NL-24-0253, License Amendment Request: Revised Technical Specifications to Adopt TSTF-554, Revised Reactor Coolant Leakage Requirements
| ML24348A233 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/13/2024 |
| From: | Coleman J Southern Nuclear Operating Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| NL-24-0253 | |
| Download: ML24348A233 (1) | |
Text
>. Southern Nuclear December 13, 2024 Docket Nos.: 52-025 52-026 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Regulatory Affairs Southern Nuclear Operating Company Vogtle Electric Generating Plant Units 3 and 4 3535 Colonnade Parkway Birmingham, AL 35243 Tel. 205.992.5000 NL-24-0253 10 CFR 50.90 License Amendment Request: Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements" Pursuant to 10 CFR 52.98(c) and in accordance with 10 CFR 50.90, Southern Nuclear Operating Company (SNC) requests an amendment to the combined licenses (COLs) for Vogtle Electric Generating Plant (VEGP) Unit 3 (License Number NPF-91) and Unit 4 (License Number NPF-92).
SNC requests adoption of TSTF-554, "Revise Reactor Coolant Leakage Requirements," which is an approved change to the Standard Technical Specifications (STS), into the VEGP Units 3 and 4 Technical Specifications (TS). The proposed amendment revises the TS definition of "LEAKAGE," clarifies the requirements when pressure boundary LEAKAGE is detected and adds a Required Action when pressure boundary LEAKAGE is identified. The change is requested as part of the Consolidated Line Item Improvement Process (CLIIP), and is identified as applicable to NUREG-2194, Standard Technical Specifications Westinghouse Advanced Passive 1000 (AP1000) Plants, Revision 1.
The enclosure provides the description, technical evaluation, regulatory evaluation (including the Significant Hazards Consideration Determination) and environmental considerations for the proposed change.
SNC requests approval of the proposed license amendment request no later than 6 months from the acceptance review determination of this request. SNC expects to implement the proposed amendment within 60 days of issuance.
This letter contains no regulatory commitments. This letter has been reviewed and determined not to contain security-related information.
In accordance with 10 CFR 50.91, a copy of this application, with enclosure, is being provided to the designated State official.
U. S. Nuclear Regulatory Commission NL-24-0253 Page 2 If you have any questions, please contact Mr. Ryan Joyce at (205) 992-6468.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on the 13th day of December 2024.
Respectfully submitted,
-~
~man Director, Regulatory Affairs Southern Nuclear Operating Company Enclosure Evaluation of Proposed Change cc:
NRC Regional Administrator, Region II NRR Project Manager - Vogtle 3 & 4 Senior Resident Inspector - Vogtle 3 & 4 Director, Environmental Protection Division - State of Georgia Document Services RTYPE: VND.LI.L00
Enclosure to NL-24-0253 Evaluation of Proposed Change License Amendment Request:
Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements"
- 1.
DESCRIPTION
- 2.
ASSESSMENT 2.1. Applicability of Safety Evaluation 2.2. Variations 2.3. Other Editorial Differences
- 3. REGULATORY ANALYSIS 3.1. No Significant Hazards Consideration Analysis 3.2. Conclusion
- 4. ENVIRONMENTAL CONSIDERATION Attachments
- 1. Licensing Basis Document Markups
- 2. Revised Licensing Basis Document Pages
- 3. Associated Technical Specifications Bases Changes (For information only)
Enclosure to NL-24-0253 Evaluation of Proposed Change
1.0 DESCRIPTION
Southern Nuclear Operating Company (SNC) requests adoption of TSTF-554, Revision 1, "Revise Reactor Coolant Leakage Requirements," which is an approved change to NUREG-2194, the Standard Technical Specifications (STS) for Westinghouse AP1000 Plants, into the Vogtle Electric Generating Plant (VEGP) Units 3 and 4 Technical Specifications (TS).
The proposed amendment revises the TS definition of "LEAKAGE" and the Reactor Coolant System (RCS) Operational LEAKAGE TS to clarify the requirements.
2.0 ASSESSMENT
2.1 Application of Safety Evaluation SNC has reviewed the safety evaluation for TSTF-554 provided to the Technical Specifications Task Force in a letter dated December 18, 2020. This review included a review of the NRC staff's evaluation, as well as the information provided in TSTF-554.
As described herein, SNC has concluded that the justifications presented in TSTF-554 and the safety evaluation prepared by the NRC staff are applicable to VEGP Units 3 and 4 and justify this amendment for the incorporation of the changes into the VEGP Units 3 and 4 TS.
2.2 Variations 2.3 SNC is not proposing any variations from the TS changes described in TSTF-554 or the applicable parts of the NRC staff's safety evaluation dated December 18, 2020.
Other Editorial Differences The proposed changes also resolve editorial inconsistencies within the affected TS.
Specifically, in the first sentence of the TS 1.1 definition of Pressure Boundary LEAKAGE, the indefinite article "a" is replaced with "an" to read, "an RCS component... "
The VEGP TS are generally written following the guidance of TSTF-GG-05-01, Revision 1, for format and content. TSTF-GG-05-01 recommends the use of articles that agree with the pronunciation of the acronyms they introduce. Furthermore, the TS includes several occurrences of the phrase "an RCS," so the change to refer to an RCS component provides consistency with the approved guidance in TSTF-GG-05-01 and the text found elsewhere in the TS.
In addition, it is noted that the NUREG-2194, Revision 1, and VEGP TS 1.1 definition of Pressure Boundary LEAKAGE, uses the word "nonisolatable" in lieu of "nonisolable,"
which is used in TSTF-554, Revision 1. While both words are undefined and are therefore subject to inconsistent interpretation as discussed in TSTF-554, they are generally considered to be the same. Accordingly, the TSTF-554 justification for removing the word "nonisolable" is equally applicable to removal of the word "nonisolatable" from VEGP TS 1.1.
E-2 of 5
Enclosure to NL-24-0253 Evaluation of Proposed Change These editorial changes have no adverse impact on the technical justification of the changes presented in TSTF-554, Revision 1. Furthermore, replacement of "nonisolable" with "nonisolatable" in the justification for the response to No Significant Hazards Consideration Analysis standard 1 does not affect the "no significant hazards consideration" finding.
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Analysis Southern Nuclear Operating Company (SNC) requests adoption of TSTF-554, "Revise Reactor Coolant Leakage Requirements," which is an approved change to the Standard Technical Specifications (STS), into the Vogtle Electrical Generating Plant Units 3 and 4 Technical Specifications (TS). The proposed amendment revises the TS definition of "LEAKAGE," clarifies the requirements when pressure boundary LEAKAGE is detected, and adds a Required Action when pressure boundary LEAKAGE is identified.
SNC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below.
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:No The proposed amendment revises the TS definition of "LEAKAGE," clarifies the requirements when pressure boundary LEAKAGE is detected and adds an ACTION when pressure boundary LEAKAGE is identified.
The proposed change revises the definition of pressure boundary LEAKAGE. Pressure boundary LEAKAGE is a precursor to some accidents previously evaluated. The proposed change expands the definition of pressure boundary LEAKAGE by eliminating the qualification that pressure boundary LEAKAGE must be from a "nonisolatable" flaw.
A new TS ACTION is created which requires isolation of the pressure boundary flaw from the Reactor Coolant System (RCS). This new ACTION provides assurance that the flaw will not result in any accident previously evaluated.
Pressure boundary LEAKAGE, and the actions taken when pressure boundary LEAKAGE is detected, is not assumed in the mitigation of any accident previously evaluated. As a result, the consequences of any accident previously evaluated are unaffected.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
E-3 of 5
Enclosure to NL-24-0253 Evaluation of Proposed Change
- 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed amendment revises the TS definition of "LEAKAGE," clarifies the requirements when pressure boundary LEAKAGE is detected and adds an ACTION when pressure boundary LEAKAGE is identified. The proposed change does not alter the design function or operation of the RCS. The proposed change does not alter the ability of the RCS to perform its design function. Since pressure boundary LEAKAGE is an evaluated accident, the proposed change does not create any new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No The proposed amendment revises the TS definition of "LEAKAGE," clarifies the requirements when pressure boundary LEAKAGE is detected and adds an ACTION when pressure boundary LEAKAGE is identified. The proposed change does not affect the initial assumptions, margins, or controlling values used in any accident analysis. The amount of LEAKAGE allowed from the RCS is not increased. The proposed change does not affect any design basis or safety limit or any Limiting Condition for Operation.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, SNC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
3.2 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.
4.0 ENVIRONMENTAL CONSIDERATION
The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in E-4 of 5
Enclosure to NL-24-0253 Evaluation of Proposed Change individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
E-5 of 5
ATTACHMENT 1 to NL-24-0253 Licensing Basis Document Markups License Amendment Request:
Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements" (This attachment consists of 4 pages, including this cover page.)
1.1 Definitions ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME LEAKAGE VEGP Units 3 and 4 Technical Specifications Definitions 1.1 The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions). The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.
LEAKAGE shall be:
- a.
Identified LEAKAGE
- 1.
LEAKAGE, such as that from seals or valve packing, that is captured and conducted to collection systems or a sump or collecting tank;
- 2.
LEAKAGE into the containment atmosphere from sources that are both specifically located and known ei-tRef-!Q_not-tG interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;
- 3.
Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System (primary to secondary LEAKAGE); or
- 4.
RCS LEAKAGE through the passive residual heat removal heat exchanger (PRHR HX) to the In-containment Refueling Water Storage Tank (IRWST).
- b.
Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.
- c.
Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE and PRHR HX tube LEAKAGE) through a nonisolatable fault in an RCS component body, pipe wall, or vessel wall.
LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.
1.1 - 3 Amendment No. ~ -#:##- (Unit 3)
Amendment No. ~-#:##- (Unit 4)
Technical Specifications RCS Operational LEAKAGE 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Operational LEAKAGE LCO 3.4.7 RCS operational LEAKAGE shall be limited to:
- a.
- b.
0.5 gpm unidentified LEAKAGE,
- c.
10 gpm identified LEAKAGE from the RCS,
- d.
150 gallons per day primary to secondary LEAKAGE through any one Steam Generator (SG), and
- e.
500 gallons per day primary to In-Containment Refueling Water Storage Tank (IRWST) LEAKAGE through the passive residual heat removal heat exchanger (PRHR HX).
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTIONS CONDITION A.
Pressure boundary A.1 LEAKAGE exists.
§.
LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.
VEGP Units 3 and 4 REQUIRED ACTION COMPLETION TIME Isolate affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> com12onent, 12i12e, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.
Reduce LEAKAGE to within limits.
3.4.7 - 1 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Amendment No. '##-# (Unit 3)
Amendment No. '##-# (Unit 4)
ACTIONS (continued)
CONDITION g
Required Action C.
and associated Completion Time not met.
00 Pressure boundary U:AKAGe exists.
OR Primary to secondary LEAKAGE not within limit.
Technical Specifications BC.1 AND BC.2 REQUIRED ACTION Be in MODE 3.
Be in MODE 5.
SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.7.1
- NOTES -
RCS Operational LEAKAGE 3.4.7 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours FREQUENCY
- 1.
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
- 2.
Not applicable to primary to secondary LEAKAGE.
Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.
- NOTE -
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
Verify primary to secondary LEAKAGE is ::; 150 gallons per day through any one SG.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 72 hours VEGP Units 3 and 4 3.4.7 - 2 Amendment No. '##-# (Unit 3)
Amendment No. '##-# (Unit 4)
ATTACHMENT 2 to NL-24-0253 Revised Licensing Basis Document Pages License Amendment Request:
Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements" (This attachment consists of 4 pages, including this cover page.)
1.1 Definitions ENGINEERED SAFETY FEATURE(ESF)RESPONSE TIME LEAKAGE VEGP Units 3 and 4 Technical Specifications Definitions 1.1 The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions). The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.
LEAKAGE shall be:
- a.
Identified LEAKAGE
- 1.
LEAKAGE, such as that from seals or valve packing, that is captured and conducted to collection systems or a sump or collecting tank;
- 2.
LEAKAGE into the containment atmosphere from sources that are both specifically located and known to not interfere with the operation of leakage detection systems;
- 3.
Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System (primary to secondary LEAKAGE); or
- 4.
RCS LEAKAGE through the passive residual heat removal heat exchanger (PRHR HX) to the In-containment Refueling Water Storage Tank (IRWST).
- b.
Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.
- c.
Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE and PRHR HX tube LEAKAGE) through a fault in an RCS component body, pipe wall, or vessel wall. LEAKAGE past seals, packing, and gaskets is not pressure boundary LEAKAGE.
1.1 - 3 Amendment No.
Amendment No.
"### (Unit 3)
"### (Unit 4)
Technical Specifications RCS Operational LEAKAGE 3.4.7 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.7 RCS Operational LEAKAGE LCO 3.4.7 RCS operational LEAKAGE shall be limited to:
- a.
- b.
0.5 gpm unidentified LEAKAGE,
- c.
10 gpm identified LEAKAGE from the RCS,
- d.
150 gallons per day primary to secondary LEAKAGE through any one Steam Generator (SG), and
- e.
500 gallons per day primary to In-Containment Refueling Water Storage Tank (IRWST) LEAKAGE through the passive residual heat removal heat exchanger (PRHR HX).
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTIONS CONDITION A.
Pressure boundary A.1 LEAKAGE exists.
B.
RCS operational B.1 LEAKAGE not within limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.
VEGP Units 3 and 4 REQUIRED ACTION COMPLETION TIME Isolate affected 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> component, pipe, or vessel from the RCS by use of a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve.
Reduce LEAKAGE to within limits.
3.4.7 - 1 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Amendment No. '##-# (Unit 3)
Amendment No. '##-# (Unit 4)
ACTIONS (continued)
CONDITION C.
Required Action and associated Completion Time not met.
Primary to secondary LEAKAGE not within limit.
Technical Specifications C.1 AND C.2 REQUIRED ACTION Be in MODE 3.
Be in MODE 5.
SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.7.1
- NOTES -
RCS Operational LEAKAGE 3.4.7 COMPLETION TIME 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 36 hours FREQUENCY
- 1.
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
- 2.
Not applicable to primary to secondary LEAKAGE.
Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.
- NOTE -
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
Verify primary to secondary LEAKAGE is ::; 150 gallons per day through any one SG.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 72 hours VEGP Units 3 and 4 3.4.7 - 2 Amendment No. '##-# (Unit 3)
Amendment No. '##-# (Unit 4)
ATTACHMENT 3 to NL-24-0253 Associated Technical Specifications Bases Changes
{For information only)
License Amendment Request:
Revise Technical Specifications to Adopt TSTF-554, "Revise Reactor Coolant Leakage Requirements" (This attachment consists of 6 pages, including this cover page.)
Technical Specifications Bases RCS Operational LEAKAGE B 3.4.7 BASES BACKGROUND (continued)
APPLICABLE SAFETY ANALYSES LCO VEGP Units 3 and 4 LCO 3.4.15, "RCS Pressure Isolation Valve (PIV) Integrity," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.
Except for primary to secondary LEAKAGE, the safety analyses do not address operational LEAKAGE. However, other operational LEAKAGE is related to the safety analyses for LOCA. The amount of LEAKAGE can affect the probability of such an event. The safety analysis for an event resulting in steam discharge to the atmosphere assumes a 300 gpd primary to secondary LEAKAGE as the initial condition.
Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leak contaminates the secondary fluid.
The FSAR Chapter 15 (Ref. 3) analyses for the accidents involving secondary side releases assume 150 gpd primary to secondary LEAKAGE in each generator as an initial condition. The design basis radiological consequences resulting from a postulated SLB accident and SGTR are provided in Sections 15.1.5 and 15.6.3 of FSAR Chapter 15, respectively.
The RCS operational LEAKAGE satisfies Criterion 2 of 1 O CFR 50.36(c)(2)(ii).
RCS operational LEAKAGE shall be limited to:
- a.
-NG--f}_Eressure boundary LEAKAGE is prohibited allo>.ved, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further RCPB deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.
B 3.4.7 - 2 Revision ~
BASES LCO (continued)
VEGP Units 3 and 4 Technical Specifications Bases RCS Operational LEAKAGE B 3.4.7
- b.
Unidentified LEAKAGE 0.5 gpm of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air F18 particulate radioactivity monitoring and containment sump level monitoring equipment, can detect within a reasonable time period. This leak rate supports leak before break (LBB) criteria. Separating the sources of LEAKAGE (i.e., LEAKAGE from an identified source versus LEAKAGE from an unidentified source) is necessary for prompt identification of potentially adverse conditions, assessment of the safety significance, and corrective action. Violation of this LCO could result in continued degradation of the RCPB, if the U:AKAGe is from the pressure boundary.
- c.
Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LeAKAGe.
Violation of this LCO could result in continued degradation of a component or system.
- d.
Primary to Secondary LEAKAGE through One SG The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 4 ). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.
B 3.4.7 - 3 Revision ~
BASES LCO (continued)
APPLICABILITY ACTIONS VEGP Units 3 and 4 Technical Specifications Bases RCS Operational LEAKAGE B 3.4.7
- e.
Primary to In-Containment Refueling Water Storage Tank (IRWST)
LEAKAGE through the Passive Residual Heat Removal Heat Exchanger (PRHR HX)
The 500 gpd limit from the PRHR HX is based on the assumption that a single crack leaking this amount would not lead to a PRHR HX tube rupture under the stress condition of an RCS pressure increase event. If leakage is through many cracks, and the cracks are very small, then the above assumption is conservative. This is conservative because the thickness of the PRHR HX tubes is approximately 60% greater than the thickness of the SG tubes.
Furthermore, a PRHR HX tube rupture would result in an isolable leak and would not lead to a direct release of radioactivity to the atmosphere.
In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.
In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.
A.1 If pressure boundary LEAKAGE exists, the affected component, pipe, or vessel must be isolated from the RCS by a closed manual valve, closed and de-activated automatic valve, blind flange, or check valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. While in this condition, structural integrity of the system should be considered because the structural integrity of the part of the system within the isolation boundary must be maintained under all licensing basis conditions, including consideration of the potential for further degradation of the isolated location. Normal LEAKAGE past the isolation device is acceptable as it will limit RCS LEAKAGE and is included in identified or unidentified LEAKAGE. This action is necessary to prevent further deterioration of the RCPB.
B.1 Unidentified LEAKAGE or identified LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This Completion Time allows time to verify leakage rates and either identify unidentified LEAKAGE or reduce LEAKAGE to within limits before the reactor must be shut down. This action is necessary to prevent further deterioration of the RCPB.
B 3.4.7 - 4 Revision ~
Technical Specifications Bases RCS Operational LEAKAGE B 3.4.7 BASES ACTIONS (continued)
SURVEILLANCE REQUIREMENTS VEGP Units 3 and 4 BC.1 and BC.2 If any pressure boundary LEAKAGE exists, or primary to secondary LEAKAGE is not within limits, or if any of the Required Actions and associated Completion Times are not met, unidentified or identified LEAKAGE cannot be reduced to 1.yithin limits 1.yithin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.
The reactor must be brought to MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and to MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This action reduces the LEAKAGE and also reduces the factors which tend to degrade the pressure boundary.
The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without ACTIONS challenging plant systems. In MODE 5, the pressure stresses acting on the RCPB are much lower, and further deterioration is much less likely.
SR 3.4.7.1 Verifying RCS LEAKAGE within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection.
Unidentified LEAKAGE and identified LEAKAGE are determined by performance of a RCS water inventory balance.
The RCS water inventory balance must be met with the reactor at steady state operating conditions. The Surveillance is modified by two Notes.
Note 1 states that this SR is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishing steady state operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.
Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer level, and reactor coolant drain tank and in-containment refueling water storage tank levels.
B 3.4.7 - 5 Revision ~
Technical Specifications Bases RCS Operational LEAKAGE B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued)
VEGP Units 3 and 4 RCS inventory monitoring via the pressurizer level changes is valid in MODES 1, 2, 3, and 4 only when RCS conditions are stable, i.e.,
temperature is constant, pressure is constant, no makeup and no letdown.
An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere F18 particulate radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These LEAKAGE detection systems are specified in LCO 3.4.9, "RCS LEAKAGE Detection Instrumentation."
Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.
The 72-hour Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents.
SR 3.4.7.2 This SR verifies that primary to secondary LEAKAGE is less or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 5. The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.
The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and reactor coolant drain tank and in-containment refueling water storage tank levels, and makeup and letdown flows.
B 3.4.7 - 6 Revision ~
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