NL-24-0387, Enclosure 4, Revision 25 to NL-24-0387, Technical Requirements Manual (TRM) Burden Reduction
| ML24297A655 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 10/23/2024 |
| From: | Southern Nuclear Company |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24298A045 | List: |
| References | |
| NL-24-0387 | |
| Download: ML24297A655 (1) | |
Text
NL-24-0387 Vogtle Electric Generating Plant, Units 1 & 2 Revision 25 to the Updated Final Safety Analysis Report, Technical Specification Bases Changes, Technical Requirements Manual Changes, 10 CFR 50.59 Summary Report, and Revised NRC Commitments Report 10 CFR 50.59 Summary Report to NL-24-0387 10 CFR 50.59 Summary Report Activity: LDCR 2022029
Title:
Technical Requirements Manual (TRM) Burden Reduction 10 CFR 50.59 Evaluation Summary:
A. This activity revises TR 13.1. 7, "Borated Water Sources - Operating," is revised by removing the requirement stating that the refueling water storage tank (RWST) is a required borated water source whenever in MODES 1, 2, 3, and 4. This revision involves changes to the TR requirement and deletion of Condition D, Surveillance Note 1, and TR Surveillance (TRS) 13.1.7.1 (and renumbering the remaining surveillances). Surveillance Note 2 is revised to clarify that the remaining surveillances for the boric acid storage tank are only required if this tank is a required borated water source, per TR 13.1.3. This activity involves an administrative change to procedures (TRM) such that Required Actions for the TRM are removed and reliance for the RWST as a borated water source is placed on RWST Technical Specification 3.5.4.
B. This activity removes TR 13.3.2, "Seismic Monitoring Instrumentation," Condition C, which provides the Actions to be performed following a seismic event involving the actuation of one or more seismic monitoring instruments. This activity includes a change to procedures (TRM) such that Required Actions to restore the instrumentation to service following a seismic event are eliminated, and only the more general actions remain to be followed whenever seismic monitoring instrumentation is determined to be nonfunctional. Removal of Condition C and the associated Required Actions does not adversely affect the capability of this instrumentation to provide its monitoring functions as described in the UFSAR.
C. This activity removes TRM requirements of TR 13.4.1, [Reactor Coolant System]
"Chemistry" requirements. The changes associated with the deletion of this TRM specification are administrative changes to procedural requirements to be taken which the same as those actions required by the Corporate PWR Primary Water Chemistry Program and the associated implementing procedures that implement the EPRI PWR Primary Water Chemistry Guidelines, which provides the basis for the same chemistry limits in the TRM.
D. This activity removes the TRM requirements of TR 13.4.2, "Pressurizer," which provides the temperature transient limits placed on the pressurizer to prevent a non-ductile failure as a result of cyclic fatigue. This activity includes a change to procedures (TRM) by deleting the limitations and associated actions identified in the TRM. Because the functionality of the pressurizer continues to be assured by the identical pressure and temperature limits imposed by the facility's Operating procedures, removal of the TRM pressurizer requirements does not involve an adverse effect on the performance or method of control of a design function as described in the Updated FSAR.
E. This activity removes the TRM requirements of TR 13.5.1, "Emergency Core Cooling System (ECCS)," TRS 13.5.1.1, which is a high-level procedure that address containment cleanliness visual inspection requirements as described in the UFSAR.
Because the same containment cleanliness visual inspection procedures are also required by TS Surveillance Requirement (SR) 3.6.7.1 and the associated implementing procedures, which implement this surveillance requirement at the same frequency as that required by the current TRS, deletion of this requirement from the TRM does not involve an adverse effect on the performance or method of control of a design function as described in the Updated FSAR.
E4-1 to NL-24-0387 10 CFR 50.59 Summary Report F. This activity removes the TRM requirements of TR 13.7.1, "Steam Generator Pressure/Temperature Limitation," which provides the pressure and temperature limits placed on the steam generator to limit pressure-induced stresses on the steam generators to below the maximum allowable fracture toughness limits. This activity includes a change to procedures (TRM) by deleting the limitations and associated actions identified in the TRM. Because the functionality of the steam generators continues to be assured by the identical pressure and temperature limits imposed by the facility's Operating procedures, removal of the TRM steam generator pressure/
temperature limitation requirements from the TRM does not involve an adverse effect on the performance or method of control of a design function as described in the Updated FSAR.
G. This activity removes the TRM requirements of TR 13.7.2, "Snubbers" requirements. The changes associated with the TRM Specification are administrative changes to procedural actions to be taken which are duplicative of other procedural actions that implement the TS definition of Operability and TS 3.0.8 for supported systems.
The activities were not identified to more than minimally increase the frequency of occurrence or consequences of an accident previously evaluated in the Updated FSAR, more than minimally increase the likelihood of occurrence or consequences of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the Updated FSAR, create the possibility for an accident of a different type or for a malfunction of an SSC important to safety with a different result than any previously evaluated in the Updated FSAR, have any impact on the integrity of the fuel cladding, reactor coolant pressure boundary, or containment, or result in a departure from a method of evaluation described in the Updated FSAR used in establishing the design bases or in the safety analyses.
Activity: SNC1249229-06
Title:
Vogtle Electric Generating Plant - Units 1 and 2 ABB-NV and WLOP Departure from Nucleate Boiling Correlation Implementation 10 CFR 50.59 Evaluation Summary:
Vogtle Units 1 and 2 are implementing the ABB non-mixing vane (NV) and Westinghouse Low Pressure (WLOP) Departure from Nucleate Boiling (DNB) correlations in lieu of the W-3 correlation to obtain more accurate Departure from Nucleate Boiling Ratio (DNBR) predictions for the heated region below the first mixing vane grid, and for low pressure and low flow conditions.
Section 4.3.8 of NEI 96-07, Revision 1 states that the following change is not considered to be a departure from a method of evaluation described in the UFSAR:
Use of a new NRC-approved methodology (e.g., new or upgraded computer code) to reduce uncertainty, provide more precise results or other reason, provided such use is (a) based on sound engineering practice, (b) appropriate for the intended application and (c) within the limitations of the applicable SER. The basis for this determination should be documented in the licensee evaluation.
The following additional guidance is provided in section 4.3.8.2 of NEI 96-07, Revision 1:
The definition of "departure... " provides licensees with the flexibility to make changes under 10 CFR 50. 59 from one method of evaluation to another provided that the new method is approved by the NRC for the intended application. A new method is approved by the NRC for intended E4-2 to NL-24-0387 10 CFR 50.59 Summary Report application if it is approved for the type of analysis being conducted, and applicable terms, conditions and limitations for its use are satisfied.
Also, according to the guidance in section 4.2.1.3 of NEI 96-07, Revision 1, a proposed activity that only involves a change to an evaluation methodology does not require evaluation against the first seven criteria of 10 CFR 50.59(c)(2). Consistent with the above guidance, implementation of the W-3 alternate DNB correlations (ABB-NV and WLOP) is, (a) based on sound engineering practice, (b) appropriate for the intended application, and (c) within the limitations of the applicable NRC SER. This change does not represent a departure from a method of evaluation that requires NRC review and approval since the application of the correlations to satisfy the DNB design basis is consistent with the NRC Safety Evaluation Report for the methods being implemented. Additionally, Technical Specifications relevant to DNB limits are unaffected by the implementation of the ABB-NV and WLOP DNB correlations.
The specific DNB correlations used, and the respective DNB limits, are not specified in the Vogtle 1 and 2 Technical Specifications. Therefore, NRC approval is not required prior to implementation of the proposed activity.
Activity: TE 1123375
Title:
Adoption of Risk Informed Methodology for Vogtle Main Steam and Feedwater Break Exclusion Region Augmented lnservice Inspection Program 10 CFR 50.59 Evaluation Summary:
A change is made to UFSAR subsections 3.6.2.1.1 and 6.6.8 to update the augmented inspection program for main steam and feedwater break exclusion region piping to adopt the risk informed methodology for piping as outlined in EPRI Topical Report 1006937, "Extension of the EPRI Risk-Informed lnservice Inspection (RI-ISi) Methodology to Break Exclusion Region (BER) Programs, Rev. 0-A for the extent of examinations. This change allows scope of the augmented program to be defined as risk informed in lieu of the 100% volumetric inservice examination of all pipe welds in the BER. This activity implements a methodology approved by the NRC for this intended application and as such, is not a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.
The NRC approved this alternate method in Safety Evaluation Report Related to "Extension of the EPRI Risk-Informed lnservice Inspection (RI-ISi) Methodology to Break Exclusion Region (BER) Programs" (EPRI 1006937, Rev. 0-A) (ML021790518). The NRC SER concluded that the methodology was applicable to all NSSS designs and all terms and conditions stipulated in the SER are met by this proposed activity.
Vogtle's 1 &2 program has been evaluated and determined to meet the requirements within report EPRI 1006937, Rev. 0-A.
Activity: SNC1358162
Title:
Unit 2 LP Turbine Monoblock Rotors 10 CFR 50.59 Evaluation Summary:
This design change replaces the existing Unit 2 Low Pressure (LP) Turbine Rotors (21301 K4001 LPA, 21301 K4001 LPB, and 21301 K4001 LPC) with new LP Turbine Rotors with a monoblock design provided by General Electric (GE). The replacement rotors' monoblock E4-3 to NL-24-0387 10 CFR 50.59 Summary Report design eliminates the wheel bore crevices, which will make the rotor less susceptible to SCC and associated failures. This doesn't have an adverse effect on design functions or procedures as described in the Updated FSAR, and does not involve tests or experiments outside the bounds of the Updated FSAR. The difference between the current Unit 2 Low-Pressure Turbine Missile Probability Analysis and the proposed change lies in the calculation of the strike and damage probability (P2 x P3). A different methodology is used to calculate P2 x P3, however, this is not a departure from the method of evaluation described in the FSAR, because the different method has been approved by NRC as described in Appendix U, "Probability of Missile Generation in General Electric Nuclear Turbines", in NUREG-1048, "Safety Evaluation Report Related to the operation of Hope Creek Generating Station, Supplement 6, July 1986.
Therefore, the revised method of evaluation used in the proposed change has generic NRC acceptance and does not require NRC approval. Note that use of this generic approval is appropriate for the intended application and is within the limitations of the applicable SER.
Therefore, in accordance with NEI 96-07, Revision 1, this is not considered a departure from a method of evaluation described in the Updated FSAR.
Activity: DCP SNC965540
Title:
U1 A-Train MSIV Actuator Replacements 10 CFR 50.59 Evaluation Summary:
Design Change Package SNC965540 will replace the existing Unit 1 Train A main steam isolation valve (MSIV) actuators with system media actuated type actuators. System media actuators use the system media (steam) as the motive force for stroking the valve. As part of the actuator replacement, the existing electro-hydraulic system and nitrogen supply to the existing actuators will be removed, which will support the site's single point of vulnerability reduction initiative.
Two aspects of the proposed design activity represent a modification to an SSC such that a design function as described in the FSAR is adversely affected. These are (1) the inability to close the MSIV in less than the required 5 seconds upon receipt of an isolation signal, and (2) the introduction of a vent line that creates a potential flow path that requires isolation to prevent the release of iodine activity accumulating in the secondary side of the steam generator following an accident that is not considered in the dose analysis.
The replacement of the MSIV actuators cannot contribute to the initiation of any accident previously evaluated in the UFSAR. Therefore, the proposed activity has no impact on the frequency of occurrence of an accident previously evaluated in the Updated FSAR.
The proposed activity does not result in an increase in likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the Updated FSAR.
To address the inability to close the MSIV in less than 5 seconds upon receipt of an isolation signal, a new MSIV closure time of 7 seconds is established for accidents initiated above 325 psia, which is within the time assumed in the accident analyses. An analysis of the applicable events for the longest MSIV closure time expected for accidents initiated at 325 psia and below demonstrate that the current analysis of record is bounding and substantiates the acceptability of a longer valve closure initiated at lower system pressures.
To address the introduction of a vent line that creates a potential flow path that requires isolation to prevent the release of iodine activity, the vent line is equipped with a solenoid valve that automatically closes after MSIV closure, isolating the flow path. To address a possible single failure of the solenoid valve, a new operator action is established to close a manual isolation E4-4 to NL-24-0387 10 CFR 50.59 Summary Report valve on the vent line. This new operator action is considered to be a new manual operator action in support of a design function credited in the safety analyses. The new operator action is reflected in plant procedures and operator training programs, and can be completed in the time required considering the aggregate affects, such as workload or environmental conditions, expected to exist when the action is required. The evaluation of the change considers the ability to recover from credible errors in performance of the action and the expected time required to make such a recover, and also considers the effect of the change on plant systems. Therefore, the proposed activity does not result in an increase in likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the Updated FSAR.
For a main steam line break accident there is no impact to the offsite doses if the Unit 1 train A MSIV actuator vent pathway remains unisolated during the 20-hour accident duration. With regard to a steam generator tube rupture accident, the vent line for the SMA actuators contains a single safety-related valve solenoid valve that is open during the closure of the MSIV and automatically closes after a delay period following the closure of the MSIV. During the time when the solenoid valve is open, steam can be released through the new vent line to the environment. This results in a new single failure case where MV3 fails to automatically close after a delay period following an SGTR. A supplemental calculation evaluates this new case in which the postulated single failure is a failure of MV3 to automatically close rather than a failure of an ARV to close. The calculation determines the time when operations will need to close the manual valve on the SMA actuator vent line so that the dose consequences in the design basis case remain bounding. An isolation time of 172 minutes after event initiation results in an offsite dose that is within what was previously calculated in the ARV single failure case.
There are redundant MSIVs in each steamline. Any single failure (i.e., malfunction) of an MSIV to close upon receipt of an isolation signal will still result in blowdown of only one steam generator at most. As discussed above, isolation of the new vent line will be achieved such that the current licensing basis cases for offsite doses remain bounding. Therefore, the proposed activity does not result in more than a minimal increase in the consequences of a malfunction previously evaluated in the Updated FSAR.
The change in MSIV closure time has been evaluated and determined to be acceptable. There are no new accident types introduced by the change in actuator. Therefore, the proposed activity does not create the possibility for an accident of a different type than any previously evaluated in the Updated FSAR.
The new system media actuators will fail to the closed position, have redundant MSIVs in each main steam line, and are seismically designed and qualified for normal and post-accident environmental conditions. Therefore, the proposed activity does not create the possibility for a malfunction of an SSC important to safety with a different result than previously evaluated in the Updated FSAR.
Applicable safety analyses have been analyzed for variable valve closure times with acceptable results, and therefore the proposed activity has no impact on the integrity of the fuel cladding and integrity of containment. The proposed activity does not modify or affect the reactor coolant pressure boundary and therefore does not impact the integrity of the reactor coolant pressure boundary.
E4-5 to NL-24-0387 10 CFR 50.59 Summary Report Activity: DCP SNC965546
Title:
U1 B-Train MSIV Actuator Eliminations 10 CFR 50.59 Evaluation Summary:
The installation of new safety train B cables in stripped flexible conduit and PVC coated flexible conduit in the north and south main steam valve rooms (MSVRs) that route to the B train solenoid valves mounted to the MSIVs (configurations that are not currently included in FSAR Table 8.3.1-4) is conservatively considered to be an adverse effect to the train separation design function for A and B safety trains and screens in for further evaluation.
Evaluation concludes that it is acceptable for stripped flexible metal, flexible metal (PVC coated), and rigid steel conduits to be in contact if containing cables 2/0 AWG or less, 480V or less.
Two aspects of the proposed design activity represent a modification to an SSC such that a design function as described in the FSAR is adversely affected. These are (1) the removal of the design function for the outboard MS IVs and addition of new manual operator action to close the outboard MSIV 15 minutes after the attempt to isolate the ruptured steam generator inboard MSIV, and (2) the introduction of a flow path through the UPC vent solenoid valve for the release of iodine activity accumulating in the steam generator secondary following an accident that is not considered in the dose analysis.
The modifications to the inboard and outboard MSIV actuators cannot contribute to the initiation of any accident previously evaluated in the UFSAR. Therefore, the proposed activity has no impact on the frequency of occurrence of an accident previously evaluated in the Updated FSAR.
The offsite dose due to leakage through the vent line associated with the main steam line break (MSLB) and steam generator tube rupture (SGTR) analyses is not affected by this activity. The proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the Updated FSAR.
The proposed activity does not result in more than a minimal increase in the consequences of a malfunction previously evaluated in the Updated FSAR.
The proposed activity does not create the possibility for an accident of a different type than any previously evaluated in the Updated FSAR.
The proposed activity does not create the possibility for a malfunction of an SSC important to safety with a different result than previously evaluated in the Updated FSAR.
The proposed activity does not have an impact on the integrity of the fuel cladding, reactor coolant pressure boundary, or containment.
The proposed activity does not impact the previously established manual action to isolate the inboard MSIV SMA common vent line as the new Train B vent path is installed upstream of this manual isolation valve.
E4-6 to NL-24-0387 10 CFR 50.59 Summary Report Activity: DCP SNC1046657
Title:
U2 B-Train MSIV Actuator Eliminations 10 CFR 50.59 Evaluation Summary:
The installation of new safety train B cables in stripped flexible conduit and PVC coated flexible conduit in the north and south main steam valve rooms (MSVRs) that route to the B train solenoid valves mounted to the MSIVs (configurations that are not currently included in FSAR Table 8.3.1-4) is conservatively considered to be an adverse effect to the train separation design function for A and B safety trains and screens in for further evaluation.
Evaluation concludes that it is acceptable for stripped flexible metal, flexible metal (PVC coated), and rigid steel conduits to be in contact if containing cables 2/0 AWG or less, 480V or less.
Two aspects of the proposed design activity represent a modification to an SSC such that a design function as described in the FSAR is adversely affected. These are ( 1) the removal of the design function for the outboard MS IVs and addition of new manual operator action to close the outboard MSIV 15 minutes after the attempt to isolate the ruptured steam generator inboard MSIV, and (2) the introduction of a flow path through the UPC vent solenoid valve for the release of iodine activity accumulating in the steam generator secondary following an accident that is not considered in the dose analysis.
The modifications to the inboard and outboard MSIV actuators cannot contribute to the initiation of any accident previously evaluated in the UFSAR. Therefore, the proposed activity has no impact on the frequency of occurrence of an accident previously evaluated in the Updated FSAR.
The new operator action, which supports a design function credited in the safety analyses, is reflected in plant procedures and operator training programs, and can be completed in the time required considering the aggregate affects, such as workload or environmental conditions, expected to exist when the action is required. The operator action relies on non-safety equipment as allowed in the Standard Review Plan (SRP). The evaluation of the change considers the ability to recover from credible errors in performance of the action and the expected time required to make such a recover, and also considers the effect of the change on plant systems. The proposed activity does not result in an increase in likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the Updated FSAR.
The offsite dose due to leakage through the vent line associated with the main steam line break (MSLB) and steam generator tube rupture (SGTR) analyses is not affected by this activity. The proposed activity does not result in more than a minimal increase in the consequences of an accident previously evaluated in the Updated FSAR.
The proposed activity does not result in more than a minimal increase in the consequences of a malfunction previously evaluated in the Updated FSAR.
The proposed activity does not create the possibility for an accident of a different type than any previously evaluated in the Updated FSAR The proposed activity does not create the possibility for a malfunction of an SSC important to safety with a different result than previously evaluated in the Updated FSAR.
The proposed activity does not have an impact on the integrity of the fuel cladding, reactor coolant pressure boundary, or containment.
E4-7 to NL-24-0387 10 CFR 50.59 Summary Report The proposed activity does not impact the previously established manual action to isolate the inboard MSIV SMA common vent line as the new Train B vent path is installed upstream of this manual isolation valve.
Activity: X6CAJ.18
Title:
Change to Steam Generator Atmospheric Relief Valve Manual Isolation Timing 10 CFR 50.59 Evaluation Summary:
This change will extend operator action time to isolate a spurious open Steam Generator Power Operated Relief Valve (also referred to as an Atmospheric Relief Valve) in a Steam Generator Tube Rupture event from 16 minutes to 30 minutes. Extending operator action time to isolate a spurious open steam generator power operated relief valve during a steam generator tube rupture event does not increase the frequency of occurrence of a malfunction or the performance of a system, structure, or component. Furthermore, this change does not impact any of the design parameters for any of the 3 fission product barriers (fuel, reactor coolant system, containment). No new equipment, features, or changes whatsoever associated with the system, structure, or component are being made, so no new malfunction types are created.
This change does cause a slight increase in steam release to the environment in an event. This slight increase in steam release has been evaluated and shown to result in a less than minimal increase in dose consequences as defined by Nuclear Energy Institute 96-07. This is because the dose increase is less than 10% of the difference between the current analyzed value and the regulatory limit (10 CFR 50.67) and is less than the guideline value communicated in Regulatory Guide 1.183 Revision 0, Table 6, for the pre-accident and concurrent spike cases.
The analysis was performed in accordance with approved methodologies for plant Vogtle. Input changes were made to reflect the isolation time of the steam generator power operated relief valve block valve to increase from 16 minutes to 30 minutes.
In conclusion, this change is commensurate with that discussed in Nuclear Energy Institute 96-07 Section 4.3.2 Example 4 (a modification to an operator action) as not requiring Nuclear Regulatory Commission approval. The effect of the change has been evaluated to result in a less than minimal increase in dose consequences. Therefore, Nuclear Regulatory Commission prior approval is not required to approve and implement this change.
Activity: RER SNC1107421
Title:
Containment Service Level 1 Coatings Update for Carboguard 890N and Amerlock 400NT 10 CFR 50.59 Evaluation Summary:
Due to obsolescence issues with the existing containment service level 1 coatings, new containment service level 1 coatings have been tested and analyzed for general use in containment. These coatings are Carboguard 890N and Amerlock 400NT. This activity will update the coatings maintenance procedures, applicable safety analyses, and licensing basis to include the new qualified containment coatings. These containment coatings are modeled in the containment integrity evaluations described in Chapter 6 of the UFSAR as well as in the containment minimum backpressure evaluation that supports the containment backpressure input into the Loss of Coolant Accident (LOCA) Peak Cladding Temperature (PCT) evaluations described in Chapter 15 of the UFSAR. Since the material properties and coatings thickness differ from those inputs used in the existing analyses, sensitivities were performed to determine the impact to the evaluations using the new coatings.
E4-8 to NL-24-0387 10 CFR 50.59 Summary Report All of the evaluated cases for LOCA and MSLB Containment Integrity peak temperature and pressure resulted in adverse changes to the results reported in the licensing basis; however, all of the design basis limits for containment pressure and temperature continue to be met.
Additionally, the environmental qualification (EQ) temperature curves continue to bound the results of the updated analyses. The EQ pressure curve for MSLB is the bounding EQ curve and is what is used for EQ of components instead of containment. This curve continues to bound the updated pressure results from the containment integrity analyses for both LOCA and MSLB. While the presented LOCA EQ curve requires updating from 36.5 psig to 36. 7 psig, it is not used for the EQ pressure in containment and is present for reference. Since the design limits continue to bound the evaluation results, the changes are minimally adverse.
The containment minimum backpressure evaluation results for LOCA resulted in an adverse impact to the LOCA peak cladding temperature (PCT) of 2°F. The new resulting PCT is 2111.1 °F which continues to be bounded by the design limit of 2200°F and the increase of 2°F is less than what is considered significant by 10 CFR 50.46. Since the design limits continue to bound the evaluation results and the change is not significant, the changes are minimally adverse.
The proposed changes will not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR because coatings are not an initiator of any accidents.
The proposed changes will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR because the updated results for containment peak temperature and pressure due to LOCA and MSLB are still bounded by the maximum EQ temperature and pressure curves.
The proposed changes will not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR and will not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR because the UFSAR Chapter 15 dose analyses are not impacted as a result of these changes.
The proposed changes will not create the possibility for an accident of a different type than any previously evaluated in the UFSAR because coatings are not an initiator of any accidents.
The proposed changes will not create the possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the UFSAR because there are no new malfunctions of SSC different from those evaluated in the UFSAR that result from the changes.
The proposed changes will not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered since the LOCA and MSLB cases for containment integrity remain below the design limits for containment. Therefore, the design basis limits are not exceeded by these changes. The LOCA PCT evaluation resulted in a non-significant increase in PCT and the resulting PCT remains below the regulatory limit imposed by 10 CFR 50.46; therefore, the fuel limits are not exceeded.
The maximum EQ temperature and pressure limits are not exceeded due to the increase in the LOCA and MSLB limiting containment temperature and pressure evaluation results, therefore the equipment relied upon to protect the nuclear fuel and the reactor coolant system are not compromised as a result of these changes.
Therefore, NRC Prior approval for these changes is not required.
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