Similar Documents at Surry |
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Category:TECHNICAL SPECIFICATIONS
MONTHYEARML18152B3591999-08-23023 August 1999 Revised Tech Specs Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in UFSAR & Includes Ref to Applicable UFSAR Section ML18152B6591999-04-28028 April 1999 Proposed Tech Specs Re Refueling Water Chemical Addition Tank Min Vol ML18152A2131999-02-16016 February 1999 Proposed Tech Specs,Consolidating AFW cross-connect Requirements by Relocation of Electrical Power Requirements from TS 3.16 to TS 3.6 ML18152B5451999-02-16016 February 1999 Proposed Tech Specs Pages,Revising Augmented Insp Requirements for Reactor Coolant Pump Flywheels ML18152B6151998-11-0404 November 1998 Proposed Tech Specs 4.6.A.1.b,re EDG Start & Load Time Testing Requirements & TS 3.16 Bases Re EDG Ratings ML18153A3311998-09-24024 September 1998 Proposed Tech Specs Modifying Testing Requirements for Reactor Trip Bypass Breaker ML18153A3351998-09-24024 September 1998 Proposed Tech Specs Pages Affected by Suppl to 960912 Resubmittal of Change Request 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[Table view] Category:TECHNICAL SPECIFICATIONS & TEST REPORTS
MONTHYEARML18152B3591999-08-23023 August 1999 Revised Tech Specs Basis Pages for TS 3.1.B,deleting Reactor Vessel Toughness Data Duplicated in UFSAR & Includes Ref to Applicable UFSAR Section ML18151A6641999-08-0606 August 1999 Rev 0 to Surry Unit 2 Cycle 16 Startup Physics Tests Rept. ML18152B6591999-04-28028 April 1999 Proposed Tech Specs Re Refueling Water Chemical Addition Tank Min Vol ML18152A2131999-02-16016 February 1999 Proposed Tech Specs,Consolidating AFW cross-connect Requirements by Relocation of Electrical Power Requirements from TS 3.16 to TS 3.6 ML18152B5451999-02-16016 February 1999 Proposed Tech Specs Pages,Revising Augmented Insp Requirements for Reactor Coolant Pump Flywheels ML18151A5511999-02-10010 February 1999 to NE-1187, Surry Unit 1,Cycle 16 Startup Physics Tests Rept. ML18152B6151998-11-0404 November 1998 Proposed Tech Specs 4.6.A.1.b,re EDG Start & Load Time Testing Requirements & TS 3.16 Bases Re EDG Ratings ML18153A3351998-09-24024 September 1998 Proposed Tech Specs Pages Affected by Suppl to 960912 Resubmittal of Change Request Re Relocation of Fire Protection Requirements from TS to UFSAR ML18153A3311998-09-24024 September 1998 Proposed Tech Specs Modifying Testing Requirements for Reactor Trip Bypass Breaker ML18152B7551998-06-19019 June 1998 Proposed Tech Specs Establishing Requirements for Use of Temporary Supply Line (Jumper) to Provide Svc Water to Component Cooling Heat Exchangers ML20249B9911998-05-0606 May 1998 Analysis of Capsule X Virginia Power Surry Unit 1 Reactor Vessel Matl Surveillance Program. W/Evaluation of Surry Unit 1 Surveillance Capsule X Results & Response to NRC RAI Re GL 92-01,rev 1,suppl 1 ML18151A1931998-05-0404 May 1998 Rev 1 to Summary of Changes to Surry Units 1 & 2 Third Interval IST Program. ML18152A3651998-03-25025 March 1998 Proposed Tech Specs Revising Station Mgt Titles to Reflect New Positions Approved by Vepc Board of Directors on 980220 ML20199B0711998-01-0505 January 1998 Rev 0 to NE-1148, Surry Unit 2,Cycle 15 Startup Physics Test Rept ML18153A3481997-12-18018 December 1997 Proposed Tech Specs Clarifying Terminology Used for Describing Equipment Surveillances Conducted on Refueling Interval Frequency.Clarification Consistent W/Info Contained in Rev 1 to NUREG-1431 ML18150A4661997-12-16016 December 1997 ISI Plan for Third Insp Interval,Vol 2,Rev 9 for Components & Component Supports,940510-040510, for Surry Power Station,Unit 2 ML18153A3941997-11-0505 November 1997 Proposed Tech Specs Re Change for Increased Enrichment of Reload Fuel ML18153A1761997-11-0505 November 1997 Proposed Tech Specs Re Temporary Svc Water Supply Line to Component Cooling Heat Exchangers ML18150A4641997-10-27027 October 1997 Risk-Informed ISI (RI-ISI) Pilot Program Submittal. ML18151A3911997-10-16016 October 1997 Rev 8 to VPAP-2103, Odcm. ML18151A7231997-08-0707 August 1997 Rev 1 to Nuclear Safety Analysis Manual Part Iv,Chapter a Probabilistic Safety Assessment Products. ML20210J5031997-07-31031 July 1997 Rev 0 to NE-1132, Surry Unit 1,Cycle 15 Startup Physics Tests Rept ML18150A4441997-06-0909 June 1997 Vol 2,Rev 8 to ISI Plan for Third Insp Interval for Components & Component Supports,Oct 14,1993-Oct 13,2003. ML18153A5231997-04-24024 April 1997 Proposed Corrected Tech Specs Pages 6.1-3 & 6.1-8 Re Relocation of Fire Protection TS to Updated Final Safety Analysis Rept ML18153A5031997-03-18018 March 1997 Proposed Tech Specs Rev to Section 4.15 for Surry Power Station to Include Pp Inadvertently Omitted from 970203 Request for Amend to Licenses DPR-32 & DPR-37 ML18153A4921997-02-0303 February 1997 Proposed Tech Specs Re Deletion of Specific ASME Section XI Code Ref ML18153A6351996-11-26026 November 1996 Proposed Tech Specs Re Removal of Record Retention Requirements,Per GL 95-06 & Administrative Ltr 95-06 ML18153A0671996-09-12012 September 1996 Proposed Tech Specs Re Relocation of Fire Protection Requirements ML18151A9761996-08-13013 August 1996 Cycle 14 Startup Physics Test Rept. W/960830 Ltr ML20134J9861996-07-30030 July 1996 /Unit 2 Fuel Assembly Insp Program ML18152A4701996-06-13013 June 1996 Cycle 13 Control Rod Performance Test Results. ML18153A6901996-04-15015 April 1996 Proposed Tech Specs,Clarifying Applicability of Quadrant Power Tilt Ration Requirements ML18153A5391996-03-21021 March 1996 Proposed Tech Specs Re Charcoal Filter Testing Clarification ML18153A5271996-03-14014 March 1996 Proposed Tech Specs,Permitting Use of 10CFR50 App J,Option B,performance-based Containment Lrt ML18153A5801996-01-30030 January 1996 Proposed Tech Specs Re Reactor Coolant Sys Liquid Sampling ML18152A0571995-12-20020 December 1995 Startup Physics Test Rept,Surry Unit 1,Cycle 14. W/960111 Ltr ML18153A6761995-11-20020 November 1995 Proposed Tech Specs Re App J Option B,performance-based Containment Leakage Rate Testing ML18151A6421995-08-0101 August 1995 Change 3 to Rev 0 to Third Interval IST Program ML18153A7141995-07-20020 July 1995 Proposed Tech Specs Establishing New Setpoint Limit for SG high-high Level & Provides More Restrictive Setting Limits for Certain Rps/Esfas Setpoints ML18153A6991995-07-14014 July 1995 Proposed Tech Specs,Providing Two H Allowed Outage Time for One RHR Pump to Accommodate Plant Safety,Emergency Power Sys Surveillance Testing & Permit Depressurizing SI Accumulators in Lieu of Accumulator Isolation ML18153A8371995-06-0808 June 1995 Proposed Tech Specs,Incorporating Revised Pressure/Temp Limits & Associated Ltops Setpoint That Will Be Valid to end-of-license ML20083C9951995-05-0808 May 1995 Rev 0 to Surry Unit 2,Cycle 13 Startup Physics Tests Rept ML18153B2301995-02-14014 February 1995 Proposed Tech Specs Re App J Testing Requirements ML18153B2131995-01-24024 January 1995 Proposed Tech Specs,Modifying as-found Test Acceptance Criterion for Pressurizer Safety Valves ML18153B1621994-11-29029 November 1994 Proposed Tech Specs Implementing Zirlo Fuel Cladding ML18153B1581994-11-22022 November 1994 Proposed Tech Specs,Deleting Unnecessary Descriptive Phrases Re Number of Cells in Station & EDG Batteries ML18153B1501994-11-10010 November 1994 Proposed Tech Specs Re Changes to TS Will Clarify SR for Reactor Protection & Engineered Safeguard Sys Instrumentation & Actuation Logic ML18153B0941994-10-11011 October 1994 Proposed Tech Specs Surveillance Frequencies for Hydrogen Analyzers ML18152A5061994-09-0606 September 1994 Proposed Tech Specs Re Mgt Safety Review Committee & Station Nuclear Safety & Operating Committee Responsibilities ML18152A1191994-08-30030 August 1994 Proposed Tech Specs to Accomodate Core Uprating 1999-08-06
[Table view] |
Text
- Attachment 2 Technical Specifications Change Surry Power Station
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. 4A . CONTAINMENT TESTS
- TS 4.4-1 Applicability Applies to containment leakage testing.
Objective To assure that leakage of the primary reactor containment and associated systems is held within allowable leakage rate limits; and to assure that periodic surveillance is performed to assure proper maintenance and leak repair during the service life of the containment.
Specification A; Periodic and post-operational integrated leakage rate tests of the containment shall be performed in accordance with the requirements of 10 CFR 50, Appendix J, "Reactor Containment Leakage Testing for Water Cooled Power Reactors."
B. Containment Leakage Rate Testing Requirements
- 1. The containment and containment penetrations leakage rate shall be demonstrated by performing leakage rate testing in accordance with 10 CFR 50 Appendix J, Option B, as modified by approved exemptions, and Regulatory Guide 1.163.
Leakage rate acceptance criteria are as follows:
- a. An overall integrated leakage rate of less than or equal to La, 0.1 percent by weight of containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, at calculated peak pressure (Pa).
- b. A combined leakage rate of less than or equal to 0.60 La for all penetrations and valves subject to Type B and C testing when pressurized to Pa.
Prior to entering an operating conditions where containment integrity is required the as-left Type A leakage rate shall not exceed 0.75 La and the combined leakage rate of all penetrations subject to Type Band C testing shall not exceed 0.6 La."
- 2. The provisions of Specification 4.0.2 are not applicable.
The leak tightness testing of all liner welds was performed during construction by welding a structural steel test channel over each weld seam and performing soap bubble and halogen leak tests.
Amendment Nos.
- The containment is designed for a maximum pressure of 45 psig. The containment is TS 4.4-2 maintained at a subatmospheric air partial pressure consistent with TS Figure 3.8-1 depending upon the cooldown capability of the Engineered Safeguards and will not rise above 45 psig for any postulated loss-of-coolant accident.
The initial test pressure for the Type A test is 4 7 .0 psig to allow for containment expansion and equalization. A review was performed to determine the effects o~
pressurizing containment above its design pressure of 45.0 psig. This review was based on the original containment test at 52 psig. During that test, the calculated stresses were found to be well within the allowable yield strength of the structural reinforcing bars, therefore performance of the Type A test at 47 psig will have no detrimental effect on the containment structure.
All loss-of-coolant accident evaluations have been based on an integrated containment leakage rate not to exceed 0.1 % of containment volume per 24 hr.
The above specification satisfies the conditions of 10 CFR 50.54(0) which stated that primary reactor containments shall meet the containment leakage test requirements set forth in Appendix J.
The limitations on closure and leak rate for the containment airlocks are required to meet the restrictions on containment integrity and containment leak rate. Surveillance testing of the airlock seals provides assurance that the overall airlock leakage will not become excessive due to seal damage during the intervals between airlock leakage tests.
References UFSAR Section 5.4 Design Evaluation of Containment Tests and Inspections of Containment UFSAR~Section 7:5:-1 *---*-*Design-Bases ofEngineered*Safeguards*Instrumentation UFSAR Section 14.5 Loss of Coolant Accident 10 CFR 50 Appendix J "Primary Reactor Containment Leakage Testing for Water Cooled Power Reactors" Amendment Nos.
- TS 6.6-11 C. Special Reports In the event that the Reactor Vessel Overpressure Mitigating System is used to mitigate a RCS pressure transient, submit a Special Report to the Commission within 30 days.
The report shall describe the circumstances initiating the transient, the effect of the PORVs or the administrative controls on the transient and any corrective action necessary to prevent recurrence.
FOOTNOTES
- 1. A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station.
- 2. This tabulation Sl!pplements the i;~quiJernents _of Section 20.2206 of 10 CFR Part 20.
- 3. A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
Amendment Nos.
- Attachment 3 Significant Hazards Consideration Determination Surry Power Station
- tO CFR 50.92 EVALUATION - BASIS FOR
- NO SIGNIFICANT* HAZARDS DETERMINATION The Nuclear Regulatory Commission has amended its regulations to provide a performance-based option for leakage-rate testing of containments. This testing option is available in lieu of compliance with the prescriptive requirements contained in Appendix J regulations. In order to implement the performance-based leakage-rate testing option the Technical Specifications must be changed to eliminate reference to the prescriptive Appendix J requirements. Therefore, Virginia Electric and Power Company (Virginia Power) is proposing a change to the Surry Technical Specifications to eliminate the current prescriptive requirements for leakage rate testing of the containment and reference Option B to 10 CFR 50 Appendix J and NRC Regulatory Guide 1.163, "Performance-Based Containment Leakage-Test Program."
This change will permit use of the performance-based surveillance testing, Option B, of 10 CFR 50 Appendix J. Specifically, operation of Surry Power Station with the proposed change will not:
- 1. Involve a significant increase in either the probability of occurrence or consequences of any accident or equipment malfunction scenario which is important to safety and which has been previously evaluated in the Updated Final Safety Analysis Report (UFSAR).
Plant systems and components will not be operated in a different manner as a result of the proposed Technical Specifications change. The proposed change permits a performance-based approach to determining the leakage-rate test frequency for the containment and containment penetrations (Type A, B, and C tests). There are no plant modifications, or changes in methods of operation. Therefore, the changes in testing intervals for the containment and containment penetrations have no affect on the probability of occurrence of a LOCA. Since the proposed change only affects the test frequency for containment and the containment penetrations, and the as-found test acceptance criteria at Surry the probability of occurrence and the consequences of an accident are not affected by the proposed changes in the leak-rate test interval.
The proposed change increases the probability of a malfunction of equrpment important to safety due to the longer intervals between leakage tests. It has been estimated that the longer test intervals will increase the overall accident risk to the public by approximately 0.7% and 2.2% (for changes in the frequency of Type A tests and Type B and C tests, respectively). However, this increase in accident risk has been judged to be insjgnificant. This increa?e ha? been reviewed and.juqged to be acceptable by the NRC as documented in NUREG-1493 and the recent rulemaking to 10 CFR 50 Appendix J.
The containment and other safety system remain operable as assumed in the accident analysis. Changing the as-found acceptance criterion to 1.0 La at Surry does not increase the probability or consequences of an accident, since the accident analysis assume a leakage rate of La for Design Basis Page 1 of 2
. .
- Accidents. The as-left Type A test acceptance criterion remains at less than
- 0. 75 La. Since the proposed changes do not affect the Limiting Conditions for Operation for the containment, the containment penetrations, or the other safety systems, the consequences of an accident are not affected by the changes in test frequency.
Therefore, the probability of an accident or consequences of an accident are not adversely affected as a result of this change.
- 2. Create the possibility of a new or different type of accident than those previously evaluated in the UFSAR.
Implementing the proposed Technical Specifications change to remove the prescriptive testing requirements and permit use of Appendix J, Option B, performance-based testing of containment and its penetrations does not create the possibility of an accident of a different type than was previously evaluated in the UFSAR. Plant systems and components will not be operated in a different manner as a result of the proposed Technical Specifications changes. Thus, the proposed Technical Specifications changes in leakage-rate test frequency do not introduce any new accident precursors or modes of operations. The containment and containment penetrations will not be operated any differently as a result of the proposed changes. Therefore, the possibility for an accident of a different type than was previously evaluated in the Safety Analysis Report is not created by the proposed Technical Specifications change.
- 3. Involve a significant reduction in a margin of safety.
The proposed Technical Specifications change, which replace the present prescriptive testing requirements with Appendix J, Option B, performance-based testing of containment and its penetrations, will continue to ensure that the existing accident analysis assumptions are maintained. The containment and containment penetrations will not be operated or tested any differently. The leakage rate test frequency is being changed as a result of the proposed change. Changing the as-found acceptance criterion to 1.0 La at Surry does not increase the consequences of an accident, since the accident analysis assume a leakage rate of La for Design Basis Accidents.
The as-left Type A test acceptance criterion remains at less than 0.75 La, which maintains the operating margin. The operational leakage-rate test acceptance criteria and the operability requirements are not being changed.
Therefore, the margin of safety as defined in the Technical Specifications bases is unaffected.