L-MT-08-043, Extended Power Uprate, Acceptance Review Supplemental Information Package 6

Extended Power Uprate, Acceptance Review Supplemental Information Package 6

ML081640435
Person / Time
Site:	Monticello
Issue date:	06/12/2008
From:	O'Connor T Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-08-043, TAC MD8398
Download: ML081640435 (18)

Text

June 12,2008 L-MT-08-043 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 Facility Operating License License No. DPR-22 Monticello Extended Power Uprate (USNRC TAC MD8398):

Acceptance Review Supplemental lnformation Packaqe 6

References:

1) NMC Letter to USNRC, "License Amendment Request: Extended Power Uprate,"

dated March 31,2008

2) NMC Letter to USNRC, "Monticello Extended Power Uprate (USNRC TAC MD8398): Acceptance Review Supplement Regarding Radiological Analysis,"

dated May 20,2008

3) NMC Letter to USNRC, "Monticello Extended Power Uprate (USNRC TAC MD8398): Acceptance Review Supplemental Information," dated May 28, 2008

4) NMC Letter to USNRC, "Monticello Extended Power Uprate (USNRC TAC MD8398): Acceptance Review Supplemental lnformation Package 3," dated May 30,2008

5) NMC Letter to USNRC, "Monticello Extended Power Uprate (USNRC TAC MD8398): Acceptance Review Supplemental lnformation Package 4," dated June 3,2008

6) NMC Letter to USNRC, "Monticello Extended Power Uprate (USNRC TAC MD8398): Acceptance Review Supplemental lnformation Package 5," dated June 5,2008 Pursuant to 10 CFR 50.90, Nuclear Management Company, LLC (NMC), requested in Reference 1 approval of amendments to the Monticello Nuclear Generating Plant (MNGP) Renewed Operating License (OL) and Technical Specifications (TS) to increase the maximum power level authorized from 1775 megawatts thermal (MWt) to 1870 MWt, an approximate five percent increase in the current licensed thermal power (CLTP). The proposed request for Extended Power Uprate (EPU) represents an 2807 West County Road 75 Monticello, Minnesota 55362-9637 Telephone: 763.295.5151 Fax: 763.295.1454

Document Control Desk Page 2 increase of approximately 12 percent above the Original Licensed Thermal Power (OLTP). The Monticello EPU application was supplemented on May 20, 2008, May 28, 2008, May 30,2008, June 3,2008 and June 5,2008 by References 2, 3,4, 5 and 6.

On May 29, 2008, the NRC staff indicated that additional information was required by the Piping and Non-Destructive Examination Branch (CPNB) to complete the acceptance review. The responses to the questions are included in Enclosure 1.

In a teleconference held June 5, 2008, the NRC staff indicated that information in addition to that submitted in References 3 and 6 would be necessary for the Electrical Engineering Branch (EEEB) to complete the acceptance review of the Monticello EPU license amendment request (LAR). Responses to the questions are included as .

NMC has reviewed the No Significant Hazards Consideration and the Environmental Consideration submitted with Reference 1 relative to the enclosed supplemental information. NMC has determined that there are no changes required to either of these sections of Reference 1.

Commitment Summarv This letter makes no new commitments and does not change any existing commitments.

I declare under penalty of perjury that the foregoing is true and correct.

Nuclear Generating Plant cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC Minnesota Department of Commerce Enclosures (2)

1. Enclosure 1, Piping and Non-Destructive Examination Branch Questions and Responses

2. Enclosure 2, Electrical Engineering Branch Questions and Responses to L-MT-08-043 Piping and Non-Destructive Examination Branch Questions and Responses

Enclosure 1 In lieu of questions, the Nuclear Regulatory Commission (NRC) Piping and Non-Destructive Examination Branch (CPNB) staff provided references showing, by comparison, information that appears to be missing from the Monticello Extended Power Uprate (EPU) application. The references were:

Susquehanna, ML071000141, dated April 10,2007 Hope Creek, ML070460243, dated February 23,2007 Vermont Yankee, ML033640138, dated December 30,2003 Browns Ferry, ML043440045, dated December 30,2004 Based on review of the above precedent, Nuclear Management Company, LLC (NMC) has provided responses to the questions asked most recently in the Susquehanna reference. The questions from the remaining references are noted at the end of this enclosure and reference back to the responses provided below.

NRC Question:

1 Identify the materials of construction for the reactor coolant pressure boundary (RCPB) piping and safe-ends. Discuss and explain the effect of the requested power uprate on the RCPB piping and safe-end materials and its impact on the potential degradation mechanisms.

NMC Response:

Appendices A and B of the NRC staff's evaluation of Monticello Nuclear Generating Plant's (MNGP) response to Generic Letter (GL) 88-01 identify the materials of construction for the reactor coolant pressure boundary piping and safe end materials at Monticello. The NRC safety evaluation (SE) was received December 19, 1989 (Reference 1). The NRC SE noted that all welds in the reactor coolant pressure boundary within the scope established by GL 88-01 are category " A . Therefore, all reactor coolant pressure boundary welds at Monticello are resistant to sensitization and Intergranular Stress Corrosion Cracking (IGSCC).

MNGP was originally designed as an ANSI B31. I plant. Modifications to the RCPB have typically used ASME Section Ill materials as shown below. Some original materials still exist.

Location Material 1 Recirculation Outlet Nozzle (NI) Safe End SA358 Type 316 Recirculafion Inlet Nozzle (N2) Safe End SA182 F316L Steam Outlet Nozzle (N3) 'safe End A5 16 Grade 70 Feed Water Nozzle (N4) Safe End A508 Class 1 Core Spray Nozzle ( ~ 5Safej End SA350 Gr. LF2QT Head Spray Cooling Nozzle (N6a) Safe SA182 Gr. F304 with Corrosion I End 1 Resistant Clad (CRC) 1 Page 1 of 7

Enclosure 1 Location Material Head Spray Cooling Spare Nozzle (N6b) SA182 Gr. F304 with Corrosion Safe End Resistant Clad (CRC)

Vent Nozzle (N7) Safe End SA182 Gr. F304 with Corrosion Resistant Clad (CRC)

Jet Pump Instrumentation Nozzle (N8A & 316 nuclear grade with 0.02%

B) Penetration Seal maximum carbon Core Differential Pressure & Liquid Control SA182 F316 or SA479 316 Nozzle Safe End (N 10)

Recirculation Piping SA376 TP 3 16* or SA358 Type 316*

• 0.02% max. carbon, 0.06-0.13%

nitrogen Main Steam Piping A106 Gr. B Feed Water Piping A106 Gr B or SA106 Gr. B or SA-333 Gr 6 Seamless, SA 420-WPL6 Core Spray Piping SA333 Gr. 6 or SA 671 GR CC70 Class 32 or SA 106 GR B RWCU Piping SA358 C1 or A-1 06B or A672, Gr B70 Jet Pump Instrumentation Piping (line SA3 12 TP3 16L sections >200°F)

CRD Nozzle N9 Weld Cap SA182 F316L Core Differential Pressure & Liquid Control SA312 TP304L or 316L Piping (line sections >200°F)

CRD Scram Discharge Volume Piping A358 or A312 GR. 304L LPCIIRHR Piping A-106B, SA-106 Gr. B, SA 358 TP316*, SA-333 Gr 6 (Seamless),

SA 671 Gr CC70 Class 32

• 0.02% max. carbon, 0.06-0.10%

nitrogen Implementation of EPU conditions at Monticello will result in an increase in:

neutron fluence, main steam and feedwater flow rate operating temperature for the feedwater system The primary material effects are increases in material fatigue usage, the potential for Irradiation Assisted Stress Corrosion Cracking (IASCC), the potential for Flow-Accelerated Corrosion (FAC), and the potential for flow induced vibration.

Page 2 of 7

Enclosure 1 These material effects are addressed as follows:

Impact on IGSCC (NRC Generic Letter 88-01 and NUREG-0313, Rev. 2) the Monticello EPU license amendment request (LAR), Reference 2, Enclosure 5 (NEDC-33322P), Section 2.2.2.

EPU effects on material fatigue usage for safe ends and piping are discussed in Reference 2, Enclosure 5, Sections 2.2.2 and 2.2.3.

The increase in fluence and the associated effect on IASCC are discussed in Reference 2, Enclosure 5, Section 2.1.4.

The increases in flow rate and temperature that result from EPU, and the associated effect on the FAC monitoring program are discussed in Reference 2, Enclosure 5, Section 2.1.6 and Reference 3, Enclosure 3.

The increases in flow rate that result from EPU and the effects on flow-induced vibration are discussed in Reference 2, Enclosure 5, Section 2.2.2 and in Reference 2, Enclosure 10.

EPU will not reduce material resistance to sensitization for IGSCC.

NRC Question:

2. ldentify the RCPB piping and safe-end components that are susceptible to intergranular stress-corrosion cracking (IGSCC). Discuss any augmented inspection programs that have been implemented and the adequacy of the augmented inspection programs in light of the EPU.

NMC Response:

As stated in Question 1 above, all subject welds at MNGP meet NUREG-0313, Rev. 2, Category "A" and are considered resistant to IGSCC. Therefore, there are no augmented inspection programs implemented at Monticello. The current Inservice Inspection (ISI) program examinations are adequate given the configuration and degradation mechanisms present.

NRC Question:

3. ldentify all flawed components including overlay repaired welds that have been accepted for continued service by analytical evaluation based on the American Society for Mechanical Engineers,Section XI rules. Discuss the adequacy o f such analyses considering the effect of the EPU on the flaws.

NMC Response:

No weld overlays have been installed to mitigate flaws within the reactor coolant pressure boundary. Furthermore, no ASME flaw evaluations have been performed on components within the reactor coolant pressure boundary as a result of indications discovered during IS1 examinations.

Page 3 of 7

Enclosure I NRC Question:

4. Identify the mitigation processes being applied at Monticello to reduce the RCPB component's susceptibility to IGSCC, and discuss the effect@) of the requested EPU on the effectiveness of these mitigation processes. For example, if hydrogen water chemistry (HWC) was applied at the plant, it would be necessary to perform the electrochemical potential measurements at the most limiting locations to ensure that the applied hydrogen injection rate is adequate to maintain the effectiveness of HWC (since oxygen content in the coolant is expected to increase due to the increased radiolysis of water from extended power uprate).

NMC Response:

As discussed in the response to Question I , the NRC SE regarding MNGP's response to Generic Letter 88-01 (Reference I ) , noted that all welds in the reactor coolant pressure boundary are Category " A . Mitigation by use of resistant materials is not impacted since EPU does not impact pressure boundary material properties as defined by Section 2.1.1 of NUREG-0313, Rev. 2.

Appendices A, B, and C of Reference 1 describe the RCPB welds that were solution heat treated or were stress improved using the induction heating stress improvement (IHSI) process. Corrosion resistant cladding (CRC) was applied to the internal surfaces of the welds in the Head Spray Nozzles and the Head Vent Nozzles. The flow, pressure, temperature and mechanical loading for most of the RCPB piping systems do not increase for EPU. Consequently, there are no changes in stress. Construction processes such as solution heat treatment, lHSl or CRC are not impacted by EPU.

Hydrogen Water Chemistry (HWC) is used at MNGP. The HWC system reduces the susceptibility of RCPB components to IGSCC in the primary system piping and improves the resistance to IGSCC in vessel internal components. The implementation of HWC further reduces the probability of degradation of pressure boundary welds to environmental effects. The HWC system was installed in accordance with the recommendations of the BWR Owners Group, "Guidelines for Permanent BWR Hydrogen Water Chemistry Installation - 1987 Revision." MNGP is a Category 2 plant using moderate HWC. Category 2 plants use the BWRVIP-112 (BWR Vessel and lnternals Project, BWR Vessel and lnternals Application (BWRVIA) Version 2.0 for Radiolysis and ECP Analysis) model to estimate the total oxidant and electrochemical potential (ECP) at various locations. Hydrogen injection rates will be increased to maintain hydrogen concentration in feedwater at a constant level and maintain ECP within acceptable limits. ECP is verified at Monticello to be <-330 mV SHE, which provides margin to the IGSCC mitigation value of -230 mV SHE. These actions will ensure that EPU will not affect the water chemistry controls used for IGSCC mitigation.

ECP probes have not been used in the recent past within MNGP's RCPB. However given the materials of construction, the use of solution heat treatment, IHSI, or CRC, and the presence of HWC with ECP verified by BWRVIA modeling, adequate mitigation Page 4 of 7

Enclosure 1 processes are in place to ensure continued acceptable performance of the RCPB welds.

Questions from the remaining references:

A. Hope Creek, ML070460243, dated 2123107,lst page of attachment 5.1 ldentify the materials of construction for the reactor coolant pressure boundary (RCPB) pipinglsafe-ends. Discuss and explain the effect of the requested power uprate on the RCPB pipinglsafe-end materials.

See the response to Question 1 above.

5.2 ldentify the RCPB pipinglsafe-end components that are susceptible to intergranular stress corrosion cracking (IGSCC). Discuss any augmented inspection programs that have been implemented and the adequacy of the augmented inspection programs in light of the EPU.

See the response to Question 2 above.

5.3 ldentify all flawed components including overlay repaired welds that have been accepted for continued service by analytical evaluation based on American Society of Mechanical Engineers (ASME),Section XI rules.

Discuss the adequacy of such analysis considering the effect of the EPU on the flaws.

See the response to Question 3 above.

5.4 ldentify the mitigation processes being applied at Hope Creek to reduce the RCPB component's susceptibility to IGSCC, and discuss the effect of the requested EPU on the effectiveness of these mitigation processes. For example, i f hydrogen water chemistry (HWC) was applied at the plant, it would be necessary to perform the electrochemical potential measurements at the most limiting locations to ensure that the applied hydrogen injection rate is adequate to maintain the effectiveness of HWC since oxygen content in the coolant is expected to increase due to increased radiolysis of water resulting from extended power uprate.

See the response to Question 4 above.

B. Vermont Yankee, ML033640138, dated 12130103, page 5 of attachment

1. Section 3.5.1 of Attachment 4 of your submittal dated September 10, 2003, provides the results of the structural evaluation of the reactor coolant pressure boundary (RCPB) piping. Provide the basis for the disposition of the first system listed in this section.

Page 5 of 7

Enclosure 1

2. ldentify the materials of construction for the Reactor Recirculation System piping and discuss the effect of the requested EPU on the material.

If other than type "A" (per NUREG 0313) material exist, discuss augmented inspection programs and discuss the adequacy of augmented inspection programs in light of the EPU.

See the response to Questions 1 and 2 above.

3.Section XI of the American Society of Mechanical Engineers (ASME)

Code allows flaws to be left in service after a proper evaluation of the flaws is performed in accordance with the ASME,Section XI rules. Indicate whether such flaws exist in the Reactor Recirculation System piping and evaluate the effect of the EPU on the flaws.

See the response to Question 3 above.

4. Discuss flaw mitigation steps that have been taken for the RCPB piping and discuss changes, if any, that will be made to the mitigation process as a result of the EPU.

See the response to Question 4 above.

C. Browns Ferry, ML043440045, dated 12/30/04, questions 1-4

1. Explain why the reactor coolant pressure boundary (RCPB) piping materials are not affected by the power uprate.

See the response to Question 1 above.

2. Identify the materials of construction for the Reactor Recirculation System piping and discuss the effect of the requested extended power uprate (EPU) on the material. If other than type "A" (per NUREG 0313) materials exist, discuss any augmented inspection programs and discuss the adequacy of augmented inspection programs in light of the EPU.

See the response to Questions 1 and 2 above.

3. Section XI of the American Society of Mechanical Engineers (ASME)

Code allows flaws to be left in service after a proper evaluation of the flaws is performed in accordance with the ASME,Section XI rules. lndicate whether such flaws exist in the Reactor Recirculation System piping and evaluate the effect of the EPU on the flaws.

See the response to Question 3 above.

Page 6 of 7

Enclosure I

4. Discuss flaw mitigation steps that have been taken for the RCPB piping and discuss changes, if any, that will be made to the mitigation process as a result of the EPU.

See the response to Question 4 above.

References:

1. USNRC Letter to NSP, "Monticello Nuclear Generating Plant - Staff Evaluation of Response to Generic Letter 88-01 (TAC No. 69146)," dated December 7, 1989.

2. NMC Letter to USNRC, "License Amendment Request: Extended Power Uprate," dated March 31, 2008.

3. NMC Letter to USNRC, "Monticello Extended Power Uprate (USNRC TAC MD8398): Acceptance Review Supplemental Information," dated May 28,2008.

Page 7 of 7 to L-MT-08-043 Electrical Engineering Branch Questions and Responses

Enclosure 2 NRC Question:

I. Provide the staff with the USAR section number that describes the AC load Study.

NMC Response:

The AC load study is described in Monticello USAR Section 8.1 0, "Adequacy of Station Electrical Distribution System Voltages."

NRC Question:

2. The licensee will provide statements that the margins discussed in the acceptance review response for the batteries will be met during the development of the modifications.

NMC Response:

In Reference 2, Enclosure 4, NMC reported the following with respect to DC battery capacity margins at Current Licensed Thermal Power (CLTP) and Extended Power Uprate (EPU) conditions:

Table 1 - Battery Margin CLTP (% Battery Margin) EPU (% Battery Margin) 125 VDC Division I Battery 10.50 15.83 250 VDC Division I Battery 23.63 20.64 125 VDC Division II Battery 20.24 26.58 1 250 VDC Division II Battery ( 2.04 8.19 1 Expected EPU electrical modifications that could impact DC loads are replacements in kind for instrument and control loads on the 125 VDC system. The additional 125 VDC loads due to these EPU modifications will not reduce the reported 125 VDC battery margin by more than five percent of the calculated capacity reported. For example, the EPU modifications will be controlled such that the remaining 125 VDC Division I battery margin is at least 10.83 percent.

Additionally, no changes to the margin for the 250V DC battery loads will result from EPU modifications.

Page 1 of 7

Enclosure 2 NRC Question:

3. For the EQ analyses, clearly state that it has been completed and that NMC has identified the equipment that is impacted by EPU conditions.

NMC Response:

The Reference 1 supplemental EPU submittal provided additional details regarding the EPU analyses relative to the qualification of electrical equipment outside containment.

The information provided in that submittal is the result of analyses that have been completed and documented in EPU task reports and supporting calculations.

Equipment impacted by EPU conditions was identified in the Reference 1 tables. A "Note" was added at the end of the tables to identify equipment impacted by EPU conditions. The note stated, "Additional supporting analysis to be performed and documented in EQ qualification file andlor equipment to be replacedlmodified prior to EPU implementation."

The Reference 1 table note was not intended to imply that the analyses necessary to identify the equipment impacted by EPU conditions have not been completed. As stated previously, and confirmed here, the necessary evaluations to identify the EQ equipment impacted by EPU conditions have been completed. The note explains that the process for final resolution of the identified EPU impact may include:

additional equipment-specific analysis to be documented in the equipment-specific qualification file, replacement or modification of a specific piece of equipment.

The process for final resolution of the identified EPU impacts (additional equipment-specific analysis, replacement or modification) is controlled in accordance with the Monticello EQ Program requirements.

A summary was included in the Reference 1 submittal. The summary concludes that analyses to determine the EPU impact are complete. It also states that the equipment-specific resolutions will be completed as controlled by the Monticello EQ Program requirements. Final resolution of identified impacts will be documented in the related equipment-specific qualification file prior to implementation of EPU in accordance with 10 CFR 50.49.

Page 2 of 7

Enclosure 2 NRC Question:

4. The licensee states the SBO analysis has been revised for EPU conditions, but does not explain what the changes are. The licensee agreed to develop a table that outlines the changes in the SBO analysis from CLTP to the EPU. The table should include the standard acceptance criteria as well as changes in assumptions.

NMC Response:

The following two tables (Table 2 and 3) capture the SBO analysis changes in regard to acceptance criteria and analysis assumptions.

Table 2 - Station Blackout (SBO) 10 CFR 50.63 Acceptance Criteria Criteria CLTP EPU Result SBO 10 CFR 50.63 criteria Yes - Yes - Adequate Core Cooling is met: reactor water reactor water maintained for the level cycles level cycles calculated coping duration "Sufficient capacity and between between (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) capability to ensure that the low-low level low-low level core is cooled and (-47") and (-47") and appropriate containment the High the HPCI trip integrity is maintained in Pressure setpoint the event of a SBO for the Coolant (+48"), which specified duration." Injection is well above (HPCI) trip the Top of setpoint Active Fuel

(+48"), which for the event is well above duration.

the Top of Active Fuel for the event duration.

Page 3 of 7

Enclosure 2 Table 2 - Station Blackout (SBO) 10 CFR 50.63 Acceptance Criteria SBO 10 CFR 50.63 criteria I Drywell 1 Drywell I Peak containment met: Pressure pressure parameters are all within 34.5 psia 42.8 psia design values for the "Sufficient capacity and calculated coping duration capability to ensure that the Suppression Suppression (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) core is cooled and Chamber Chamber appropriate containment Pressure Pressure integrity is maintained in 34 psia 41.3 psia the event of a SBO for the specified duration." Drywell shell Drywell temperature airspace 271.5"F temperature 268.4"F (shell temperature bounded is by the gas temperature value)

Suppression Suppression Chamber Chamber airspace airspace temperature temperature 134°F 178.9"F Peak Peak Suppression Suppression Pool Pool Temperature Temperature 151.2 175.5 Page 4 of 7

Enclosure 2 3 - Changes to SBO Assumptions from CLTP to EPU Tab~ l e Assumption CLTP EPU Comment - ~

Initial Reactor 1904 MWt with 2004 MWt with The CLTP power of 1904 MWt Power 100 day decay GE 14 End of was intended to bound the first heat Cycle (EOC) (24 EPU target power level for month) decay CLTP, which was ultimately heat reduced to 1775 MWt.

The EOC decay heat is conservative and bounds the NUMARC 87-00 guidance of 100 days for the decay heat parameter.

Analytical MAAP 3.OB The suitability of SHEX-O6A for Method Station Blackout containment response applications is addressed in the Monticello EPU License Amendment Request (LAR) (Reference 3),

Enclosure 1.

Occurs due to a Preset to occur This preset MSIV signal adds Isolation Valve Group 1 Primary a t t = O conservatism to the (MSIV) closure Containment containment response because signal Isolation of the increased heat transfer (Reactor Water to the torus.

Low-Low Level) as the This assumption change is simulation consistent with NUMARC 87-00 progresses (at Section 2.4.2 guidance for the approximately plant-specific response to MSIV 28 seconds). closure.

Recirculation 70 gpmlpump 18 gpmlpump The CLTP value of 70 gpm per Pump Leakage pump was a conservative generic leakage value that had been developed as part of the Monticello response to NU REG-0737 on recirculation 1 pump seal leakage. See MNGP USAR Section 4.3.2.2.5.

I The value of 18 gpm is recommended by Appendix J of I

I HPCl only Page 5 of 7

Enclosure 2 Table 3 - Changes to SBO Assumptions from CLTP to EPU Assumption CLTP EPU Comment HPCl suction The analytical The model has The existing SBO model is a

, source model assumes been changed conservative simplification that modeling and a torus only to more does not account for HPCl via operator HPCl suction accurately the CST as the preferred actions source reflect automatic injection system. Plant Condensate emergency procedures direct Storage Tank the operator to use the CST (CST) to torus HPCl suction sources if HPCl suction available.

transfers.

Adequate CST inventory is available, and CST suction lowers overall SBO risk as the net positive suction head (NPSH) available to the HPCl pump is greater than with torus suction and improves HPCl reliability.

This approach is consistent with the NUMARC 87-00 guidance on CST usage during an SBO given in Sections 4.2.1 and 4.3.1(5).

Automatic Low-low reactor Low-low reactor A reactor low-low level signal is HPCl initiation level level or high assured in an SBO event signal drywell pressure regardless of the non-SRV (safety relief valve) reactor coolant inventory loss assumptions. The high drywell pressure automatic initiation HPCl signal is also possible and is sensitive to increased non-SRV reactor coolant inventory loss assumptions.

The non-SRV reactor coolant inventory loss assumptions are artifacts of the SBO methodology, and the actual non-SRV reactor coolant inventory loss is likely to be less.

I The analysis now includes Page 6 of 7

Enclosure 2 Table 3 - Changes to SBO Assumptions from CLTP to EPU Assumption CLTP EPU Comment separate cases for each automatic HPCl initiation signal to conservatively account for both outcomes. The low-low reactor level HPCl start response case causes a more severe containment response, and this response is used in determining the torus temperature margin. The high drywell pressure HPCl start response case causes an extra HPCl loading cycle (greater DC loads) and is used in determining battery margin.

References:

1. NMC Letter to USNRC, "Monticello Extended Power Uprate (USNRC TAC MD8398): Acceptance Review Supplemental Information Package 5," dated June 5,2008

2. NMC Letter to USNRC, "Monticello Extended Power Uprate (USNRC TAC MD8398): Acceptance Review Supplemental Information," dated May 28,2008

3. NMC Letter to USNRC, "License Amendment Request: Extended Power Uprate," dated March 31, 2008 Page 7 of 7