L-MT-13-033, Extended Power Uprate: Secy 11-0014 Use of Containment Accident Pressure - Responses to Requests for Additional Information

Extended Power Uprate: Secy 11-0014 Use of Containment Accident Pressure - Responses to Requests for Additional Information

ML13085A033
Person / Time
Site:	Monticello
Issue date:	03/21/2013
From:	Schimmel M A Xcel Energy, Northern States Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-13-033, TAC MD9990, SECY 11-0014
Download: ML13085A033 (28)

Text

ENCLOSURE 2 CONTAINS PROPRIETARY INFORMATION

-WITHHOLD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH 10 CFR 2.390 SXcelEnergy Monticello Nuclear Generating Plant 2807 W County Rd 75 Monticello, MN 55362 March 21, 2013 L-MT-13-033 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Monticello Nuclear Generating Plant Docket 50-263 Renewed License No. DPR-22 Monticello Extended Power Uprate: SECY 11-0014 Use of Containment Accident Pressure -Responses to Requests for Additional Information (TAC MD9990)

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 Janua 21, 2010.(ADAMS Accession No. ML1 00280558)

,\ 00 3) Letter from J G Giitter (NRC) to T J O'Connor (NSPM), 'Su bject: 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)

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Document Control Desk Page 2 5) 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, Sections 6.6.4 and 6.6.7 (TAC Nos. MD9990 and ME3145)," L-MT-12-107, dated November 30, 2012. (ADAMS Accession No.MLI123380435)

6) Email from T Beltz (NRC) to J Fields (NSPM), "Monticello

-Requests for Additional Information Supporting the EPU Containment Accident Pressure and MELLLA+ Reviews (MD9990)," dated February 15, 2013. (ADAMS Accession No. ML13050A021)

7) 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)

8) 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 References 4 and 5, 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 (ADAMS Accession No. ML102590196).

NSPM has previously 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.

Document Control Desk Page 3 In Reference 6, the NRC sent a request for additional information (RAI) concerning the letters sent by NSPM in References 4 and 5. The RAI included 10 separate questions.

Enclosure 1 to this letter provides the NSPM response to NRC RAIs 1 -4, and 7 -9.Enclosure 2 to this letter provides responses to NRC RAIs 5, 6 and 10. The responses in Enclosure 2 were prepared by General Electric-Hitachi (GEH) and contain proprietary information.

Enclosure 3 to this letter is a non-proprietary version of the responses provided in Enclosure 2.Enclosure 4 contains a GEH affidavit executed to support withholding Enclosure 2 from public disclosure.

The affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4).

NSPM requests that the proprietary information in Enclosure 2 be withheld from public disclosure in accordance with 10 CFR 2.390(a)4, as authorized by 10 CFR 9.17(a)4.

Accordingly, it is respectfully requested that the information which is proprietary to GEH be withheld from public disclosure in accordance with 10 CFR 2.390.Correspondence with respect to the copyright or proprietary aspects of GEH information or the supporting GEH affidavit in Enclosure 4 should be addressed to L C Dolan, Manager, Regulatory Compliance, GE-Hitachi Nuclear Energy Americas LLC, 3901 Castle Hayne Rd, Wilmington, NC 28401.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 7 and 8 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.

Document Control Desk Page 4 Summary of Commitments This letter makes no new commitments and no revisions to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct.Executed on: March2_/, 2013)Aark kA. SchgimmelZ, kSite Vice-President Monticello Nuclear Generating Plant Northern States Power Company-Minnesota Enclosures (4)cc: Administrator, Region III, 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-13-033 ENCLOSURE 1 MONTICELLO NUCLEAR GENERATING PLANT RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION FROM THE CONTAINMENT AND VENTILATION BRANCH QUESTIONS 1, 2, 3, 4, 7, 8 AND 9 This enclosure provides responses from the Northern States Power Company, a Minnesota corporation (NSPM), doing business as Xcel Energy, to a request for additional information (RAI) provided by the Nuclear Regulatory Commission (NRC) on February 15, 2013. This enclosure covers questions 1, 2, 3, 4, 7, 8 and 9.The NRC RAIs contain references to documents previously submitted by NSPM. Below are the NRC references referred to in the NRC RAIs. These evaluations cover both the Monticello Nuclear Generating Plant (MNGP) extended power uprate (EPU) and maximum extended load line limit plus (MELLLA+)

license amendment requests.NRC REFERENCES

1. NSPM letter to NRC dated September 28, 2012, "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)," ADAMS Accession No.ML12276A057

2. NSPM letter to NRC dated November 5, 2008, "License Amendment Request Extended Power Uprate (TAC MD9990)," ADAMS Accession No. ML08323011

3. NSPM letter to NRC dated July 13, 2009, "Monticello Extended Power Uprate: Response to NRC Containment and Ventilation Review Branch (SCVB) Requests for Additional Information (RAI) dated March 19, 2009 and March 26, 2009 (TAC No. MD9990)," ADAMS Accession No. ML092170404

4. NSPM letter to NRC dated November 30, 2012, "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)5. NSPM letter to NRC dated January 21, 2010, Attachment 3 of L-MT-10-003,"MELLLA Plus Safety Analysis Report Proprietary" Page 1 of 13 L-MT-13-033 Enclosure 1 References provided in the NSPM responses are provided at the end of Enclosure 1.The NRC questions are provided below in bolded font and the NSPM response is provided in the normal font.NRC Question 1. In Reference 1, NSPM provided the following responses:

Response to 6.6.5 states: "The EOPs also include a caution that identifies to the operators that if containment pressure falls below 7 psig (the value required for CAP based on use of NPSHr3%) then this may result in inadequate NPSH. This NPSHr value does not change for EPU." Response to 6.6.1 states: "The uncertainties in NPSHr included in the staff's guidance address the possibility that conditions during the NPSHr vendor tests could be different than those seen by the pumps during operation at the plant, effectively increasing the NPSHr values. The differences could arise due to pump inlet temperature variation, pump inlet geometry variation, dissolved gas evolution, and increase in mechanical wear ring clearance." The response to 6.6.5 is in conflict with response to 6.6.1. The response to 6.6.1 addresses the uncertainty in the vendor tested value of Net Positive Suction Head (NPSH) required at 3-percent pump head drop (NPSHr3%)

and states that the required NPSH seen by the pumps during operation could be different.

The Nuclear Regulatory Commission (NRC) staff position is that in the Emergency Operating Procedure (EOP) caution statement, the required NPSH for the pump should be NPSH Required Effective (NPSHreff), where NPSHreff = (1 + uncertainty) x NPSHr3%, instead of simply NPSHr3%.Please revise the EOP caution statement or justify the use of NPSHr3% in the EOPs.NSPM Response: The existing EOP caution will be revised to identify to the operators that if containment pressure falls below 8.6 psig that inadequate NPSH may exist. This value corresponds to the amount of containment pressure required for NPSHreff with a design basis accident (DBA) loss of coolant accident (LOCA).Page 2 of 13 L-MT-13-033 Enclosure 1 NRC Question 2. In Reference 1, NSPM Response to 6.6.6, states the following: "Worst case conditions are at an NPSHr value that is between NPSHr3%and the inception point for cavitation, NPSHrO%." Please explain what are the worst case conditions in the stated NPSHr range?NSPM Response: The NSPM statement being addressed in RAI 2 was made in reference to a discussion of the cavitation erosion rate for the MNGP Residual Heat Removal (RHR) pump.The analysis of the cavitation erosion rate for the MNGP RHR pump is discussed in BWROG-TP-12-012 Task 4 -Operation in the Maximum Erosion Rate Zone (CVDS Pump). For the MNGP RHR pump; the erosion rate for various flow rates is shown on Figure 2 of this report. This shows that the maximum erosion rate occurs in the range of 1.75 to 2.5 times the ratio of NPSHa/NPSH3%.

This report was previously provided to the NRC on October 24, 2012 by the Boiling Water Reactor Owners Group (BWROG)1.NRC Question 3. In Reference 1, NSPM Response to 6.6.7, under heading "Monitoring During Normal Operation", states the following: "A computer point is provided that continually calculates the N2 mass in containment and provides a computer alarm in the control room if the N2 mass is too low or too high. The low inventory alarm corresponds to the minimum noncondensable gas mass assumed for the ECCS pump NPSH analysis.

Calculated values below the minimum assumed in the NPSH analysis will result in operator action to declare the ECCS pumps inoperable." Please describe the action to be taken after the operator declares that the emergency core cooling system (ECCS) pumps are inoperable.

In case the plant operation at full power continues with the ECCS pumps declared inoperable, please justify plant operability.

I Letter BWROG-12051 from Frederick P. "Ted" Schiffley, II to Joe Golla (NRC), "

Subject:

Submittal of BWROG Technical Reports Addressing the NRC Guidance Document on the Use of Containment Accident Pressure (CAP) in Analyzing Emergency Core Cooling System (ECCS) and Containment Heat Removal System Pump Performance in Postulated Accidents," dated October 24, 2012 (ADAMS Accession No. ML12300A286).

Page 3 of 13 L-MT-13-033 Enclosure 1 NSPM Response: MNGP containment nitrogen (N2) inventory is maintained above the low inventory alarm point. The computer alarm response procedure has an alarm at 14,225 Ibm and a warning at 14,500 Ibm. Analyses of primary containment pressure response following a LOCA for ECCS pump NPSH used an input of 14,225 Ibm for minimum initial non-condensable mass.The required operator action for low N2 inventory (indicative of a containment leak) is states: If Primary Containment N2 inventory decreases to less than 14,225 Ibm, then declare all RHR and Core Spray pumps inoperable and evaluate the following:

Tech Specs: 3.5.1 (ECCS -Operating)

If all ECCS pumps are inoperable under Technical Specification (TS) 3.5.1.0, the required action is to enter TS Limiting Condition for Operation 3.0.3 immediately, which requires initiation of a reactor shutdown.NRC Question 4. In Reference 1, NSPM Response to 6.6.7, Item (3), does not specify an appropriate time limit for performing the correction in case the containment leakage rate (determined by the on-line test and monitoring described in response to 6.6.7 item (2)) exceeds the acceptance limits. TS 3.6.1.1 specifies that containment is operable if it's Surveillance Requirement of visual inspection and integrated leak rate testing is met, which is performed every 15 years during an outage. This would imply that the containment is operable even if the on-line test and monitoring gives unacceptable results.Under the condition when the containment leakage is not corrected and with the plant in Modes 1, 2 or 3, please justify proper operation of the ECCS and containment heat removal pumps without containment accident pressure (CAP) (or without adequate NPSH) during design basis or non-design basis accidents.

NSPM Response: NSPM concurs that containment is operable as long as TS 3.6.1.1 is satisfied.

NSPM's statements in Reference 1, section 6.6.7(2) were not directed at TS 3.6.1.1 but rather to describe a method to ensure that no leakage greater than 30 La occurs to ensure sufficient required NPSH is available for ECCS pumps. The method defined in the response provided in section 6.6.7 does not have sufficient instrument accuracy to measure leakage rates to satisfy TS or 10 CFR 50 Appendix J criteria.

The on-line leakage rate test method described does have sufficient repeatability to identify a trend from normal operation that could be indicative of a developing leak and allows for early investigation into the cause.Page 4 of 13 L-MT-13-033 Enclosure 1 As shown in the response to RAI 3 above and Reference 1, Section 6.6.7(2), alternate methods to monitor and control plant operation to support verification of required ECCS pump NPSH are prescribed in plant procedures.

However, little operational data for the on-line leakage rate test discussed in the Reference 1, section 6.6.7(2) response exists.Therefore, it is appropriate to define containment operability based on this method of estimating containment leakage to use an acceptance criterion based on comparison to that leak rate that is sufficient to lose containment accident pressure needed for adequate NPSH margin as determined in Reference 1, section 6.6.7(1).

The evaluation in section 6.6.7(1) showed that a leak of greater than 30 La (228 scfm) will result in loss of margin for NPSHreff.

This would indicate that all RHR and Core Spray (CS) pumps are inoperable when a leak of greater than 30 La is present.Upon NRC approval of EPU, NSPM will implement plant procedure changes to require the on-line leakage test to meet the acceptance criterion of 150 scfm. This will satisfy the Reference 1, section 6.6.7(3) time limit to ensure the plant does not operate when the actual containment leakage rate exceeds the leakage rate requirement.

If the acceptance criteria is not met, the plant will immediately enter TS 3.5.1.0 (ECCS -Operating) for multiple ECCS injection/spray subsystems inoperable.

TS 3.5.1.0 requires immediate entry into TS LCO 3.0.3. TS LCO 3.0.3 requires that action be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in: a. MODE 2 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />;b. MODE 3 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and c. MODE 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.NRC Question 7. In Reference 1, NSPM Response to Section 6.6.7, states: "Section 6.6.4 above addresses possible loss of containment isolation that could compromise containment integrity for an Appendix R event." (a) Neither Reference I nor NSPM Response to Section 6.6.4 in Reference 4 addresses Appendix R Fire induced failure of associated circuits that could result in a loss of containment integrity due to containment venting.Please provide results of safe shutdown analysis showing that adequate NPSH will be available for the residual heat removal (RHR) and core spray (CS) pumps under a loss of containment integrity due to fire induced multiple spurious operations (MSO), or justify that a loss of containment integrity cannot occur under a Appendix R Fire scenario.Page 5 of 13 L-MT-13-033 Enclosure 1 NSPM Response: Although NSPM does not consider multiple spurious operations to be included in the MNGP licensing basis, multiple spurious operations (MSOs) were evaluated in accordance with the applicable guidance of Regulatory Guide (RG) 1.189, Revision 2,"Fire Protection for Power Plants" and NEI 00-01, Revision 2, "Guidance for Loss of Remote Shutdown Capability During a Control Room Fire". This evaluation did include consideration of impact of MSOs that could result in loss of containment integrity due to containment venting.A description of the MSO process utilized at MNGP is provided below. In summary with respect to CAP: MSO scenarios and combinations of MSO scenarios (combined per guidelines of NEI 00-01 Revision 2) that could challenge Appendix R required CAP, were precluded from occurring through modifications and configuration changes as described below. Those scenarios and combinations of scenarios that were identified as applicable to MNGP, but not precluded from occurring, were demonstrated by analysis to show that they would not pose a challenge to Appendix R required CAP. In other words, NPSH is maintained, because sufficient CAP is available, for the pumps used in mitigation of any Appendix R events.The MSO evaluation included the use of an expert panel to identify the applicable generic MSOs (NEI 00-01 Appendix G), as well as plant specific MSOs. As a result of this process, several modifications and a configuration change were performed to preclude fire-induced MSOs from adversely affecting safe shutdown:* Modified the containment vent and purge air operated valves (to preclude loss of CAP) as well as several motor operated valves (to address Information Notice (IN)92-18 and diversion path/loss concerns).

A total of nine (9) valves were modified.* Modified the RHR minimum flow bypass valve (to preclude potential pump damage under alternative shutdown scenario), and modified two breakers (to preclude non-synchronous paralleling of power supplies).

• Modified two valves in the High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) test-return path (to preclude potential loss of suppression pool inventory).

• Configuration change removed control power fuses (closing) to preclude remote closing of 480VAC cross-tie breaker.In addition to the completion of the modifications described above, NSPM also performed a penetration-by-penetration evaluation using the penetrations identified in MNGP Updated Safety Analysis Report (USAR) Table 5.2-3a, to validate that potential diversion flow paths out of primary containment had been properly addressed.

Page 6 of 13 L-MT-13-033 Enclosure 1 Finally, NSPM performed a series of thermal-hydraulic and supporting analyses to demonstrate that MSOs would not adversely affect safe shutdown of MNGP. These analyses were performed using the limiting alternative shutdown scenario, while considering combinations of four spuriously operating components (NEI 00-01 guidance for combining MSOs). The analyses demonstrated the following goals and objectives:

1. Voluntarily addressed the "risk" of MSOs occurring in the alternative shutdown scenario using the guidance of NEI 00-01, Revision 2, Appendices D and G.2. Validated expert panel assumptions (e.g., small leakage paths <1" in dia.), and addressed the concern of the effect of combining small leakage paths under MSO conditions.

3. Validated that the alternative shutdown timeline remains valid under MSO conditions.

4. Validated that MSO effects on containment accident pressure and suppression pool temperature would not adversely affect ECCS pumps' available NPSH. This demonstrates that there would be sufficient NPSHa to mitigate an Appendix R event under limiting MSO conditions.

5. Demonstrated that fuel damage would not occur under the most severe MSO conditions.

The thermal hydraulic analysis identified the need for the following:

• Modification to preclude multiple spurious operation of drywell spray valves. This modification has been completed.

• Modification to preclude multiple spurious operation of the main steam line drain valves. This modification is currently being installed.

NRC Question (b) Reference 4, NSPM Response to 6.6.4 states that the effect of MSO was evaluated after transferring to the alternate shutdown system (ASDS)panel as per the guidance of NEI 00-01 Revision 2. The response does not address the scenarios listed in the NEI document Table G-1 under"Decay Heat Removal" which are applicable to MNGP. Please list the applicable scenarios and provide their NPSH margin results. Please note that scenarios 4r, 4s, and 4t discuss the potential fire-induced impact on CAP and possible NPSH loss which should be addressed.

Justify why the remaining scenarios listed for BWR-3 under "Decay Heat Removal" are not applicable for MNGP.Page 7 of 13 L-MT-1 3-033 Enclosure 1 NSPM Response: The applicable scenarios from NEI 00-01, Table G-1 under "Decay Heat Removal" were considered and the limiting NPSH margin results are provided in Reference 4 Table 6.6.4-1. The impact of scenarios 4r, 4s, and 4t were considered for the potential fire-induced impact on CAP and possible NPSH loss. These scenario evaluations are described in more detail below:* Scenario 4r -NPSH loss due to spurious initiation of containment sprays Reference 4, Table 6.6.4-1 included consideration of spurious initiation of containment spray as part of the evaluation listed for Cases 2C1, 2C3 and 2D1.* Scenario 4s -NPSH loss, Spurious opening of Containment Vent not including rupture disc.Reference 4, Table 6.6.4-1 included consideration of containment venting as part of the evaluation listed. Case 2C4 includes failure of the drywell floor drain and the drywell equipment drain sump valves modeled as a vent of containment.

Cases 2B3, 2C2, 2C3 and 2E3 include failure of the drywell equipment drain sump valves which are modeled as a vent of containment.

As noted above, the purge and vent valves were modified to preclude susceptibility to spurious operation.

These are the normal containment vent valves.* Scenario 4t -NPSH loss, Spurious opening of the drywell floor drain sump valves Reference 4, Table 6.6.4-1 included consideration of Case 2C4 which includes failure of the drywell floor drain and the drywell equipment drain sump valves modeled as a vent of atmosphere from containment.

Drywell sump valves are identified in Table 6.6.4-1 as "DW Leakage." The following decay heat removal generic MSOs, applicable to BWR-3 (taken from Table G-1 of NEI 00-01 Revision 2), were identified as "not applicable" to MNGP by the MSO expert panel. Note that scenario 4b was subsequently determined to be applicable to MNGP as described below: The expert panel originally identified the following MSO as "not applicable" to MNGP, citing that the diversion flow to drywell spray was expected to have only a minor effect on suppression pool level. However, the thermal-hydraulic analysis that was performed to verify this assumption determined that the evaporative cooling effect produced by the diversion to drywell sprays could significantly affect NPSH.Therefore, the drywell spray valves were modified to preclude this diversion path:* Scenario 4b -Failure due to diversion of suppression pool decay heat removal through 16B & 17B.The expert panel identified the following two MSOs as "not applicable" to MNGP, citing the fact that HPCI and RCIC are not credited for operation post-fire.

Note that the penetration-by-penetration evaluation independently considered and verified Page 8 of 13 L-MT-13-033 Enclosure 1 proper disposition of all potential MSOs that could divert inventory away from primary containment.

In the case of HPCI and RCIC, modifications to the test/return valves were performed to preclude inventory loss to the Condensate Storage Tank (CST). Additionally, inventory loss through spurious operation of the HPCI and RCIC condensate drains (steam and liquid), was considered and conservatively modeled in the thermal-hydraulic analysis.* Scenario 41 -Isolation Condenser

-Spurious closure of Steam line isolation valves (IV-39-07, 08, 09 or 10) results in failure of decay heat removal. Failure to isolate for pipe breaks.* Scenario 4m -Isolation Condenser

-Condensate return isolation valve failure to move/remain in correct position.The expert panel identified the following four MSOs as "not applicable" to MNGP as there is no isolation condenser.

Note that the penetration-by-penetration evaluation independently considered and verified proper disposition of all potential MSOs that could divert inventory away from primary containment.

By performing an independent analysis, plant specific adaptations of the generic MSOs were considered.

• Scenario 4n -Isolation Condenser

-Failure of Vent to main steam line valve (IV-05-02 and 03) to close or hot short(s) that keeps both valves open results in loss of inventory." Scenario 40 -Isolation Condenser

-Failure of Vent to main steam line valves (IV-05-01, 11, 12, 04) to close results in loss of inventory.

• Scenario 4p -Isolation Condenser

-Spurious opening of normally-closed Vent to torus valves (BV-05-05 and 07) results in loss of inventory.

• Scenario 4q -Isolation Condenser

-Spurious closure of Cross-connect valve (BV-60-13) on makeup line results in loss of cross-connect capability of makeup source.>- The expert panel identified the following MSO as "not applicable" to MNGP, citing that the design basis for MNGP assumes that all pumps aligned to the torus start simultaneously.

Note that the thermal-hydraulic analysis considered spurious starts of non-credited ECCS pumps and their effects on suppression pool temperature and NPSH.* Scenario 4w -Mark I containment with Torus Ring Header: too much flow through the ring header due to spurious operation of multiple pumps (more than allowed by design) from the ring header.Reference 4 provides the requested information for the impact of Appendix R induced failure of associated circuits that could result in a loss of containment integrity.

The Reference 4 assessments of multiple spurious operations (MSOs) are the limiting fire induced failures of associated circuits that could impact containment integrity.

Page 9 of 13 L-MT-13-033 Enclosure 1 Reference 4, Table 6.6.4-1 listed the combinations of MSOs that were considered to verify that the credited ECCS pumps would have sufficient NPSH to mitigate an Appendix R event. Table 6.6.4-1 included the MSOs for either vent paths or other system spurious operations.

In some cases failures other than venting were more limiting then other postulated vent paths from containment and these were therefore evaluated.

In conclusion, NSPM evaluated the Appendix R Fire induced failures of associated circuits that could result in a loss of containment integrity due to containment venting.The most limiting combination of postulated vent paths or other system spurious operation was presented in the Reference 4 response.NRC Question 8. In Reference 1, NSPM Response to 6.6.7 mentions "drywell Continuous Air Monitor (CAM)".Is this the same system as described in USAR Section 5.2.2.7, "Containment Atmosphere Monitoring System"?NSPM Response: The "Drywell Continuous Air Monitor (CAM)" and the "Containment Atmosphere Monitoring System" are separate systems but are related. Both systems are connected as they share portions of piping as shown in USAR Section 15 drawings NH-91197 and NH-46162.The Containment Atmosphere Monitoring System is described in USAR Section 5.2.2.7 and is shown on drawing NH-91197.

This system is required to assess the degree of core damage during a beyond design-basis accident.

MNGP also has a Drywell Continuous Air Monitor (CAM) (see drawing NH-46162) that provides an early warning of abnormal reactor coolant leakage into the Drywell by continuously monitoring for changes in particulate activity in the Drywell atmosphere.

Components upstream of the system also provide for obtaining grab particulate, iodine, noble gas and tritium samples which are analyzed in the radiochemistry laboratory.

NRC Question 9. In Reference 1; Figure 6.6.6-5 caption states: "Long-term RHR NPSH Margin"; graph title states: "Core Spray Pump B Long-term DBA LOCA NPSH".Please provide the correct graph for Long-term Residual Heat Removal (RHR) NPSH margin.Page 10 of 13 L-MT-1 3-033 Enclosure 1 NSPM Response: Figure 6.6.6-5 erroneously indicated that the data was applicable to the MNGP Core Spray B pump. This error is reconciled below as replacement curves for Reference 1, Figure 6.6.6-5 -Long-Term RHR NPSH Margin are provided below. This reflects margin at a pump flow rate of 4178 gpm. Figures for RHR pumps B and C are provided below.RHR pump B Long term DBA- LOCA NPSH-. NPSHa --- NPSHr3% x 1.211 40.00 35.00 30.00 25.00 20.00 15.00 10.00 5.00 0.00 Seconds Page 11 of 13 L-MT-1 3-033 Enclosure 1 RHR pump C Long term DBA LOCA NPSH I-- NPSHa --NPSHr3%x 1.21 40 35 30 25 o20 Li-15 10 5 0 100 1000 10000 Seconds 100000 1000000 Page 12 of 13 L-MT-13-033 Enclosure 1 REFERENCES

1. 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)

2. Not Used 3. Not Used 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, Sections 6.6.4 and 6.6.7 (TAC Nos. MD9990 and ME3145)," L-MT-12-107, dated November 30, 2012. (ADAMS Accession No. ML123380435)

Page 13 of 13 L-MT-13-033 ENCLOSURE 3 GE-MNGP-AEP-3272 R1, ENCLOSURE 2 -NON-PROPRIETARY GEH RESPONSES TO RAIS SUPPORTING THE EPU CAP REVIEW NRC QUESTIONS 5, 6 AND 10 6 pages follow ENCLOSURE 2 GE-MNGP-AEP-3272 RI GEH Response to RAIs Supporting the EPU CAP Review Non-proprietary Information

-Class I (Public)NON-PROPRIETARY INFORMATION NOTICE This is a non-proprietary version of the Enclosure I of GE-MNGP-AEP-3272 RI which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here [[ I].

GE-MNGP-AEP-3272 Ri Non-proprietary Information-Class I (Public)Enclosure 2 Page 2 of 6 Requests for Additional Information Supporting the EPU Containment Accident Pressure and MELLLA+ Reviews RAI 5 In Reference 2, Enclosure 5, Section 2.6.5, under heading "Small Steam Line Break Accident (SBA)".Please confirm that 0.01 ft 2 is the limiting break area for the net positive suction head (NPSH)analysis.GEH Response -RAI 5"Reference 2, Enclosure 5" refers to NEDC-33322P, Safety Analysis Report for Monticello Nuclear Generating Station Extended Power Uprate, (Reference 6). This 0.01 ft 2 SBA break size was selected from a review of the SBA analyses performed to establish the equipment environmental qualification (EQ) drywell temperature profile for the MNGP EPU. These analyses included steam line break sizes of 0.5 ft 2 , 0.1 ft 2 and 0.01 ft 2.The peak predicted suppression pool temperature for the 0.01 ft 2 SBA was 27F higher than the peak predicted suppression pool temperature obtained from either of the other two cases. Additionally, because of the small break size, this case maximized the direct transfer of vessel and decay heat energy to the suppression pool with SRV discharges which in turn minimized the energy transferred to the drywell airspace with the break flow. This minimized the drywell pressure response which also minimized the predicted wetwell pressure response.

As a result, a bounding conservative condition (minimum wetwell airspace pressure with maximum suppression pool temperature) was obtained with the selection of a 0.01 ft 2 SBA break size for use in evaluating NPSH margins.Therefore, it is concluded that 0.01 ft 2 SBA break is the limiting break area for NPSH analysis.

GE-MNGP-AEP-3272 RI Non-proprietary Information-Class I (Public)Enclosure 2 Page 3 of 6 RAI 6 In Reference 3, NSPM response to NRC RAI No 7, eighth paragraph, for the SBA analysis for NPSH.Please describe the alternate method used to simulate the one-pump variable K-value for evaluation of the containment response.GEH Response -RAI 6 Te This ensured an accurate prediction of the wetwell pressure response for the NPSH evaluation.

GE-MNGP-AEP-3272 RI Non-proprietary Information-Class I (Public)Enclosure 2 Page 4 of 6 RAI 10a In Reference 5, Section 4.1.1, second paragraph states: "The peak drywell temperatures for the current licensed operating domain and the MELLLA+operating domain are 291'F and 2907F, respectively." (a) Please define "current licensed operating domain' in terms of value of thermal power and the operating domain.GEH Response -RAI 10(a)"Reference 5" in this question refers to NEDC-33435P, Safety Analysis Report for Monticello Maximum Extended Load Line Limit Analysis Plus, (Reference 7). The "current licensed operating domain" corresponds to the region labeled as "Current Licensed Operating Boundary including MELLLA Operating Domain" in Figure 1-1 of Reference 7, with 100% of Current Licensed Thermal Power defined as 2004 MWt.

GE-MNGP-AEP-3272 RI Non-proprietary Information-Class I (Public)Enclosure 2 Page 5 of 6 RAI 10b (b) Reference 1, Table 2.6-1 gives a peak drywell temperature under current licensed thermal power as 3357F and under EPU maximum extended load line limit (MELLLA) conditions as 338°F. These temperatures do not match with the temperatures given in the above statement.

Please remove the discrepancy or justify the difference.

GEH Response -RAI 10(b)"Reference 1, Table 2.6-1" in this question refers to NEDC-33322P (Reference 6). In this report Current Licensed Thermal Power corresponds to the pre7EPU value of 1775 MWt, and EPU corresponds to 2004 MWt. The peak drywell temperatures reported under EPU/MELLLA conditions in Table 2.6-1 of NEDC-33322P were obtained from a long-term containment response calculation for a small steam line break accident (SBA) at 102% of EPU power and 100% core flow with the SHEX code. [[]] Therefore, this SHEX SBA calculation bounds all power/flow conditions with EPU/MELLLA and also bounds the extended operating conditions with MELLLA plus.The drywell temperatures given in the above statement, taken from Section 4.1.1 of NEDC-33435P, Safety Analysis Report for Monticello Maximum Extended Load Line Limit Analysis Plus, (Reference 7), were obtained from the short-term DBA-LOCA recirculation suction line break (RSLB) analyses performed with the M3CPT/LAMB models at EPU and MELLLA plus conditions respectively.

These were intended to show the effect of MELLLA plus vs. EPU/MELLLA on the predicted RSLB peak drywell temperature.

A separate SHEX SBA analyses was not performed for MELLLA plus because the results obtained for the EPU, and reported in NEDC-33322P, are bounding and would therefore not be impacted by the MELLLA plus operating conditions.

To summarize, the peak drywell temperatures of 335°F and 338°F reported in Table 2.6-1 of NEDC-33322P represent the maximum values at the pre-EPU power (1775 MWt) and EPU power 2004 MWt respectively, and were obtained for the limiting event for peak drywell temperature which is a steam break. The value of 338°F represents the maximum expected drywell temperature for the range of possible break sizes and types and for the range of operating conditions for EPU including EPU/MELLLA plus. The values for peak drywell temperatures shown in Section 4.1.1 of Reference 7, are from the RSLB analysis at the EPU power (2004 MWt) for EPU/MELLLA and EPU/MELLLA plus. The RSLB is limiting for peak drywell pressure but not peak drywell temperature, and therefore the peak drywell temperatures from the RSLB analyses are included in Section 4.1.1 of NEDC-33435P for information purposes only.

GE-MNGP-AEP-3272 RI Non-proprietary Information-Class I (Public)Enclosure 2 Page 6 of 6 References

1. NSPM letter to NRC dated September 28, 2012, "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)," ADAMS Accession No. ML12276A057.

2. NSPM letter to NRC dated November 5, 2008, "License Amendment Request: Extended Power Uprate (TAC MD9990)," ADAMS Accession No. MNL0832301 11.3. NSPM letter to NRC dated July 13, 2009, "Monticello Extended Power Uprate: Response to NRC Containment and Ventilation Review Branch (SCVB) Request for Additional Information (RAI) dated March 19, 2009, and March 26,2009 (TAC No. MD9990)," ADAMS Accession No. ML092170404.

4. NSPM letter to NRC dated November 30, 2012, "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).5. NSPM letter to NRC dated January 21, 2010, Attachment 3 of L-MT-10-003, "MELLLA Plus Safety Analysis Report Proprietary." 6. GE Hitachi Nuclear Energy, NEDC-33322P, "Safety Analysis Report For Monticello Nuclear Generating Station Extended Power Uprate," Revision 3, October 2008.7. GE Hitachi Nuclear Energy, NEDC-33435P, "Safety Analysis Report for Monticello Maximum Extended Load Line Limit Analysis Plus," Revision 1, December 2009.

L-MT-13-033 ENCLOSURE 4 GENERAL ELECTRIC -HITACHI AFFIDAVIT FOR WITHHOLDING PROPRIETARY INFORMATION 3 pages follow GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, Linda C. Dolan, state as follows: (1) 1 am the Manager, Regulatory Compliance of GE-Hitachi Nuclear Energy Americas LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in Enclosure 1 of GEH letter, GE-MNGP-AEP-3272 RI, "GEH Response to Requests for Additional Information Supporting EPU CAP Review," dated March 14, 2013. The GEH proprietary information in Enclosure 1, which is entitled "GEH Response to RAIs Supporting the EPU CAP Review, GEH Proprietary Information

-Class III (Confidential)" is identified by a dark red dotted underline inside double square brackets.

[LThis..s.entence.is.an.examp..,e t 3)]]. In each case, the superscript notation , refers to Paragraph (3) of this affidavit that provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 U.S.C. Sec. 552(b)(4), and the Trade Secrets Act, 18 U.S.C.Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualifies under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F.2.d 871 (D.C. Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F.2.d 1280 (D.C. Cir. 1983).(4) The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. Some examples of categories of information that fit into the definition of proprietary information are: a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over GEH or other companies.

b. Information that, if used by a competitor, would reduce their expenditure of resources or improve their competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.c. Information that reveals aspects of past, present, or future GEH customer-funded development plans and programs, that may include potential prQducts of GEH.d. Information that discloses trade secret or potentially patentable subject matter for which it may be desirable to obtain patent protection.

Affidavit for GE-MNGP-AEP-3272 RI Page I of 3 GE-Hitachi Nuclear Energy Americas LLC (5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to the NRC in confidence.

The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, not been disclosed publicly, and not been made available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary or confidentiality agreements that provide for maintaining the information in confidence.

The initial designation of this information as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure are as set forth in the following paragraphs (6) and (7).(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, who is the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or who is the person most likely to be subject to the terms under which it was licensed to GEH. Access to such documents within GEH is limited to a "need to know" basis.(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation.

Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary or confidentiality agreements.

(8) The information identified in paragraph (2) above is classified as proprietary because it contains a proprietary method developed by GEH for calculating a small break analysis.The results of this analysis are used to support Monticello's extended power uprate and maximum extended load line limit analysis license applications.

This analysis is part of the GEH extended power uprate and maximum extended load line limit analysis methodology.

Development of the extended power uprate and maximum extended load line limit analysis methodologies and the supporting analysis techniques and information and their application to the design, modification and processes were achieved at a significant cost to GEH.The development of the evaluation methodology along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset.(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities.

The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.the value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply Affidavit for GE-MNGP-AEP-3272 R1 Page 2 of 3 GE-Hitachi Nuclear Energy Americas LLC the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH. The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.Executed on this 1 4 th day of March, 2013.Linda C. Dolan Manager, Regulatory Compliance GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Rd Wilmington, NC 28401 Linda.dolan@ge.com Affidavit for GE-MNGP-AEP-3272 R1 Page 3 of 3