L-MT-12-107, Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus License Amendment Requests: Supplement to Address Secy 11-0014 Use of Containment Accident Pressure, Sections 6.6.4 and 6.6.7
| ML123380435 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 11/30/2012 |
| From: | Schimmel M Xcel Energy, Northern States Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-12-107, SECY 11-0014, TAC MD9990, TAC ME3145 | |
| Download: ML123380435 (15) | |
Text
XcelEnergy° Monticello Nuclear Generating Plant 2807 W County Rd 75 Monticello, MN 55362 November 30, 2012 L-MT-12-107 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, rDC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed License No. DPR-22 Monticello Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus License Amendment Requests: Supplement to Address SECY 11-0014 Use of Containment Accident Pressure, Sections 6.6.4 and 6.6.7 (TAC Nos. MD9990 and ME3145)
References:
- 1) Letter from T J O'Connor (NSPM) to Document Control Desk (NRC),
"License Amendment Request: Extended Power Uprate (TAC MD9990)," L-MT-08-052, dated November 5, 2008. (ADAMS Accession No. ML083230111)
- 2) Letter from T J O'Connor (NSPM) to Document Control Desk (NRC),
"License Amendment Request: Maximum Extended Load Line Limit Analysis Plus," TAC ME3145, L-MT-10-003, dated January 21, 2010.
(ADAMS Accession No. MIL100280558)
- 3) Letter from J G Giitter (NRC) to T J O'Connor (NSPM), "
Subject:
Monticello Nuclear Generating Plant - Linking of the Proposed Extended Power Uprate Amendment and the MELLLA+ Amendment (TAC NOS. MD9990 AND ME2449)," dated November 23, 2009.
(ADAMS Accession No. ML093160816)
- 4) Letter from M A Schimmel (NSPM) to Document Control Desk (NRC),
"Monticello Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus License Amendment Requests: Supplement to Address SECY 11-0014 Use of Containment Accident Pressure (TAC Nos. MD9990 and ME3145)," L-MT-12-082, dated September 28, 2012. (ADAMS Accession No. ML12276A057)
A-00(
Document Control Desk Page 2
- 5) Letter from T J O'Connor (NSPM) to Document Control Desk (NRC),
"Monticello Extended Power Uprate: Response to NRC Reactor Systems Branch and Nuclear Performance & Code Review Branch Request for Additional Information (RAI) dated January 16, 2009 (TAC No. MD9990)," L-MT-09-017, dated March 19, 2009. (ADAMS Accession No. ML090790388)
- 6) Letter from T J O'Connor (NSPM) to Document Control Desk (NRC),
"Monticello Extended Power Uprate: Updates to Docketed Information (TAC MD9990)," L-MT-10-072, dated December 21, 2010. (ADAMS Accession No. ML103570026)
- 7) Letter from M A Schimmel (NSPM) to Document Control Desk (NRC),
"Monticello Extended Power Uprate: Supplement to Revise Technical Specification Setpoint for the Automatic Depressurization System Bypass Timer (TAC MD9990)," L-MT-12-091, dated October 30, 2012.
(ADAMS Accession No. ML12307A036)
Pursuant to 10 CFR 50.90, the Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, requested in Reference 1 an amendment to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications (TS) to increase the maximum authorized power level from 1775 megawatts thermal (MWt) to 2004 MWt. This is also known as an extended power uprate (EPU).
Also pursuant to 10 CFR 50.90, NSPM requested in Reference 2 an amendment to the MNGP Renewed OL and TS to allow operation within the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) operating domain.
The Nuclear Regulatory Commission (NRC) permitted these two license amendment requests to be linked in Reference 3.
In Reference 4, NSPM submitted a detailed assessment of the MNGP Residual Heat Removal (RHR) and Core Spray (CS) pumps' ability to meet the uncertainties and margins described in SECY 11-0014. NSPM has taken credit for containment accident pressure (CAP) for the MNGP Emergency Core Cooling System (ECCS) analyses.
Credit for CAP is part of the licensing basis for the MNGP core cooling analyses as presented in both the EPU and MELLLA+ License Amendment Requests (LARs)
(References 1 and 2) discussed above.
However, Reference 4 identified that SECY 11-0014, Enclosure 1, section 6.6.4, Assurance that Containment Integrity is not Compromised, and 6.6.7, Assurance of no Pre-existing leak, were not completely addressed and this separate letter would be provided to complete the response to SECY 11-0014.
Document Control Desk Page 3 to this letter provides the balance of the response to SECY 11-0014, , sections 6.6.4 and 6.6.7. The analysis described in Enclosure 1 of this letter assumes the operating conditions of EPU and MELLLA+ including the most limiting events. Finally, the analysis concludes that the MNGP ECCS pumps can reliably perform their required design functions to mitigate the consequences of an Appendix R event for the required mission time while meeting the definition for Net Positive Suction Head required (NPSHr) provided in SECY-11-0014, Enclosure 1. provides a correction to the Appendix R event, time to cold shutdown analysis provided in an EPU RAI response (Reference 5, Enclosure 1, NRC RAI 2.8.4.4-2). Reference 5 erroneously reported that the time to cold shutdown was 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. NSPM recalculated the Appendix R event time to cold shutdown, and determined that the correct value should be 44.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. This is still well within the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to cold shutdown permitted by 10CFR50, Appendix R.
The supplemental information provided herein does not change the conclusions of the No Significant Hazards Consideration and the Environmental Consideration evaluations provided in Reference 1 as revised by References 6 and 7 for the Extended Power Uprate LAR. Further, the supplemental information provided herein does not change the conclusions of the No Significant Hazards Consideration and the Environmental Consideration evaluations provided in Reference 2 for the MELLLA+ LAR.
In accordance with 10 CFR 50.91(b), a copy of this application supplement, without enclosures is being provided to the designated Minnesota Official.
Summary of Commitments This letter makes no new commitments. This letter completes the actions required by a commitment associated with the MELLLA+ LAR. In letters L-MT-09-100 and L-MT 003, NSPM committed to resolve the CAP issue in the same manner as the issue is resolved for the delayed EPU amendment. The analysis included herein provides analysis of CAP assuming EPU and MELLLA+ conditions. Therefore, this letter, along with Reference 4, fully satisfies this commitment and the commitment may be closed.
Document Control Desk Page 4 I declare under penalty of perjury that the foregoing is true and correct.
Executed on: November 3- 0 , 2012 Mark A. Schimmel Site Vice-President Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosures (2) cc: Administrator, Region Ill, USNRC (w/o enclosures)
Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC (w/o enclosures)
Minnesota Department of Commerce (w/o enclosures)
L-MT-12-107 ENCLOSURE I EVALUATION OF THE RESIDUAL HEAT REMOVAL AND CORE SPRAY PUMPS FOR THE MONTICELLO NUCLEAR GENERATING PLANT WHEN APPLYING THE GUIDANCE OF SECY 11-0014, SECTIONS 6.6.4 AND 6.6.7
1.0 INTRODUCTION
Northern States Power - Minnesota (NSPM) has taken credit for containment accident pressure (CAP) for the Monticello Nuclear Generating Plant (MNGP) Emergency Core Cooling System (ECCS) analyses. NSPM also took credit for CAP in both the Extended Power Uprate (EPU) and Maximum Extended Load Line Limit Analysis Plus (MELLLA+)
License Amendment Requests (LARs) (References E1-1 and E1-2).
In Reference E1-3, NSPM submitted a detailed assessment of the MNGP Residual Heat Removal (RHR) and Core Spray (CS) pumps' ability to meet the uncertainties and Net Positive Suction Head (NPSH) margins described in SECY 11-0014. However, the letter identified that SECY 11-0014 (Reference 5), Enclosure 1, section 6.6.4, Assurance that Containment Integrity is not Compromised, and 6.6.7, Assurance of no Pre-existing leak, were not completely addressed and this evaluation would be provided to complete the response to SECY 11-0014.
SECY 11-0014 sections 6.6.4 and 6.6.7 are concerned that circuit issues associated with an Appendix R fire could result in a loss of containment integrity from containment venting, and an eventual loss of CAP.
This enclosure along with Reference E1-3 concludes that the MNGP ECCS pumps can reliably perform their required design functions to mitigate the consequences of accidents and events for the required mission time while using appropriate uncertainties defined for NPSH required effective (NPSHreff) for the Design Bases Accident - Loss of Coolant Accident (DBA-LOCA) and NPSH required (NPSHr) for other events. The ECCS pumps meet the guidance contained in SECY-1 1-0014, Enclosure 1.
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L-MT-12-107
2.0 BACKGROUND
The MNGP Fire Protection Program has been established to minimize the likelihood of fires, ensure the capability to shutdown the reactor and maintain it in a safe shutdown condition, and minimize radioactive releases to the environment in the event of a fire.
The Fire Protection Program implements the philosophy of defense-in-depth protection against the hazards of fire and its associated effects on equipment important to safety by:
o Preventing fires from starting o Rapidly detecting, controlling and promptly extinguishing fires that do occur o Providing protection for structures, systems, and components important to safety so that a fire not promptly extinguished by fire suppression activities will not prevent safe shutdown of the plant.
Appendix J of the MNGP Updated Safety Analysis Report (USAR) describes the fire protection related organizational responsibilities, administrative and technical controls, fire suppression and detection systems, fire hazards analyses, and the post-fire safe shutdown methods, which comprise the Fire Protection Program. Included within the MNGP Fire Protection Program is the design and licensing bases for the 10CFR50, Appendix R program.
10CFR50.48(b), which became effective on February 17, 1981, required all nuclear power plants licensed for operation prior to January 1, 1979 to meet the requirements of Sections Ill.G, Ill.J, and 111.0 of Appendix R to 10CFR50.
For Appendix R events, the NRC approved use of CAP in license amendment 139 dated June 2, 2004 (Reference E1-5). In the amendment the NRC recognized that use of CAP is required for adequate available NPSH for the low-pressure emergency core cooling system pumps following a LOCA, reactor vessel isolation, and Appendix R fire.
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L-MT-12-107 3.0 DETAILED ANALYSES SECY 11-0014 (Reference E1-4), Enclosure 1, Section 6.6 provides a breakdown of the guidance that the NRC has provided with respect to determining uncertainty and margins for relying on CAP in accident and transient analysis. The detailed analyses provided below demonstrate the margin available in the NPSH analysis for the RHR and CS pumps during Appendix R events that require mitigation actions from the alternate shutdown panel.
In the discussion below, the NRC guidance from SECY 11-0014, Enclosure 1, Section 6.6 is provided in italics and then followed by NSPM's response to the NRC guidance.
6.6.4 Assurance that Containment Integrity is not Compromised It should be demonstratedconservatively that, for the plant examined, loss of containmentintegrity from containment venting, circuitissues associatedwith an Appendix R fire, or othercauses cannot occur or that they would occur only after use of containment accidentpressure is no longer needed.
NSPM Response to 6.6.4 The NRC guidance asks for an evaluation concerning loss of containment integrity resulting from containment venting and circuit issues associate with an Appendix R fire or other causes. The response below focuses on circuit issues associated with an Appendix R fire. Containment integrity losses resulting from other causes are addressed in Reference E1-3, sections 6.6.4 and 6.6.7.
Circuit Issues Associated with Appendix R:
Appendix R Containment Accident Pressure (CAP) Bounding Fire Scenario:
The bounding CAP scenario for Appendix R is a fire occurring in a 10CFR50 Appendix R, III.G.3 - alternative shutdown area (cable spreading room or control room). For alternative shutdown areas, MNGP credits the use of the alternative shutdown system (ASDS) panel and the BWR alternate shutdown cooling method (ASCM).
For ASCM, the reactor is depressurized using the Auto Depressurization System (ADS). When the reactor is depressurized to below the shutoff head of the pump, a CS pump or RHR pump (in the LPCI mode) is then used to flood the vessel.
Safety/relief valves are then used to discharge the heated reactor water to the suppression pool. The head developed by the RHR pump is used to hold the safety/relief valves open. Water discharged to the torus is cooled with the suppression pool cooling mode of the RHR system. This water can then be re-injected into the reactor by aligning the pump suction to the suppression pool.
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L-MT-12-107 Consideration of Multiple Spurious Operations (MSOs) in the Alternative Shutdown Fire Scenario:
MNGP safety evaluation report (SER) dated September 11, 1985, Alternative Shutdown Design, established the licensing basis for MNGP's compliance to the applicable provisions of 10CFR50 Appendix R, sections III.G.3 and III.L. In part, the SER established that MNGP need only consider spurious operation of one worst-case non-high-low pressure interface. Per supporting technical information submitted on December 16, 1983, the worst-case spurious operation for MNGP was demonstrated and defined to be one Safety Relief Valve (SRV) failing open.
Although MNGP's SER for alternative shutdown remains the licensing basis for the plant, for conservatism, the effect of MSOs on CAP in the Appendix R bounding fire scenario was considered for this evaluation. Per the guidance of NEI 00-01, Revision 2, "Guidance for Post Fire Safe Shutdown Circuit Analysis," and Regulatory Guide 1.189, Revision 2, "Fire Protection for Nuclear Power Plants," the effect of MSOs were evaluated after transferring to the ASDS panel. When combining MSO scenarios, a combined maximum of four components spuriously operating was considered.
To assess the most limiting combination of MSOs for reactor vessel response, NSPM modeled the MNGP reactor coolant pressurized boundary with a GOTHIC Rev. 7.2b model. This modeling allowed for assessment of drywell heat loads and energy release to the suppression pool, among other assorted leak rates influencing coolant inventory. These parameters were then used in a containment response GOTHIC model to assess available NPSH for ECCS pumps. GOTHIC is a safety related code that performs best estimate thermal hydraulic evaluations. Since the Appendix R cases analyzed in this evaluation are not design basis events, GOTHIC is suitable for the purpose intended. The base GOTHIC model for evaluating MNGP Appendix R events was developed by the NRC. However, model alterations were required to indicate MSO occurrences to adequately perform the NPSH margin evaluations.
The evaluation considered the potential for a MSO scenario, consisting of a combined maximum of four components spuriously operating. The impact from MSOs were derived based on NEI 00-01, "Guidance for Post-Fire Safe-Shutdown Circuit Analysis" (Revision 2 issued May 2009). In RIS 2005-30, "Clarification of Post-Fire Safe-Shutdown Circuit Regulatory Requirements," the NRC endorsed NEI 00-01 as an acceptable method for performing circuit analysis.
All MSOs were evaluated to assess the worst case combination of any four components spuriously operating on containment accident pressure and suppression pool temperature with the corresponding impact on ECCS available NPSH.
The evaluation considered 13 cases for NPSH margin for the CS pumps (the most limiting ECCS pump). NPSH margin is defined as NPSH available (NPSHa) -
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L-MT-12-107 NPSHr. NPSHr is defined as the NPSH at a specific pump flow rate that results in a 3 percent drop in Total Developed Head (TDH). The cases were divided into the following five sets of cases with the results as shown in Table 6.6.4-1 below:
Table 6.6.4 Cases evaluated with Results for Appendix R event NPSHr verification Case SRV HPCI/RCIC System System System Minimum NPSH No. Failures? Drains Open? Failure #1 Failure #2 Failure #3 Mnrgin (1t)
Set A: Maximum RPV Leakage Pathway Component Failures 2AI No No CRD RWCLJ 6.1
.Set B: CRD Leakage Pathway Component Failures 2B11 No No CRD 2 RHR Pumps 6.1 2B2 No No CRD I RHR Pump WW Spray 4.2 2B3 No No CRD DW Leakage 6.1 Set C: No RPV Leakage Pathway Component Failures 2C 1 No No 3 RHR Pumps WW Spray 1.7 2C2 No No 2 RHR Pumps DW Leakage 5.7 2C3 No No DW Leakage WW Spray I RHR Pump 3.8 2C4 No No DW Leakage DW Leakage 5.9 Set D: Recirculation Pump Seal Leakage Pathway Component Failures 2I2 No No Recirc. Seal WW Spray 2.0 2D2 No No Recirc. Seal WW RHR Pump 5.5 Set E: RWCU Leakage Pathway Component Failures 2E] No No RWCU 2 RHR Pumps 5.7 2E2 No No RWCU I RHR Pump WW Spray 1.9 2E3 No No RWCU DW Leakage 5.7 As can be seen in Table 6.6.4-1, Case 2C1 provided the most limiting results. This case is discussed further below.
Case 2C1: At 600 seconds into the event, 3 RHR non credited pumps stait and the wetwell spray valve opens. The system failures continue for the duration of the event. The credited RHR pump also starts. Heat is added to the suppression pool from all 4 RHR pumps. The identification of these failures is shown in Table 6.6.4-1.
Figure 6.6.4-1 below, shows the NPSH margin results as a function of time for all the 2C cases.
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L-MT-12-107 Figure 6.6.4-1 - NPSH Margin for the "C" Cases NPSH Margin, "C" Cases, No RPV Leakage Pathway Component Failures 25 20 15 U-10 5
0 0.00E+00 5.00E+04 1.00E+05 1.50E+05 2.OOE+05 2.50E+05 3.00E+05 Seconds Conclusion of Evaluation:
All 13 MSO cases evaluated met the acceptance criteria that the NPSHa is greater than NPSHr, with the most limiting case demonstrating a margin of 1.7 feet.
SECY 11-0014, Enclosure 1, section 6.6.4 states that circuit issues associated with an Appendix R fire cannot occur or that they would occur only after use of containment accident pressure is no longer needed. Demonstration of NPSH margin for the CS pump with up to four components spuriously operating provides an acceptable alternative to the NRC guidance provided in SECY 11-0014, Enclosure 1, section 6.6.4.
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L-MT-12-107 6.6.7 Assurance of no Pre-existing leak Licensees and applicantsshould consider a loss of containment isolation that could compromise containmentintegrity. Possible losses of containment integrity include containment venting requiredby procedures or loss of containment isolation from a postulatedAppendix R fire. It should be demonstrated conservatively that, for the plant examined, loss of containment integrity from these causes cannot occur or that they would occuronly after use of containment accident pressureis no longer needed.
To reduce the likelihood of a preexisting leak, licensees proposing to use containment accidentpressure in determining NPSH margin should do the following:
(1) Determine the minimum containment leakage rate sufficient to lose the containment accidentpressure needed for adequate NPSH margin.
(2) Propose a method to determine whether the actual containmentleakage rate exceeds the leakage rate determined in (1) above. For inerted containments, this method could consist of a periodic quantitative measurement of the nitrogen makeup performed at an appropriatefrequency to ensure that no unusually large makeup of nitrogen occurs. Monitoring oxygen content is anothermethod. For subatmosphericcontainments, a similarprocedure might be used.
(3) Proposea limit on the time interval that the plant operates when the actual containment leakage rate exceeds the leakage rate determined in (1) above.
NSPM Response to 6.6.7 In Reference E1-3, NSPM provided a complete response to section 6.6.7 with the exception of the Appendix R event. Section 6.6.4 above addresses possible loss of containment isolation that could compromise containment integrity for an Appendix R event. See the response to 6.6.4 for Appendix R Fire evaluation of containment integrity. The information provided in this letter supplements the response provided in Reference E1-3.
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L-MT-12-107
4.0 CONCLUSION
This enclosure demonstrates that there is adequate margin for the RHR and CS pumps to perform their design functions under Appendix R events (called special event in SECY 11-0014). The Appendix R events described herein satisfy the SECY 11-0014 guidance. For Appendix R events, MNGP's design bases require consideration of one spurious operation; however, NSPM evaluated four components spuriously operating to demonstrate that during an Appendix R fire, the limiting case MSOs will not affect the conclusion that sufficient CAP will be available for safe shutdown of MNGP following an Appendix R fire event.
Therefore, NSPM has concluded that use of CAP is justified and available during an Appendix R event. The analyses above demonstrate margin exists to NPSHr during the Appendix R limiting case when MSOs are included.
Combining this analysis with the analysis provided in Reference E1-3, NSPM concludes that the MNGP ECCS pumps can reliably perform their required design functions to mitigate the consequences of accidents and events for the required mission time while using appropriate uncertainties defined for NPSHreff for DBA-LOCA and NPSHr for other events. The ECCS pumps meet the guidance provided in SECY-1 1-0014, .
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L-MT-12-107
5.0 REFERENCES
El-1 Letter from T J O'Connor (NSPM) to Document Control Desk (NRC), "License Amendment Request: Extended Power Uprate (TAC MD9990)," L-MT-08-052, dated November 5, 2008. (ADAMS Accession No. ML083230111)
E1-2 Letter from T J O'Connor (NSPM) to Document Control Desk (NRC), "License Amendment Request: Maximum Extended Load Line Limit Analysis Plus," TAC ME3145, L-MT-10-003, dated January 21, 2010. (ADAMS Accession No. ML100280558)
E1-3 Letter from M A Schimmel (NSPM) to Document Control Desk (NRC), "Monticello Extended Power Uprate and Maximum Extended Load Line Limit Analysis Plus License Amendment Requests: Supplement to Address SECY 11-0014 Use of Containment Accident Pressure (TAC Nos. MD9990 and ME3145)," L-MT 082, dated September 28, 2012. (ADAMS Accession No. ML12276A057)
E1-4 SECY 11-0014, Use of Containment Accident Pressure in Analyzing Emergency Core Cooling System and Containment Heat Removal System Pump Performance in Postulated Accidents, dated January 31, 2011. (ADAMS Accession No. ML102780586)
E1-5 Letter from L M Padovan (NRC) to T J Palmisano (NMC), "Monticello Nuclear Generating Plant - Issuance of Amendment RE: Revised Analyses of Long-Term Containment Response and Net Positive Suction Head (TAC No. MB7185),
dated June 2, 2004. (ADAMS Accession No. ML041540568)
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L-MT-12-107 ENCLOSURE 2 CORRECTION OF APPENDIX R - TIME TO COLD SHUTDOWN VALUE
Background
In Reference E2-1, Northern States Power Company, a Minnesota corporation (NSPM),
doing business as Xcel Energy, stated to the following in response to NRC Request for Additional Information (RAI) 2.8.4.4-2:
The design basis Appendix R event utilizes the ASCM [Alternate Shutdown Cooling Method] to attain cold shutdown. This Appendix R alternateshutdown analysis was performed using this criteriaat EPU conditions which demonstrates that the reactor can be cooled to cold shutdown (<200°tF) in 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. This analysis is discussed in more detail in PUSAR [Power Uprate Safety Analysis Report] Section 2.5.1.4.
The RAI response also indicated that NRC Regulatory Guide 1.139 referred to the Appendix R analysis, whereas, the NRC indicated that this document has been withdrawn and its use is no longer necessary.
Discussion NSPM discovered that the time to cold shutdown value of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> provided is in error.
It was determined that the calculation used the maximum allowable cooldown rate of 100°F/hour to arrive at a time to cold shutdown value of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />.
A new Appendix R design bases calculation was performed using GOTHIC 7.2b as the evaluation tool. The calculation used a cooldown rate based on actual plant parameters and included the following:
o Worst case EPU conditions, including 2044 MWt (102% of EPU) o Initial suppression pool inventory is minimized o Initial suppression pool temperature is maximized o Core Spray flow rate throttled to 2700 gpm o Residual Heat Removal (RHR) pump flow rate is 4000 gpm - single pump cooling o Service water temperature of 90°F is applied to the RHR heat exchanger.
Since, the Appendix R design bases event does not require the assumption of multiple spurious operations (MSOs), the evaluation was performed without the use of additional component failures. The results of this calculation produced a maximum value of 44.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> to cold shutdown (at 200 0 F). The acceptance value for time to cold shutdown is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as stated in the MNGP USAR and 10CFR50, Appendix R.
The discussion in the PUSAR (Reference E2-2, Enclosure 5), section 2.5.1.4 does not contain any discussion of the time to cold shutdown following an Appendix R event.
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L-MT-12-107 The method of achieving cold shutdown (via the ASCM) is not changed from that presented in the PUSAR section 2.5.1.4.
In addition, the corrected time to cold shutdown calculation also looked at the worst case conditions (indicated above) and added to that an assumption of three non-credited (not performing any cooling function) RHR pumps starting and adding heat to the suppression pool to account for the worst case combination of any four components spuriously operating - MSOs - on containment accident pressure and suppression pool temperature. The results of this calculation produced a maximum value of 57.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to cold shutdown (at 200 0 F). The acceptance value for time to cold shutdown is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> as stated in the MNGP USAR and 10CFR50, Appendix R.
Regarding use of RG 1.139, previously an analysis was performed to evaluate the ability of the ASCM to achieve a temperature of 212°F in less than 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> using design bases criteria. The analysis was performed to satisfy NRC RG 1.139. RG 1.139 has subsequently been withdrawn by the NRC and therefore, this requirement no longer exists. However, the analysis was completed for information only as a means to demonstrate margin. The calculation produced a maximum time required to achieve a cold shutdown temperature (212 0 F) of 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br />.
Conclusion Based on the above the statement in response to RAI 2.8.4.4-2 should be revised to read as follows:
The design basis Appendix R event utilizes the ASCM to attain cold shutdown. This Appendix R alternateshutdown analysis was performed using this criteriaat EPU conditions which demonstrates that the reactorcan be cooled to cold shutdown
(<200'F)in 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br />. This analysis is discussed in more detail in PUSAR Section 2.5.1.4.
The discussion in the PUSAR section 2.5.1.4 does not contain any discussion of the time to cold shutdown following an Appendix R event and therefore, no revision to this information is required. References to compliance with RG 1.139 are eliminated as this document's use is no longer necessary.
References:
E2-1 Letter from T J O'Connor (NSPM) to Document Control Desk (NRC), "Monticello Extended Power Uprate: Response to NRC Reactor Systems Branch and Nuclear Performance & Code Review Branch Request for Additional Information (RAI) dated January 16, 2009 (TAC No. MD9990)," L-MT-09-017, dated March 19, 2009. (ADAMS Accession No. ML090790388)
E2-2 Letter from T J O'Connor (NSPM) to Document Control Desk (NRC), "License Amendment Request: Extended Power Uprate (TAC MD9990)," L-MT-08-052, dated November 5, 2008. (ADAMS Accession No. ML083230111)
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