NL-22-0223, License Amendment to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting Condition for Operation (LCO) Value
| ML22181B145 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/30/2022 |
| From: | Gayheart C Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-22-0223 | |
| Download: ML22181B145 (22) | |
Text
Cheryl A. Gayheart Regulatory Affairs Director 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5316 cagayhea@southernco.com June 30, 2022 Docket Nos. 50-348 NL-22-0223 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Units 1 and 2 License Amendment to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance Limiting Condition for Operation (LCO) Value Pursuant to the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (CFR), Southern Nuclear Operating Company (SNC) hereby requests a license amendment to Joseph M. Farley Nuclear Plant (FNP) Unit 1 Renewed Facility Operating License NPF-2 and Unit 2 Renewed Facility Operating License NPF-8. The proposed amendment revises the as-found setpoint low side tolerance for the pressurizer safety valves (PSVs) described in Technical Specification (TS) 3.4.10, Pressurizer Safety Valves. The as-found setpoint low side tolerance is requested to be changed from -1%
(> 2460 psig) to -2.5% (> 2423 psig). This change in the as-found setpoint low side tolerance is needed to reduce an unnecessarily restrictive Limiting Condition for Operation (LCO). The proposed change does not affect the PSV setpoint of 2485 psig or the high side tolerance of +1%. The change does not affect the as-left setting of the PSVs at + 1%
of the lift setpoint. The change will not impact the reliability of the PSVs or adversely impact their ability to perform their safety function.
The enclosure provides a basis for the proposed change. Attachments 1 and 2 contain marked-up TS pages and revised TS pages, respectively. Attachment 3 contains the marked-up TS Bases for information only.
SNC requests approval of the proposed amendment within 12 months of completion of the NRCs acceptance review with an implementation period of 60 days.
In accordance with 10 CFR 50.91, a copy of this application, including attachments, is being provided to the designated Alabama Official.
This letter contains no regulatory commitments. If you have any questions, please contact Ryan Joyce at 205.992.6468.
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Joseph M. Farley Nuclear Plant - Units 1 and 2 License Amendment to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance LCO Value ENCLOSURE Basis for Proposed Changes to NL-22-0223 Basis for Proposed Changes E-1 1.0
SUMMARY
DESCRIPTION Southern Nuclear Operating Company (SNC) hereby requests a license amendment to Joseph M. Farley Nuclear Plant (FNP) Unit 1 Renewed Facility Operating License NPF-2 and Unit 2 Renewed Facility Operating License NPF-8. The proposed amendment revises the as-found setpoint low side tolerance for the pressurizer safety valves (PSVs) described in Technical Specification (TS) 3.4.10, Pressurizer Safety Valves. The as-found setpoint low side tolerance is requested to be changed from -1% (> 2460 psig) to -2.5% (> 2423 psig). This change in the as-found setpoint low side tolerance is needed to reduce an unnecessarily restrictive Limiting Condition for Operation (LCO). The proposed change does not affect the PSV setpoint of 2485 psig or the setpoint high side tolerance of +1%. The change does not affect the as-left setting of the PSVs at + 1% of the lift setpoint. The change will not impact the reliability of the PSVs or adversely impact their ability to perform their safety function.
2.0 DETAILED DESCRIPTION 2.1 Existing System Design The PSVs provide, in conjunction with the Reactor Protection System, overpressure protection for the Reactor Coolant System (RCS). The PSVs are totally enclosed pop type, spring loaded, self-actuated valves with backpressure compensation. The PSVs are designed to prevent the system pressure from exceeding the system Safety Limit (SL), 2735 psig, which is 110% of the design pressure.
Because the PSVs are totally enclosed and self-actuating, they are considered independent components. The relief capacity for each valve, 345,000 lb/hr, is based on postulated overpressure transient conditions resulting from a complete loss of steam flow to the turbine. This event results in the maximum surge rate into the pressurizer, which specifies the minimum relief capacity for the PSVs. The discharge flow from the PSVs is directed to the pressurizer relief tank. This discharge flow is indicated by an increase in temperature downstream of the PSVs or increase in the pressurizer relief tank temperature or level. The upper limit of overpressure protection is based upon the peak surge into the pressurizer of the reactor coolant produced as a result of turbine trip under full load, assuming no reactor trip. The self-actuated safety valves are sized on the basis of steam flow from the pressurizer to accommodate this surge at a setpoint of 2485 psig and a total accumulation of 3 percent. Note that no credit is taken for the relief capability provided by the power operated relief valves (PORVs) during this surge.
Overpressure protection is required in Modes 1, 2, 3, 4, and 5. However, in Mode 4, with one or more RCS cold leg temperatures less than or equal to the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the Pressure and Temperature Limits Report (PTLR), and Mode 5 and Mode 6 with the reactor vessel head on, overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP) System." The upper and lower pressure limits are based on the
+/- 1% tolerance requirement for lifting pressures above 1000 psig. The lift setting is for the ambient conditions associated with Modes 1, 2, and 3. This requires either that the valves be set hot or that a correlation between hot and cold settings be established.
The pressurizer safety valves are part of the primary success path and mitigate the effects of postulated accidents. Operability of the safety valves ensures that the RCS to NL-22-0223 Basis for Proposed Changes E-2 pressure will be limited to 110% of design pressure. The consequences of exceeding the American Society of Mechanical Engineers (ASME) pressure limit could include damage to RCS components, increased leakage, or a requirement to perform additional stress analyses prior to resumption of reactor operation.
2.2 Reason for the Proposed Change This change is proposed to reduce an unnecessarily restrictive LCO. FNP has PSVs manufactured by Crosby. There have been six instances since 2015 where the PSVs were tested and found outside the +/-1% tolerance limits. All out of tolerance test results were outside the low end of the setpoint tolerance (-1%). These as-found results have prompted the generation of licensee event reports (LERs) in accordance with 10 CFR 50.73(a)(2)(i)(B), Any operation or condition prohibited by the plants Technical Specifications... The as-found results did not exceed -3% of the pressure setpoint.
Based on the lift pressure meeting the Inservice Test (IST) program requirements, no IST scope expansion testing was needed. Since the as-found result was lower than the allowed value in the TS, the condition did not have an adverse impact on its over-pressurization function. This is within the safety analysis assumptions that are credited for PSVs, and the plant remained bounded by the accident analyses in the Final Safety Analysis Report (FSAR). Setpoint drift was determined to be the cause of the PSVs lifting low out of tolerance. The PSVs are performing within the design analysis assumptions. Therefore, generating a LER for a PSV that is performing satisfactorily within the design analysis assumptions becomes an unnecessary burden for both the licensee and the NRC.
2.3 Description of the Proposed Change TS 3.4.10 requires the three PSVs in each Unit to be Operable during Modes 1, 2, and 3, and Mode 4 with all RCS cold leg temperatures > the LTOP System applicability temperature specified in the PTLR. LCO 3.4.10 establishes the lift setting limits for the PSVs.
Current TS LCO 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift settings 2460 psig and 2510 psig.
Proposed TS LCO 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift settings 2423 psig and 2510 psig.
The current lift settings reflect a + 1% tolerance around the nominal lift setpoint of 2485 psig. The requested lift setting will reflect a +1%, -2.5% tolerance around the existing nominal lift setpoint of 2485 psig.
The current associated TS Surveillance Requirement (SR) 3.4.10.1 states: Verify each pressurizer safety valve is OPERABLE in accordance with the INSERVICE TESTING PROGRAM. Following testing, lift settings shall be within +/- 1%. This SR will remain unchanged.
to NL-22-0223 Basis for Proposed Changes E-3
3.0 TECHNICAL EVALUATION
The three PSVs are set to open at the RCS design pressure (2485 psig), and within the American Society of Mechanical Engineers (ASME) specified tolerance, to avoid exceeding the maximum design pressure SL, to maintain accident analyses assumptions, and to comply with ASME requirements. The current upper and lower pressure tolerance limits are based on the
+/-1% tolerance requirements (ASME, Boiler and Pressure Vessel Code,Section III) for lifting pressures above 1000 psig. The PSVs are part of the primary success path and mitigate the effects of postulated accidents. Operability of the PSVs ensures that the RCS pressure will be limited to 110% of design pressure.
A -2.5% tolerance on the nominal PSV set pressure of 2485 psig corresponds to a minimum lift pressure of 2423 psig.
Accident Analyses All accident and safety analyses in the FSAR that require PSV actuation assume operation of three PSVs to limit increases in RCS pressure. The overpressure protection analysis is also based on operation of three PSVs. Accidents that could result in overpressurization if not properly terminated include:
- a. Uncontrolled rod withdrawal from full power;
- b. Loss of reactor coolant flow;
- c. Loss of external electrical load;
- d. Loss of normal feedwater;
- e. Loss of all AC power to station auxiliaries; and
- f. Locked rotor.
Detailed analyses of the above transients are contained in the FSAR. PSV actuation is required in events c, d, and e (above) to limit the pressure increase. These analyses, as presented in the FSAR, focus on limiting overpressure of the RCS. Since the proposed changes to the TS do not include either a setpoint change or a change in the setpoint high side tolerance for the PSVs, their ability to respond to an increase in RCS pressure during the accidents described above is not changed. The PSVs will continue to provide their safety function as analyzed in the FSAR.
There are several safety analyses in the FSAR that model a negative (low) PSV tolerance.
These analyses could be affected by a change in the setpoint low side PSV tolerance and therefore are evaluated. They are shown below.
Accident FSAR Section Computer Code Loss of external electrical load - minimum departure from nucleate boiling ratio (DNBR) case and maximum main steam system pressure case 15.2.7 RETRAN Feedline break 15.4.2 RETRAN-02 Inadvertent operation of the emergency core cooling system 15.2.14 LOFTRAN Rod cluster control assembly withdrawal at power
- DNBR cases 15.2.2 LOFTRAN to NL-22-0223 Basis for Proposed Changes E-4 A -3% PSV tolerance is prescribed by both the LOFTRAN and RETRAN base deck calculation notes and is explicitly modeled in each of the safety analyses. A review of these analyses has confirmed that the non-LOCA safety analyses of record as described in the FSAR utilize a -3%
PSV tolerance. This bounds the requested -2.5% PSV tolerance. The results of the analyses show that the applicable acceptance criteria continue to be met for all events. The PSVs will continue to provide their safety function as analyzed in the FSAR.
A Large Break Loss-of-Coolant Accident (LBLOCA) is a depressurization transient due to the postulated break, so the RCS does not reach a pressure that would require the PSVs to open.
The PSVs are not modeled in the FNP Units 1 and 2 LBLOCA analysis of record. Therefore, the LBLOCA analysis of record is not impacted by a change in the PSV setpoint low side tolerance.
The Small Break Loss-of-Coolant Accident (SBLOCA) and Post-LOCA Long Term Cooling (LTC) design basis analyses are depressurization transients that do not model the PSVs. LOCA hydraulic forces design basis analyses are short-term transients that last only a few seconds and do not model the PSVs. With the presence of a postulated break modeled in each of these analyses, the RCS will not re-pressurize to a pressure that would require the PSVs to open.
Therefore, the SBLOCA, Post-LOCA LTC, and LOCA hydraulic forces analyses of record are not impacted by a change in the PSV setpoint low side tolerance.
Additionally, the following design basis analyses for the containment and radiological analysis are not impacted by the change in the PSV setpoint low side tolerance. The PSVs are either not modeled or not credited so any change in the setpoint tolerance would not affect these analyses.
Long-term LOCA mass and energy (M&E) releases for containment integrity Short-term LOCA M&E releases for subcompartment analyses Long-term steamline break (SLB) M&E releases inside containment for containment integrity Long-term SLB M&E releases outside containment for compartment response Short-term SLB M&E releases for subcompartment analyses Steam release for dose Steam generator tube rupture The SLB M&E releases inside and outside containment result in a depressurization of the primary side of the RCS. PSV inputs to LOFTRAN and RETRAN reflect typical positive tolerance values. However, the PSV inputs to the analyses of the SLB M&E releases inside or outside containment have no impact on the transient response since the RCS is depressurizing.
Additionally, PSVs are not modeled in the calculation of steam releases for dose. Therefore, the PSV setpoint low side tolerance change to -2.5% does not impact the dose analyses for FNP Units 1 and 2.
Margin Between PORVs and PSVs The pressurizer is equipped with two types of devices for pressure relief: pressurizer safety valves and PORVs. The PORVs are air operated valves that are controlled to open at a specific set pressure when the pressurizer pressure increases and close when the pressurizer pressure decreases. Each unit has two PORVs, each with a set pressure of 2335 psig. The functional design of the PORVs is based on maintaining pressure below the Pressurizer PressureHigh to NL-22-0223 Basis for Proposed Changes E-5 reactor trip setpoint following a step reduction of 50% of full load with steam dump. In addition, the PORVs minimize challenges to the PSVs.
The impact of the proposed change on the adequacy of margin between the opening of the pressurizer PORVs and the opening of the PSVs was evaluated. The pressurizer PORVs are designed to lift prior to the PSVs.
The calculated uncertainty associated with opening the pressurizer PORVs was determined to be +/-48.1 psi. Both pressurizer PORVs receive input from uncompensated pressure signals.
Conservatively applying the pressurizer PORV instrument uncertainty (i.e., +/-48.1 psi, -1.5 psi) to the nominal actuation setpoint of the pressurizer PORV setpoint (i.e., 2335 psig), results in an actuation band between 2285.4 psig and 2383.1 psig.
When compared to the current PSV lift setpoint and associated setpoint tolerance (i.e., 2485 psig +/-1% resulting in an actuation band of 2460 and 2510 psig), the margin between the highest possible actuation of a pressurizer PORV (i.e., 2383 psig) and the lowest possible actuation of a PSV (i.e., 2460 psig) is 77 psi.
When compared to the PSV lift setpoint and proposed low side tolerance (i.e., 2485 psig +1%,
-2.5%, resulting in an actuation band of 2423 and 2510 psig), the margin between the highest possible actuation of a pressurizer PORV (i.e., 2383 psig) and the lowest possible actuation of a PSV (i.e., 2423 psig) is 40 psi.
Although the proposed change results in a reduction in the margin between opening the pressurizer PORVs and opening the PSVs, the resulting margin is still sufficient to ensure that the PORVs will actuate prior to the PSVs.
Margin Between High Pressure Reactor Trip and PSVs It is appropriate to ensure that the Reactor Trip System (RTS) trips the reactor prior to PSV opening. As discussed in the bases for the Pressurizer Pressure - High RTS function, this helps minimize challenges to the PSVs since a reactor trip will mitigate most pressure transients before pressure increases to the PSV setting. A -2.5% tolerance on the nominal PSV set pressure of 2485 psig corresponds to a minimum lift pressure of 2423 psig. This lift pressure remains above the nominal high pressurizer pressure reactor trip setting of 2385 psig. The calculated uncertainty associated with the Pressurizer Pressure - High RTS function was determined to be +28.8 psi. The setpoint uncertainty was determined using a Farley-specific setpoint report utilizing a statistical methodology. The Westinghouse setpoint uncertainty calculation methodology was selected because it was used at many Westinghouse PWRs, conformed to industry practices such as ISA Standard S67.04, 1987, Setpoints for Nuclear Safety-Related Instrumentation Used in Nuclear Power Plants, and was previously approved by the NRC. The method identifies the distinct uncertainty components, determines the associated uncertainty allowance, and combines the uncertainties statistically to determine the total channel uncertainty. Then, for each protection function, the total channel uncertainty is compared to the total allowance (i.e., the difference between the SAL and the Technical Specifications trip setpoint) to demonstrate margin.
Applying the Pressurizer Pressure - High RTS instrument uncertainty (i.e., +28.8 psi) to the nominal actuation setpoint of the Pressurizer Pressure - High RTS setpoint (i.e., 2385 psig),
to NL-22-0223 Basis for Proposed Changes E-6 results in a high actuation limit of 2414 psig. This demonstrates that the Pressurizer Pressure -
High RTS trip will actuate prior to the PSV opening.
PSV opening prior to a reactor trip is judged to be unlikely even with a -2.5% PSV tolerance since this would require a large pressure transient to occur with the maximum pressurizer pressure reactor trip uncertainties and maximum negative side PSV tolerance. Such a transient would generally be mitigated by the control systems (e.g., pressurizer PORVs) or actuate another RTS function before PSVs would open. An as found -2.5% PSV setpoint tolerance does not prevent the RTS functions, including the high pressurizer pressure function, from providing their design basis protection.
4.0 REGULATORY EVALUATION
4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.
Criterion 15 - Reactor Coolant System Design The reactor coolant system and associated auxiliary, control, and protection systems are designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences.
Conformance The reactor coolant system and associated auxiliary, control, and protection systems are designed to ensure the integrity of the reactor coolant pressure boundary with adequate margins during normal operation and during Condition I and Condition II transients. The system boundary accommodates loads due to the safe shutdown earthquake during normal operation including normal operational transients (Condition II) within upset condition code stress limits. The system boundary accommodates loads due to the safe shutdown earthquake combined with loads due to piping failures such as circumferential pipe ruptures of reactor coolant pipes at junctures with equipment nozzles and connecting pipes at junctures to reactor coolant piping (Condition IV), without propagation of failure to remaining reactor coolant system loops, steam power conversion system, or other piping or equipment needed for emergency cooling. The components of the reactor coolant system and associated fluid systems are designed in accordance with appropriate ASME and ANSI codes. The protection system is designed in accordance with IEEE-Std 279. Overpressure protection is provided by automatic controls, pressure relief valves, and code safety valves. The selected design margins include operating transient changes due to thermal lag, coolant transport times, pressure drops, system relief valve characteristics, and instrumentation and control response characteristics.
Criterion 31 - Fracture Prevention of Reactor Coolant Pressure Boundary The reactor coolant pressure boundary is designed with sufficient margin to assure that when stressed under operating, maintenance, testing, and postulated accident conditions the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is minimized. The design reflects consideration of service temperatures and other conditions of the boundary material under operating, to NL-22-0223 Basis for Proposed Changes E-7 maintenance, testing, and postulated accident conditions and the uncertainties in determining material properties, the effects of irradiation on material properties, residual, steady-state and transient stresses, and size of flaws.
Conformance FNP is designed to conform to the intent of Criterion 31. The reactor coolant pressure boundary is designed so that, for all transients, normal, upset, and faulted, the reactor coolant pressure boundary behaves in a nonbrittle manner. The Units were designed for 650° and 2500 psia. The normal service temperature and pressure are 550°F and 2250 psia. The reactor pressure vessels were designed in accordance with Section III of the ASME Boiler and Pressure Vessel Code, which considers cyclic loading, defect characterization, minimum material toughness, and maximum allowable stresses.
Material selection and testing were in accordance with the Summer 1970 Addenda of the ASME Code. Normal operating limits are calculated based on the toughness properties of the ferritic components of the pressure boundary in accordance with nonmandatory Appendix G,Section XI of the ASME Boiler and Pressure Vessel Code.
Postulated accident conditions are analyzed using the concepts of fracture mechanics technology with which one quantitatively considers defect size and geometry, applied stresses (residual, thermal, and membrane), and the material properties. The effects of irradiation are considered in the generation of the heatup and cooldown limit curves, analysis of post operational tests, and for analyses of all other transients and postulated accidents. A reactor vessel materials radiation surveillance program is performed in accordance with ASTM E-185.
These General Design Criteria (GDC) continue to be met because the design bases accidents that experience an overpressure condition in the RCS were evaluated and found to limit the RCS pressure to less than 110% of the design pressure.
Additionally, Standard Review Plan 5.2.2, Overpressure Protection, requires RCS safety valves to be designed with sufficient capacity to limit RCS pressure to less than 110% of RCS pressure boundary design pressure during the most severe abnormal operational transient. The capacity of the PSVs is not changed and will continue to provide sufficient overpressure protection.
4.2 Precedent In 2004, the NRC approved a similar request from the Braidwood and Byron Nuclear Stations (References 1 and 2). This request also modified the high and low side PSV setpoint tolerances as well as changing one of the PSV setpoints.
This precedent is similar to the proposed change. However, this proposed change is a narrower scope than the precedent. This change requests a change only to the PSV setpoint low side tolerance. No change is requested to the PSV setpoint high side tolerance and no change is requested for the PSV setpoint. This results in fewer analyses potentially being affected and less discussion of the change.
4.3 No Significant Hazard Consideration Determination Analysis Southern Nuclear Operating Company (SNC) has evaluated the proposed changes to the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has to NL-22-0223 Basis for Proposed Changes E-8 determined that the proposed changes do not involve a significant hazards consideration.
The proposed amendment revises the as-found pressurizer safety valve (PSV) setpoint low side tolerance described in TS 3.4.10, Pressurizer Safety Valves. The as-found PSV setpoint low side tolerance is requested to be changed from -1% (> 2460 psig) to
-2.5% (> 2423 psig). This change in the as-found PSV setpoint low side tolerance is needed to reduce an unnecessarily restrictive Limiting Condition for Operation (LCO).
The proposed change does not affect the PSV setpoint of 2485 psig or the PSV setpoint high side tolerance of +1%. The change does not affect the as-left setting of the PSVs at +1% of the lift setpoint. The change will not impact the reliability of the PSVs or adversely impact their ability to perform their safety function.
As required by 10 CFR 50.91(a), the SNC analysis of the issue of no significant hazards consideration is presented below:
- 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change, modifying the as-found PSV setpoint low side tolerance in the LCO for the PSVs, does not change the design function or operation of the PSVs and it does not change the way the PSVs are maintained, tested, or inspected. The PSVs are not accident initiators, they operate in response to the pressurization of the Reactor Coolant System (RCS). They limit the pressure of the RCS to less than the allowable American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section III, during an accident or transient. Analyses were performed of peak pressure events, which are evaluated against the RCS limit. Action of the PSVs is required to mitigate the consequences of these events. The change in the PSV setpoint low side tolerance was explicitly considered in the analysis of these events. The RCS pressure remained below the required limits with this change considered. Therefore, this change does not impact the ability of the PSVs to perform their safety function during evaluated accidents.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 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 change, modifying the as-found PSV setpoint low side tolerance in the LCO for the PSVs, does not change the PSVs design function to maintain RCS pressure below the RCS pressure Safety Limit of 2735 psig during design basis accidents nor does it affect the PSVs ability to perform this design function. The proposed change does not require any modification to the plant or change equipment operation or testing. It also does not create any credible new failure mechanisms, malfunctions, or accident initiators that would cause an accident not previously considered.
to NL-22-0223 Basis for Proposed Changes E-9 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 change, modifying the as-found PSV setpoint low side tolerance as described in the LCO for the PSVs, does not involve a significant reduction in the margin of safety in maintaining RCS pressure below the Safety Limit of 2735 psig during design basis accidents. The PSVs maintain an adequate safety margin assuming the change in the PSV setpoint low side tolerance. The response of the PSVs would maintain an adequate safety margin to the RCS pressure Safety Limit of 2735 psig.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
4.4 Conclusion Based on the considerations discussed above, SNC concludes: (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 Commissions 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.
5.0 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 Part 20, and 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 effluent that may be released offsite, or (iii) a significant increase in 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.
6.0 REFERENCES
- 1. Letter from Exelon Generation to NRC, dated June 27, 2003, Request for a license Amendment to Revise the Pressurizer Safety Valve Lift Settings (ML031810344)
- 2. Letter from NRC to Exelon Generation, dated August 26, 2004, Issuance of Amendments re: Pressurizer Safety Valve Setpoints (ML042250516)
Joseph M. Farley Nuclear Plant - Units 1 and 2 License Amendment to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance LCO Value Proposed Technical Specification Changes (Marked-up Page)
Pressurizer Safety Valves 3.4.10 Farley Units 1 and 2 3.4.10-1 Amendment No. 225 (Unit 1)
Amendment No. 222 (Unit 2) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LCO 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift settings 242360 psig and 2510 psig.
APPLICABILITY:
MODES 1, 2, and 3, MODE 4 with all RCS cold leg temperatures > the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the PTLR.
NOTE----------------------------------------------
The lift settings are not required to be within the LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions. This exception is allowed for 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
NOTE------------
RICT entry is not permitted for this loss of function Condition when a pressurizer safety valve is intentionally made inoperable.
One pressurizer safety valve inoperable.
A.1 Restore valve to OPERABLE status.
15 minutes OR In accordance with the Risk Informed Completion Time Program
Joseph M. Farley Nuclear Plant - Units 1 and 2 License Amendment to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance LCO Value Revised Technical Specification Page
Pressurizer Safety Valves 3.4.10 Farley Units 1 and 2 3.4.10-1 Amendment No. (Unit 1)
Amendment No. (Unit 2) 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.10 Pressurizer Safety Valves LCO 3.4.10 Three pressurizer safety valves shall be OPERABLE with lift settings 2423 psig and 2510 psig.
APPLICABILITY:
MODES 1, 2, and 3, MODE 4 with all RCS cold leg temperatures > the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the PTLR.
NOTE----------------------------------------------
The lift settings are not required to be within the LCO limits during MODES 3 and 4 for the purpose of setting the pressurizer safety valves under ambient (hot) conditions. This exception is allowed for 54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> following entry into MODE 3 provided a preliminary cold setting was made prior to heatup.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
NOTE------------
RICT entry is not permitted for this loss of function Condition when a pressurizer safety valve is intentionally made inoperable.
One pressurizer safety valve inoperable.
A.1 Restore valve to OPERABLE status.
15 minutes OR In accordance with the Risk Informed Completion Time Program
Joseph M. Farley Nuclear Plant - Units 1 and 2 License Amendment to Revise Technical Specification 3.4.10, Pressurizer Safety Valves to Decrease Low Side Setpoint Tolerance LCO Value Proposed Technical Specification Bases Changes (Mark-up) for Information Only
Pressurizer Safety Valves B 3.4.10 Farley Units 1 and 2 B 3.4.10-1 Revision 63 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.10 Pressurizer Safety Valves BASES BACKGROUND The pressurizer safety valves provide, in conjunction with the Reactor Protection System, overpressure protection for the RCS. The pressurizer safety valves are totally enclosed pop type, spring loaded, self actuated valves with backpressure compensation. The safety valves are designed to prevent the system pressure from exceeding the system Safety Limit (SL), 2735 psig, which is 110% of the design pressure.
Because the safety valves are totally enclosed and self actuating, they are considered independent components. The relief capacity for each valve, 345,000 lb/hr, is based on postulated overpressure transient conditions resulting from a complete loss of steam flow to the turbine.
This event results in the maximum surge rate into the pressurizer, which specifies the minimum relief capacity for the safety valves. The discharge flow from the pressurizer safety valves is directed to the pressurizer relief tank. This discharge flow is indicated by an increase in temperature downstream of the pressurizer safety valves or increase in the pressurizer relief tank temperature or level.
Overpressure protection is required in MODES 1, 2, 3, 4, and 5; however, in MODE 4, with one or more RCS cold leg temperatures the Low Temperature Overpressure Protection (LTOP) System applicability temperature specified in the PTLR, and MODE 5 and MODE 6 with the reactor vessel head on, overpressure protection is provided by operating procedures and by meeting the requirements of LCO 3.4.12, "Low Temperature Overpressure Protection (LTOP)
System."
The upper and lower pressure limits areis based on the +/- 1%
tolerance requirement (Ref. 1) for lifting pressures above 1000 psig.
The lower limit (- 2.5%) is bounded by the +/- 3% of valve nameplate set-pressure from the ASME OM Code (Ref. 4). The lift setting is for the ambient conditions associated with MODES 1, 2, and 3. This requires either that the valves be set hot or that a correlation between hot and cold settings be established.
The pressurizer safety valves are part of the primary success path and mitigate the effects of postulated accidents. OPERABILITY of the safety valves ensures that the RCS pressure will be limited to 110% of design pressure. The consequences of exceeding the (continued)
Pressurizer Safety Valves B 3.4.10 Farley Units 1 and 2 B 3.4.10-2 Revision 0 BASES BACKGROUND American Society of Mechanical Engineers (ASME) pressure limit (continued)
(Ref. 1) could include damage to RCS components, increased leakage, or a requirement to perform additional stress analyses prior to resumption of reactor operation.
APPLICABLE All accident and safety analyses in the FSAR (Ref. 2) that require SAFETY ANALYSES safety valve actuation assume operation of three pressurizer safety valves to limit increases in RCS pressure. The overpressure protection analysis (Ref. 3) is also based on operation of three safety valves. Accidents that could result in overpressurization if not properly terminated include:
- a. Uncontrolled rod withdrawal from full power;
- b. Loss of reactor coolant flow;
- c. Loss of external electrical load;
- d. Loss of normal feedwater;
- e. Loss of all AC power to station auxiliaries; and
- f. Locked rotor.
Detailed analyses of the above transients are contained in Reference 2. Safety valve actuation is required in events c, d, and e (above) to limit the pressure increase. Compliance with this LCO is consistent with the design bases and accident analyses assumptions.
Pressurizer safety valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
LCO The three pressurizer safety valves are set to open at the RCS design pressure (2500 psia), and within the ASME specified tolerance, to avoid exceeding the maximum design pressure SL, to maintain accident analyses assumptions, and to comply with ASME requirements. The upper and lower pressure tolerance limits areis based on the +/- 1% tolerance requirements (Ref. 1) for lifting pressures above 1000 psig. The lower limit (- 2.5%) is bounded by the +/- 3% of valve nameplate set-pressure from the ASME OM Code (Ref. 4). The limit protected by this Specification (continued)
Pressurizer Safety Valves B 3.4.10 Farley Units 1 and 2 B 3.4.10-3 Revision 72 BASES LCO is the reactor coolant pressure boundary (RCPB) SL of 110% of (continued) design pressure. Inoperability of one or more valves could result in exceeding the SL if a transient were to occur. The consequences of exceeding the ASME pressure limit could include damage to one or more RCS components, increased leakage, or additional stress analysis being required prior to resumption of reactor operation.
APPLICABILITY In MODES 1, 2, and 3, and portions of MODE 4 when all RCS cold leg temperatures are > the LTOP System applicability temperature specified in the PTLR, OPERABILITY of three valves is required because the combined capacity is required to keep reactor coolant pressure below 110% of its design value during certain accidents.
MODE 3 and portions of MODE 4 are conservatively included, although the listed accidents may not require the safety valves for protection.
The LCO is not applicable in MODE 4 when one or more RCS cold leg temperatures are the LTOP System applicability temperature specified in the PTLR or in MODE 5 because LTOP is provided.
Overpressure protection is not required in MODE 6 with reactor vessel head detensioned.
Normally demonstration of the safety valves lift settings will occur during shutdown and will be performed in accordance with the provisions of the ASME Code for Operation and Maintenance of Nuclear Power Plants.
The Note allows entry into MODES 3 and 4 with the lift settings outside the LCO limits. This permits testing and examination of the safety valves at high pressure and temperature near their normal operating range, but only after the valves have had a preliminary cold setting. The cold setting gives assurance that the valves are OPERABLE near their design condition. Only one valve at a time will be removed from service for testing. The 54 hour6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> exception is based on 18 hour2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> outage time for each of the three valves. The 18 hour2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> period is derived from operating experience that hot testing can be performed in this timeframe.
Pressurizer Safety Valves B 3.4.10 Farley Units 1 and 2 B 3.4.10-4 Revision 96 BASES ACTIONS A.1 With one pressurizer safety valve inoperable, restoration must take place within 15 minutes. The Completion Time of 15 minutes reflects the importance of maintaining the RCS Overpressure Protection System. An inoperable safety valve coincident with an RCS overpressure event could challenge the integrity of the pressure boundary. Alternatively, a Completion Time can be determined using the Risk Informed Completion Time Program (Ref. 5).
Condition A is modified by a Note. The Note states RICT entry is not permitted for this loss of function Condition when a pressurizer safety valve is intentionally made inoperable. The RICT program cannot be used for voluntary removal of systems or components which would result in a loss of safety function. The RICT program is only applicable for a situation where a pressurizer safety valve is found inoperable. The Risk Informed Completion Time for this Condition cannot exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If a pressurizer safety valve is intentionally made inoperable, it must be restored to OPERABLE status within 15 minutes.
B.1 and B.2 If the Required Action of A.1 cannot be met within the required Completion Time or if two or more pressurizer safety valves are inoperable, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least 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 4 with any RCS cold leg temperatures the LTOP System applicability temperature specified in the PTLR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 challenging plant systems. With any RCS cold leg temperatures at or below the LTOP System applicability temperature specified in the PTLR, overpressure protection is provided by the LTOP System. The change from MODE 1, 2, or 3 to MODE 4 reduces the RCS energy (core power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by three pressurizer safety valves.
Pressurizer Safety Valves B 3.4.10 Farley Units 1 and 2 B 3.4.10-5 Revision 96 BASES SURVEILLANCE SR 3.4.10.1 REQUIREMENTS Pressurizer safety valves are to be tested in accordance with the requirements of the ASME OM Code (Ref. 4), which provides the activities and Frequencies necessary to satisfy the SRs. No additional requirements are specified.
The pressurizer safety valve setpoint is +/- 1%+ 1%, - 2.5% for OPERABILITY; however, the valves are reset to +/- 1% during the Surveillance to allow for drift.
REFERENCES
- 1. ASME, Boiler and Pressure Vessel Code,Section III.
- 2. FSAR, Chapter 5.2, 5.5, 15.2, 15.3 and 15.4.
- 3. WCAP-7769, Rev. 1, June 1972.
Joseph M. Farley Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos. 225 and 222 Regarding Implementation of NEI 06-09, Risk-Informed Technical Specifications Initiative 4B, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0-A (EPID L-2018-LLA-0210), dated August 23, 2019 (ADAMS Accession No. ML19175A243).