2.
- 15) This LOO was not evaluated in Reference 2. Reference 4 requires that the Residual Heat Removal System be retained .in the Technical Specifications.
l 1
l I
l
TABLE 2A NORTH ANNA UNIl 2 POST ACCIDENT MONITORING INSTRUMENTATION REGULATORY GUIDE 1.97 TYPE A VARIABLES
- 1. Steam generator narrow range level
- 2. Steam generator pressure
- 3. Core exit temperature (incore thermccouples) 4 .- Reactor coolant system (RCS) cold leg temperature
- 5. RCS hot leg temperature
- 6. RCS wide range pressure
- 7. Emergency condensate storage tank level
- 8. Refueling water storage tank level
- 9. Pressurizer power operated relief valve (PORV) position indication
- 10. Pressurizer level
- 11. Containment intermediate range pressure
- 12. Reactor coolant system subcooling margin monitor (degree of subcooling)
- 13. Containment sump wide range level
- 14. Containment area radiation I
- 15. Auxiliary feedwater flow
]
l l
~
Technical Specification Screening Form TABLE 3 NDRTH ANNA UNIT 1 (1)TECHNICALSPECIFICATION Boron Dilution, RCS Flow TS LOCATION LCO 3.1.1.3.1 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect.
WT and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (R) A process variable that is an initial condition WT of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission ,
product barrier.
X (3) A structure, system or component that is part of WT the primary success path and which functions or 4 actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "NO" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
Technical Specification Serrening Ferm TABLE 3 (Continued) .
(3) DISCUSSIONS This specification identifies the minimum flow rate of reactor coolant .I
~
through the reactor coolant system whenever a reduction in the system's boron concentration is being made. The specification applies to all modes of operation. A flow rate of at least 3000 gpm will circulate an equivalent reactor coolant system volume in approximately 30 minutes.
This provides adequate mixing, prevents stratification and ensures gradual change of reactivity. i The objective of th.is specification is to ensure that the rate of reactivity change, associated with boron reductions, will be within specified limits for operator recognition and control. This limit on the flow through the reactor coolant system is not necessary to mitigate the consequences of a DBA or Transient.
The lower limit on the flow rate of reactor coolant is not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
The flow limit 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. This specification does not satisfy criterion 2.
The limit on flow in the reactor coolant system is not part of the primary success path which 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. This specification does not satisfy criterion 3.
Based on the PRA Sumary Report contained in Appendix A, the Boron Dilution, RCS Flow has not been identified as a significant risk contributor.
(4) CONCLUSION This current Technical Specification is included in the revised 4 Technical Specification document.
)
X This current Technical Specification may be relocated to another controlled document. 1 1
Technical Specification Scretning Form TABLE 3 (Contin 9ed)
NORTH ANNA UNIT 1 (1) TECHNICAL SPECIFICATION Boron Dilution, Valve Position TS LOCATION LCO 3.1.1.3.2 (2) EVALUATION BASED ON POLICY STATENENT CRITERIA ,
Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, "TEF T and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition WT of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of-or presents a challenge to the integrity of a fission product barrier. 1 X (3) A structure, system or component that is part of Y ~~F the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical i Specifications. 1 If the answer to all three of the above questions is "NO" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
1 l
- - - - - - - - - - - - - - - - - - - - _ - - - - - - - - - )
q Technical Specification Scr ening Form i
1 TABLE 3 (Continued)
I q
(3) DISCUSSIONS J
The primary grade water, which is supplied from the primary water tanks, i is employed for various makeup and flushing operations and for adjusting I the reactor coolant boron concentration for reactivity control. To )
prevent unplanned dilution of the boron concentration in the reactor j coolant system, it is necessary to control the isolation of the primary grade water from the borated water in the reactor coolant system.
This specification provides the administrative control for the isolation of primary grade water from the regular, borated Chemical and Volume Control System water source from the volume control tank to the reactor coolant reystem. The specification regulates the o>ening/ closing of the isolation valves between primary grade water and tie Chemical and Volume Control System borated water source.
The specification does not address instrumentation that is used to detect and/or indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1. a l
In the analysis of the Uncontrolled Boron Dilution Event it is assumed that uncontrolled dilution flow cannot take place from the primary water tanks to the RC System in the shutdown modes. Certainty for this requires positive isolation of the relevant flow paths with strict control of the operation of the isolation valves. Therefore, this specification represents 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. This specification does satisfy criterion 2 and must, accordingly, be retained in the Technical Specifications.
The isolation valves are not a structure, system or component that is part of the primary success path and which functions or actuates to mitigate a DBA or Transient. This specification does not satisfy criterion 3. )
l (4) CONCLUSION X This current Technical Specification is included in the revised Technical Specification document.
This current Technical Specification may be relocated to another controlled document. J l
l
Technical Specification Scraenino Form TABLE 3 (Continued) 1 1
NORTH ANNA UNIT 1 (1) TECHNICAL SPECIFICATION Boric Acid Transfer Pumps Shutdown TS LOCATION LCO 3.1.2.5 (2) EVALUATION BASED ON PCLICY STATENENT CRITERIA ,
1 Is the Technical Specification applicable to: 1 X (1) Installed instrumentation that is used to detect, WT and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition YT of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of
. YT the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical i Specifications.
If the answer to all three of the above questions is *NO" and the system / component has not been shown to be a significant risk contributor, ,
the Technical Specification requirements may be relocated to another l controlled document. ;
L !
l l
\
y i
Technical Specification Screening Form TA%E 3 (Continued)
(3) DISCUSSIONS l J '
The boric acid transfer pumps are part of the boration subsystem o: the Chemical and Volume Control System. The boration subsystem provides the -
means to speet one of the functional requirements of the Chemical and ,
Volume Control System, i.e., to control the chemical neutron absorber f (boron) concentration in the Reactor Coolant System and to help maintain the shutdown margin. To accomplish this functional requirement, the current specifications require a source of borated water and one or more flow paths to inja.ct this borated water into the Reactor Coolant System.
The boration subsystem is not assumed to be operable to mitigate the consequences of a DBA or Transient. In the case of a malfunction of the Chemical and Volume Control System, which cause a boron dilution event, ;
the automatic response, or that required by the operator, is to close the i appropriate valves in the Reactor Nakeup System. This action is required before the shutdown margin is lost. Operation of the boration subsystem .
I is not assumed to mitigate this event.
The boric acid transfer p' umps in the boration subsystem are not installed 1 instrumentation that is used to detect and indicate in the control room a j significant abnormal degradation of the reactor coolant pressure '
boundary. This specification does not satisfy criterion 1. .
The boric acid tranufer pumps in the boration subsystem are not process 1 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. This specification does not satisfy criterion 2.
The boric acid transfer pumps in the boration subsystem are not part of the primary success path and which function or actutte to mitigate a DBA or Transient, that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not satisfy criterion 3.
Based on the PRA Summary Report contained in Appendix A, the Boric Acid Transfer Pumps, Shutdown have not been identified as a significant risk contributor.
(4) CONCLUSION This current Technical Specification is included in the revised Technical Specification document.
1 This current Technical Specification may be relocated to another controlled document.
Tcchnical Specification Screening Form TABLE 3 (Continued)
NORTH ANNA UNIT 1 (1) TECHNICAL SPECIFICATION Boric Acid Transfer Pumps, Operating.
TS LOCATION LCO 3.1.2.6 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, WT and indicate in the control room, a significant l abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition WT of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3)Astructure,systemorcomponentthatispartof
~TET ~" W the primary success path and which functions or actuates to mitigate a Design Basis Accident or J Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical i Specifications. ;
If the answer to all three of the above questions is "NO" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
l i
Technical Specification Screening Form TABLE 3(Continued)
(3) DISCUSSIONS The boric acid transfer pumps are part of the boration subsystem of the Chemical and Volume Control System. The boration subsystem provides the means to meet one of the functional requirements of the Chemical and Volume Control System, i.e., to control the chemical neutron absorber (boron) concentration in the Reactor Coolant System. To accomplish this functional requirement, the current specifications require a source of berated water and one or more flow paths to inject this borated water into the Reactor Coolant System.
The boration subsys' tem is not assumed to be operable to mitigate the consequence of a DBA or Transient. In the case of a malfunction of the Chemical and Volume Control System, which causes a boron dilution event, the automatic response, or that required by the operator, is to close the appropriate valves in the reactor makeup system. Operation of the boration subsystem is not assumed to mitigate this event.
The boric and transfer pumps in the boration subsystem are not installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reacte coolant pressure g boundary. This specification does not satisfy criterion 1. [
The boric acid transfer pumps in the boration subsystem are not process variables that are initial conditions of a DBA or Transient Analysis that either assume the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not satisfy criterion 2.
The beric acid transfer pumps in the boration subsystem are not part of the primary success path and which 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. This specification does not satisfy criterion 3.
Based on the PRA Summary Report contained in Appendix A, the Boric Acid Transfer Pumps, Operating have not been identified as a significant risk contributor.
(4) CONCLUSION I
This current Technical Specification is included in the revised Technical Specification document.
X This current Technical Specification may be relocated to another controlled document.
Technical Specification Scraening Form TABLE 3(Continued) ;
I NORTH ANNA UNIT 1 (1) TECHNICAL SPECIFICATI0d A;cident Monitoring Instrumentation.
i TS LOCATION LCO 3.3.3.6 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, W-F and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition W ~" F of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of WT the primary success path and.which functions or actuates to mitigate a Design Basis Accident.or Transient that either assumes'the failure of or presents a challenge.to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "N0" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
-4B-
c- .
Technical Specification Serenning Form TABLE 3 (Continued)
(3) DISCUSSIONS The Accident Monitoring Instrumentation ensures that sufficient information is available following an accident to allow the operator to verify the response of automatic safety systems, and to take preplanned manual actions to accomplish a safe shutdown of the plant.
l The Accident Monitoring Instrumentation is not intended to be a leading
! indicator of RCS leakage. Although accident monitoring instruments l respond to the consequences of a LOCA, the instruments captured by l
criterion 1 are instruments which give some indication of RCS leakage j l prior to the LOCA event itself. The Accident Monitoring Instrumentation {'
is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary prior to a LOCA event. The Accident Monitoring Instrumentation does not satisfy criterion 1.
The Accident 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 barrier. The Accident Monitoring Instrumentation does not l satisfy criterion 2.
Specific Accident Monitoring Instrumentation provides the operator with the information needed to perform the required manual actions to bring the plant to a stable condition following an accident. This instrumentation is a component that is part of the primary success path and which 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. Specific Accident Monitoring Instrumentation satisfies criterion 3.
Reference 2 stipulates that Post-Accident Monitoring Instrumentation -
that satisfies the definition of Type A variables in Regulatory Guide 1.97 meets criterion 3 and should be retained in Technical Specifications. The North Anna Unit 1 Type A variable instruments are listed in Table 1A and will be retained in the technical specifications. )
Reference 2 also stipulates that non-Type A, Category 1 instruments are to be evaluated for inclusion in the technical specifications. The evaluation for the North Anna Unit I non-Type A, Category 1 instruments follows.
- 1. Neutron Flux (Gamma-Metrics)
Neutron Flux is identified in Regulatory Guide 1.97 as a Type B, Category 1 variable. In the North Anna Emergency Operating Procedures (EOP), neutron flux is the specified means to verify maintenance of the subtriticality safety function, and is to be continually monitored during E0P usage. Indication of significant ti
(_ .
Technical Soecificatibn Screenino Form TABLE 3 (Continued) post-trip power generation (>5% of rated power) results in entry of a Function Restoration Procedure (FRP) designed to ensure adequate shutdown reactivity-Based on the significance of this variable in the E0P's, neutron flux will be included in the new Technical Specifications.
The Gamma-Metrics channels are not currently included in the Technical Specifications. In May of 1988 Gamma Metrics reported a problem with cable leakage leading to potential inoperability of their excore neutron flux monitoring system in accordance with 10 CFR Part 21 (Ref. 1). Virginia Power prepared Justification for Continued Operation (JCO) 89-06, as allowed for in USNRC Generic Letter 88-07, based on the availability of adequate administrative controls to ensure that the subtriticality safety function is performed. The JC0 is intended to cover operation through the end of Unit 1, Cycle 8 and Unti. 2. Cycle 7, or a subsequent outage of sufficient duration when a qualified cable is available. During the refueling outages following these fuel cycles, the Gamma-Metrics system would be repaired using environmentally qualified cable seal kits which will be available at that time.
Virginia Power will therefore add neutron flux (Gamma-Metrics) to the North Anna MERITS Technical Specifications after this instrumentation is returned to a fully operable status.
REFERENCE
- 1. Letter from C. L. Lingren, Vice President-Gamma-Metrics, to Director, Office of Inspection and Enforcement, USNRC, pursuant to 10 CFR 21.21(3), May 10, 1988.
The Emergency Procedures make use of a number of RVLIS setpoints related to RCS inventory control, indication of Inadequate Core Cooling (ICC) and related applications. These include:
- Indication of ICC - separate setpoints are defined for various combinations of operating reactor coolant pumps.
An alternate to RCS subcooling and pressurizer level as a safety injection (SI) initiation criterion. ,
A means of controlling charging flow if pressurizer level l indication is not available.
A means of controlling voiding in the upper head region during natural circulation, or to determine if such a void exists.
- A means of determining if residual system (RHR) operation will be effective (i.e., collapsed liquid level at or above the top of the hot legs).
I Technical Sp:cification Screening Form The detection of ICC represents a potential near-term breach of the fuel cladding integrity. Use of RVLIS indication in the Emergency Procedures includes some situations which are within the scope of the design basis accidents (e.g., small break LOCA). RVLIS indication is used for decisions related to top level (Red Path) critical safety function restoration for core cooling. Based on the above, this instrumentation will be included in the technical specifications.
i
- 3. Containment Isolation Valve Position Indication The Emergency Procedures make use of this indication as part of an immediate response to a reactor trip with safety injection actuated.
The operator is directed to confirm containment isolation (Phase A) as an immediate response to any safety injection. If the position indicators indicate any valves are not closed, then manual action is to be taken to close them. Similarly, for containment pressures in excess of the CLS Hi-Hi setpoint, Phase B isolation is to be verified or manually performed, based on operator reference to the control board status lights.
Failure of this indication or failure of the operator to respond to this indication could result in a release path for radioactive materials to escape to the environment. However, a double failure would have to occur. For the design basis events (no double failure) there are Type A variables (containment pressure, sump level, containment area radiation monitors) which provide the operator with information required to perform actions which ensure the containment integrity critical safety function during a design basis accident.
Therefore, this instrumentation should not be included in the technical specifications.
- 4. Radiation Concentration In Primary Coolant -
Regulatory Guide 1.97 defines the purpose of monitoring this variable as " detection of breach" (i.e., of the fuel cladding). The Emergency Operating Procedures do not base any decisions or actions on this variable. This information is not required to take appropriate actions to ensure the integrity of any fission product barrier. This variable does NOT meet Selection Criterion 3 and should not be included in the technical specifications.
- 5. Containment Hydrogen Concentration l
North Anna Unit 1 LCO 3.6.4.1, Hydrogen Analyzers, specifies the l requirements for containment hydrogen concentration instrumentation.
l This LCO will be retained in the technical specifications. It would be redundant to re-state these requirements in the Accident Monitoring Instrumentation technical specification, LCO 3.3.3.6.
Therefore, the Hydrogen Analyzer requirements will not be repeated in LCD 3.3.3.6.
l I
l
Technical Specification Screenino Form
- 6. Steam Generator Level - Wide Range The North Anna Emergency Operating Procedures use steam generator level wide range as an indicator of steam generator dryout and as a criterion for establishing bleed and feed cooling of the reactor coolant system. Loss of steam generator level does not, in and of itself, represent an approach to a breach of clad or RCS integrity.
This instrumentation does provide information required to perform a manual action which directly preserves a critical safety function ,
(i.e., heat sink), but the situations being addressed by these actions (loss of all inventory in steam generators NOT identified as faulted) are outside the scope of the design basis accidents.. Top level (Red Path) actions for restoration of the heat sink safety-function depend on Type A variables (narrow range steam generator level, auxiliary feedwater flow), and not wide range steam generator level. Therefore, this instrumentation should not be included in the technical specifications.
- 7. Containment Press. ore - Wide R'ange Regulatory Guide 1.97 defines the purpose of monitoring this variable as " detection of potential for or actual breach; accomplishment of mitigation". The Emergency Operating Procedures do not base any decisions or actions on this variable. All ar.tions related to containment pressure indication are based on the intermediate range l containment pressure channel, which is a Type A variable for North Anna. Wide range information is not required to take appropriate actions to ensure the integrity of any fission product barrier. This variable does NOT nieet Selection Criterion 3 and should not be included in the technical specifications.
There are several accident monitoring instruments currently in the technical specifications that were not addressed by the above evaluations; their evaluation follows.
A. Boric Acid Tank Solution Level The Emergency Operating Procedures do not base any decisions or actions on this parameter. This parameter is not listed in Regulatory Guide 1.97. This parameter is not a Type A or Category 1 variable for North Anna. )
It will be relocated to an6 her controlled ;
document. j B. Pressurizer PORV Block Valve Position Indicato_t This parameter is not listed in Regulatory Guide 1.97. This parameter is not a Type A or Category 1 variable for North Anna. It will be relocated to another controlled document.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ ./
Technical Specification Scrarning Form C. Pressurizer Safety Valve Position Indicator This' parameter is not a Type A or Category I variable for North Anna. It is a Type D, Category 2 variable. It will be relocated to another controlled document.
D. Containment Water Level - Narrow Range This parameter is not a Type A or Category 1 variable for North Anna. It is a Type B, Category 2 variable. It will be relocated to another controlled document. The Containment Water Level - Wide Range is a Type A variable and will be retained in the technical specifications.
As part of the PRA study contained in Appendix A, the impact of individual accident monitoring instruments on risk was reviewed. No case was identified in which accident monitoring instruments had been explicitly modeled in an accident scenario. Therefore, no accident monitoring instruments were identified as significant risk contributors.
(4) CONCLUSION X This current Technical Specification is included in the revised Technical Spect'ication document.
~~~ This current Technical Specification may be relocated to another controlled document.
I Tachnical Specification Scraning Form TABLE 3(Continued)
N_0RTH ANNA UNIT 1-(1) TECHNICAL SPECIFICATION Primary to Secondary Leakage.
TS LOCATION LCO 3.4.6.3 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect.
TE!i T and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition TET T of.a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of TEF- -' F the primary success path and which functions or-actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "N0" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
Technical Specification Scraening Form TABLE 3 (Continued)
(3) DISCUSSIONS The basic specification (LCO 3.4.6.2) on primary to secondary leakage limits the leakage to 1 GPM total for all three steam generators and 500 gallons per day through any one steam generator. This specification (LCO 3.4.6.3) provides additional primary to secondary leakage limits which have been imposed for operation above 50% of rated thermal power.
This specification does not contain requirements for installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
The limits on primary to secondary leakage which establish the initial !
conditions for DBAs such as the steam generator tube rupture and main steamline break are specified in LCO 3.4.6.2. Operational leakage is an input variable to the offsite dose calculations associated with certain Design Basis Accidents which result in the release of secondary coolant.
However, it is not a process variable that is controlled by the operator. The limits on this variable which are assumed in the analyses are maintained by a separate LCO (3.4.6.2) which is being retained.
Therefore this specification does not satisfy criterion 2.
Reactor Coolant System operational leakage is not a structure, system or component that is part of the primary success path and which 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. This specification does not satisfy criterion 3.
Based on a review of available literature, primary to secondary leak rate surveillance is not modeled as a contributing factor which limits the likelihood or severity of any accident sequence cormonly found to
. dominate offsite health effects, and has not been identified as a 1 significant risk contributor.
(4) CONCLUSION l This current Technical Specification is included in the revised l Technical Specification document. ,
X This current Tecnnical Specification may be relocated to another l controlled document.
i
________________-_.1__-_ --- --
.j
Technical Specification Scresnino Form TABLE 3 (Continued)
NORTH ANNA UNIT 1-(1) TECHNICAL SPECIFICATION Primary to Secondary Leakage Detection Systems.
TS LOCATION LCO 3.4.6.4 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, YE5- - RU- and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition YE5- N0 of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of "YE5- - RU- the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical Specifications. .l l
If the answer to all three of the above questions is "NO" and the i system / component has not been shown to be a significant risk contributor, i the Technical Specification requirements may be relocated to another !
controlled document, j
i i
i
__-__-_----_-_m
Technical Specification Screenino Form x
TABLE 3 (Continued) 1 (3) DISCUSSIONS l The purpose of the instrumentation required by this Specification is the
( '
early detection of steam generator tube degradation via small (of the order of 100 gpd) primary to secondary leaks. This will provide ample time for controlled actions to reduce power prior to a major tube-failure. ,
i This instrumentation was installed specifically to provide a sensitive means of early detection of rapidly propagating fatigue cracks in the U-bend region of the tubes. This failure mode was identified as the l cause of a tube rupture experienced on Unit 1 in. July 1987.. Extensive modifications and engineering analyses have been performed to reduce the potential for such fatigue induced tube failures. These have included a detailed engineering analysis to identify specific tubes suscept%1e to this mechanism, tube stabilization measures, reduction of local flow velocities.in the tube bundle region via hydraulic restriction of the recirculating secondary coolant (installation of downcomer flow resistance plates), and a preventive plugging program. As.such, the conditions which led to the-incorporation of the primary to secondary leakage detection system have essentially been mitigated.
1 This specification does not contain requirements for installed I instrumentation that is used to detect and indicate in the control room a j significant abnormal degradation of the reactor coolant pressure' boundary. Pressure boundary leakage is defined ~in NUREG-0452 J l
(Westinghouse Standardized Technical Specifications):
)
"1.19 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall."
The instrumentation covered in LCO 3.4.6.4 does not measure pressure 3 boundary leakage, and therefore is not used for detection of pressure 1 boundary degradation. There is other instrumentation, used to perform reactor coolant system leakage detection, which is being retained in the Technical Specifications. This includes the containment atmosphere particulate radioactivity monitoring system, the containment atmosphere gaseous radioactivity monitoring system, and the containment sump
)
discharge flow measurement system. .In addition, LCO 3.4.6.2 limits i primary to secondary leakage to 1 gpm total and 500 gpd from any one .
generator. Adequate procedures and instrumentation will remain operable '
to meet the surveillance requirements to support this specification.
i i
l l
Tnchnical Specification Scraening Form 4
.LCO 3.4.6.4 does not satisfy criterion 1.
The Primary to Secondary Leakage Detection Systems 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. The Primary to Secondary Leakage Detection Systems do not satisfy criterion 2.
The Primary to Secondary Leakage Detection Systems are not systems that are part of the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These systems do not actuate any automatic function or provide-any information leading to an operator action assumed in the design basis accident analysis. The Primary to Secondary Leakage Detection Systems do not satisfy criterion 3.
Based on a review of available literature, Primary to Secondary Leakage Detection Systems are not modeled as a contributing factor which limits the likelihood or severity of any accident sequence commonly found to dominate offsite health effects, and have not been identified as a significant risk contributor.
(4) CONCLUSION This current Technical Specification is included in the revised Technical Specification document.
X This current Techniedl Specification may be relocated to another controlled document.
i i
I
f .
1 l
Technical Specification Screening Form TABLE 3 (Continued)
NORTH ANNA UNIT 1 (1)_ TECHNICAL SPECIFICATION Steam Turbine Assembly TS Location LCO 3.7.1.6 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, WT and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure-boundary.
X (2) A process variable that is an initial condition of a WT Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of T NU the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions in "YES", then the Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "NO" and the system /
component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
59-
Technical Specification Scra ning Form l TABLE 3 (Continued)
(3)-DISCUSSIONS This specification provides the control of the structural integrity of the turbine assembly. The structural integrity of the steam turbine l assembly shall be demonstrated at least once per 40 months, during shutdown. At the least once per 10 years the turbine must be disassembled and all normally inaccessible parts inspected.
, Demonstration of the structural integrity of the steam turbine assembly ensures an acceptable low probability of generation of turbine missile (s) which could damage safety related equipment.
The specification does not address instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
No process variable is involved 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. This specification does not satisfy criterion 2.
The steam turbine assembly is not a structure, system or component that is part of the primary success path and which 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. This specification does not satisfy criterion 3.
Based on the PRA Summary Report contained in Appendix A, the integrity of the turbine assembly has not been identified as a significant risk contributor.
(4) CONCLUSION This current Technical Specification is included in the revised Technical Specification document.
X This current Technical Specification may be relocated to another controlled document.
Tschnical Specification Scrnning Form TABLE 3 (Continued)
NORTH ANNA UNIT 1 (1) TECHNICAL SPECIFICATION Settlement of Class 1 Structures TS LOCATION tCO 3.7.12.1 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, WT and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition WT of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or
. presents a challenge to the integrity of a fission l product barrier.
X (3) A structure, system or component that is part of W '"ND the primary success path and which functions or !
actuates to mitigate a Design Basis Accident or-Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "NO" and the system / component has not been shown to be a significant risk contributor, j the Technical Specification requirements may be relocated to another l controlled document.
{
l
}
i
Tochnical Specification Scrmning Form
. TABLE 3(Continued)
.(3) DISCUSSIONS This Technical Specification addresses the monitoring of the settlement of class 1 structures. The program was developed in 1975 after numerous communications with the NRC. The testing approach, sequences and allowable settlements were included in the Technical Specifications.
The Technical Specifications call for an engineering review, evaluation of the consequences and submittal of report to NRC when the settlement exceeds 75 percent of allowable limits. This ensures adequate time to implement corrective actions to prevent reaching 100 percent of allowable limits, at which time a step-wise plant shut down would have to be initiated.
The monitoring and engineering review of the settlement of class 1 structures do not represent installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
The surveillance of the settlement of class I structures does not represent a process variable that is an initial condition of a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not satisfy criterion 2.
-The surveillance of class I structures is not a structure, system or component that is part of the primary success path and which functions to mitigate a Design Basis Accident or Transient. This specification does not satisfy criterion 3.
Based on the PRA Sumary Report contained in Appendix A, this specification has not been identified as a significant risk contributor. l (4) CONCLUSION This current Technical Specification is included in the revised Technical Specification docu:nent. l X This current Technical Specification may be relocated to another controlled document.
l i
._.__- _ _ _ U
Technical Specification Scrc ning Form TABLE 3(Continued)
NORTH ANNA UNIT 1 (1) TECHNICAL SPECIFICATION Groundwater Level, Service Water Reservoir l TS LOCATION MO3.7.13 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA l .{
l Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, '
TET T and indicate in the control room, a significant '
abnormal degradation of the re ctor coolant pressure boundary.
X (2) A process variable that is an initial condition WT of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or i presents a challenge to the integrity of a fission i product barrier.
X (3) A structure, system or component that is part of W ~" T D'- the primary success path and which functions or actuates to mitigate a Design Basis Accident or i Tretsient that either assumes the failure of or i presents a challenge to the integrity of a fission !
product barrier.
! If the answer to any one of the above questions is "YES", then the !
Technical Specifications shall be included in the new h hnical Specifications. ,
If the answer to all three of the above questioniis "NO" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another ,
controlled document. '!
1 l
l_ _ _ _ _ _ _ - - _ -
q
- T:chnical Specification Scr:ening Form ;
TABLE 3(Continued) i (3) DISCUSSIONS ,
This Technical Specification addresses the requirements for the monitoring of water level and water drainage from the service water !
reservoir. The actions to be implemented when the specification is exceeded consist of engineering studies with reporting of relevant findings and corrective actions to the NRC. Hence, the Technical Specification is outlining a long term surveillance program with no immediate actions short of the initiation of an engineering evaluation and reporting of the findings. This will ensure reservoir integrity and full compatibility between plant operation and the availability of service water. .'
This specification does not involve installed instrumentation that is used to detect and indicate in the control room significant abnormal 3 degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1. {-
This specification does not involve 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. This specification does not satisfy' criterion 2.
The specification is not a structure, system or component that is part of the primary success path and which functions to mitigate a DBA or Transient that either assures the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this 1, specification does not satisfy criterion 3.
Based on the PRA Summary Report contained in Appendix A, this specification has not been identified as a significant risk contributor.
(4) CONCLUSION This current Technical Specification is included in the revised Technical Specification document.
1 This current Technical Specification may be relocated to another 4 controlled document. ;
4 l.
l
.I i Technical Speci'fication Screaning Form TABLE 4 NORTH ANNA UNIT 2 (1) TECHNICAL SPECIFICATION Boron Dilution, Valve Position TS LOCATION LCO 3.1.1.3 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA I Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, W F' and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition W N0 of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of W ND the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "NO" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
1 l
l l
I J
l
Technical Specification Screening Form TABLE 4 (Continued)
(3) DISCUSSIONS The primary grade water, which is supplied frem the primary water tanks, is employed for various makeup and flushing operations and for adjusting the reactor coolant boron concentration for reacMvity control. To prevent unplanned dilution of the boron concentration in the reactor coolant system, it is necessary to control the isolation of the primary grade water from the borated water in the reactor coolant system.
This specification provides the administrative contrc,1 for the isolation of primary grade water from the regular, borated Chemical and Volume Control System water source from the volume control tank to the reactor coolant system. The specification regulates the opening / closing of the isolation valves between primary grade water and the Chemical and Volume Control System borated water source.
The specification does not address instrumentation that is used to detect and/or indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
In the analysis of the Uncontrolled Boron Dilution event it is assumed that uncontrolled dilution flow from the primary water tanks to the RC System cannot take place in the shutdown modes. Certainty for this requires positive isolation of the relevant flow paths with strict control of the operation of the isolation valves. Therefore, this specification represents 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. This specification does satisfy criterion 2 and must, accordingly, be retained in the Technical Specifications.
The isolation valves are not a structure, system or component that is part of the primary success path and which functions or actuates to-mitigate a DBA or Transient. This specification does not satisfy criterion 3.
(4) CONCLUSION !
X This current Technical Specification is included in the revised Technical Specification document.
This current Technical Specification may be relocated to another controlled document.
1
1 Technical Specification Screening Form TABLE 4(Continued)
NORTH ANNA UNIT 2 (1) TECHNICAL SPECIFICATION Accident Monitoring Instrumentation.
I TS LOCATION LCO 3.3.3.6 l 1
(2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification aprflicable to:
X (1) Installed instrumentation thct is used to detect, W NO and indicate in the control roem, a significant "
abnormal degradation of the reactor coolant pressere boundary.
X (2) A process variable that is an initial condition W N0 of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of W NO the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical i Specifications. l If the answer to all three of the above questions is "NO" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document, i
p Technical Specification Screening Form i
TABLE 4 (Continued) 1 (3) DISCUSSIONS The Accident Monitoring Instrumentation ensures that sufficient information is available following an accident to allow the operator to .
verify the response of automatic safety systems, and to take preplanned I manual actions to accomplish a safe shutdown of the plant. I The Accident Monitoring Instrumentation is not intended to be a leading indicator of RCS leakage. Although accident monitoring instruments respond to the consequences of a LOCA, the instruments captured by criterion 1 are instruments which give some indication of RCS leakage prior to the LOCA event itself. The Accident Monitoring Instrumentation is not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary prior to a LOCA event. The Accident Monitoring Instrumentation does not satisfy criterion 1.
1 The Accident 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 barrier. The Accident Monitoring Instrumentation does not satisfy criterion 2.
Specific Accident Monitoring Instrumentation provides the operator with the information needed to perform the required manual actions to bring the plant to a stable condition following an accident. This instrumentation is a component that is part of the primary success path and which 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. Specific Accident Monitoring Instrumentation satisfies criterion 3.
1 Reference 2 stipulates that Post-Accident Monitoring Instrumentation that satisfies the definition of Type A variables in Regulatory Guide 1.97 meets criterion 3 and should be retained in Technical Specifications. The North Anna Unit 2 Type A variable instruments are listed in Table 2A and will be retained in the technical ~ specifications.
Reference 2 also stipulates that non-Type A, Category 1 instruments are to be evaluated for inclusion in the technical specifications. The ,
evaluation for the North Anna Unit 2 non-Type A, Category 1 instruments follows.-
- 1. Neutron Flux (Gamma-Metrics) l l
Neutron Flux is identified in Regulatory Guide 1.97 as a Type B, Category 1 variable. In the North Anna Emergency Operating ;
Procedures (EOP), neutron flux is the specified means to verify maintenance of the suberiticality safety function, and is to be l
! continually monitored during EOP usage. Indication of significant l post-trip power generation (>5% of rated power) results in entry of a Function Restoration Procedure (FRP) designed to ensure adequate shutdown reactivity. '
Technical Specification Screenina Form TABLE 4 (Continued)
Based on the significance of this variable in the E0P's, neutron flux will be included in the new' Technical Specifications.
The Gamma-Metrics channels are not currently included in the Technical Specifications. In May of 1988 Gamma Metrics reported a problem with cable leakage leading to potential inoperability of their excore neutron flux monitoring system in accordance with 10 CFR Part 21 (Ref. 1). Virginia Power prepared Justification for Continued Operation (JCO) 89-06, as allowed for in USNRC Generic Letter 88-07, based on the availability of adequate administrative controls to ensure that the subtriticality safety function is performed. The JC0 is intended to cover operation through the end of Unit 1, Cycle 8 and Unit 2, Cycle 7 or a subsequent outage of sufficient duration when a qualified cable is available. During the refueling outages foilowing these fuel cycles, the Gamma-Metrics system would be repaired using environmentally qualified cable seal kits which will be available at that time.
Virginia Power will therefore add neutron flux (Gamma-Metrics) to the North Anna MERITS Technical Specificatic.ns after this instrumentation is returned to a fully operable status.
REFERENCE
- 1. Letter from C. L. Lingren, Vice President-Gamma-metrics, to Director, Office of Inspection and Enforcement, USNRC, pursuant to 10 CFR 21.21(3), May 10, 1988.
The Emergency Procedures make use of a number of RVLIS setpoints related to RCS inventory control, indication of Inadequate Core Cooling (ICC) and related applications. These include:
- Indication of ICC - separate setpoints are defined for various combinations of operating reactor coolant pumps.
- An alternate to RCS subcooling and pressurizer level as a safety injection (SI) initiation criterion.
A means of controlling charging flow if pressurizer level indication is not available.
A means of controlling voiding in the upper head region during natural circulation, or to determine if such a void exists.
A means of determining if residual system (RHR) operation will be effective (i.e., collapsed liquid level at or above the top of the hot legs).
Technical Specification Screenino Form The detection of ICC represents a potential near-term breach of the fuel cladding integrity. Use of RVLIS indication in the Emergency Procedures includes some situations which are within the scope of the design basis accidents (e.g., small break LOCA). RVLIS indication is used for decisions related to top level (Red Path) critical safety function restoration for core cooling. Based on the above, this
. Instrumentation will be included in the technical specifications.
- 3. Containment Isolation Valve Position Indication The Emergency Procedures make ute of this indication as part of an immediate response to a reactor trip with safety injection actuated.
l The operator is directed to confirm containment isolation (Phase A)
! as an immediate responsfi to any safety injection. If the position indicators indicate any valves are not closed, then manual action is to be taken to close therm. Similarly, for containment pressures in excess of the CLS Hi-H1 setpoint, Phase B isolation is to be verified or manually performed, based on operator reference to the control board status lights.
Failure of this indication or failure of the operator to respond to this indication could result in a release path for radioactive l
materials to escape to the environment. However, a double failure l would have to occur. For the design basis events (no double failure) there are Type A variables (containment pressure, sump level, I containment area radiation monitors) which provide the operator with information required to perform actions which ensure the containment integrity critical safety function during a design basis accident.
Therefore, this instrumentation should not be included in the l' technical specifications.
- 4. Radiation Concentration In Primary Coolant Regulatory Guide 1.97 defines the purpose of monitoring this variable as " detection of breach" (i.e., of the fuel cladding). The Emergency Operating Procedures do not base any decisions or actions on this ,
variable. This information is not required to take appropriate l actions to ensure the integrity of any fission product barrier. This !
variable does NOT meet Selection Criterion 3 and should not be included in the technical specifications.
- 5. Containment Hydrogen Concentration North Anna Unit 2 LCO 3.6.4.1, Hydrogen Analyzers, specifies the requirements for containment hydrogen concentration instrumentation.
This LCO will be retained in the technical specification. It would be redundant to re-state these requirements in the Accident >
Monitoring Instrumentation technical specification, LCO 3.3.3.6.
Therefore, the Hydrogen Analyzer requirements will not be repeated in LCO 3.3.3.6.
q Technical Soecificat16n Screening Form
- 6. Steam Generator Level - Wide Range The North Anna Emergency Operating Procedures use steam generator !
level wide range as an indicator of steam generator dryout and as a criterion for establishing bleed and feed cooling of the reactor coolant system. Loss of steam generator level does not, in and of itself, represent an approach to a breach of clad or RCS integrity.
This instrumentation does provide information required to perform a manual action which directly preserves a critical safety function l
(i.e., heat sink), but the situations being addressed by these actions (loss of all inventory in steam generators NOT identified as j faulted) are outside the scope of the design basis accidents. Top {
level (Red Path) actions for restoration of the heat sink safety {
function depend on Type A variables-(narrow range steam generator j level, auxiliary feedwater flow), and not wide range steam generator level. Therefore, this instrumentation should not be included in the technical specifications.
- 7. Containment Pressure - Wide Range Regulatory Guide 1.97 defines the purpose of monitoring this variable as " detection of potential for or actual breach; accomplishment of mitigation". The Emergency Operating Procedures do not base any decisions or actions on this variable. All actions related to containment pressure indication are based on the intermediate range containment pressure channel, which is a Type A variable for North Anna. Wide range information is not required to take appropriate actions to ensure the integrity of any fission product barrier. This variable does NOT meet Selection Criterion 3 and should not be included in the technical specifications.
There are several accident monitoring instruments currently in the technical specifications that were not addressed by the above evaluations; their evaluation follows.
A. Boric Acid Tank Solution level The Emergency Operating Procedures do not base any decisions or actions on this parameter. This parameter is not listed in Regulatory Guide 1.97. This parameter is not a Type A or Category 1 variable for North Anna. It will be relocated to another controlled document.
B. Pressurizer PORV Block Valve Position Indicator This parameter is not listed in Regulatory Guide 1.97. This parameter is not a Type A or Category 1 variable for North Anna. It will be relocated to another controlled document.
Technical Specification Screenina Form C. Pressurizer Safety Valve Position Indicator This parameter is not a Type A or Category 1 variable for North Anna. It is a Type D,' Category 2 variable. It will be relocated to another controlled document.
D. Containment Water Level - Narrow Range This parameter is not a Type A or Category 1 variable for North Anna. It is a Type B, Category 2 variable. It will be relocated to another controlled document. The Containment Water Level - Hide Range is 3 Type A variable and will be retained in the technical specifications.
As part of the PkA study contained in Appendix A, the impact of individual accident monitoring instruments on risk was reviewed. No case was identified in which accident monitoring instruments had been explicitly modeled in an accident scenario. Therefore, no accident monitoring instruments were identified as significant risk contributors. .
(4) CONCLUSION
_X_ This current Technical Specification is included in the revised Technical Specification document.
This current Technical Specification may be relocated to another controlled document.
l l
Technical Sp2cification Screening Form TABLE 4 (Continued)
L NORTH ANNA UNIT 2 (1) TECHNICAL SPECIFICATION Primary to Secondary Leakage.
l TS LOCATION LCO 3.4.6.3 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, W NO and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition WW of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or, presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of W NO the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or-presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the- 1 Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "N0" and the .
system / component has not been shown to be a significant risk contributor, (
the Technical Specification requirements may be relocated to another l controlled document. l 4
i i
i
i i
Technical Specification Scrcening Form TABLE 4 (Continued)
(3) DISCUSSIONS The basic specification (LCO 3.4.6.2) on primary to secondary leakage limits the leakage to 1 GPM total for all three steam generators and 500 gallons per day through any one steam generator. This specification (LCO 3.4.6.3) provides additional primary to secondary leakage limits which have been imposed for operation above 50% of rated thermal power.
This specification does not contain requirements for installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
The limits on primary to secondary leakage which establish the initial ,
conditions for DBAs such as the steam generator tube rupture and main '
steamline break are specified in LCO 3.4.6.2. Operational leakage is an input variable to the offsite dose calculations associated with certain fesign Basis Accidents which result in the release of secondary coolant. 1 However, it is not a process variable that is controlled by the operator. The limits on this variable which are assumed in the analyses i are maintained by a separate LCO (3.4.6.2) which is being retained. j Therefore this specification does not satisfy criterion 2. j Reactor Coolant System operational leakage is not a structure, system or component that is part of the primary success path and which 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. This specification does not satisfy criterion 3.
Based on a review of available literature, primary to secondary leak rate surveillance is not modeled as a contributing factor which limits the likelihood or severity of any accident sequence commonly found to dominate offsite health effects, and has not been identified as a significant risk contributor.
(4) CONCLUSION This current Technical Specification is included in the revised ,
Technical Specification document.
X This current Technical Specification may be relocated to another ;
controlled document.
1
Technical Specification Screening Form TABLE 4 (Continued)
NORTH ANNA UNIT 2 (1) TECHNICAL' SPECIFICATION Primary to Secondary Leakage Detection Systems.
TS LOCATION LC0 3.4.6.4 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is.used to detect, W NO and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition W N0 of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of WT the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a. fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "N0" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
Technical Specification Screening Form TABLE 4 (Continued)
(3) DISCUSSIONS The purpose of the instrumentation required by this Specification is the early detection of steam generator tube degradation via small (of the order of 100 gpd) primary to secondary leaks. This will provide ample time for controlled actions to reduce power prior to a major tube failure.
This instrumentation was installed specifically to provide a sensitive means of early detection of rapidly propagating fatigue cracks in the U-bend region of the tubes. This failure mode was identified as the cause of a tube rupture experienced on Unit 1 in July 1987. Extensive modifications and engineering analyses have been performed to reduce the potential for such fatigue induced tube failures. These have included detailed engineering analysis to identify specific tubes susceptible to this mechanism, tube stabilization measures, reduction of local flow velocities in the tube bundle region via hydraulic restriction of the recirculating secondary coolant (installation of downcomer flow resistance plates), and a preventive plugging program. As such, the conditions which led to the incorporation of the primary to secondary leakage detection system have essentially been mitigated.
This specification does not contain requirements for installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. Pressure boundary leakage is defined in NUREG-0452 (Westinghouse Standardized Technical Specifications):
"1.19 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall." ,
The instrumentation covered in LCO 3.4.6.4 does not measure pressure boundary leakage, and therefore is not used for detection of pressure boundary degradation. There is other instrumentation, used to perform reactor coolant system leakage detection, which is being retained in the Technical Specifications. This includes the containment atmosphere particulate radioactivity monitoring system, the containment atmosphere gaseous radioactivity monitoring system, and the containment sump discharge flow measurement system. In addition, LCO 3.4.6.2 limits primary to secondary leakage to 1 gpm total and 500 gpd from any one generator. Adequate procedures and instrumentation will remain operable to meet the surveillance requirements to support this specification.
LCO 3.4.6.4 does not satisfy criterion 1.
The Primary to Secondary Leakage Detection Systems 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. The Primary to Secondary Leakage Detection Systems do not satisfy criterion 2.
(
Technical Specification Screening Form The Primary to Secondary Leakage Detection Systems are not systems that are part of the primary success path and which functions or actuates to mitigate'a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. These systems do not actuate any automatic function or provide any information leading to an operator action assumed in the design basis accident analysis. The Primary to Secondary Leakage Detection Systems do not satisfy criterion 3.
Based on a review of available literature, Primary to Secondary Leakage Detection Systems.are not modeled as a contributing factor which limits the likelihood or severity of any accident sequence commonly found to dominate offsite health effects, and have not been identified as a significant risk contributor.
(4) CONCLUSION This current Technical Specification is included in the revised Technical Specification document.
_ X_ This current Technical Specification may be relocated to another controlled document.
Technical Specification Screening Form TABLE 4 (Continued)
NORTH ANNA UNIT 2-(1) TECHNICAL SPECIFICATION Steam Turbine Assembly TS Location LC0 3.7.1.6 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, W NO and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition of a W NO Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of W NO the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions in "YES", then the Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "NO" and the system / l component has not been shown to be a significant risk contributor, the !
Technical Specification requirements may be relocated to another ;
controlled document. ;
I l
j
)
. r Technical Specification Screening Form TABLE'4 (Continued)
(3) DISCUSSIONS ,
This specification provides the control of the structural integrity of the turbine assembly. The structural integrity of the steam turbine assembly shall be demonstrated at least once per 40 months, during shutdown. At the least once per 10 years the turbine must be disassembled and all normally inaccessible parts inspected.
Demonstration of the structural integrity of the steam turbine assembly ensures an acceptable. low probability of generation of turbine missile (s) which could damage safety-related equipment.
The specification does not address instrumentation that is used to detect and indicate in the contro1 room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
No process variable is involved that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a I challenge to the integrity of a fission product barrier. This specification does not satisfy criterion 2.
The steam turbine assembly is not a structure, system er component that is part of the primary success path and which 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. This specification does not satisfy criterion 3.
Based on the PRA Summary Report contained in Appendix A, the integrity of the turbine assembly has not been identified as a significant risk contributor.
(4) CONCLUSION This current Technical Specification is included in the revised Technical Specification document.
X This current Technical Specification may be relocated to another controlled document.
I Technical Specification Screening Form TABLE 4 (Continued) l NORTH ANNA UNIT 2
-(1) TECHNICAL SPECIFICATION Settlement of Class 1 Structures TS LOCATION LCD 3.7.12.1 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrumentation that is used to detect, TET F and indicate in the centrol room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition WT of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or co:r.ponent that is part of W NO the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical 3 Specifications.
If the answer to all three of the above questions is "NO" and the ,
system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
.]
1 Technical Specification Screenina Form i
TABLE 4 (Continued)
(3) DISCUSSIONS ,
This Technical Specification addresses the monitoring of the settlement of class 1 structures. The program was developed in 1975 after numerous !
communications with the NRC. The testing approach, sequences and allowable settlements were included in the Technical Specifications.
l The Technical Specifications call for an engineering review, evaluation of the consequences and sutaittal of report to NRC when the settlement exceeds 75 percent of allowable limits. This ensures adequate time to implement corrective actions to prevent reaching 100 percent of allowable limits, at which time a step-wise plant shut down would have to be initiated.
The monitoring and engineering review of the settlement of class 1 structures do not represent installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
The surveillance of the settlement of class I structures does not represent a process variable that is an initial condition of a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not satisfy criterion 2.
The surveillance of class I structures is not a structure, system or component that is part of the primary success path and which functions to mitigate a Design Basis Accident or Transient. This specification does not satisfy criterion 3.
Based on the PRA Summary Report contained in Appendix A, this specification has not been identified as a significant risk contributor (4) CONCLUSION This current Technical Specification is included in the revised Technical Specification document.
X This current Technical Specification may be relocated to another controlled document.
q Technical Specification Screening Form TABLE 4 (Continued)
NORTH ANNA UNIT 2 (1) TECHNICAL SPECIFICATION -Groundwater Level, Service Water . Reservoir TS LOCATION LCO 3.7.13 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrument'ation that is used to detect, WT and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition WT of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of WT the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of cr presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical Specifications.-
If the answer to all three of the above questions is "NO" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
Technical Specification Screening Form i
TABLE 4 (Continued)
(3) DISCUSSIONS ,
This Technical Specification addresses the requirements for the monitoring of water level and water drainage from the service water l reservoir. The actions to be implemented when the specification is exceeded consist of engineering studies with reporting of relevant findings and corrective actions to the NRC, Hence, the Technical Specification is outlining a long term surveillance program with no immediate actions short of the initiation of an engineering evaluation and reporting of the findings. This will ensure reservoir integrity and full compatibility between plant operation and the availability of service water.
This specification does not involve installed instrumentation that is used to detect and indicate in the control room significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
This specification does not involve 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. This specification does not satisfy criterion 2.
The specification is not a structure, system or component that is part of the primary success path and which functions to mitigate a DBA or Transient that either assures the failure of or presents a challenge to the integrity of a fission product barrier. Therefore, this specification does not satisfy criterion 3.
{
1 Based on the PRA Summary Report contained in Appendix A, this specification has not been identified as a significant risk contributor. ;
(4) CONCLUSION l
~~ This current Technical Specification is included in the revised I Technical Specification document.
X This current Technical Specification may be relocated to another controlled document.
1 Technical Specification Screening Form TABLE 4 (Continued) ]
NORTH ANNA UNIT 2- I (1) TECHNICAL SPECIFICATION Normally De-energized Power Circuits l'
TS LOCATION LCO 3.8.2.7 (2) EVALUATION BASED ON POLICY STATEMENT CRITERIA Is the Technical Specification applicable to:
X (1) Installed instrunientation that is used to detect, W NO and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.
X (2) A process variable that is an initial condition W N0 of a Design Basis Accident (DBA) or Transient Analyses that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
X (3) A structure, system or component that is part of W NO the primary success path and which functions or actuates to mitigate a Design Basis Accident or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
If the answer to any one of the above questions is "YES", then the Technical Specifications shall be included in the new Technical Specifications.
If the answer to all three of the above questions is "NO" and the system / component has not been shown to be a significant risk contributor, the Technical Specification requirements may be relocated to another controlled document.
l 1
i Technical Specification Screening Form TABLE 4 (Continued)
(3) DISCUSSIONS This specification specifies the A.C. power circuits which must be de-energized inside the containment during operation in Modes 1, 2, 3 and 1
- 4. The equipment which is subjected to these requirements are the loop '
stop valves, the loop bypass stopvalves and various equipment associated with refuelings and operations in the reactor cavity such as cranes, RCC fixture changer and stud tensioner. The de-energization of A.C. power circuits inside the primary containment minimizes the potential for a fault in a component inside the containment from potentially damaging the electrical penetrations. However, the de energization is not required for normal operation or accident mitigation.
This specification does not contain requirements for installed instrumentation that is used to detect and indicate in the control room a significant abnormal degradation of the reactor coolant pressure boundary. This specification does not satisfy criterion 1.
The de-energization of A.C. power inside the containment help to preserve the assumptions of the safety analysis by enhancing proper equipment operation. However, it is not a process variable that is an initial condition of a DBA or a transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. This specification does not satisfy criterion 2.
The de-energization of A.C. circuits inside the primary containment provides equipment and distribution system protection from faults or improper operation of other protective devices in addition to that provided by the design of the distribution system. This specification does not contain requirements for a structure, system or component that is part of the primary success path and which 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. This specification does not satisfy criterion 3.
Based on the PRA Summary Report contained in Appendix A, this i specification has not been identified as a significant risk contributor.
(4) CONCLUSION This current Technical Specification is included in the revised Technical Specification document.
X This current Technical Specification may be relocated to another controlled document. )
l ,
l
l l
APPENDIX A l
PROBABILISTIC RISK ASSESSMENT (PRA) j EVALUATION OF THE NORTH ANNA TECHNICAL SPECIFICATIONS l
i l
l l
i
- ------------___..___m.____
i l
PROBABILISTIC RISK ASSESSMENT (PRA)
EVALUATION OF THE NORTH ANNA TECHNICAL SPECIFICATIONS >
A.I OBJECTIVE l This appendix documents the Probabilistic Risk Assessment (PRA) evaluation of l l the North Anna Technical Specifications in order to support the implementation of the MERITS Program. In response to the Nuclear Regulatory Commission (NRC)
Interim Policy Statement on Technical Specification Improvements for. Nuclear i Power Reactors, the. Westinghouse Owners Group and Virginia Power have jointly-agreed to identify the North Anna Power Station as the lead plant in the implementation of the MERITS Program. Per the Interim Policy Statement, Technical Specifications must be evaluated from a PRA point of view. Thus, the purpose of this analysis is to determine if the parameters, components, or systems addressed by the North Anna Technical Specifications, have been modeled within the available literature on risk insights and PRA studies, and whether they are of prime importance in limiting the likelihood or severity of accident sequences commonly found to dominate offsite health ~ effects.
A.2 EVALUATION BASES Three criteria in addition to the PRA evaluation process are included in the Interim Policy Statement for determining which specifications are to be retained in the Technical Specifications. These three criteria plus the PRA evaluation were applied to the Standard Technical Specifications (STS) for Westinghouse Pressurized Water Reactors (NUREG 0452, Revision 4 and draft Revision 5) and documented in Reference I. If none of.the three criteria was identified as a constraint for a given specification, then that specification was identified as a candidate for relocation to another controlled document-.
Specifications identified for possible relocation were then evaluated using PRA experience to determine if the parameters, components, or systems addressed by the specifications have been modeled within the available literature on risk insights and PRA studies, and whether they are of prime importance in limiting the likelihood or severity of accident sequences commonly found to dominate offsite health effects. If these specifications were found not to be of pr:me importance in limiting the likelihood or severity of accident sequences that dominate offsite health effects, then these specifications could be relocated to another controlled document.
A-I
The results of the MERITS criteria application to the Standard Technical ,
Specifications were submitted to the NRC by the Westinghouse Owners Group.
The NRC then issued a position statement on the results identifying which specifications in the STS must be retained or may be relocated. These NRC determinations (Ref. 2) formed the basis or starting point for the application of the Interim Policy Statement criteria and the _ PRA evaluation to the North Anna Technical Specifications.
In the main section of this report, the North Anna Tec!.nical Specifications-were compared to the STS. For those specifications with comparable requirements, the disposition followed the NRC position statement. For those requirements unique to North Anna or inconsistent with the'STS, the-Interim Policy Statement criteria were applied. Table A-1 identifies th's North Anna plant specific Technical Specification relocation candidates for which a PRA study is required. As directed by the NRC Interim Policy Statement, these must now be evaluated using PRA experience to determine if the parameters, components, or systems addressed by the specifications have been modeled within the available literature on risk insights and PRA studies, and whether they are of prime importance in limiting the likelihood or severity of I accident sequences commonly found to dominate offsite health effects. If these specifications are found not to be of prime importance in limiting the likelihood or severity of events that dominate offsite health effects,I then they could be relocated to another controlled document. In this way, a decision to retain or relocate each of the Technical Specifications for North Anna Units 1 and 2 was obtained.
The evaluation of the -isk impact of the North Anna technical specifications that are relocation candidates from Table A-1 is based on the following:
A. It is assumed that any technical specification that is relocated will be transferred to other documents subject to control by the utility.
1 1
1 i
A-2
i i
B. The risk criteria used in determining the disposition of a technical specification are the following:
- 1. If the technical specification contains constraints of orime importance in limiting the likelihood or severity of the accident sequences commonly found to dominate offsite health effects, it should be retained.
- 2. If the technical specification covers items modeled in a dominant sequence but has an insignificant impact on the probability or severity of that sequence, it may be relocated to another controlled document.
- 3. If the technical specification is not modeled in risk dominant sequences, it may be relocated to another controlled document.
C. The measures related to risk used in this evaluation are core melt frequency and offsite health effects and are consistent with the Interim Policy Statement on Technical Specifications and the Safety Goal and Severe Accident Policy Statements.
D. The numerical risk evaluation criteria used to determine if a sequence is risk dominant are:
- 1. For core melt, any sequence whose frequency is commonly found to be greater than 1x10(-6) per reactor year is maintained as a possible dominant sequence as a conservative first cut. This is roughly 2% of the total core melt frequency of 5x10(-5) for typical PRAs. Each l specific sequence found to be greater than Ix10(-6) per. reactor year is then evaluated based on the offsite health effects.
- 2. For offsite health effects, any sequence whose frequency is commonly found to be greater than 1x10(-7) per reactor year is considered to be a dominant risk sequence for this analysis. This criterion is in agreement with the present NRC position in the Safety Goal Policy for a goal of 1x10(-6) for a total frequency of severe offsite release, and no greater than lx10(-7) for an individual sequence (Ref. 3 & 4).
A-3
s E. Tables A-2 and A-3 list representative sequences for all identified types of initiating events considered in risk assessments for two different plants (Ref. 5 and 6).
Table A-2-list sequences for a Westinghouse three-loop plant like North Anna. And Table A-3 list sequences for a subatmospheric containment plant similar to North Anna. These tables have been reviewed for consistency with NRC sponsored PRA programs documented in NUREG's 3301, 1150,-4550, 4551, 4624, and 4700 (Ref. 7-12) and the results have been found to be consistent. For example, these NRC sponsored PRA programs have identified similar accident sequences, such as: transients with loss of support systems leading to RCP seal LOCAs, loss of offsite power with failure to recover power, LOCAs with failure of injection or recirculation, and loss of support systems (vital AC/DC, service water and component cooling -
water).
Systems identified in Tables A-2 and A-3 that contribute significantly to risk as defined in item D above are listed in Tables A-2a, A-2b, A-3a, and A-3b. These identified systems were used to screen the requirements of the technical specifications identified in Table A-1 as possible candidates for possible relocation. Those specifications whose requirements were relevant to these systems, sequences, and initiating events were further evaluated for risk dominance. The remaining specifications were evaluated on the basis of risk insights from references that were not formal full scale risk assessments. If the requirements of a technical specification were not found to be modeled in any reference and no significant issues were identified from a review of the risk insights, the conclusion was that it' '
did not contain constraints of orime importance to limiting the likelihood l or severity of sequences commonly found to dominate offsite health effects.
F. Certain instrumentation technical specifications, such as Accident Monitoring Instrumentation were determined to meet Criterion 3 and therefore will be retained. For the Accident Monitoring specifications, a review of the available PRA information was conducted to determine if I
A-4
the risk significance of individual instrumer.ts contained in this specification could be determined. No specific cases were identified in which single monitors had been explicitly modeled in an accident scenario. On the basis of this review, it was concluded that the available PRA information is not sufficiently detailed to determine the risk significance of individual accident monitoring instruments in the North Anna specifications.
A.3 METHODOLOGY The formal process used in evaluating the risk significance of the technical' specificatier, items is outlined below:
A. The requirements of the technical specifications were screened against the dominant sequences and initiating events of Tables A-2, A-2a, A-2b, A-3, A-3a, and A-3b i,o identify if they were pertinent to risk dominant sequences. The conclusion of this screening was stated at the beginning of the " COMMENTS" section of the review form discussed below.
B. The references used for the STS PRA evaluation (Ref. 9) were reviewed based on the North Anna design and the specifications requiring a PRA evaluation identified in Table A-1. Those references identified as providing possible PRA insights applicable to the North Anna evaluation were selected to be used in the PRA evaluation process. A list of all reference documents used in this evaluation is provided in section A.5.
C. PRA review forais (pages A-19 to A-29) were developed which formalized the PRA review. These review forms contain:
- 1. The number and title of the technical specification;
- 2. A description of the technical specification requirement;
- 3. The potential safety effect of the technical speciffr.ation requirement; I l l
l A-5
_ _ - __- ____ U
- 4. The reference documents that were utilized to support the comments and conclusions on the technical specification requirement;
- 5. Comments, wt.ich contain a discussion of the information that was used as the basis to arrive at the conclusion;
- 6. A' conclusion as to whether the technical specification should be retained or relocated.
D. Each of the North Anna specifications listed in Table A-1 was screened to determine which reference documents addressed the requirements of that specification. All of the documents listed in section A.5 were considered as part of this review. The screening of references was performed as discussed in Section A.2 under item E.
If there were no reference documents found as a basis for evaluating a specification it was indicated on the review form by entering "None" under the PRA- Design Basis Reference column. If constraints of a specification were not modeled in any formal risk assessments, the references which were reviewed to confirm this were listed.
E. A PRA review form was completed for each North Anna technical specification that did not meet at least one of the Interim Policy Statement screening criteria (Table A-1). These review forms, containing the information and the reference documentation that were used to arrive at the PRA conclusion are documented in this appendix.
F. As a result of the PRA screening, a conclusion was reached as to whether the technical specification:
- 1. Should be retained as a technical specification;
- 2. May be relocated to other administrative 1y controlled documents.
A-6
I I
This methodology is based on the PRA methodology presented in the MERITS ~j Program Criteria Application report (Ref.1). However, several changes were made to the original PRA methodology. The PRA results used to identify dominant sequences and system for a large dry containment were replaced with PRA results from Westinghouse 3-loop plants to be more representative of North {
Anna. No senior review group was established since the specifications requiring review were significantly less in number and fairly similar to the Standard Technical Specifications. Only formal PRA assessments were used as references.
A.4 PRA EVALUATION RESULTS The results of the PRA evaluation are shown in Table A-4. A " Retain" entry in the PRA evaluation column indicates the technical specification requirements contain restraints of orime importance in limiting the likelihood or severity of accident sequences commonly found to dominate oJfsite health effects. A
" Relocate" entry in the PRA evaluation column indicates that the technical specification requirements do not contain constraints of orime'importance in limiting the likelihood or severity of the accident sequences commonly found to dominate offsite health effects.
I 4
l I
A-7 '
4
A.5 REFERENCES
- 1. WCAP-11618, " Methodically Engineered, Restructured, and Improved Technical Specifications," MERITS Program - Phase II, Task 5, Criteria Application.
- 2. NRC Letter, T.Murley to W.Wilgus (B&W Owners Group Chairman), May 9,1988, (Documenting the NRC Staff review of each Owners Group's Criteria Application Topical Report.
- 3. 10CFR Part 50, " Policy Statc ient on Severe Accidents."
- 4. 10CFR Pa?t 50, " Policy Statement on Safety Goals for the Operation of Nuclear Power Plants."
- 5. WCAP-10590, PUN Probabilistic Safety Study," 1984.
- 6. " Millstone Unit 3 Probabilistic Safety Study," Northeast Utilities Company, August 1983.
- 7. NUREG-1150, " Reactor Risk Reference Document," U. S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, February 1987.
- 8. NUREG/CR-3301, " Catalog of PRA Dominant Accident Sequence Information,"
June 1983.
- 9. NUREG/CR 4550, Volume 3, " Analysis of Core Damage Frequency Estimation from Internal Events: Surry Unit 1," November 1986.
- 10. NUREG/CR-4551, Volume 1, " Evaluation of Severe Accident Risk and the Potential for Risk Reduction: Surry Power Station Unit 1," February 1987.
- 11. NUREG/CR-4624, Volume 3, " Radionuclides Release Calculations for Selected Severe Accident Scenarios PWR Subatmospheric Containment Design: Surry Unit 1," July 1986.
- 12. NUREG/CR-4700, Volume 1, " Containment Event Analysis for Postulated Severe Accidents: Surry Power Station Unit 1," February 1987.
- 13. NUREG/CR-4550, Volume 5, " Analysis of Core Damage Frequency Estimation from Internal Events: Sequoyah Power Station, Unit 1," February 1987.
- 14. NUREG/CR-4550, Volume 7, " Analysis of Core Damage Frequency Estimation from Internal Events: Zion Unit 1," October 1986.
- 16. "v'ogtle Electric Generating Plant, Inadvertent Boron Dilution Analysis,"
Final Report, Rev. 2.
- 17. "A Probabilistic Safety Analysis of Boron Dilution Events at Millstone-Unit 3," Northeast Utilities Service Company.
l A-8
- - - - - - - - - - - - - - - - - - - - - 1
- 18. Zion Probabilistic Safety Study," Commonwealth Edison Company, September 1981.
- 19. NUREG-0460, " Anticipated Transients Without Scram for Light Water Reactors," Volumes 1-4, 1978-80.
- 20. 10CFR Part 50.62, " Anticipated Transients Without Scram (ATWS)."
- 21. NUREG-0800, " Standard Review Plan."
- 22. NRC Regulatory Guide 1.115, " Protection Against Low Trajectory Turbine Missiles," Revision 1, July 1977.
- 23. NRC Standard Review Plan 3.5.1.3, " Turbine Missiles," Rev. 2, July 1981.
- 24. " Indian Point Probabilistic Safety Study," Power Authority of the State of New York, Consolidated Edison Company of New York, Inc.,1982.
25.10CFR Part 50, Appendix A (General Design Criteria 4), " Environmental and Missile Design Bases."
- 26. WCAP-10161, " Evaluation of Impact of Reduced Testing of Turbine Valves,"
Westinghouse Electric Corporation, September 1982.
- 27. WCAP-ll525, "Probabilistic Evaluation of Reduction in Turbine Valve Test Frequency," Westinghouse Electric Corporation, June 1987.
- 28. NUREG/CR-2728, " Interim Reliability Evaluation Program Procedures Guide,"
1983.
- 29. NUREG/CR-2300, "Probabilistic Risk Assessment Procedures Guide," January 1983.
- 30. NUREG/CR-2815, "Probabilistic Safety Assessment Guide," Rev. 1, August 1985.
- 31. IDCOR Technical Report 86.3Al, " Individual Plant Evaluation Methodology for Pressurized Water Reactors," April 1987 A-9
TABLE A-1 NORTH ANNA PLANT SPECIFIC TECHNICAL SPECIFICATION RELOCATION CANDIDATES RE0VIRING PRA EVALUATION Tech-Spec STS-Rev 5 NRC Criteria-Unit Number Number Title Results Results 1 3.1.1.3.1 None Boron Dilution, Not RCS Flow Reviewed Relocate 1 3.1.2.5 None Boric Acid Transfer Not Pumps, Shutdown Reviewed Relocate 1 3.1.2.6 None Boric Acid Transfer Not Pumps, Operating Reviewed Relocate 1,2 3.7.1.6 None Steam Turbine Not Relocate Assembly Reviewed 1,2 3.7.12.1 None Settlement of Class 1 Not Structures Reviewed Relocate 1,2 3.7.13 None Groundwater Level, Not Service Water Reservoir Reviewed Relocate 2 3.8.2.7 None Normally De-energized Not Power Circuits Reviewed Relocate l I A-10
v .
TABLE A-2 DOMINANT ACCIDENT SE0VENCES FOR A WESTINGHOUSE 3-LOOP PLANT Rank Witn Mean Annual .1ean Annual Respect To Core Melt Release Core Melt Seauence Description Freauency Freauency 1 Small LOCA: Failure of Sump Valves 3.0E-06 <1.0E-07.
2,3,6, Transient: Small LOCA and Loss of 21,24, Support Systems Component Cooling, ESF 27,53 Actuation, Service Water, or AC Power 8.2E-06 >1.0E-07 4,7 Large/ Medium LOCA: Failure of:
Accumulators 3.4E-06 <1.0E-07 5 Interfacing Systems LOCA: 1.9E-06 >1.0E-07 8 Medium LOCA: Failure of High Pressure Injection and Low Pressure Injection and Secondary Depressurization 1.0E-06 <1.0E-07 9,13, Transient: Steam Leak, Steam Generator 20 Tube Rupture, Failure of Residual Heat J Removal (Operator Action or Equipment) 2.0E-06 <1.0E-07 10 Loss of Vital AC/DC: Small LOCA and Loss of Support Systems AC Power or ESF Actuation <l.0E-06 >1.0E-07 14 Loss of Component Cooling: Small LOCA and Loss of Support Systems Service Water, ESF Actuation, or AC Power <1.0E-06 >1.0E-07 16 Transient: Steam Leak and Loss of One l Train of Support Systems Component i Cooling, ESF Actuation, AC Power or Service Water <l.0E-06 >l.0E-07 19,67, Loss of Vital AC/DC: Steam Leak and 68 Loss of One Train of Support Systems Service Water, Component Cooling, ESF Actuation, or AC Power <1.0E-06 >1.0E-07 22 Loss of Offsite Power: Loss of One Train of Support Systems Component Cooling, Service Water, AC Power, or ESF Actuation <1.0E-06 >1.0E-07 l 25 Small LOCA: Loss of Support Systems Component Cooling, ESF Actuation, Service Water, or AC Power <1.0E-06 >1.0E-07 '
A-11
r .
TABLE A-2 DOMINANT ACCIDENT SEQUENCES FOR A WESTINGHOUSE 3-LOOP PLANT (Continued)
Rank With Mean Annual Mean Annual Respect To Core Melt Release Core Melt Seouence Description Freouency Freouency 26 Loss of Vital AC/DC: Small LOCA and Loss of Support Systems Component Cooling, Service Water;, AC Power or ESF Actuation <1.0E-06 >1.0E-07 39 Loss of Vital AC/DC: Steam Leak and Loss of Support Systems Component Cooling, Service Water, AC Power or ESF Actuation <1.0E-06 >1.0E-07 44 Small LOCA: Loss of One Train of Support Systems ESF Actuation, Service Water Component Cooling or AC Power and J Failure of Controlled Primary Cool- '
down and Cont. Fan Coolers <1.0E-06 >1.0E-07 45,57 Transient: Small LOCA and Loss of One Train of Support Systems Component l
Cooling, ESF Actuation, Service Water or AC Power and failure of Controlled Primary Cooldown and Cont. ian Coolers < l . 0 E-06 >1.0E-07 46,60 Transient: Steam Leak, Small LOCA and Loss of Support Systems Component Cooling, ESF Actuation, AC Power or Service Water <l.0E-06 >1.0E-07 50 Loss of Offsite Power: Loss of One l Train of Support Systems AC Power, ESF Actuation, Service Water or Component
, Cooling, and Failure of Cont. Fan i' Coolers and Controlled Primary Cooldown <1.0E-06 >1,0E-07 l 63 Loss of Offsite Power: Steam Leak and Loss of One Train of Support Systems Service Water, Component Cooling, ESF Actuation, or AC Power <1.0E-06 >1.0E-07 65 Transient: Steam Leak, Steam Generator Tube Rupture, Loss of Support Systems Component Cooling, ESF Actuation, AC Power or Service Water and Failure :
of Residual Heat Removal (Operator Action or Equipment) <l.0E-06 >1.0E-07 A-12
y; .
l TABLE A-2 DOMINANT ACCIDENT SEQUENCES FOR A WESTINGHOUSE 3-LOOP PLANT (Continued)
General Notes
- 1. Only those accident sequences with a coremelt frequency >1x10(-6)-
and/or result in a serious release frequency >1x10(-7) are shown.
- 2. In order to minimize the number of individual accident sequences listed, similar sequences were combined based on initiating event categories or. equivalent system fa; lures.
- 3. Since the reference Westinghouse 3-loop analysi:; did not include a containment analysis, the frequency of serious release was based on previous PRA experience and reviewing each individual accident sequence. For accident sequences without containment failure (i.e.,
containment cooling available), it was assumed such sequences would not result in a serious release. Although other possible failures could occur that would lead to containment failure, such as, failure of contaiment isolation, failure of containment cooling is the most probable. For accident sequences that lead to a steam generator tube rupture with a steam leak, it was assumed that the resulting release outside containment would not constitute a serious release as defined in previous risk assessment studies.
l l
A-13
1 TABLE A-2a SYSTEM FAILURES CONTRIBUTING'TO COREMELT FOR ACCIDENT SEQUENCES > 1x10f-6)/vear FOR A WESTINGHOUSE 3-LOOP PLANT l Dominant Plant System Failures Accumulators / Low Pressure Injection RHR-Inlet Valves Failure RHR Sump Valves ESF Actuation High Pressure Injection Containment Fan Coolers Secondary Side Steam Relief Valves Recirculation (SI and/or RHR pumps)
Component Cooling Water Service Water TABLE A-2b SYSTEM FAILURES CONTRIBUTING TO 0FFSITE HEALTH EFFECTS FOR ACCIDENT SE00ENCE$ > lx10f-7)/vear FOR A WESTINGHOUSE 3-LOOP PLANT
- Dominant Plant System Failures ESF Actuation l Containment Fan Coolers Recirculation (SI and/or RHR pumps)
Component Cooling Water Service Water l RHR Inlet Valves failure (leading to an. interfacing systems LOCA) l
- Latent Fatalities >I000 ,
A-14
TABLE A-3 DOMINANT ACCIDENT SEQUENCES FOR A SUBATMOSPHERIC CONTAINMENT PLANT Rank With Mean Annual Mean Annual Respect To Core Melt Release Core Melt Seouence Description Freouency Freouency 1 Seismic: Loss of offsite power 9.08E-06 7.25E-06
- 2. Medium LOCA: Failure of High Pressure Recirculation 3.80E <1.0E-07 3 Loss of Vital DC Bus 1 of 2: Failure of Auxiliary Feedwater and Bleed / Feed Cooling 2.20E-06 <1.0E-07 4 Loss of Vital AC Bus 1 or 2: Failure of Auxiliary feedwater and High Pressure Recirculation 1.98E-06 <l.0E-07 5 Loss of Vital AC Bus 3 or 4: Failure of Auxiliary Feedwater and of High Pressure Recirculation 1.98E-06 <l.0E-07 6 Interfacing Systems LOCA: Failure of RHR Inlet Valves 1.90E-06 6.9E-06 7 Loss of Off ite Power: Failure of Both Diesel Generators, to Recover Power in 6 Hours and Quench Spray Recovery 1.65E-06 1.6E-06 8 Loss of Offsite Power: Steam Leak, Failure of One ESF Bus, Auxiliary Feedwater and Bleed through PORV's 1.63E-06 <l.0E-07 9 Steam Line Break Outside Containment:
Failure to Isolate Main Steam Line and Primary Bleed through PORV's 1.54E-06 <l.0E-07 10 Small LOCA: Failure to Control Primary Depressurization and High Pressure Recirculation 1.40E-06 <l.0E-07 11 Large LOCA: Failure of Low Pressure l Recirculation 1.30E-06 <l.0E-07 12 Medium LOCA: Failure of Accumulators 1.10E-06 <1.0E-07 13 Loss of Offsite Power 1.10E-06 <1.0E-07 14 Loss of Vital AC Bus 1 or 2: Failure of Auxiliary Feedwater and Primary Bleed and Feed <1.0E-06 <1.0E-07 A-15
___ _ _ _ _ _ .i
7-TABLE A-3 DOMINANT ACCIDENT SEQUENCES FOR A SUBATMOSPHERIC CONTAINMENT PLANT (Continued)
Rank With Mean Annual Mean Annual i Respect To Core Melt Release Core Melt- Seauence Description Freauency Freauency 15 Loss of Vital AC Bus 3 or 4: <1.0E-06 <1.0E-07 21 Loss of Vital AC Bus 1 or 2: Failure of Opposite Train ESF Cabinet, Auxiliary Feedwater, Bleed / Feed Cooling, and Quench Spray <1.0E-06 7.2E-07 22 Primary to Secondary Power Mismatch:
Failure of Both ESF Cabinets, Auxiliary Feedwater, Bleed / Feed Cooling, and Quench Spray <1.0E-06 6.1E-07 27 Reactor Trip: Failure of Both ESF 4 Cabinets, Auxiliary Feedwater, Bleed / (
Feed Cooling, and Quench Spray
<1.0E-06 4.8E-07 33 Turbine Trip: Failure of Both ESF Cabinets, Auxiliary Feedwater, Bleed /
Feed Cooling, and Quench Spray <1.0E-06 3.7E-07 42 Primary to Secondary Power Mismatch:
Coincident Station Blackout, Small LOCA, Failure of High Pressure Injection, Secondary Depressurization and Low Pressure Injection, and Quench Spray Recovery <1.0E-06 2.4E-07 48 Reactor Trip: Coincident Station Blackout, Small LOCA, Failure of High Pressure Injection, Secondary Depressurization and Low Pressure Injection, and Quench Spray Recovery <1.0E-06 1.9E-07 56 Turbine Trip: Coincident Station Blackout, Small LOCA, Failure' of High Pressure Injection, Secondary Depressurization and Low Pressure Injection, and Quench Spray Recovery <1.0E-06 1.5E-07 72 Loss of Vital AC Bus 1 or 2: Failure of Auxiliary Feedwater, High Pressure Recirculation, Containment Recirculation Spray <l.0E-06 <1.0E-07 A-16 L - _ ---- _ ---- __ _ _ _
7-TABLE A-3a SYSTEM FAILURES CONTRIBUTING TO COREMELT FOR ACCIDENT SE00ENCES > 1x10(-6)/vear FOR A SUBATMOSHPERIC CONTAINMENT PLANT Dominant Plant System Failures Accumulators Auxiliary Feedwater System Diesel Generators Diesel Generators (Seismic)
Electrical Power ESF Bus Feed and Bleed (Primary PORVs and High Pressure Injection)
High Pressure Recirculation Low Pressure Recirculation Main Steam Line Isolation Primary PORVs RHR Inlet Valves I l
l TABLE A-3b SYSTEM FAILURES CONTRIBUTING TO 0FFSITE HEALTH EFFECTS FOR l ACCIDENT SE00ENCES > 1x10f-71/vear FOR A SUBATMOSHPERIC CONTAINMENT PLANT 1
Dominant Plant System Failures Auxiliary Feedwater System ;
Containment / Quench Spray Diesel Generators ;
Diesel Generators (Seismic)
ESF Cabinets Feed and Bleed (Primary PORVs and High Pressure Injection)
High Pressure Recirculation RHR Inlet Valves
- Latent Fatalities >1000 !
I A-17
Il .I s
TABLE A-4 PRA EVALUATION OF THE NORTH ANNA PLANT SPECIFIC -
TECHNICAL SPECIFICATION RELOCATION CANDIDATES !
Tech-Spec NRC Criteria PRA f
APP 1 MDjl Number Title Results Results Results Eggg 1 3.1.1.3.1 Boron Dilution, Not RCS Flow Reviewed Relocate Relocate A-19 1 3.1.2.5 Boric Acid Transfer Not i.
Pumps, Shutdown Reviewed Relocate Relocate A-21 l 1 3.1.2.6 Boric Acid Transfer- Not Pumps, Operating Reviewed P,elocate Relocate A-21 1,2 3.7.1.6 Steam Turbine Not Assembly Reviewed Relocate Relocate A 1,2 3.7.12.1 Settlement of Class 1 Not Structures Reviewed Relocate Relocate A-25 1,2 3.7.13 Groundwater Level, Not Service Water Reservoir Reviewed Relocate Relocate A-26 2 3.b.2.7 Normally De-energized Not ,,
Power Circuits Reviewed Relocate Relocate A-27 )
1 l
A-18
r _ .
PRA TECHNICAL SPECIFICATION REVIEW FORM NORTH ANNA POWER STATION i
TECHNICAL SPECIFICATION: Unit 1 3.1.1.3.1 - BORON DILUTION, RCS FLOW DESCRIPTION OF RE0VIREMENT: This specification identifies the minimum flow rate of reactor coolant through the reactor coolant system whenever a reduction in the system's boron concentration is being made. The objective is-to ensure that the rate of reactivity' change, associated with boron reductions, will be within specified limits for operator recognition and control. This specification applies to all modes of operation.
POTENTIAL EFFECT: Boration, Shutdown REFERENCE DOCUMENTS UTILIZED:
PRA DESIGN BASIS RISK Ref. 18: NUREG-0460 Ref. 7: NUREG-1150 Reactor (ATWS Analysis Studies) Reference Document Ref. 19: Zion PSS COMMENTS: The requirements of this technical specification are.nqt risk dominant based on the core melt and health effects screening criteria provided in Section A.2 under item D parts l'and 2. Further discussion of insights based on review of the reference documents is provided below.
Boration: NUREG-0460 (ATWS Analysis Studies), NUREG-1150 Reactor Reference Document, and the Zion PSS assume that plants can borate to subcritici.il within 10 minutes after an ATWS event. For the case where boration is modeled, its failure is not a dominant contributor to risk. A review of these studies ~also shows that the failure probabilities for boration are dominated by human errors, not system failures. This function is not significant to dominant risk sequences.
Shutdown: The Zion PSS was the only study reviewed which modeled shutdown after steambreaks/feedbreaks and steam generator tube ruptures. The Zion PSS models shutdown as an extreme consideration and shutdown failures were not dominant risk contributors.
, Probabilistic boron dilution analysis is also part of the licensing basis for i numerous plants. A minimum reactor coolant flow is required for effective boration and dilution, but is not required for mitigation of an inadvertent dilution event. Mitigation only requires termination of the dilution prior to criticality. Reactor coolant flow is not significant to dominant risk j - sequences.
l l
I l
A-19
l PRA TECHNICAL SPECIFICATION REVIEW FORM NORTH ANNA POWER STATION TECHNICAL SPECIFICATION: UNIT 1 3.1.1.3.1 - BORON DILUTION, RCS FLOW i i
(continued)
CONCLUSION: This technical specification does not contain constraints of prime importance in limiting the likelihood or. severity of the accident i
sequences commonly found to dominate risk.
RETAIN IN RELOCATE FROM TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATIONS NO YES l
l l
A-20
PRA TECHNICAL SPECIFICATION REVIEW FORM NORTH ANNA POWER STATION TECHNICAL SPECIFICATION: Unit 1 3.1.2.5 - BORIC ACID TRANSFER PUMPS, SHUTDOWN Unit 1 3.1.2.6 - BORIC ACID TRANSFER PUMPS, OPERATING DESCRIPTION OF RE0VIREMENT: The boric acid transfer pumps are part of the boration subsystem of the Chemical and Volume Control System. The functional requirement of the boration subsystem is to control the boron concentration in the reactor coolant system and to help maintain shutdown margin. - For modes 5 and 6, this ability to adjust boron concentration is necessary to provide a sufficient Shutdown Margin to compensate for xenon decay and cooldown from 200F to 140F. For modes 1, 2, 3, and 4 this ability to adjust boron concentration is necessary to provide a sufficient Shutdown Margin to compensate for xenon decay and cooldown to 200F. To meet this requirement, these specifications require a source of borated water and one or more flow paths to inject this borated water into the reactor coolant system.
POTENTIAL EFFECT: Boration, Shutdown, Operator Action Times During Inadvertent Boron Dilution Accidents REFERENCE DOCUMENTS UTILIZED:
PRA DESIGN BASIS RISK Probabilistic Boron Dilution Ref. 6: Millstone 3 PSS Analysis Ref. 7: NUREG-1150 Reactor Risk l Ref. 15: South Texas Project Reference Document Ref. 16: Vogtle Ref. 9: NUREG-4450 Vol. 3 Surry Ref. 17: Milestone 3 Ref. 13: NUREG-4450 Vol. 5 Sequoyah i Ref. 18: NUREG-0460 Ref. 14: NUREG-4450 Vol. 7 Zion !
(ATWS Analysis Studies) Ref. 19: Zion PSS COMMENTS: The requirements of this technical specification are not risk I dominant based on the core melt and health effects screening criteria provided in Section A.2 under item D parts 1 and 2. Further discussion of insights based on review of the reference documents is provided below. )
I Boration: NUREG-0460 (ATWS Analysis Studies), NUREG-1150 Reactor Reference l Document, and the Zion PSS assume that plants can borate to suberitical within !
10 minutes after an ATWS event. For the case where boration is modeled, its failure is not a dominant contributor to risk. A review of these studies also shows that the failure probabiFties for boration are dominated by human errors, not system failures.
- s function is not significant to dominant !
risk sequences.
Shutdown: The Zion PSS was the only study reviewed which modeled shutdown after steambreaks/feedbreaks and steam generator tube ruptures. The Zion PSS '
models shutdown as an extreme consideration and shutdown failures were not dominant risk contributors.
A-21
E .
PRA TECHNICAL SPECIFICATION REVIEW FORM NORTH ANNA POWER STATION TECHNICAL SPECIFICATION: Unit 1 3.1.2.5 - BORIC ACID TRANSFER PUMPS, SHUTDOWN Unit 1 3.1.2.6 - BORIC ACID TRANSFER PUMPS, OPERATING (continued)
Ooerator Action Times Durino Inadvertent Boron Dilution Accidents: In order to meet operator action time requirements of the Standard Review Plan (Rev. 2 and greater), and 10CFR Part 50.62, " Anticipated Transients Without Scram (ATWS)." (Ref. 20 & 21) PRA has been used to define maximum credible dilution flow rates and to determine the necessary operator actions and alarm setpoints. Boron dilution initiating events are dominanted by mechanical failures and operator errors which result in the flow of unborated water to the reactor coolant system. Failures of the boric acid transfer pumps and the boric acid storage subsystem are considered in the Probabilistic Boron Dilution Analyses referenced above and the resulting accident sequences are shown not to be dominant contributors to risk. In addition, PRAs have typically demonstrated or assumed that boron-dilution events are not_ dominant risk contributors.
CONCLUSION: This technical specification does not contain constraints of prime importance in limiting the likelihood or severity of the accident sequences commonly found to dominate risk.
RETAIN IN RELOCATE FROM TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATIONS N0 YES l
A-22
L PRA TECHNICAL SPECIFICATION REVIEW FORM NORTH ANNA POWER STATION TECHNICAL SPECIFICATION: Unit 1 & 2, 3.7.1.6 - STEAM TURBINE ASSEMBLY DESCRIPTION OF RE0VIREMENT: These specifications provide control of the structural integrity of the turbine assembly by requiring periodic inspection (a 40 month and a 10 year inspection interval). These inspections are required to prevent turbine structural failures which may result in the generation of potentially damaging missiles from the turbine. A secondary effect of a structural failure would be a secondary system steam leak.
POTENTIAL EFFECT: Missile Ejection due to Turbine Structural Failures and Possible Damage to Safety Related Equipment, Secondary System Steam Leak REFERENCE DOCUMENTS UTILIZED:
PRA DESIGN BASIS RISK l Turbine Missile Reports Ref. 5: PUN PSS Ref. 22: Reg. Guide 1.115 Ref. 6: Millstone 3 PSS Ref. 23: SRP 3.5.1.3 Ref. 19: Zion PSS Ref. 25: 10CFR Part 50 App. A Ref. 24: Indian Point 2 & 3 PSS Ref. 26: WCAP-10161 Ref. 27: WCAP-1152S COMMENTS: The requirements of this technical specification are not risk dominant based on the core melt and health effects screening criteria provided in Section A.2 under item D parts 1 and 2. Further discussion of insights based on review of the reference documents is provided below.
Missile E.iection due to Turbine Structural Failures and Possible Damaae to Safety Related Eouioment: Turbine structural failures are not an event in any PRA accident sequence resulting in core damage or fission product release.
Moreover, the most likely cause of turbine missiles being generated is from turbine overspeed not from structural failures. Given that a structural failure occurs, a missile (turbine disk) may be ejected through the turbine casing. The ejected missile may strike safety related equipment and damage the equipment so that it can no longer perform its required function.
The frequency of a turbine missile causing damage to a safety related system is the product of: the frequency of turbine failure, the probability of a missile being generated, the probability of the missile striking the safety related equipment, and the probability that the equipment will be damaged.
PRAs that have included loss of a safety related component or system due to l turbine missile damage have generally found that the contribution is not a l significant event in the quantification of core damage frequency. The actual l frequencies of turbine missile ejection and subsequent safety system damage are, in general, so low that they have little or no impact on core damage '
l frequency. PRAs such as the Zion PSS have made this conclusion.
A-23 -
I
__ _d
7 ,
< i l
l l
PRA TECHNICAL SPECIFICATION REVIEW FORM j NORTH ANNA POWER STATION
! TECHNICAL SPECIFICATION: Unit 1 & 2, 3.7.1.6 - STEAM TURBINE ASSEMBLY (continued)
Previous studies have calculated a turbine missile strike frequency of lx10(-6) per year and a damage frequency to safety related structures of 1x10(-7) per year. These numbers are used as a plant design basis since the damage frequency of 1x10(-7) per year is low enough to be an acceptable level of risk according to Reg. Guide 1.115.
Recent probabilistic studies, WCAP 10161 and 11525, have shown that the
, frequency of turbine missile ejection is less than the acceptable risk level I
of lx10(-5) per year as given by the NRC for this event. These studies are based on turbine overspeed events which are caused by more likely failures, such as the turbine control valves rather than turbine structural failures.
Secondary System Steam Leak: As a result of a structural turbine failure, such as a rupture of the turbine casing, a secondary system steam leak could occur. Given a steam leak, there are multiple plant features that would act l to mitigate this event. The turbine control (or governor) valves and the turbine stop valves would close. In addition, the main steam isolation valves would close on low steam pressure. Due to these type of plant features, PRAs have shown that secondary system steam breaks or leaks, regardless of how they occur, are not dominant contributors to core damage or offsite health effects.
CONCLUSION: This technical specification does not contain constraints of prime importance in limiting the likelihood or severity of the accident sequences commonly found to dominate risk.
RUAIN IN RELOCATE FROM TECHNICAL $ SPECIFICATIONS TECHNICAL SPECIFICATIONS NO YES i
A-24
l i
PRA TECHNICAL SPECIFICATION REVIEW FORM NORTH ANNA POWER STATION TECHNICAL SPECIFICATION: Unit 1 & 2, 3.7.12.1 - SETTLEMENT OF CLASS 1 STRUCTURES DESCRIPTI_0N OF RE0VIREMENT: This specification addresses the monitoring of the settlement of Class I structures and covers testing approach sequences and allowable settlements. The specifications require an engineering review, evaluation of the consequences and submittal of a report to the NRC when the settlement exceeds 75 percent of the allowable limits. This ensures edequate time to implement corrective actions to prevent reaching 100 percent of the allowable limits, at which time a step-wise plant shutdown would have to be initiated.
POTENTIAL EFFECT: Potential Distruption of Safety /Non-Safety Plant Functions REFERENCE DOCUMENTS UTILIZED:
PRA DESIGN BASIS RISK None Ref. 5: PUN PSS Ref. 6: Millstone 3 PSS Ref. 19: Zion PSS Ref. 24: Indian Point 2 & 3 PSS COMMENTS: The requirements of this technical specification are not risk dominant based on the core melt and health effects screening criteria provided in Section A.2 under item D parts 1 and 2. Further discussion of insights based on review of the reference documents is provided below.
Potential Distruotion of Safety /Non-Safety Plant Functions: Settlement of plant structures, except that caused by an external event such as an earth-quake is not modeled in PRAs. Moreover, settlement of plant structures is a slow process which would be. detectable by inspections and testing before plant safety or non-safety functions would be disrupted. None of the PRAs reviewed modeled or considered this failure mechanism as a dominant contributor to core
. damage or risk.
CONCLUSION: This technical specification does not contain constraints of prime importance in limiting the likelihood or severity of the accident sequences commonly found to dominate risk.
RETAIN IN RELOCATE FROM TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATIONS NO YES
~
A-25
I PRA TECHNICAL SPECIFICATION REVIEW FORM NORTH ANNA POWER STATION TECHNICAL SPECIFICATION: Unit 1 & 2 3.7.13 - GROUNDWATER LEVEL, SERVICE WATER RESERVOIR n DESCRIPTION OF RE0VIREMENT: This specification addresses the monitoring of water level and water drainage from the service water reservoir. The plant specifications require engineering studies and reporting of relevant findings and corrective actions to the NRC. Hence, this specification provides a long term surveillance program with no immediate actions short of the initiation of ;
an engineering evaluation and reporting of the findings. This specification is intended to monitor the seepage and drainage from the service water reservoir and to prevent failure of the reservoir.
POTENTIAL EFFECT: Loss of Service Water Cooling REFERENCE DOCUMENTS UTILIZED:
PRA DESIGN BASIS RISK None Ref. 5: PUN PSS Ref. 6: Millstone 3 PSS Ref. 19: Zion PSS Ref. 24: Indian Point 2 & 3 PSS COMMENTS: The requirements of this technical specification are not risk dominant based on the core melt and health effects screening criteria provided in Section A.2 under item D parts 1 and 2. Further discussion of insights based on review of the reference documents is provided below.
Loss of Service Water Coolino: Loss of service water cooling due to internal events, such as, pipe break or component failure, and external events, such as, fire, flooding or earthquake is modeled in nearly all PRAs. However, loss of service water due to seepage or drainage has not been modeled since it is a slow process which would be detectable by inspections and testing before. all service water cooling capability was lost. None of the PRAs reviewed modeled !
or considered this failure mechanism as a dominant contributor to core damage or risk. l CONCLUSION: This technical specification does not contain constraints of prime importance in limiting the likelihood or severity of the accident sequences commonly found to dominate risk.
RETAIN IN RELOCATE FROM TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATIONS NO YES l
A-26
7 .
PRA TECHNICAL SPECIFICATION REVIEW FORM i NORTH ANNA POWER STATION ;
TECHNICAL SPECIFICATION: Unit 2, 3.8.2.7 - NORMALLY DE-ENERGIZED POWER )
CIRCUITS :
DESCRIPTION OF RE0VIREMENT: This specification specifies the power circuits I which must be de-energized during operation in Modes 1, 2, 3 and 4. !
POTENTIAL EFFECT: Inadvertent Component Operation Effecting System Performance l REFERENCE DOCUMENTS UTILIZED:
PRA DESIGN BASIS RISK Ref. 9: NUREG-4550 Vol. 1 Surry Ref. 5: PUN PSS )
Ref. 28: NUREG-2728 IREP Guide Ref. 6: Millstone 3 P5S Ref. 29: NUREG-2300 PRA Procedures Ref. 19: Zion PSS Ref. 30: NUREG-2815 PSA Procedures Ref. 24: Indian Point 2 & 3 PSS !
Ref. 31: IDCOR 86.3Al IPEM COMMENTS: The requirements of this technical specification are not risk dominant based on the core melt and health effects screening criteria provided in Section A.2 under item D parts I and 2. Further discussion of insights based on review of the reference documents is provideo below.
Inadvertent Comnonent Operation Effectina System Performance: The circuits !
covered by this specification are for plant equipment that is not used during !
normal operation or for accident mitigation. However, equipment covered by this specification could possibly effect safety /non-safety system performance if operated. PRAs and PRA procedures, a referenced above, do not model operation of equipment that would be normally de-energized during normal plant operation since the effect on system performance is negligible Moreover, these type of faults would be quickly detected through system operability checks, and would be easily corrected in the event the systen was required to perform its intended function. No PRAs have identified this mechanism as a dominant contributor to core damage or risk.
CONCLUSION: This technical specification does not contain constraints of prime importance in limiting the likelihood or severity of the accident sequences commonly found to dominate risk.
RETAIN IN RELOCATE FROM TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATIONS NO YES A-27
- - - _ _ _ _ _ _