ML040420418

Responses to NRC Clarification Questions on Responses to Requests for Additional Information Regarding License Amendment Request 195, Stretch Power Uprate

ML040420418
Person / Time
Site:	Kewaunee
Issue date:	01/30/2004
From:	Coutu T Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-04-012, TAC MB9031
Download: ML040420418 (51)

Text

NMmmittedto~uC<>- ' Kewaunee Nuclear Power Plant Operated by Nuclear Management Company, LLC January 30, 2004 NRC-04-012 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 KEWAUNEE NUCLEAR POWER PLANT DOCKET 50-305 LICENSE No. DPR-43 Responses To NRC Clarification Questions On Responses To Requests For Additional Information Regarding License Amendment Request 195, Stretch Power Uprate For Kewaunee Nuclear Power Plant (TAC NO. MB9031)

References:

1) Letter NRC-03-057 from Thomas Coutu to Document Control Desk, "License Amendment Request 195, Application for Stretch Power Uprate for Kewaunee Nuclear Power Plant," dated May 22, 2003.

2) Letter from Thomas Coutu to Document Control Desk, "Responses to Requests for Additional Information and Supplemental Information Regarding License Amendment Request 195, Stretch Power Uprate For Kewaunee Nuclear Power Plant," dated November 5, 2003.

3) Letter from Thomas Coutu to Document Control Desk, "Responses To NRC Clarification Questions On Responses To Requests For Additional Information Regarding License Amendment Request 195, Stretch Power Uprate For Kewaunee Nuclear Power Plant," dated December 15, 2003.

4) Letter from John Lamb (NRC) to Thomas Coutu (NMC), Kewaunee Nuclear Power Plant - Review Of License Amendment Request No.

195, Stretch Power Uprate (TAC NO. MB9031)," dated January 26, 2004 In accordance with the requirements of 10 CFR 50.90, Nuclear Management Company, LLC (NMC) submitted license amendment request (LAR) 195 (Reference 1) for a stretch power uprate of six percent. The stretch power uprate would change the operating license and the associated plant Technical Specifications (TS) for the Kewaunee Nuclear Power Plant (KNPP) to reflect an increase in the rated power from 1673 MWt to 1772 MWt.

N490 Highway 42 Kewaunee, Wisconsin 54216-9511 Telephone: 920.388.2560

Docket 50-305 NRC-04-01 2 January 30, 2004 Page 2 On November 5, 2003 and December 15, 2003, NMC responded to requests for additional information from the Nuclear Regulatory Commission (NRC) regarding the proposed stretch power uprate (References 2 and 3). Subsequent to NMC's response, the NRC staff has requested clarification on some of the responses submitted in NMC's November 5th and December 1 5 th letters. This letter and enclosure contains the NMC responses to the NRC requests for clarification.

Enclosure 1 contains the clarification questions from the NRC. These questions were derived from emails and conference calls with the NRC staff. Enclosure 2 contains NMC's response to these clarification questions. Enclosure 3 is Main Steam piping velocities. Enclosure 4 contains the list of regulatory commitments with updated due date status and one additional commitment. Enclosure 5 contains NMC's response to the NRC staffs' questions contained in reference 4.

These responses do not change the Operating License or Technical Specifications for the KNPP, nor do they change any of the proposed changes to the Operating License or Technical Specifications in reference 1. The responses do not change the no significant hazards determination, the environmental considerations, the requested approval date, or the requested implementation period originally submitted in reference 1.

In accordance with 10 CFR 50.91, a copy of this letter, with attachments, is being provided to the designated Wisconsin Official.

Summary of Commitments This letter makes the following new commitments:

1) NMC will provide NRC with a status update and schedule for resolution of GL 96-06 water hammer issues by April 2, 2004.

If there are any questions or concerns associated with this response, contact Mr. Gerald Riste at (920)388-8424 I declare under penalty of perjury that the foregoing is true and correct.

Executed on January 30, 2004.

Thomas Coutu Site Vice-President, Kewaunee Plant Nuclear Management Company, LLC GOR

Docket 50-305 NRC-04-01 2 January 30, 2004 Page 3

Enclosures:

(5) cc:- Administrator, Region ll, USNRC Project Manager, Kewaunee Nuclear Power Plant, USNRC Senior Resident Inspector, Kewaunee Nuclear Power Plant, USNRC Electric Division, PSCW

ENCLOSURE 1 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 30, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

Responses to NRC Clarification Questions on Responses to Requests for Additional Information Regarding LAR 195 NRC Clarification Questions on Responses to Requests for Additional Information 4 Pages Follow

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 1 NRC EMEB Branch Questions

1. Clarify the pipe stress Table 1 in RAI response #55 - Explain why the "Loading Condition" for the Condensate System is "Sustained + Thermal" and the loading condition for Feedwater, Bleed Steam, and Heater Drains is only "Thermal."

2. For RAI question 57 on pump and valve programs, provide more details on how NMC evaluated that the valves were acceptable for uprated conditions, giving examples of typical evaluations for different pump or valve types (e.g., relief valves, MSIVs). Also briefly indicate; 1) why there are no relief valve setting changes or pump change-outs, 2) whether the MSIV has a minimum close time limit, if the increased steam flow could reduce the MSIV closure time, and why the design and operating experience with the MSIV supports a finding that the closure time will not be affected, 3) whether the ambient temperature increase has been evaluated for the GL 95-07 pressure locking and thermal binding issue, and 4) whether Regulatory Commitment #16 on MOV temperature effects had been completed

3. Based on the environmental qualification (EQ) calculations submitted in NMC's December 15, 2003 letter, what is the licensee's overall conclusion concerning the results.

4. RAI question 13 questioned the acceptability of the overdutied circuit breakers on KNPP Buses 1-1 through 1-4. Enclosure J of the November 5 2003 was an evaluation performed on this condition. The NRC staff request information associated with any prior NRC review of this condition and impact of the power uprate on the evaluations conclusions.

5. RAI questions 10 and 16 are associated with the operation of the feedwater and condensate pumps under reduced voltage conditions and into the motors service factor rating. In the November 5, 2003 and December 15, 2003 responses it appears to the staff that the IEEE 666 standard and NEMA MG-1 are being inappropriately combined.

The staff requested additional information for supporting evidence that the overloading of these motors based on their service factors wouldn't affect the operation of these motors. The licensee responded to the NRC staff's RAI but the staff still requires further clarification. Describe for the staff the affect of the power uprate (e.g. reduced voltage and use of the pumps service factor) on the probability of a plant trip or transient.

NRC SPLB Branch Questions

1. For BOP systems, the evaluations in Section 4 of the Licensing Report (BOP Interface by Westinghouse) and Section 8 (BOP system evaluations by SWEC) do not have comprehensive, integrated conclusions.

A. Provide specific conclusions for BOP systems (focus on MS, Condensate, FW, AFW, and TG) addressing the impact of uprated flows, pressures, temperatures, etc.

B. Why does the Westinghouse design criteria for MS safety valves having 100%

normal steam flow capacity not have to be met?

C. Justify how we can operate the FW heaters beyond their design rating and summarize the inspection results.

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 2 D. Explain why the bolting changes were required for the HP and LP turbines and conclude why the upgraded bolts are acceptable.

E. Explain the regulatory basis of the FAC program.

F. Explain the basis of the FW control valve Cv.

2. The Licensee's response to RAI question 4 (November 5, 2003 Letter) stated that the requirements associated with borating the RCS to cold shutdown, utilizing the RWST, with and without letdown in service is being evaluated for any required procedure changes. This evaluation, and update of any required procedures, will be completed prior to implementation of the SUR. This is considered part of our Regulatory Commitment #12. The NRC staff needs to know the results of this evaluation before it can issue a safety evaluation on the acceptability of this amendment request. Please submit a summary of the results of this evaluation.

3. The NRC staff has some additional questions concerning the stretch power uprate.

A. From WCAP 16040-P, page 8-7 "Most calculated MS flow velocities at power uprate do not exceed the previously accepted recommended maximum values for continuous operation." What are calculated and acceptable values?

B. How did/will the licensee ensure new Feed Reg Valve (FRV) Trim is acceptable?

C. How does the licensee ensure sufficient NPSH for the Feedwater pumps? How much margin does the licensee expect in pump suction pressure?

D. Is the Feedwater Isolation Valve (FWIV) acceptable for Stretch Uprate containment response?

E. How is the KNPP Flooding Analysis impacted by the power uprate?

F. Describe the affect of the power uprate on the KNPP missile analysis.

G. Describe the affect of the power uprate on the Turbine overspeed trip settings (reference WCAP 16040-P, p8-12).

H. From WCAP 16040-P, the licensee stated, "The exhaust end loading at the uprated conditions increases. At this loading the exhaust flow guide bolting may be affected.

If desired, upgraded bolting can be installed." What actions did the licensee do regarding turbine exhaust flow guide bolting?

4. In the aftermath of the Three Mile Island, Unit 2, accident the NRC requested, as a part of NUREG 0737, an evaluation of the auxiliary feedwater system. Please describe the basis for the present condensate storage tank level found in KNPP technical specifications and describe how KNPP conforms to the Standard Review Plan, section 10.4.9, branch technical position ASB10-1, item B.1

5. Describe the results of the impact of the power uprate on the GL 96-06 results NRC SPSB Branch Questions

1. Please describe the assumptions made regarding the behavior of drops formed by the break flow in high energy line break calculations done for structural analysis (Section 8.5

I1 Docket 50-305 NRC-04-012 January 30, 2004 Enclosure 1, Page 3 of WCAP 16040-P) and equipment qualification calculations (Section 8.9 of WCAP 16040-P).

2. What single failure assumptions are made in the analyses to support TS 3.3.c.1.A.3?

3. Please verify that the assumption that the break mitigation hardware reduces the break discharge to 50% of the no load flow in the cubicle pressurization analyses is consistent with the current Kewaunee licensing basis. If it is not, please provide a justification for this assumption NRC SRXB Branch Questions

1. In the November 5, 2003 response, number 34, NMC stated that calculations demonstrate that even if upper plenum injection is established as late as 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> after the LOCA (SBLOCA), the vessel boron concentration will still be 4 weight % under the boron precipitation point, assuming a saturation pressure of 35 psia. Other nuclear plants have stated that they will reach the boron saturation point in approximately 6-9 hours (Adams Accession NO. ML032720538). Please discuss this inconsistency and any actions necessary to ensure boric acid precipitation is not a concern.

2. The NRC requests additional information concerning the elimination of the DNBR rod bow penalty for application to KNPP Cycle 26. Specifically, the NRC requests detailed information be provided about the way the FDH burndown was calculated during the cycle and the way the corresponding DNBR credit was calculated.

NRC SPSB-C Branch Question

1. With regard to the waste gas decay tank rupture and volume control tank rupture analyses, you state that in absence of guidance in RG 1.183, you have assumed that the offsite dose acceptance criterion is the same as given for the fuel handling accident (i.e.,

6.3 rem TEDE). However, in the staff's safety evaluation (issued March 17, 2003) for implementation of the alternative source term, this acceptance criterion was not used by the NRC staff in its finding. Kewaunee was originally licensed with a site boundary dose acceptance criterion of 500 mrem whole body for these accidents. Branch Technical Position ETSB 11-5, "Postulated Radioactive Releases Due to a Waste Gas Tank System Leak or Failure," which is located in SRP Section 11.3, "Gaseous Waste Management Systems," also presents a dose acceptance criterion of 500 mrem whole body. The curie content of these tanks is controlled so that 10 CFR 20.1301 dose limits to individual members of the public are not exceeded in the event of a release.

A. Please explain how your assumed acceptance criterion of 6.3 rem TEDE was chosen.

B. Please explain how your assumed dose acceptance criterion of 6.3 rem TEDE for these accidents is consistent with Kewaunee's original licensing and design basis, and Branch Technical Position ETSB 11-5, or why the difference should be found acceptable.

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 4 C. Please explain how the assumed accident dose acceptance criterion interacts with control of radioactivity in the tanks and the 10 CFR 20 dose limits for individual members of the public.

NRC EEIB Branch Question:

1. NMC's December 15, 2003 response to NRC item 9 described the setpoint methodology used for the KNPP. The NRC staff requires additional information associated with how channel operability is determined

ENCLOSURE 2 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 30, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

NMC Responses to NRC Clarification Questions on Responses to Requests for Additional Information Regarding LAR 195 NMC Responses to NRC Clarification Questions 26 Pages Follow

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 1 NRC EMEB Questions

1. Clarify the pipe stress Table 1 in RAI response #55 - Explain why the "Loading Condition" for the Condensate System is "Sustained + Thermal" and the loading condition for Feedwater, Bleed Steam, and Heater Drains is only "Thermal."

Response

Per USAS B31.1.0-1967 piping code, applicable to KNPP BOP piping systems, the following applies to the allowable stress for thermal expansion (Sa):

Sa = f (1.25 Sc + 0.25 Sh) equation (1) from code section 102.3.2 or, when considering the sustained stress margin per Section 102.3.2(d)

Sa = f (1.25 Sc + 0.25 Sh + Sh - Ssus) = 1.25(Sc + Sh) - Ssus Where: Sc = material allowable stress at minimum cold temperature Sh = material allowable stress at maximum hot temperature f = stress range reduction factor (equal to 1.0 in this case)

Ssus = sum of the calculated sustained stresses For the calculated thermal expansion stress (SE) to be acceptable, the following condition has to be met:

SE < Sa, where Sa is determined with or without the sustained stress margin When considering the sustained stress margin, the following applies:

SE < 1.25(Sc + Sh) - Ssus or shown differently, SE + Ssus < 1.25 (Sc + Sh)

For the Feedwater, Bleed Steam, and Heater Drains piping systems, the SE stress levels were all less than Sa (without need for consideration of the sustained stress margin), and were therefore acceptable under the Thermal Loading Condition alone. For the condensate piping, the calculated thermal expansion stress was slightly above the Sa value without including the sustained stress margin, but was less than the Sa value, and therefore acceptable, when the sustained stress margin was included. Hence, for condensate, the condition showing compliance to the Sustained + Thermal Loading Condition (SE + Ssus) was shown in Table 1.

Docket 50-305 NRC-04-012 January 30, 2004 , Page 2

2. For RAI question 57 on pump and valve programs, provide more details on how NMC evaluated that the valves were acceptable for uprated conditions, giving examples of typical evaluations for different pump or valve types (e.g., relief valves, MSIVs). Also briefly indicate; 1) why there are no relief valve setting changes or pump change-outs, 2) whether the MSIV has a minimum close time limit, if the increased steam flow could reduce the MSIV closure time, and why the design and operating experience with the MSIV supports a finding that the closure time will not be affected, 3) whether the ambient temperature increase has been evaluated for the GL 95-07 pressure locking and thermal binding issue, and 4) whether Regulatory Commitment #16 on MOV temperature effects had been completed

Response

As indicated previously, the KNPP motor operated valve (MOV) and air operated valve (AOV)

Programs assume the maximum system design pressures, temperature, and expected valve differential pressures. The maximum expected valve differential pressure includes the momentum pressure associated with the maximum system flow capability.

In the current programs, valve differential pressure (AP) is based on either the system safety valve setting or the total shutoff head of the pumps in the system and includes the momentum pressure of the flow through the valve. The system flow capacity is based on the highest flow capable through the system valves or the capacity of the pumps in the system.

The 6% power uprate makes no changes to the settings of any system relief valves nor does it require change out of any of the pumps within any of the systems. Changes to relief valve settings are not required since the post uprate system operating pressures remain the same or are slightly less than the pre-uprate system operating pressures. Pump changes are not required since the required increase in system flows remain within the original pump design capability. The only valves or piping configuration changes required to be changed as part of the uprate was the main feedwater regulating valves. The trim was changed to increase the Cv to allow passing the required uprate flow with a reduced delta P. Although the flow capability of these valves has increased, the design flow of the feedwater system is based on the flow capacity of the main feedwater pumps which have not been changed.

The MOV Program also takes into consideration any changes in ambient conditions resulting from a high energy line break (HELB) incident. Increases in ambient conditions tend to reduce the available MOV motor torque. The Program MOVs inside containment assume the containment design temperature and pressure which was shown to bound the uprate post accident conditions. The valves outside containment have been reviewed for changes in post uprate HELB temperatures. The review indicates the MOV motor torque will remain adequate following a post uprate HELB incident, however, certain calculations need to be revised. These revisions are currently in progress and as indicated in LAR 195 commitment 16, will be completed prior to implementation of the power uprate (see LAR 195 Commitment 16).

Some specific examples of AOVs evaluated are:

• SD-3A/B, Main Steam PORVs - These valves are evaluated at hot shutdown conditions (1005 psig - 5471F) which do not change due to power uprate. These valves are evaluated assuming a AP of 1050 psid at rated flow.

Docket 50-305 NRC-04-012 January 30, 2004 , Page 3

• AFW-2A/B, Motor Operated Aux Feedwater pump discharge valves - These valves are evaluated based on a maximum AP of 1200 psid with an auxiliary feedwater pump discharge pressure of 1700 psi at rated pump flow less recirculation flow. These conditions do not change as a result of the power uprate.

• RB-V1 - V4, Reactor Building Vent Valves - These valves are evaluated based on the maximum expected system flow at shutdown conditions with a maximum expected AP of approximately 3.9 psid. These shutdown conditions do not change as a result of the power uprate.

• MS-1 A/B, MSIVs - These valves are reverse acting check valves and are evaluated for a maximum AP of 1138 psid, based on the main steam safety valve settings, which are not changing for uprate. The calculated increased AP for the MSIVs associated with the uprate steam flow is approximately 0.4 psid. This small increase in AP and the increased uprate steam flow will tend to enhance the closing of these valves and ensure the required maximum 5 secondl closing time is met at uprate conditions. There is no minimum closing time for the MSIVs. The valves are designed to close under steam line break accident flow conditions that are not impacted by the 6% power uprate.

Some specific examples of MOVs evaluated are:

• MS-1 OOA/B, Steam Generator A/B steam to the turbine driven Aux Feedwater pump isolation- These valves along with the MS-1 02, Turbine driven Aux Feedwater pump main steam isolation, are evaluated based on a maximum AP of 1085 psid with flow conditions associated with a faulted steam generator, main steam line rupture, and a steam generator tube rupture. The 1085 psid AP corresponds to the setpoint of the lowest steam generator safety valve, and the flows are limited by the flow capability of the installed valves, neither of which change due to power uprate.

• AFW 1OA/B, Aux Feedwater Pump discharge crossover valves - These valves are evaluated based on a maximum AP of 1451 and 1453 respectively, based on normal no load steam pressure of 1005 psig. None of the conditions associated with the valve AP change as a result of the power uprate.

• FW-12A/B, Steam Generator A/B feedwater isolation valves - These valves are evaluated based on the maximum AP based on the discharge pressure of the condensate pumps operating with a recirculation flow of 2200 gpm. The power uprate required no changes to the condensate pumps or the recirculation flow rate, therefore the power uprate does not impact the operation of these valves.

As indicated in our previous response, issues such as the valve thermal binding and pressure locking (GL 95-07) and thermally induced over pressurization of water filled piping inside containment (GL 96-06) were determined not to be impacted, because the new post uprate conditions are bounded by existing program boundary conditions. Both programs assume containment design basis pressure and temperature values for post accident conditions. The uprate containment post accident conditions remain bounded by the design temperature and pressure values therefore, these programs remain unaffected.

KNPP USAR, revision 18, section 14.2.5

3 Docket 50-305 NRC-04-012 January 30, 2004 Enclosure 2, Page 4

3. Based on the environmental qualification (EQ) calculations submitted in NMC's December 15, 2003 letter, what is the licensee's overall conclusion concerning the results.

Response

All environmentally qualified (EQ) electrical equipment at KNPP has been evaluated for the resulting environmental conditions resulting from the 6% stretch power uprate and will remain environmentally qualified.

4. RAI question 13 questioned the acceptability of the overdutied circuit breakers on KNPP Buses 1-1 through 1-4. Enclosure J of the November 5, 2003 was an evaluation performed on this condition. The NRC staff request information associated with any prior NRC review of this condition and impact of the power uprate on the evaluations conclusions.

Response

WCAP-1 6040-P NSSS/BOP Licensing Report discusses this on page 8-74, under the heading "4160 VAC Switchgear", it states "There were no new equipment installed or modifications to equipment that changed the short circuit characteristics of the installed equipment at the medium voltage level." This statement means that there is no change in the available three phase or asymmetrical fault current at any plant 4160 Volt bus due to the uprate. The same page says that the bus 1-1 through 1-4 overduty condition "...was previously evaluated and found acceptable." This statement refers to an NRC letter to WPSC dated Jun 2, 1992; see 3.1.1 on page 3, titled "Fault Duty on 4.16kV System".

5. RAI questions 10 and 16 are associated with the operation of the feedwater and condensate pumps under reduced voltage conditions and into the motors service factor rating. In the November 5, 2003 and December 15, 2003 responses it appears to the staff that the IEEE 666 standard and NEMA MG-1 are being inappropriately combined. The staff requested additional information for supporting evidence that the overloading of these motors based on their service factors wouldn't affect the operation of these motors. The licensee responded to the NRC staff's RAI but the staff still requires further clarification. Describe for the staff the affect of the power uprate (e.g. reduced voltage and use of the pumps service factor) on the probability of a plant trip or transient.

Response

The concern here falls under the category of a Condition II event. The KNPP USAR definition (pg. 14.0-2) is "Condition II occurrences include incidents, any one of which may occur during a calendar year for a particular plant." Feedwater System Malfunction is listed as a Condition II incident. Failure of a Main Feedwater or Condensate Pump motor could therefore cause a Condition II event.

KNPP has operated for 29 years. In that time there has been one Main Feedwater Pump motor failure and one Condensate Pump motor failure.

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 5 As you stated in RAI #10, the life of the [motor] insulation is reduced by 50% if the temperature is increased by 8-12 degrees C [above motor nameplate full load temperature]. As noted in our previous response, the condensate pump motor temperature is expected to increase by about 2.3 degrees C and the feedwater pump motor by about 3.9 degrees C due to uprate.

Conservatively assuming these motors are operating at full load temperature continuously prior to uprate, the impact of the added 4 degrees C should be less than a 25% reduction in motor insulation life. Therefore the motors should operate for 29 years x 0.75 = 21.75 years between failure (at worst) after uprate. The motor failure is much less frequent than once a calendar year, therefore such failures would remain Condition II incidents and there would not be a more than minimal increase in the frequency of the Feedwater System Malfunction.

A Probabilistic Risk Assessment (PRA) evaluation was performed on the effect of a 25%

increase in the feedwater pump motor failure rate, which showed that the change in KNPP Core Damage Frequency (CDF) was less than the 1.OE-6 Delta-CDF criteria for a significant risk impact. In addition, there was no change in the calculated Large Early Release Frequency (LERF) due to this increase in motor failure rate.

I Docket 50-305 NRC-04-01 2 January 30, 2004 Enclosure 2, Page 6 NRC SPLB Branch Questions

1. For BOP systems, the evaluations in Section 4 of the Licensing Report (BOP Interface by Westinghouse) and Section 8 (BOP system evaluations by SWEC) do not have comprehensive, integrated conclusions.

A. Provide specific conclusions for BOP systems (focus on MS, Condensate, FW, AFW, and TG) addressing the impact of uprated flows, pressures, temperatures, etc.

Response

Main Steam (MS) - The Main Steam system was evaluated for expected power uprate pressure, temperature, and flow velocity conditions. Power uprate will result in a Main Steam system flow rate increase of approximately 8.4 percent with a steam generator steam outlet operating pressure increase of approximately 7.6 percent. The MS system design pressure of 1100 psia bounds the power uprate operating condition of 797 psia. Additionally, the highest normal operating pressures and temperatures occur at no load conditions, which are not affected by the power uprate. The Main Steam pressures and temperatures associated with power uprate are bounded by the designs of the MS system and its components with margin, and are therefore acceptable.

Most calculated Main Steam system flow velocities at power uprate do not exceed the previously accepted recommended maximum values for continuous operation. The exception is that the calculated flow velocities through the 18- and 8-inch diameter lines to the steam dump valves exceed the recommended maximum value (Stone & Webster standard practice) at current and power uprate conditions. However, since the Steam Dump Systems are operated in an intermittent manner and at conditions less than full power (that is, during plant start up and cooldown, and sudden load reductions), which will not change with the implementation of power uprate, the piping velocities are acceptable for the power uprate. Therefore, the Main Steam system flow velocities at power uprate remain acceptable.

The Main Steam (MS) System is included in KNPP's FAC program. The FAC Program owner has a corrective action item to update the FAC program for all impacted systems, including Main Steam, for the stretch uprate (see LAR 195, Regulatory Commitment Number 2).

Condensate and Feedwater (CD & FW) - A hydraulic flow model of the Condensate and Feedwater Systems was used to evaluate system performance under normal and transient conditions. The model analyses included runs to simulate existing plant conditions, and to evaluate plant power uprate normal and transient conditions. The expected system pressures and temperatures were obtained from the power uprate heat balances and the hydraulic model calculation. These values were compared to the design specifications. The system design pressure envelops the system pressures at power uprate, both during normal operating and transient conditions.

With the exception of the piping from feedwater heater 14 to the FW pumps suction, all piping for power uprate is bounded by the design temperature. For this piping section, power uprate conditions show a temperature of 366 0F, 16 0F above the design temperature of

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 7 3500F. The KNPP Piping Specification, however, allows a maximum temperature of 6500F for the same pipe at a maximum pressure of 400 psi. As this pressure is not exceeded at power uprate conditions, this section of pipe is considered acceptable. This design discrepancy was resolved and documented within the KNPP corrective action program.

The valve trim in the feedwater regulating valves (FRVs) was determined to be undersized to support the required normal and abnormal operating requirements at uprated conditions.

During the R26 refueling outage in April 2003 the trim in the FRVs was changed out as recommended by the uprate evaluation.

The new operating points for the condensate pumps are within design parameters for flow, head, and NPSH. Therefore, the condensate pumps have sufficient head and capacity margin to support the power uprated plant conditions.

The new operating point for the FW pumps is within the design curve limits for flow, head, and NPSH. Therefore, the FW pumps have sufficient head and capacity margin to support the power uprated plant conditions.

The feedwater heater analysis is discussed in a separate question below.

Auxiliary Feedwater (AFW). - No AFW performance changes (pressure, temperature, or flow) are required as a result of the power uprate. Therefore, acceptable AFW flow performance has been demonstrated for the power uprate conditions, since the existing AFW performance does not change.

Higher uprate decay heat during the four hour Station Blackout coping period necessitated an increase in Condensate Storage Tank (CST) usable volume from 39,000 gallons to 41,500 gallons. Therefore, with the required change in the CST inventory, the AFW system is acceptable for uprate. This inventory change was included in the requested change to Tech Spec 3.4.c for the stretch uprate.

Turbine Generator - Siemens Westinghouse, turbine generator original equipment manufacture (OEM), performed the evaluations of the turbine and generator to support a nuclear steam supply system (NSSS) power of 1780 MWt (1772 MWt + 8 MWt reactor coolant pump (RCP) heat). An evaluation of the impact of the 7.4% uprate conditions on the existing turbine missile analysis and torsional analyses showed the existing missile and torsional analyses remain bounding at uprated conditions.

The turbine evaluations identified that the volumetric steam flow for a 7.4% power uprate with a Tave of 5620F and predicted steam generator pressure (715 psig) could not be accommodated by the existing high pressure (HP) turbine nozzle block without modification.

It was determined that increasing Tave to 5720 F would reduce the HP turbine steam volumetric flow requirement to within the capacity of the existing nozzle block and also retain a minimal margin to a valves-wide-open (VWO) condition.

The turbine evaluations contained the OEM recommendation to reduce the maximum allowable mechanical overspeed settings from 111% to 109% and the electronic redundant overspeed trip (ROST and AEH) from 111.5% to 109.5% to ensure the turbine generator design overspeed of 120% is not exceeded at the 7.4% uprated power. In addition, the

Docket 50-305 NRC-04-012 January 30, 2004 Enclosure 2, Page 8 Load Drop Anticipator (LDA) settings would also need to be appropriately revised for the 7.4% uprate conditions.

During the R26 refueling outage in April 2003, the necessary changes to the turbine overspeed instrumentation were implemented. The current mechanical overspeed setting is within this specified range and did not require adjustment.

The OEM recommendations also included changing out selected HP Turbine horizontal flange bolts and low pressure (LP) Turbine coupling bolts to a higher strength material to minimize the potential of a HP Turbine horizontal flange leak and bending of some of the LP Turbine coupling bolts at the uprate power. The required bolting changes were accomplished during the R26 refueling.

The OEM evaluation of the main generator and generator support systems showed the generator and supporting systems are capable of supporting the full 7.4% uprate, provided the generator is operated within the existing generator capability curve and the Service Water temperature to the hydrogen coolers does not exceed 86 0F. Note that the maximum design service water supply temperature is 80 0F. If the generator hydrogen cooling valve is full open, the manual bypass valve may need to be throttled open to augment flow to the Hydrogen Coolers to maintain hydrogen gas temperatures within normal operating limits.

B. Why does the Westinghouse design criteria for MS safety valves having 100%

normal steam flow capacity not have to be met?

Response

The Westinghouse design criteria for the MS safety valves to have 100% rated power steam flow capacity was a conservative criterion for construction of new plants. It was for the original design to provide significant margin to ensure that main steam pressure did not exceed 110% of the steam generator shell-side design pressure. This original design criterion no longer applies to the uprated condition as long as the B31.1 code requirement for protection of the steam generator shell is met. The safety valves provide only approximately 98.6% of the maximum uprated steam flow, however, they continue to meet the over pressure protection requirement (as stated in Sections 4.2.4.1.1 and 6.2 of the Licensing Report).

C. Justify how we can operate the FW heaters beyond their design rating and summarize the inspection results.

Response

Based on feedwater heater drain data taken from heat balances developed for a 7.4%

power uprate, the evaluation identified the potential that the current power (1650 MWt) flow and pressure conditions (rho V squared) associated with the cascading inlet drain lines to some feedwater heaters (11, 12, and 15) exceeded the Heat Exchange Institute (HEI) recommended values. At uprated conditions, only the predicted rhoV2 values for feedwater heater 11 show any appreciable increase. Exceeding the HEI recommended values on these feedwater heaters can potentially result in increased erosion to the inlet nozzles and tubes at the area of the drain inlets. Ultrasonic inspections of the B Loop LP feedwater heaters (11/12B and 13B) and the A and B Loop HP feedwater heaters (14A/B and 15A/B)

I Docket 50-305 NRC-04-01 2 January 30, 2004 Enclosure 2, Page 9 were performed during the R26 refueling outage in April 2003. These inspections did not identify any evidence of abnormal tube erosion in the areas associated with the heater drain inlets. Since the heater inspections did not identify any areas of concern in the feedwater heater drain inlet areas, the feedwater heaters are considered acceptable to support the uprate to 1772 MWt. Additional inspections will be performed following uprate to identify any subsequent tube erosion and degradation (see LAR 195 Commitment Number 4).

D. Explain why the bolting changes were required for the HP and LP turbines and conclude why the upgraded bolts are acceptable.

Response

The turbine generator vendor, Siemens Westinghouse, recommended changing out selected HP Turbine horizontal flange bolts and LP turbine coupling bolts to a higher strength material to minimize the potential of a HP turbine horizontal flange leak and bending of some of the LP turbine coupling bolts at the uprate power. Upgraded bolts were deemed acceptable due to following the vendor's recommendations and installation instructions.

E. Explain the regulatory basis of the FAC program.

Response

NRC Generic Letter 89-08 requested verification that formalized procedures or administrative controls have been implemented to ensure continued implementation of an erosion/corrosion monitoring program. KNPP subsequently committed to initiate and issue procedures and administrative controls for a Pipe Inspection Program. KNPP routinely performs inspections to detect piping degradation caused by Flow Accelerated Corrosion (FAC) or other erosion/corrosion mechanisms. The EPRI software "CHECMATE" is currently used for analyzing pipe inspection data (see LAR 195 Commitment Number 2).

F. Explain the basis of the FW control valve Cv.

Response

To ensure stable feedwater control with constant speed feedwater pumps, the pressure drop across the FRVs at 100% power should be approximately 1.5 times the feedwater system dynamic losses from the feedwater pump discharge through the steam generators. In addition, to meet the requirements for load rejection, adequate margin should be available in the FRVs at full load conditions to permit a condensate and feedwater system delivery of 96% of rated flow with a 100 psi pressure increase above the full load pressure with the FRVs fully open. A hydraulic flow model of the Condensate and Feedwater systems was used to evaluate system performance under normal and transient conditions. This model was used to evaluate the FRV Cv and determine the expected pressure drop across the FRVs under uprated feedwater flow conditions. The analysis used in the hydraulic model indicated that the current valve Cv would be undersized and would not be adequate to support power uprate normal and abnormal operational requirements. A FRV trim with a larger Cv was chosen which met the flow and pressure drop requirements above (see LAR 195 Commitment Number 3).

k -

Docket 50-305 NRC-04-012 January 30, 2004 Enclosure 2, Page 10

2. The Licensee's response to RAI question 4 (November 5, 2003 Letter) stated that the requirements associated with borating the RCS to cold shutdown, utilizing the RWST, with and without letdown in service is being evaluated for any required procedure changes. This evaluation, and update of any required procedures, will be completed prior to implementation of the SUR. This is considered part of our Regulatory Commitment #12. The NRC staff needs to know the results of this evaluation before it can issue a safety evaluation on the acceptability of this amendment request. Please submit a summary of the results of this evaluation.

Response

A calculation has been performed to demonstrate that following the uprate to 1772 MWt, the Appendix R requirement requiring the ability to borate the RCS to the hot shutdown (HSD)

Xenon free minimum technical specification required shutdown margin (SDM) within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> using the RWST as the boration source continues to be satisfied. KNPP assumes letdown can be restored during response to a fire however, the calculation demonstrates that with or without letdown established, the RWST volume required to borate the RCS to the required shutdown margin can be added well within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. In addition, the calculation demonstrates that if letdown is not established, sufficient Pressurizer volume exists to allow the required RWST volume to be added and still remain within the available Pressurizer level indication. No changes to the existing Appendix R shutdown procedures are required for the stretch uprate.

3. The NRC staff has some additional questions concerning the stretch power uprate.

A. From WCAP 16040-P, page 8-7 "Most calculated MS flow velocities at power uprate do not exceed the previously accepted recommended maximum values for continuous operation." What are calculated and acceptable values?

Response

The calculated steam velocities are tabulated in Calculation C11386, Main Steam, and are attached. On Tables 5A- 8B, (Enclosure 3) the Stretch Uprate column is labeled as HB1 07.

The recommended maximum value is 250 feet per second, from a Stone and Webster standard practice.

B. How did/will the licensee ensure new Feed Reg Valve (FRV) Trim is acceptable?

Response

Following installation, post maintenance testing showed acceptable performance. This included stroke testing, and pressure drop and valve position monitoring. Current operation has shown that the FRV is performing as designed. Additional monitoring will be performed during the stretch uprate power ascension to ensure performance remains as expected.

C. How does the licensee ensure sufficient NPSH for the Feedwater pumps? How much margin does the licensee expect in pump suction pressure?

Docket 50-305 NRC-04-01 2 January 30, 2004 Enclosure 2, Page 11

Response

The low pressure feedwater heater bypass valve (C-13) opens at a feedwater pump suction pressure of 220 psig to ensure sufficient NPSH. This setpoint remains acceptable for stretch uprate conditions and will not change.

The current and predicted pump values are given in the following table.

Table 3.2.3-3b Feedwater Pump Parameters (Normal Operation)

Pump Design 6% Uprate 7.4% Uprate Parameters Conditions Conditions Conditions Flow (pm) 8,000 8,700 8,800 Head (ft) 2,130 2,090 2,075 NPSHa (t) 373 332 310 NPSHr (ft)105 115 115 BHP 4,800 5,100 5,150 Efficiency 84% 85% 86%

In addition, feedwater pump operation will be monitored during the stretch uprate power ascension to ensure operation remains as expected.

D. Is the Feedwater Isolation Valve (FWIV) acceptable for Stretch Uprate containment response?

Response

The FWIV is credited in the main steam line break (MSLB) accident analyses when the single active failure (SAF) is the failure of a feedwater regulating valve (FRV) to close. The timing assumption for FWIV closure is 90 seconds. For Stretch Uprate, we have shown that the MSLB FRV SAF cases are acceptable at 0%, 30%, 70%, and 100% power, with the most limiting case being from full power.

E. How is the KNPP Flooding Analysis impacted by the power uprate?

Response

The plant internal flooding impact was evaluated for the HELBs as a result of the power uprate.

The flooding due to moderate energy lines was reviewed and was determined not to be impacted, since the moderate energy system operating conditions (pressure and temperature) do not change so as to result in any new HELBs, and the inventory sources for flooding have not increased. Therefore, the internal flooding evaluation was limited to existing HELBs for the impact on the submergence levels identified in the EQ plan. The FW line break outside containment is the bounding break for determining the maximum flood (submergence) levels.

The areas flooded and the associated submergence levels are based on specified break locations determined by the HELB criteria contained in the USAR Appendix 1OA and Fluor Power Services letter KPS-6601, Kewaunee Nuclear Power Plant Submergence Elevations-High-Energy Line Breaks. The break locations for the FW lines are not impacted since the locations of peak pipe stress are not impacted and the physical line arrangements (terminal ends, branch connections, etc.) have not changed. These inputs are not impacted and are

Docket 50-305 NRC-04-01 2 January 30, 2004 Enclosure 2, Page 12 bounded by the existing evaluation. In addition, the maximum flood levels in all compartments, except the basement levels, are controlled by physical plant features such as doors and curb plates. Therefore, the submergence levels in these areas are not impacted by the power uprate.

Results: The flooding (submergence) levels remain the same and are not impacted as a result of the power uprate.

F. Describe the affect of the power uprate on the KNPP missile analysis.

Response

Systems that are located inside and outside the reactor containment are potential targets of internally generated missiles. High-pressure Reactor Coolant System equipment, which could be the source of missiles is suitably protected either by the concrete wall enclosing the reactor coolant loops or by the concrete operating floor to block any passage of missiles to the containment walls. The steam drum, which forms an integral part of the steam generator, represents a mass of steel which provides protection from missiles originating in the section of the containment within the shield wall and below the operating floor. A concrete structure is provided over the RCCA drive mechanisms to block any missiles generated from a hypothetical fracture of an RCCA housing.

The missile shield structures inside Containment were analyzed using the conservation of momentum method. The penetration depth was calculated by the Ballistic Research Laboratories Formula for steel and concrete.

Missile protection is designed to the following criteria:

a) The Reactor Containment Vessel is protected from loss of function due to damage by such missiles as might be generated in a loss-of-coolant accident for break sizes up to and including the double-ended severance of a main coolant pipe (DBA).

b) The Engineered Safety Features Systems and components required to maintain containment integrity are protected against loss of function due to damage by missiles.

During the detailed plant design, the missile protection concepts necessary to meet these criteria were developed. These concepts are:

a) Components of the reactor coolant system are examined to identify and to classify missiles according to size, shape and kinetic energy for purposes of analyzing their effects b) Missile velocities are calculated considering both fluid and mechanical driving forces, which could act during missile generation.

c) The structural design of the missile shielding takes into account both static and impact loads.

d) The Reactor Coolant System is surrounded by reinforced concrete and steel structures designed to withstand the forces associated with double-ended rupture of a main coolant pipe, and designed to stop these missiles.

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 13 The types of missiles for which protection is provided are:

• Valve Stems Valve bonnets Instrument thimbles Various types and sizes of nuts and bolts Complete RCCA drive mechanisms, or parts thereof Piece of pipe Pressurizer valves, instrumentation thimbles and heaters Removable slabs, blocks and partitions were evaluated to determine whether they could become possible missiles. These items were reviewed to assure that the functions of safety related structures and equipment would not be affected during an OBE or tornado.

Removable slabs or plates that were found to be potential missiles are anchored to Class I structures by means of bolts and anchor bolts. Concrete inserts and bolts provided for removing the slabs, are designed with lifting capacities equal to two times their normal design stress capabilities.

Turbine Missiles - It has been determined (Ref. 2) that turbine missiles could cause some localized damage to protective concrete barriers or structures. However, as the maximum turbine overspeed is being maintained below 120% (see Engineering Report section for turbine generator), the effects of a turbine missile after uprate will not result in any significant damage.

The reactor thermal power uprate will not adversely impact the evaluations performed for internally generated missiles or their conclusions. The uprate will not increase the number of postulated missiles that require consideration nor increase the equipment required to safely shutdown the plant following the generation of a missile.

Question F

References:

1. "Kewaunee Nuclear Power Plant Updated Safety Analysis Report," Rev. 17, Section 5.2, Containment System Structure Design

2. "Kewaunee Nuclear Power Plant Updated Safety Analysis Report," Rev. 17, Appendix B9, Turbine Missile Effects G. Describe the affect of the power uprate on the Turbine overspeed trip settings (reference WCAP 16040-P, p 8-12).

Response

Results indicate that with the present trip settings, the unit would overspeed to 121% at the uprated conditions, which slightly exceeds the 120% overspeed design. Therefore, an adjustment to the overspeed trip setting is recommended. The recommended uprate settings as stated in the Siemens-Westinghouse report are as follows:

Docket 50-305 NRC-04-012 January 30, 2004 , Page 14 OVERSPEED CURRENT NEW TRIP SETPOINT SETPOINT MECHANICAL 111% 109%

EHC 111.5% 109.5%

ROST 111.5% 109.5%

The mechanical overspeed trip setting already met the new setpoint requirement and did not require any adjustment. The electro-hydraulic control (EHC) and redundant overspeed trip (ROST) setpoints were adjusted during the April 2003 refueling outage to meet the new setpoint requirements (see LAR 195, Regulatory Commitment Number 6).

H. From WCAP 16040-P, the licensee stated, "The exhaust end loading at the uprated conditions increases. At this loading the exhaust flow guide bolting may be affected. If desired, upgraded bolting can be installed." What actions did the licensee do regarding turbine exhaust flow guide bolting?

Response

Siemens Westinghouse (TG Vendor) issued AIB 8804 which discusses low pressure turbine exhaust flow guide bolt failures due to stress corrosion and fatigue cracking. KNPP has not implemented the prescribed recommendation of Westinghouse's AIB 8804 for the following reason.

Every other refueling outage, an LP turbine is disassembled, during which an evaluation is performed on the top half bolting of the flow guides (which must be removed for disassembly),

and a sample of the bottom half bolting. This evaluation reveals any bolt deterioration due to corrosion and/or erosion that may have occurred during operation, and determines the need for bolt replacement. If an LP is not disassembled during the outage, a crawl-through is performed and the bolts are subject to a visual inspection that may identify the need for bolt replacement. It is much more economical to perform these inspections than to do the bolt replacement modification. KNPP has not had an exhaust flow guide bolt failure, and we are confident that our inspection and maintenance process is satisfactory in minimizing the potential for future bolt failures

4. In the aftermath of the Three Mile Island, Unit 2, accident the NRC requested, as a part of NUREG 0737, an evaluation of the auxiliary feedwater system. Please describe the basis for the present condensate storage tank level found in KNPP technical specifications and describe how KNPP conforms to the Standard Review Plan, section 10.4.9, branch technical position ASB10-1, item B.1

Response

Condensate Storage Tank Inventory KNPP's original Technical Specifications required the condensate storage tanks (CST's) to contain 75,000 gallons of water. The specified minimum water supply in the condensate storage tanks (75,000 gallons) was sufficient for two hours at hot shutdown plus cooling of the

I Docket 50-305 NRC-04-01 2 January 30, 2004 Enclosure 2, Page 15 plant to the cold shutdown condition and maintenance of this condition for an additional four hours. Unlimited replenishment of the condensate storage supply is available from Lake Michigan. Post Three Mile Island, Unit 2, (TMI-2) accident the NRC staff,', acknowledged this CST level requirement in the staffs' description of the KNPP AFW system.

Subsequent to the NRC staff's description of the AFW system at Kewaunee, several technical specification (TS) changes have occurred to the required CST volume. Starting at 75,000 gallons the TS was changed to 30,000 gallons in 1983, then changed to the present requirement of 39,000 gallons in 1993.

KNPP license amendment 4912] changed the CST volume requirement from 75,000 gallons to 30,000 gallons. Wisconsin Public Service Corporation (WPSC) requested 31 a level reduction from 75,000 to 10,000 gallons. The bases for 10,000 gallons was the time required for the operators to switch the auxiliary feedwater (AFW) pumps' suction from the CST's to the service water system supply (i.e. Lake Michigan). In the NRC staff's safety evaluation for amendment 49, the staff agreed with WPSCs rational for the 10,000 gallons but stated that post TMI-2 short and long term recommendations GS-5 and GL-3, identified in NUREG 061 1, require that the AFW system be capable of operation for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a complete loss of all alternating-current (a-c) power sources (what is now called a station blackout]. The TS basis, included with this amendment, stated that this specified minimum water supply in the condensate storage tanks was sufficient for ninety minutes at hot shutdown plus a suitable margin to prevent loss of net positive suction head prior to switching suction to the service water system.

KNPP license amendment 9714] changed the minimum CST volume from 30,000 gallons to 39,000 gallons, the current CST volume requirements. In this SE the NRC staff stated that in a November 20, 1990, Safety Evaluation (SE) addressing Kewaunee's implementation of the Station Blackout (SBO) Rule, the NRC staff recommended a TS amendment to increase the minimum required CST inventory. The licensee determined that a minimum of 39,000 gallons in the CSTs satisfies the 4-hour SBO coping duration requirement to remove heat. Additionally, the KNPP TS basis states that this specified minimum water supply in the condensate storage tanks is sufficient for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of decay heat removal. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> are based on the Kewaunee site-specific station blackout coping duration requirements.

Condensate Storage Tank Response

References:

I'l Letter from Darrell G. Eisenhut (NRC) to Eugene R. Mathews (WPSC), "NRC Requirements for Auxiliary Feedwater Systems at Kewaunee Plant," dated September 21, 1979, enclosure, 1 page 2.

(21Letter from Marshall Grotenhuis (NRC) to C.W. Geisler (WPSC), dated April 28, 1983.

131Letter from C.W. Geisler (WPSC) to H.R. Denton (NRC), "Proposed Amendment No. 51 to the KNPP Technical Specifications," dated December 20, 1982 (41Letter from Allen G. Hansen (NRC) to C.A. Schrock (WPSC), "Amendment NO. 97 to Facility Operating License NO. DPR-43 (TAC NO. M84062)," dated February 1,1993.

Auxiliary Feedwater System requirements NUREG 0800, Standard Review Plan, "Section 10.4.9, "Auxiliary Feedwater Systems," (AFW)

Rev. 2, dated July 1981, contain the NRC acceptance criteria in question for the KNPP AFW system.Section II, item 5.b states:

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 16

11. ACCEPTANCE CRITERIA Acceptability of the design of the auxiliary feedwater system, as described in the applicant's safety analysis report (SAR), is based on specific general design criteria and regulatory guides. Listed below are the specific criteria used in this SRP section as they relate to the AFWS.

5. General Design Criteria 34 and 44, to assure:

b. Redundancy of components so that under accident conditions the safety function can be performed assuming a single active component failure. (This may be coincident with the loss of offsite power for certain events.) Branch Technical Position ASB 10-1 as it relates to AFW pump drive and power supply diversity shall be used in meeting these criteria.

Standard Review Plan Branch Technical Position ASB 10-1, item B.1 states:

The auxiliary feedwater system should consist of at least two full-capacity independent systems that include diverse power sources.

10 CFR 50 Appendix A, "General Design Criteria (GDC) states the following for GDC 34 and 44.

Criterion 34--Residual heat removal. A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

Criterion 44--Cooling water. A system to transfer heat from structures, systems, and components important to safety, to an ultimate heat sink shall be provided.

The system safety function shall be to transfer the combined heat load of these structures, systems, and components under normal operating and accident conditions.

Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

KNPP accident analysis demonstrates that GDC 34 and 44 are met. NMC has performed accident and transient analysis for all events that credit the AFW system. In these analyses the only accident or transient for which unacceptable results are achieved with only one AFW pump is the loss of normal feedwater accident (LONF). The LONF event requires two AFW pumps to achieve acceptable results; this still meets the GDC 34 and 44 requirements.

KNPP's initial license required two AFW pumps to be operable. However, due to scenarios where a depressurization of the steam generator (e.g. feed line break or steam line break (SLB)) were to occur concurrent with a single active failure, three AFW pumps were needed to provide feedwater to the intact steam generator."I In the safety evaluation for the addition of the three AFW pump operability requirement the NRC staff 21 concurred with this justification. This license amendment also added the requirement that if one AFW pump was out-of-service the

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 17 licensee had 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to return the pump to an operable status or the reactor shall be shutdown and cooled below 3500 F using normal operating procedures.

After the TMI-2 accident the NRC issued requirements 3 1 for the auxiliary feedwater system for the Kewaunee Nuclear Power Plant identified during the course of the NRR Bulletins and Orders Task Force review of operating reactors in light of the accident. This letter contains two type of requirements; 1) generic requirements applicable to most Westinghouse-designed operating plants, and 2) plant-specific requirements applicable only to Kewaunee. Included in the letter is recommendation GS-1 (generic, short-term) that stated that the licensee should propose modifications to the Technical Specifications to limit the time that one AFW system pump and its associated flow train and essential instrumentation can be inoperable. The letter recommended 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with a subsequent action time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This letter did not state a basis for this recommendation.

The WPSC responses to this recommendation (GS-1) stated that no proposed modification to the KNPP TS associated with the AFW pumps is warranted. WPSC stated that the AFW system at KNPP is designed in accordance with general design criteria (GDC) 41 among others, the redundant motor driven pumps are supplied power from safeguards buses, two of the three AFW pumps are required to be available, and if two AFW pumps are unavailable for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the plant is required to be shutdown in a shorter time period then the NRC recommended.

Therefore, no action is warranted. The NRC respondedt 51 that this was unacceptable and that WPSC should propose a TS modification. WPSC responded 61 stating the NRC's recommendation to require three auxiliary feedwater pumps to be operable whenever the reactor coolant system is above 3500F is not provided with technical justification and that KNPP's current TS comply with 10CFR50.36 for what needs to be included in TS. WPSC further stated that when technical justification is received it would be reviewed and responded to accordingly.

Subsequently, NUREG 0737, "Clarification of TMI Action Plan Requirements," was issued (October 31, 1980r7) which contains an item related to an evaluation of the AFW system; item II.E.1.1, "AFW System Evaluation." This item included an action to perform a deterministic review of the AFW system using the acceptance criteria of Standard Review Plan Section 10.4.9 and associated Branch Technical Position ASB 10-1 as principal guidance. For the KNPP this item was closed out by an NRC safety evaluation8 (SE) concluding that the licensee's response to short term recommendation, GS-1, was acceptable.

In the NRC's SE for GS-1, two reasons were stated for the justification for requiring all three AFW pumps to be included in the TS. Although not specifically stated it appears these reasons are also the justifications for two full-capacity independent systems that include diverse power sources. These reasons are 9l :

a. An order of magnitude increase in system reliability is noticed for a three pump system versus a two pump system. And with the Kewaunee specification the system must be considered as having only two pumps since a third pump can be down for maintenance for an indefinite period, and

b. If the turbine driven pump is the one not available, then the system will not have power diversity and the plant will not be able to remove decay heat in the event of a complete loss of all AC power sources [station blackout].

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 18 The KNPP TS were modified to include three AFW pumps but not for the same reasons the NRC stated. Wisconsin Public Service Corporation (WPSC), because of concerns identified during an NRC vital area analysis, agreed'll to modify KNPP TS to include all three AFW pumps, which was completed when KNPP TS amendment 45 was implemented. The basis for the three AFW pumps was stated in the NRC's SE1 '1 as being necessary to meet the circumstances following a rupture of a steam pipe situation. Subsequently, in the NRC's letter dated December 16, 1982, the NRC closed out generic short-term recommendation GS-1.

To ensure adequate flow capability for a station blackout the turbine driven AFW pump was required to be full capacity, which is true for all accident analyses for the present KNPP operating power level. At uprated power (1772 MWth) the only accident where the turbine driven AFW pump is not full capacity is the loss of normal feedwater (LONF) event. Additionally, if the reactor coolant pumps, which are assumed to supply 8 MWth to the reactor coolant system, were tripped, the turbine driven AFW pump would supply adequate flow to achieve acceptable results in the LONF analysis. For the station blackout (SBO) analysis, the rational given for the power diversity (i.e. the turbine driven AFW pump has full capacity) is still met for uprated power. In a SBO event the reactor coolant pumps are tripped.

Therefore, at 1772 MWth, the KNPP TS still requires three AFW pumps to be operable, meets the requirements of GDC's 34 and 44. Additionally, for events where the reactor coolant pumps are tripped, such as station blackout, KNPP two full capacity systems with diverse power sources, meeting the intent of the standard review plan branch technical position guidelines for diverse power sources.

Auxiliary Feed Water System

References:

I Letter from E.R. Mathews (WPSC) to S.A. Varga (NRC), "Auxiliary Feedwater System," dated October 1 1981.

12; Letter from Marshall Grotenhuis (NRC) to Eugene R. Mathews (WPSC), dated June 2,1982, KNPP License Amendment 45.

13] Letter from Darrell G. Eisenhut (NRC) to Eugene R.Mathews (WPSC), "NRC Requirements for Auxiliary Feedwater Systems at Kewaunee Plant," dated September 21, 1979.

[4] Letter from E.R. Mathews (WPSC) to Darell G. Eisenhut (NRC), "NRC Requirements for Auxiliary Feedwater System at Kewaunee Plant," dated October 30, 1979.

[5] Letter from Steven A. Varga (NRC) to Eugene R. Mathews (WPSC), Kewaunee Nuclear Power Plant:

Additional Information Requirements for the Auxiliary Feedwater System," dated August 18, 1980.

[61Letter from E.R. Mathews (WPSC) to Steven A. Varga (NRC), "Additional Information for Auxiliary Feedwater System," dated October 17, 1980.

[7] Introductory letter from Darrell G. Eisenhut (NRC) to all licensees, "Post TMW-Requirements," dated October 31, 1980.

[8] Letter from Steven A. Varga (NRC) to C.W. Giesler (WPSC), "Kewaunee Nuclear Power Plant - Safety Evaluation Input on the Implementation of Recommendations for the Auxiliary Feedwater Systems," dated December 16,1982.

[91Letter from Steven A. Varga (NRC) to C.W. Giesler (WPSC), "Kewaunee Nuclear Power Plant - Safety Evaluation Input on the Implementation of Recommendations for the Auxiliary Feedwater Systems," dated December 16, 1982, safety evaluation page 2.

[101 Letter from E.R. Mathews (WPSC) to S.A. Varga (NRC), Auxiliary feedwater System," dated October 1,1981

['ll Letter from Marshall Grotenhuis (NRC) to Eugene R. Mathews (WPSC), dated June 2,1982

Docket 50-305 NRC-04-012 January 30, 2004 Enclosure 2, Page 19

5. Describe the results of the impact of the power uprate on the GL 96-06 results.

Response

A) GL 96-06 Thermal Over pressurization of Isolated Water-Filled Piping Systems:

Generic Letter 96-06 was issued to address (among other concerns) the concern of thermally induced over pressurization of isolated water-filled piping sections in containment that could jeopardize the ability of accident-mitigating systems to perform their safety functions and also could lead to a breach of containment integrity via bypass leakage.

The potential for a thermally induced over-pressurization of isolated water-filled piping systems in containment is not impacted by the uprate since:

1) The volumes of isolated water-filled piping sections in containment are not being changed by uprate conditions

2) The design basis accident (DBA) containment temperature and pressure response values for the 7.4% uprate conditions are bounded by the values that were used for the current KNPP GL 96-06 thermal over pressurization analysis.

Therefore the current GL 96-06 analysis and operability determination for thermally induced over pressurization of isolated water-filled piping sections inside containment remain bounding and valid for operation at stretch power uprate conditions.

B) GL 96-06 Two-Phase Flow in Service Water System (SWS) Discharge Piping for Containment Fan Coil Units (CFCUs):

The CFCU's operation for accident conditions was based on the design containment temperature and pressure limits. The limiting condition for the CFCUs is the SW exit temperature from the cooling coils. In order to avoid two-phase flow in the lines downstream of the CFCUs, the CFCU SWS exit temperature must be limited such that the corresponding saturation pressure is below the actual system pressure in these lines. The GL 96-06 concern is that two-phase flow could result in higher back- pressure conditions that could reduce the SW flow to the CFCU's below the minimum required flow of 800 gpm during accident conditions.

Based on the current operability analysis and SWS configuration the saturation pressure for the maximum SW temperature in the exit line from the CFCUs is below the static pressure at this location for all containment temperatures up to and including the 2680 F design temperature.

The static pressure, based on the current return line arrangement, is sufficient to prevent flashing and two-phase flow in these lines.

The potential for flashing and two-phase flow of the SW in the outlet/drain lines from the CFCUs was evaluated for stretch power uprate conditions. The current GL 96-06 analysis was based on containment being at the containment design pressure and temperature limit values, no CFCU tube side fouling (ideal heat exchanger to maximize heat input to SWS), and a CFCU SWS flow of 800 gpm.

Docket 50-305 NRC-04-012 January 30, 2004 , Page 20 The containment analysis for the 7.4% uprate shows that the maximum containment temperature and pressure are less than the containment pressure and temperature design limits. The power uprate containment conditions are therefore bounded by the conditions assumed in the GL 96-06 two- phase flow analysis.

The current GL 96-06 analysis for two phase flow conditions in SWS piping downstream of the CFCU's is therefore bounding and remains valid for stretch power uprate operation.

C) CFCU SW Water hammer:

Generic Letter 96-06 was issued for licensees to determine if the SWS lines to and from the CFCU's were susceptible to water hammer during accident conditions.

An evaluation of the current GL 96-06 operability was performed for the power uprate. The potential for water hammer is not impacted by the power uprate since:

1) The containment temperature and pressure values for the DBA's for the 7.4% uprate conditions are bounded by the values that were used for the current KNPP GL 96-06 operability analysis.

2) Based on sensitivity studies performed in the current operability analyses different containment temperature profiles were shown to not significantly impact the analysis results.

GL 96-06 analyses have been performed for KNPP by a vendor under the vendor's quality assurance program using the EPRI methodology. The new GL 96-06 analyses are currently being reviewed by NMC and may need to be revised. The new analysis results bound the stretch power uprate plant operating conditions and demonstrate that the SWS is operable for the GL 96 -06 water hammer issue. The new analyses using the EPRI methodology also show that power uprate will have no impact on the GL 96-06 water hammer results.

Due to the need to complete our review of the vendor analysis and the potential need for revision to the vendor analysis, NMC will provide NRC with a status update and schedule for submittal of the resolution of GL 96-06 water hammer issues by April 2, 2004.

Therefore based on the evaluation performed for stretch power uprate conditions of the current GL 96-06 operability analysis and the results from the new analysis with the EPRI methodology, the SWS GL 96-06 water hammer operability analysis is bounding and valid for operation at stretch power uprate conditions.

NRC SPSB Branch Questions

1. Please describe the assumptions made regarding the behavior of drops formed by the break flow in high energy line break calculations done for structural analysis (Section 8.5 of WCAP 16040-P) and equipment qualification calculations (Section 8.9 of WCAP 16040-P).

Docket 50-305 NRC-04-012 January 30, 2004 , Page 21

Response

For power uprate structural analysis (section 8.5 of WCAP 16040-P) it was concluded that the initial conditions for the current structural analysis that is based on a high energy line break from no load conditions (1020 psia main steam system pressure) are the same initial condition assumed for the stretch power uprate.

For the power uprate equipment qualification calculations (section 8.9 of WCAP 16040-P) the high energy line break (HELB) mass and energy releases were conservatively determined with no water entrainment (no water drops) for the entire break spectrum included in the analysis.

This assumption maximizes the temperature of the atmosphere in the room and thus the temperature of the equipment for environmental qualification.

2. What single failure assumptions are made in the analyses to support TS 3.3.c.1.A.3?

Response

The single failure of one emergency diesel generator to start and load one of the two trains of containment engineered safeguards equipment (one train of containment engineered safeguards equipment consists of one train of containment spray and two containment fan coil units) is the single failure assumption in the containment integrity analyses This single failure assumption supports TS 3.3.c.1.A.3.

3. Please verify that the assumption that the break mitigation hardware reduces the break discharge to 50% of the no load flow in the cubicle pressurization analyses is consistent with the current Kewaunee licensing basis. If it is not, please provide a justification for this assumption.

Response

The current Kewaunee licensing basis cubicle pressurization analysis, documented in the Kewaunee Nuclear Power Plant Updated Safety Analysis Report, Rev. 18, Appendix 10A, Postulated Pipe Failure Analysis, assumes that the break mitigation hardware reduces the pipe break discharge flow rate to 50 percent of the no load flow rate.

NRC SRXB Branch Questions

1. In the November 5, 2003 response, number 34, NMC stated that calculations demonstrate that even if upper plenum injection is established as late as 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> after the LOCA (SBLOCA), the vessel boron concentration will still be 4 weight % under the boron precipitation point, assuming a saturation pressure of 35 psia. Other nuclear plants have stated that they will reach the boron saturation point in approximately 6-9 hours (Adams Accession NO. ML032720538). Please discuss this inconsistency and any actions necessary to ensure boric acid precipitation is not a concern.

.1 4 Docket 50-305 NRC-04-012 January 30, 2004 Enclosure 2, Page 22

Response

As explained in our responses to RAI #'s 33 and 34, the emergency operating procedures (EOP) direct the operator to verify low-head recirculation (via the vessel upper head injection lines) at RCS pressures below 150 psig. The subject calculations conservatively assume a boric acid solubility limit that corresponds to an RCS at saturated conditions at 20 psig (35 psia).

This boric acid solubility limit used at 20 psig was 38 w/o and was taken from experimental data (with 4% uncertainty margin). The boric acid solubility limit typically used for non-UPI plants that must consider the LBLOCA scenario (full RCS depressurization) is 23.53 w/o with 4%

uncertainty margin. The difference between 38 w/o and 23.53 w/o solubility limit explains the cited difference between the Kewaunee calculation and the calculations for non- UPI plants that result in 6-9 hours (Adams accession No. ML03270538).

KNPP is a two-loop Westinghouse reactor plant with upper plenum injection (i.e., the low head safety injection (SI) pumps inject emergency core cooling water directly into the reactor vessel upper plenum). The shutoff head of the low head SI pumps is approximately 150 psig.

Therefore, as soon as the reactor coolant system (RCS) pressure falls below about 150 psig, then the low head Si pumps would be capable of injecting cooling water into the RCS upper plenum. For a large-break loss of coolant accident (LOCA) the reactor coolant system (RCS) depressurizes quickly to below the shutoff head of the low head SI pumps and therefore, ECCS flow is quickly injected into the reactor vessel upper plenum directly above the reactor core at high flow rates to address the potential boron precipitation concern.

For small-break LOCAs the RCS does not immediately depressurize. However, the KNPP Integrated Plant Emergency Operating Procedures (IPEOPs) direct the operators to perform a post-LOCA cooldown and depressurization of the RCS. It is expected that the initial operator response to a small-break LOCA would be completed and an RCS cooldown started within about one hour. The cooldown rate is limited to a maximum of 100 degrees F per hour due to reactor vessel integrity concerns. It is expected that the RCS would be cooled down and depressurized to well below the low head Si pumps' shutoff head of 150 psig within a total time of approximately 6 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This will result in upper plenum injection of ECCS water directly into the reactor vessel above the reactor core, which addresses the potential boron precipitation concern for small-break LOCAs.

2. The NRC requests additional information concerning the elimination of the DNBR rod bow penalty for application to KNPP Cycle 26. Specifically, the NRC requests detailed information is provided about the way the FDH burndown was calculated during the cycle and the way the corresponding DNBR credit was calculated.

Response

Based on discussions with NRC, NMC has decided to withdraw its request for approval of the non-standard rod bow methodology that was used to evaluate the rod bow DNBR penalties at 1772 MWth reactor thermal power. Instead the standard Westinghouse rod bow methodology including the 2.6% rod bow DNBR penalty has been applied to KNPP C 26 core thermal hydraulic analyses at the stretch uprate (SUR) conditions. DNBR margins were shown to be sufficient for C 26 SUR operation with the standard rod bow methodology and standard rod bow penalty of 2.6% applied. Therefore the thermal hydraulic design of the Cycle 26 reload at

Docket 50-305 NRC-04-012 January 30, 2004 , Page 23 conditions up to and including the SUR meets the applicable TS and COLR limits and remains bounded by the safety analyses.

NRC SPSB-C Branch Question

1. With regard to the waste gas decay tank rupture and volume control tank rupture analyses, you state that in absence of guidance in RG 1.183, you have assumed that the offsite dose acceptance criterion is the same as given for the fuel handling accident (i.e.,

6.3 rem TEDE). However, in the staff's safety evaluation (issued March 17, 2003) for implementation of the alternative source term, this acceptance criterion was not used by the NRC staff in its finding. Kewaunee was originally licensed with a site boundary dose acceptance criterion of 500 mrem whole body for these accidents. Branch Technical Position ETSB 11-5, "Postulated Radioactive Releases Due to a Waste Gas Tank System Leak or Failure," which is located in SRP Section 11.3, "Gaseous Waste Management Systems," also presents a dose acceptance criterion of 500 mrem whole body. The curie content of these tanks is controlled so that 10 CFR 20.1301 dose limits to individual members of the public are not exceeded in the event of a release.

A. Please explain how your assumed acceptance criterion of 6.3 rem TEDE was chosen.

B. Please explain how your assumed dose acceptance criterion of 6.5 rem TEDE for these accidents is consistent with Kewaunee's original licensing and design basis, and Branch Technical Position ETSB 11-5, or why the difference should be found acceptable.

C. Please explain how the assumed accident dose acceptance criterion interacts with control of radioactivity in the tanks and the 10 CFR 20 dose limits for individual members of the public.

Response

A & B. The appropriate KNPP offsite dose limit for the gas decay tank rupture and volume control tank rupture analyses is 0.5 rem TEDE, consistent with the original licensing basis and Branch Technical Position ETSB 11-5. The site boundary dose reported for the gas decay tank rupture and volume control tank rupture is 0.1 rem TEDE, which is below the limit. KNPP is not requesting that a higher limit be accepted.

C. At the time of KNPP's original license the 10CFR20 limit for individual members of the public was 0.5 rem in a year. This earlier version of 10CFR20 is also the basis for the Branch Technical Position ETSB 11 -5 limit of 0.5 rem for the short-term dose due to a tank failure. Therefore, although the current version of 10CFR20 identifies a limit of 0.1 rem, the KNPP limit of 0.5 rem remains applicable. The calculation of the tank inventories is described in Section 7.6 of the licensing report (Attachment 4 of LAR 195[11).

Docket 50-305 NRC-04-012 January 30, 2004 , Page 24 NRC EEIB Branch Question:

1. NMC's December 15, 2003 response to NRC item 9 described the setpoint methodology used for the KNPP. The NRC staff requires additional information associated with how channel operability is determined.

Response

WCAP-15821-P, Rev.0 (Ref. 1) provides the basis for the COLR OTDT and OPDT reactor trip setpoints. Note the following differences in terminology:

1) The Kewaunee technical specifications do not have an allowable value and technical specification setpoints are referred to as limiting safety system settings or instrument setting limits.

2) The Kewaunee Core Operating Limits Report (COLR) does not have an Allowable Value. The Kewaunee COLR setpoints are referred to as setpoints.

The total instrument channel uncertainty [or Channel Statistical Allowance (CSA)] is calculated with all measured and unmeasured instrument channel uncertainties. The total instrument channel uncertainty + margin are subtracted from a defined safety analysis limit (SAL) to determine the Technical Specification or COLR setpoint. A more conservative Kewaunee plant setting is used to prevent a violation of the Technical Specification or COLR setpoint.

KNPP instrumentation for the OPDT indicates in percent of full power reactor coolant loop delta T (AT). The span of the instrumentation is 0 - 150%. For the COLR OPDT reactor trip setpoint, WCAP-1 5821 -P, Rev.0 shows a CSA of 4.2% of span (see Table 3-8 (Ref. 1)), a margin of 0.1 % of span (see Table 3-8 (Ref. 1)), and the K4 SAL of 1.16 (see Table 3-10 (Ref. 1)). The corresponding COLR OPDT K4 is 1.095 (see Table 3-10 (Ref. 1)). This results in a positive margin of 0.065 (1.16 - 1.095) between the SAL and the COLR OPDT Trip Setpoint, which is >

the total CSA of 4.2% of span.

Percent of Span to % Full Power AT Indicated Value

% of Span Span  % Full Power AT Indication Value 4.2 150% 6.3 0.1 150% 0.15 4.33 150% 6.5 Therefore, as there is margin between the assumed safety analysis limit and the Technical Specification or COLR value, after accounting for all uncertainties for a given protection function, there will be an increase in margin when addressing a subset of the uncertainties.

I Docket 50-305 NRC-04-012 January 30, 2004 Enclosure 2, Page 25 Using the above information for the OPDT instrument the determination of the setpoint and operability value can be further explained. The SAL is 1.16. This means that the operators would see a value of 116% on their OPDT setpoint indication if this were the OPDT setpoint. To determine the OPDT setpoint the CSA and margin are subtracted from the SAL (1.16 - 0.063 -

0.0015 = 1.0955 [or - 1.0951). This is the value that is entered into the KNPP COLR for the OPDT setpoint (K4). To determine the plant trip setpoint additional margin is subtracted from the COLR value. This value (margin) is a historical number that has been demonstrated by experience to ensure the COLR value will not be exceeded (0.0125). Adding this margin yields a plant trip setpoint of 1.0825 or 108.25% full power AT, as indicated in the control room.

NUREG 1431, Standard Technical Specifications Westinghouse Plants," revision 2 states that there is some point beyond which the device would have not been able to perform its function due, for example, to greater than expected drift. This value needs to be specified in the technical specifications in order to define OPERABILITY of the devices and is designated as the Allowable Value. NMC determines an allowable value but not by the same method used in ISA 67.04 and does not currently utilize this "Allowable Value" for the determination of operability for the KNPP. NMC uses the plant setpoint +/- instrument accuracy value for the determination of operability.

For operability and calibration tolerance of the OPDT trip function at KNPP NMC uses the COLR/TS value and the instrument accuracy, respectively (see Figure 1). For example, an instrument and control (&C) technician is performing a calibration of an OPDT channel. The I&C technician finds the K4 setpoint to be 1.085. The l&C technician will leave the setpojnt as found because it is within 0.5% span (0.5% span is instrument accuracy or 0.5% x 150 =

0.0075) of plant setpoint of 1.0825. Thus if the l&C technician finds the setpoint to be within 1.0825 +/- 0.0075 the setpoint is not adjusted, it is left as is. If the as found setpoint is outside the

+/-0.5% instrument accuracy the setpoint is adjusted to be within that instrument accuracy band and the channel determined to be inoperable until adjusted within the accuracy band. If an as found setpoint is outside of the instrument accuracy band NMC also performs an evaluation of the condition. If the setpoint is found to be within the KNPP allowable band the instrument is determined to meet its design function.

Docket 50-305 NRC-04-01 2 January 30, 2004 , Page 26 Figure 1 ISA S67.04 Method 3 Kewaunee Method Analytical Analytical -

Limit Umit (SAL) 0 E

Allowable ii, Value Nominal Trip COLRITS Value Setpoint _ margin Allowable Value g V

_r to __

Instrument Accuracy <<ant Setpoint Normal -

Operation Calculation 1 (Calc.1) adds/subtracts the measured uncertainties and unmeasured uncertainties plus margin from the analytical limit.

Calculation 2 (Calc.2) adds/subtracts the measured uncertainties.

Calculation 3 (Calc.3) adds/subtracts the measured uncertainties plus a margin (margin may be as low as zero)

Although NMC uses the instrument accuracy in determining operability of the instrument, the plan is to expand the operability band to the KNPP allowable value. Thus the instrument would be considered operable until the setpoint exceeded the KNPP allowable value. Additionally there are some instruments at KNPP that the value for Calc. 3 and/or Calc. 2 has not been determined. For these instruments NMC uses the TS/COLR value, adds/subtracts a historical value that has been shown to prevent exceeding the TS/COLR value and bases operability on the instrument accuracy band.

Thus further margin is added between the analytical limit and the actual plant setpoint.

TS Setting Limits cannot be changed without NRC approval per current processes and regulations.

Ill Letter from Thomas Coutu (NMC) to Document Control Desk (NRC), License Amendment Request 195, Application for Stretch Power Uprate for Kewaunee Nuclear Power Plant," dated May 22, 2003 (Adams Accession NO. ML031530424)

ENCLOSURE 3 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 30, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

NMC Responses to NRC Clarification Questions on Responses to Requests for Additional Information Reqarding LAR 195 KNPP Main Steam Piping Velocities 8 Pages Follow

TABLE 5A Main Steam Piping Velocities Steam Generator A to High Pressure Turbine Stop Valve Heat Balance Description HB101 HB102 l HB103 HB104 HB105 l HB106 HB107 Steam Generator Power, MWt 1656 1757 1757 1757 1780 1780 1780 Steam Generator Pressure, psla 744.25 770.75 785.75 80 .75 770.75 785.75 800.75 Piping Pipe d Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Description StzelSchedule Inches t/min It/sec ft/min ft/sec ft/min ft/sec Rt/min t/sec ft / min ft/sec R/ min ft/sec Rt/min ft/sec StPim Inlt torA Outet nto J1.5 oo9742rmi.w. 28.0516 8,313 138.6 8,569 142.8 8,398 140.0 8,233 137.2 8,696 144.9 8,522 142.0 8.355 139.2 31.5"00 Piping Inlet to 31.5"00 rPbn outtet 31'0D 1.451"mrinwal 28.078 8,382 139.7 8,646 144.1 8,470 141.2 8,301 138.3 8.777 146.3 8,598 143.3 8,426 140.4 Pump ranch 30 OD/0e9742-0wFl 28.0516 8,416 140.3 8,683 144.7 8,505 141.8 8.334 138.9 8,815 146.9 8,635 143.9 8,461 141.0 TOWPump Branch to_30____4 r A m p BrInch to 3e"x 240 30I0.9742min wal 28.0516 8,417 140.3 8.684 144.7 8,506 141.8 8.335 138.9 8,816 146.9 8,636 143.9 8,462 141.0 30' a 24' Crosato SteamiDump Branch "(4111'Une toAtmospher) 3r00 10.9742 mi well 28.0516 8,427 140.4 8,695 144.9 8,516 141.9 8,345 139.1 8,827 147.1 8,646 144.1 8,472 141.2 Steam Dump Branch (lIa Un to Atmosphere) to Steam Dump Branch 300OD1.9742rin wall 28.0516 8,508 141.8 8,785 146.4 8.602 143.4 8,425 140.4 8,922 148.7 8,735 145.6 8,556 14Z6 ir Llno to Condenser)_

SteamDump Branch (18' Line lo Condenser) to 2VEquazing Un 30ODIO.9742rminwan 28.0516 8,340 139.0 8,621 143.7 8,430 140.5 8,247 137.4 8,762 146.0 8,567 142.8 8,381 139.7 20M quaizing Un to "Branch to 30O10.9742rmhwa 28.0516 8,375 139.6 8,660 144.3 8,467 141.1 8.281 138.0 8,803 146.7 8,606 143.4 8,417 140.3 US StoBranc to MSRs to HP BralveMRSt I Turie Trun 30"0D 011974minwal 28.0516 7,832 130.5 8,103 I 135.0 7,908 131.8 7,721 128.7 8.243 137.4 8,044 134.1 7,854 130.9 Calculation C11386, Rev. 0 Main Steam System - Power Uprate Assessment MS Preenure Drop - 19 VetSummary final.xis Sm Gen A Page 28 of 35

TABLE 5B Main Steam Piping Velocities Steam Generator B to High Pressure Turbine Stop Valve Heat Balance Description HB101 HB102 HB103 HB104 HB105 HB106 HB107 Steam Generator Power, MWt 1656 1757 1757 1757 1780 1780 1780 Steam Generator Pressure. psia 736.75 763.25 778.25 793.25 763.25 778.25 793.25 Piping Pipe d Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Description SizeISchodule Inches ft/rmin Rt/sec ft/rin ft/sec ft/ in ft/sec ft/rmin ft/sec ft/rin ft/sec ft/rmin ft/sec ft/ min ft/sec SteamOnrtto0 eutbt to 3.5 0 3O0.942rinwal 28.0516 8.834 147.2 9.103 151.7 8,919 148.7 8,743 145.7 9,238 154.0 9,052 150.9 8,872 147.9 Pipi5g OD___Piping_ I*____ __Inlet_ _ ___ ___

Ptptgn Outeit 31'D011 l4elminwad 28.078 8,919 148.7 9,199 153.3 9,009 150.2 8.827 147.1 9,338 155.6 9,146 152.4 8,961 149.3 3.m Piping OutiettoT AFW O0 .974r min w 28.0516 8,959 149.3 9,242 154.0 9.050 150.8 8,866 147.8 9,383 156.4 9.188 153.1 9,001 150.0 PumA rFnW Pump_____Branch_ _ _ _to_ _30__ __ ___24_ _

TDAFW Pump Drenchto S0 x24 300010.9742minwa 28.0516 8,960 149.3 9,243 154.0 9,051 150.9 8,867 147.8 9,384 156.4 9,189 153.2 9,003 150.0 3r x 24' Cross to Steam Dump I MSR 30O 10.974r min wat 28.0516 8,972 149.5 9,256 154.3 9,064 151.1 8,879 148.0 9.397 156.6 9,202 153.4 9,015 150.2 Branch (tirStm Dump to Condensor_

Steam DumpIMSR Branch (I Line 30 0D 09742'mhwat 28.0516 8.775 146.3 9,060 151.0 8.862 147.7 8,672 144.5 9.205 153.4 9,003 150.1 8,810 146.8 to Condenser) to SteamDump Branch 3' tir lne to Atmosphere)

Steam Dump Branch t8' Une to 30' OD I0.974r min wall 28.0516 8,801 146.7 9,089 151.5 8,889 148.1 8,697 145.0 9,235 153.9 9,031 150.5 8,836 147.3 Atmoshpere) to 20 Equalizing Una 20' Equalizing Un. to HP Turbine 30 D 0.9742' mh 28.0516 8,839 147.3 9,131 152.2 8,928 148.8 8,734 145.6 9.279 154.6 9,073 151.2 8,875 147.9 IStop Vs "e_ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ __ _ _

Calculation C11386, Rev. 0 Main Steam System - Power Uprate Assessment MS PressureDrp - HB Vet Sunaryfialxl SboGe B Page 29 of 35

TABLE 6A Atmospheric Dump Vaive Piping Velocities Steam Generator A Heat Balance Description HB101 HB102 HB103 HB 104 HB105 HB106 HB107 Steam Generator Power, MWt 1656 1757 1757 1757 1780 1780 1780 Steam Generator Pressure, psia 740.50 767.00 782.00 797.00 767.00 782.00 797.00 Piping Pipe d Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Description Size/Schedule Inches fl/min ft / sec ft/min ft/sec ft/min ft / sec ft/min ft/sec ft/min ft/sec ft/min ft/sec ftmin ft/sec Steam G ntaor A Outt to 31.5 oo 3000 O 0Io.s74rmin wa t 28.0516 8,572 142.9 8,835 147.2 8,657 144.3 8.487 141.4 8.966 149.4 8,786 146.4 8,612 143.5 Piping Intet__ _ __ _ _ _

Mit O ipn Iet 3100114er1mhinwal 28.078 8,648 144.1 8.920 148.7 8,738 145.6 8,562 142.7 9,056 150.9 8,870 147.8 8.691 144.9 31.r 00 Piping Outlet to TOAFW Pump Brtnch 30 ODI0.9742mni wat 28.0516 8,685 144.7 8,959 149.3 8,775 146.3 8,598 143.3 9,096 151.6 8,909 148.5 8,729 145.5 DAFw Pump ncho1x24 s 30 0.974rmina 28.0516 8,686 144.8 8,960 149.3 8,776 146.3 8,599 143.3 9,097 151.6 8,910 148.5 8,729 145.5 30' x 24' Ceossto SteamDump 30 ODI 0.974r min wet. 28.051s 8,696 144.9 8,972 149.5 8,787 146.5 8.609 143.5 9.109 151.8 8.921 148.7 8,740 145.7 Oranch ( t Uns to Atmosphere) .

From SG A Main Hadr to FCV468A 1rODISchPO 16.124 11,967 199.4 12.356 205.9 12.096 201.6 11,846 197.4 12,548 209.1 12,284 204.7 12,031 200.5 Takeoff__ ___

FCV46A Takeoff to FcV46s Takoff t ODI SdBea 18.124 8,123 135.4 8,399 140.0 8,215 136.9 8,040 134.0 8,534 142.2 8,349 139.1 8,170 136.2 FCV468B Takoff to FcV48C Ttakeoff 18001SDScho 16.124 4,067 67.8 4,205 70.1 4,113 68.6 4,025 67.1 4,273 71.2 4,180 69.7 4,090 68.2 1tHeedsrtoFCV468A r0OISeh80 7.625 18,162 302.7 18,777 3129 18.368 306.1 1 7,977 299.6 19.081 318.0 18666 311.1 18.267 304.5 I Headerto FCV4ss OD I Sch 80 7625 18,185 303.1 18.802 313.4 18,392 306.5 18,000 300.0 19108 318.5 18.691 311.5 18.291 304.9 tH*d ftoFcv4ate rOD/Sch80 7.62 18,191 303.2 18,809 313.5 18,399 306.6 18.005 300.1 19,115 318.6 18,698 311.6 18,297 3050 Calculation C11386. Rev. 0 Main Steam System - Power Uprate Assessment MS Pressue Drop-H Vei Sumary taxs S0 A Arn Drp Page 30 of 35

TABLE 68 Atmospheric Dump Valve Piping Velocities Steam Generator B Heat Balance Description HB101 HB102 HB103 I HB104 HB105 l HB106 l HB107 Steam Generator Power, MWt 1656 1757 1757 1757 1780 1780 1780 Steam Generator Pressure, psla 740.50 767 .00 782.00 797.00 767.00 782.00 797.00 Piping Pipe d Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Description SizdSehduie Inches ft/min ft/sec ft/min ft/sec ftImin ft/sec ft/min ft/sec t/min t / sec ft/Rmin ft/sec Imin ft/sec SteamGentoB Outt to 315' 00 30'OD109742'aI 28.0516 8,572 142.9 8,835 147.2 8,657 144.3 8,487 141.4 8,966 149.4 8,786 146.4 8,612 143.5 3i~n 3_______ ________PipingInletto 3IP1" Out0pt tt 31-Oet1 4trinwal 28.078 8,649 144.2 8,922 148.7 8,739 145.6 8,563 142.7 9,057 150.9 8,871 147.9 8,693 144.9 31.5" 00 Piping Outlet to TOAFW 3'01094 i a 4. 4. ,3 4 Pump BrmnchO 30'ODIO974 rd 28.0516 8,687 144.8 8,961 149.4 8,777 146.3 8,600 143.3 9,098 151.6 8,911 148.5 8,730 145.5 To AFW Pump Branch to 30' x 24' 31I0010.9741 mhlwl 28.0516 8,688 144.8 8,963 149.4 8,778 146.3 8.601 143.3 9,099 151.7 8.912 148.5 8,731 145.5 38 x 24 cro to Steam Dump I SR 30'0 I 0974rmin wsl 28.0516 8,698 145.0 8,974 149.6 8,789 146.5 8,611 143.5 9,111 151.9 8.923 148.7 8,743 145.7 Branch (O' StmDump to Conde nser)

SteamDump I MSR Branch (I S' Line to Condens r) to Steam Dump Brnd 30 0D0 .974?r minwal 28.0516 8,790 146.5 9,075 151.3 8.884 148.1 8,701 145.0 9.217 153.6 9,023 150.4 8.837 147.3 (ir Line to Atmosphere)

FromoSG aiDnpeader to tmpercDump Head.r lODISch 80 1r 16.124 12,007 200.1 12,399 206.7 12,137 202.3 11,885 198.1 12.594 209.9 12,328 205.5 12,072 201.2 IS Atmosphndc Dump Header to 1rODISrh80 16.124 4,047 67.5 4,183 69.7 4,093 68.2 4,006 66.8 4.251 708 4,159 69.3 4,070 67.8 FCV4TSC TakeoffI

-8BAtmosphericDump Hesder to 1r OD I Sh 80 16.124 8,095 134.9 8,366 139.4 8,186 136.4 8.012 133.5 8,501 141.7 8,317 138.6 8.141 135.7 FCV4T88 Takeoff__ _ __ _ _ _ _ _ _ _

FCV478B Takeoff to FCV478A Takeoff Ir OD I Sch 80 16.124 4,052 67.5 4,189 69.8 4,098 68.3 4,011 66.8 4,256 70.9 4,164 69.4 4,075 67.9 18'HeedertoFCV478C rODISch8O 7.625 18,104 301.7 18,712 311.9 18,307 305.1 17,919 298.7 19,013 316.9 18,602 310.0 18,207 303.4 18' Header to FCV478B rODISceh80 7.625 18,120 302.0 18,730 312.2 18,324 305.4 17.936 298.9 19,032 317.2 18.620 310.3 18,224 303.7 r Hederto FCV4aA rODISch 80 7625 18126 302.1 18,737 312.3 18.331 305.5 17.941 299.0 19.040 317.3 18.626 310.4 18.230 303.8 Calculation C11386, Rev. 0 Main Steam System - Power Uprate Assessment MS Pressure Oep HB Vo Sunmmay fmiel.xts SG Abs Dmp Page 31 of 35

TABLE 7A Condenser Dump Valve PIping Velocities Steam Generator A Heat Balance Description HB101 HB102 HB103 HB104 HB105 HB106 HB107 Steam Generator Power, MWt 1656 1757 1757 1757 1780 1780 1780 Steam Generator Pressure, psla 740.50 767.00 782.00 797.00 767.00 782.00 797.00 Piping pipe d Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Description Sze/Schedul, inches ftmin ft/sec ft/min ft / sec ft / min ft/sec ft/min ft / sec ft/rmin ft/sec ft / min ft/sec ft / min ft / sec Steam OsnoratorA Outlet to 31.5 00 390 0.974rminwo 28.0516 8,572 142.9 8,835 147.2 8,657 144.3 8,487 141.4 8,966 149.4 8.786 146.4 8,612 143.5 Piping Inlet____

PiS 'O Piping Inlet to 31O.StOD 3tD0011.e1mnwall 28.078 8,648 144.1 8.920 148.7 8,738 145.6 8,562 142.7 9,056 150.9 8,870 147.8 8.691 144.9 Oplng 311p5'g1 _u___ttoT____

1ut5 00 Piping OutOt to TOAFW 3-O0.974r minw 28.0516 8,685 144.7 8,959 149.3 8,775 146.3 8,598 143.3 9,096 151.6 8.909 148.5 8,729 145.5 Pum rAncWPumpranchto_30__24 TDCAFWPump Branch to 38x24 30-oo10.974rmh waFi 28.0516 8,686 144.8 8,960 149.3 8,776 148.3 8,599 143.3 9,097 151.6 8.910 1485 8,729 145.5 U0 x 24 Cross to Steam Dump 30 ODI0.974rmhnwa 28.0516 8,696 144.9 8,972 149.5 8,787 146.5 8,609 143.5 9,109 151.8 8,921 148.7 8,740 145.7 Branch (1S Una to Atmosphere)

Stexm Dump Branch 18' Un to 30 0O 0.974rmhwall 28.0516 8,786 146.4 9,071 151.2 8,881 148.0 8,698 145.0 9,213 153.6 9,019 150.3 8,833 147.2 Atmosphere) to SteamDump Branch 118Ln to Condenser) _

From S0 A Maln Header to Takeoff to 18`OI Sch80 16.124 10,803 180.0 11,166 186.1 10.924 182.1 10,693 178.2 11,348 189.1 11.100 185.0 10,864 181.1 FCV4M4 TkeotttoFCV44to6'tSR 18 OD I Sh 80 16.124 7,355 122.6 7,614 126.9 7,443 124.0 7.279 121.3 7,742 129.0 7,568 126.1 7,401 123.3 r SR Takeoff to TkofH to t8ODISch 80 16.124 7,362 122.7 7,622 127.0 7,450 124.2 7,286 121.4 7,751 129.2 7.576 126.3 7.408 123.5 FCV484CI Takeoff to FCV484C to Takeoff to 80D Sch 80 16.124 3,685 61.A 3,815 63.6 3,729 62.1 3,647 60.8 3,880 64.7 3,792 63.2 3,708 61.8 8' Takeoff to FCV484A 8' 00 Sch 80 7.625 16,480 274.7 17,064 284.4 16,678 278.0 16,310 271.8 17,354 289.2 16,960 282.7 16,584 276.4 ti' Tkeoft to FCV4848 e OD I Sh o 7.625 16,443 274.1 17.023 283.7 16,640 277.3 16.274 271.2 17,311 288.5 16,920 282.0 16.546 275.8 i' Takeoff to FCVt4C e'00ISch80 7625 16,476 274.6 17,060 284.3 16.674 277.9 16.306 271.8 17,349 289.2 16.956 282.6 16,580 276.3 Calculation C11386, Rev. 0 Main Steam System - Power Uprate Assessment MS Pressure Drop - HBVat Summary finat.ts SG A CondDrmp Page 32 of 35

4 .

t-i TABLE 7B Condenser Dump Valve Piping Velocities Steam Generator B Heat Balance Description HB101 l HB102 l HB103 l HB104 HB105 HB106 HB107 Steam Generator Power, MWt 1656 1757 1757 1757 1780 1780 1780 Steam Generator Pressure, psia 740.50 767.00 782.00 797.00 767.00 782.00 797.00 Piping Pipe d Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Description SizeSchedule Inches ft/min It/ sec ft/min ft/sec ft/min ft/sec ft/min ft/sec fIt/ min ft/sec ft/min ft/sec R ft / min ft/sec Stam Generator B Outle to 3.5'OD 3 00 0S742 minwa 23.0516 8.572 142.9 8,835 147.2 8,657 144.3 8,487 141.4 8,966 149.4 8,786 146.4 8.612 143.5 Piping ___

PiP. Outpit 31ODI4AS1mih wad 28.078 8,649 144.2 8,922 148.7 8,739 145.6 8,563 142.7 9,057 150.9 8,871 147.9 8,693 144.9 iump inch 30 OD 0.974r minwon 28.0518 8,687 144.8 8.961 149.4 8,777 146.3 8.600 143.3 9.098 151.6 8,911 148.5 8,730 145.5 CDAFWPpan chtoODx 24' 3o979T4r minwa 280518 8.688 144.8 8,963 149.4 8,778 146.3 8,601 143.3 9.099 151.7 8,912 148.5 8,731 145.5 3 x 24' Cross to Steam Dump I SR 30 00 0.974r mh wol 28.0516 8,698 145.0 8,974 149.6 8,789 146.5 8.611 143.5 9.111 151.9 8,923 148.7 8,743 145.7 Brnch (I18'Stm Dump to Condenser .

From SGB Main Haderto Takeoffto I WO Sch so 16.124 10,641 177.4 10.987 183.1 10.756 179.3 10,534 175.6 11.159 186.0 10.924 182.1 10.698 178.3 Takeoff to FCV44F to 6 M5R Taknoft 1S 00 / Sch80 18.124 7.204 120.1 7,446 124.1 7,285 121.4 7,130 118.8 7.566 126.1 7,402 123.4 7,245 120.7 r RTakeoffto TakreotFCV44E 180015ch80 18.124 7.211 120.2 17,454 124.2 7.292 121.5 7.137 119.0 7.574 126.2 7,410 123.5 7.252 120.9 Takeoff to FCV454E to Takeoff to 1S O I Sch 80 18.124 3.609 60.1 3,731 62.2 3,650 60.8 3,572 59.5 3,791 63.2 3,709 61.8 3.630 60.5 FCV4540 _ _ _ _ _ _ _ _ _ _

6" Takeoff to FCV484 S'OO/Sch8o 7.625 16,141 269.0 16,687 278.1 16.324 272.1 15,976 266.3 16,958 282.6 16,588 276.5 16,234 270.6

' Takeoff to FCV454E e 00 Sch0 7.625 16,137 269.0 16.683 278.0 16,320 272.0 15.973 266.2 16,953 282.6 16,584 276.4 16,230 270.5 S' Takeoffto FCV45F 7.825 16106 268.4 16.648 277.5 16,288 271.5 15,942 265.7 16,917 28.0 16.550 275.8 16,198 2700 Calculation C11386, Rev. 0 Main Steam System - Power Uprate Assessment MS PressureDrop . HBVetSumnmary final.al 50B CmodDmp Page 33 of 35

TABLE 8A SRV Piping Velcocitles Steam Generator A Heat Balance Description l HB101 HB102 HB103 l HB104 HB105 HB106 l HB107 Steam Generator Power, MWt 1656 1757 1757 1757 . 1780 1780 1780 Calculated Pressure (c SG A, psla 1177.3 1182.3 1182.3 1201.6 1203.2 1203.3 1203.3 Piping Pipe d Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Description Size/Schtdule Inches ft/rmin ft/see ft/min ft/sec fR/min t / sec ft/mn ft/sec ft/rmin ft/sec ft/rmin ft/sec ft/rmin ft/sec Steam Genator A Outt to 31.5' /

3rOD 0.974 mi 28.051l 5142 85.7 5,474 91.2 5,476 91.3 5,376 89.6 5,443 90.7 5,445 90.8 5,447 90.8 31.r oDPping InI to 31.50OO 31rODter11mbinwal 28.078 5.155 85.9 5,491 91.5 5,493 91.6 5,392 89.9 5,460 91.0 5,462 91.0 5,464 91.1 Ptnoutlet l _ _ _ _ _ __ _ _ _ _ _ _

31.5 O Pping Outlt toTO AFW 3 00 10.974 mn wl 28.0516 5.169 86.2 5,508 91.8 5,509 91.8 5,408 90.1 5,476 91.3 5,478 91.3 5,480 91.3 Pump Branchi_ _ _ __ _ _ _ _ _ _ _ _ _

To AFW Pump Branch to S x 24 30' ODI0.9742rihw 28.0516 5,170 86.2 5.508 91.8 5.510 91.8 5,408 90.1 5,477 91.3 5.478 91.3 5,480 91.3 Cross _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

SGAk30x 24C to o23.22400D I Sch 80 2152 3,502 58.4 3,731 62.2 3,732 62.2 3.664 61.1 3,710 61.8 3,711 61.9 3,712 61.9 S023-2 Takeoff to 23-1 Takeoff 24-O IISch 80 21.52 1,752 29.2 1,867 31.1 1,868 31.1 1,833 30.6 1,856 30.9 1,857 31.0 1,858 31.0 S023-2 Takeoff to S023-2 WOO I Sch 80 5.761 24,543 409.0 26.153 435.9 26,162 436.0 25.679 428.0 26,004 433.4 26.013 433.6 26.022 433.7 S023-1 Takeoff to SD23-1 OoI Sc 80 5.761 24,544 409.1 26.155 435.9 26,164 436.1 25,681 428.0 26,006 433.4 26,015 433.6 26,024 433.7 so A 30 x 24 Cross to SD23.3 24 00 ISdl 80 21.562 5,252 87.5 5,596 93.3 5,598 93.3 5,495 91.6 5,565 92.7 5,567 92.8 5,569 92.8 SD23 3Takeof" to SD23-4 Takeoff 24 OD I Sch80 21.562 3,507 58.5 3,738 62.3 3,739 62.3 3,670 61.2 3,716 61.9 3.718 62.0 3.719 62.0 SD23-4 Takeoff to S23- Takeoff 24 00 1 Sch 80 21.562 1,754 29.2 1,869 31.2 1,870 31.2 1,835 30.6 1,859 31.0 1,859 31.0 1,860 31.0 SD23-3 Takeoff to S023-3 S' 00! S s6o 5.761 24,564 409.4 26,179 436.3 26,188 436.5 25,704 4284 26 ,030 433.8 26,039 434.0 26,048 434.1 SD23.4 Takeoff to S023.4 WOO Sch 80 5.761 24,570 409.5 26,186 436.4 26,195 436.6 25,711 428.5 26,037 434.0 26,046 434.1 26,055 434.3 5023-5 Takeof to S023 ero'ISch8O 5.761 24,571 409.5 26,18 436.5 26,197 436.6 25,71 4285 26,039 434.0 26,048 434.1 26,057 434.3 Calculation C11386, Rev. 0 Main Steam System - Power Uprate Assessment MS Pressure Drop. H8 Ve Su ry final.,s SG A SRV Page 34 of 35

TABLE B SRV Piping Velocities Steam Generator B Heat Balance Descripton HBtO1 HB102 HB103 HB104 HB105 HB106 HB107 Steam Generator Power, MWt 1656 1757 1757 1757 1780 1780 1780 Calculated Pressure c0 SG B. psia 1,177.4 1,182.4 1,182.4 1,201.7 1,203.3 1,203.4 1,203.4 Piping Pipe d Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Velocity Description SimaSchedule Inches fl/mh ft/sec ft/min ft/sec fl/ min nt/sec Rt/min ft/sec ft/mh n ft/sec ft/mn ft/sec ft/min ft/sec Steam Ganarator BOutt to 31.5'00 30ODI0.9742 min 28.0516 5,141 85.7 5,474 91.2 5,475 91.3 5,376 89.6 5,443 90.7 5,444 90.7 5,446 90.8 F'ptgn Outrt 0 31-et 0D /1.41 mh 28.078 5.155 85.9 5,491 91.5 5,493 91.5 5,392 89.9 5,460 91.0 5,462 91.0 5,464 91.1 31.5- 00 Piping Oulet to TDAFW 3 00 10.9742mn wal 25.0510 5,169 86.2 5,508 91.8 5,509 91.8 5.408 90.1 5.476 91.3 5.478 91.3 5,480 91.3 Pump Branch _ _ _ _ _ _ _ _ _ _ I__ _ _ _ _ I__ I _ _ _

TOAFW Pump Brnchto 0' x 24' 30'OD10.974rminwall 28.0516 5,170 86.2 5.508 91.8 5,510 91.8 5,408 90.1 5,477 91.3 5,478 91.3 5,480 91.3 SGB 30' x 24' Cross to 023- 24'0D1Sch80 21.62 3,502 1,171.3 3,731 1,175.5 3,732 1,175.5 3,664 1,194.9 3,710 1,196.3 3,711 1,196.4 3,712 1,196.4 Takeoff _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

50239 Takeoff to S2310 Takeoff 24'OIScheo 21.562 1,752 1,170.6 1,867 1,174.7 1,868 1,174.7 1,833 1,194.1 1.856 1,195.5 1,857 1,195.6 1,858 1,195.6 SD23S Takeoffto S029 tr OD I Sch eo 5.761 24,543 1,170.6 26,153 1,174.7 26162 1,174.7 25,679 1,194.1 26.004 1,195.5 26.013 1,195.6 26.022 1,195.6 5023.10 Takeoff to S02310 6' ODI Sch 80 5.761 24,544 1,170.5 26,155 1,174.6 26.164 1,174.6 25.681 1,194.0 26,006 1,195.4 26,015 1,195.5 26.024 1,195.5 SO A 30'x 24' Cross to S0238 24-00IScheo 21M2 5,252 1,171.3 5,596 1,175.5 5.598 1,175.5 5,495 1,194.9 5,565 1,196.3 5,567 1,196.4 5,569 1,196.4 Takeoff__ _ ___ _ ___ _ _ _ _ _ _

S023-8 TakeoffS o237Takoff 24 00 I Sch80 2152 3,507 1,169.7 3,738 1,173.7 3,739 1,173.7 3,670 1,193.1 3,716 1,194.5 3,718 1,194.5 3,719 1,194.6 S023.7 Takeoffto S2 Takoff 24- OD I Sch80 21562 1,754 1,169.5 1.869 1,173.4 1,870 1,173.4 1,835 1,192.8 1,859 1,194.2 1,859 1,194.3 1,860 1,194.3 S023-8Takeoffto 2S4 6,OO ISc80 5.761 24,564 1,169.7 26,179 1,173.7 26188 1,173.7 25,704 1,193.1 26,030 1,194.5 26,039 1,194.5 26.048 1,194.6 S023-7 Takeoff to S0237 r ODI Sh o 5.761 24,570 1,169.5 26,16 1,95 4 5 1,173. 25,711 1,1928 2,037 1,194.2 26,046 1,194.3 26,055 SO23OTakeofftoSO234 r3oScheo 5.761 24,571 1,169.4 26,188 1,173.3 26,197 1,173.3 25.713 1,192.8 26,039 1,194.1 26,048 1,194.2 26.057 1 Calculation C11386, Rev. 0 Main Steam System - Power Uprate Assessment MS Presure Drop-- BVe Summnary finatl. G B SRV Page 35 of 35

ENCLOSURE 4 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 30, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

NMC Responses to NRC Clarification Questions on Responses to Requests for Additional Information Regarding LAR 195 REGULATORY COMMITMENTS UPDATE 3 Pages Follow

- **4 -

Docket 50-305 NRC-04-012 January 30, 2004 Enclosure 4, Page 1 LIST OF REGULATORY COMMITMENTS (Update)

The following table identifies those actions committed to by NMC made in association with LAR 195. Any other statements in this submittal or prior submittals associated with LAR 195 are provided for information purposes and are not considered to be regulatory commitments.

The purpose of restating these commitments is to provide the NRC staff with an up-to-date status of commitment completion and those for which actions are still required. The changes to this table are that commitments 9 and 11 have been completed.

Commitment Description Due Date/Event Section 7.1.2 of attachment 4 states that an evaluation of the thermal and hydraulic safety analyses for the Framatome ANP fuel demonstrates that the DNBR design basis is met for the Framatome Prior to 1 ANP fuel in Cycle 26. This evaluation has been performed for cycle implementation of 26 reload safety evaluation, the COLR, and the associated the stretch power 10 CFR 50.59. All documents will be revised, as appropriate, for uprate.

the stretch power uprate to address Framatome fuel DNBR design basis.

Sections 8.3.3.3 and 8.3.7.3 of attachment 4 state the increase in flow rate and velocities, as well as the changes in operating Prior to the next 2 pressures and temperatures, will be incorporated into the KNPP scheduled refuelin FAC program as part of the power uprate implementation. The outae FAC program models will be updated prior to the next program o inspections scheduled for the next refueling outage.

Section 8.3.3.3, 8.3.15, 8.6.4, 8.6.5, and 8.6.6 of attachment 4 all 3 refer to the feedwater control valve trim modification. This Complete.

modification was completed during the R26 refueling outage.

Section 8.3.3.3 of attachment 4 describes the following in regards to the feedwater heaters:

a. Initial inspection and analyses will establish a baseline prior to a Complete the stretch uprate implementation. These baseline inspections and . p analyses were completed during the R26 refueling outage.

4

b. An inspection and monitoring program will be established to b. Prior to the next monitor potential heater degradation at the stretch power uprate b rui ote conditions. An inspection program will be developed based on the refueling outage.

baseline inspection results and using programs and processes in place at KNPP. This will be completed prior to the next refueling outage.

-. "d* -

Docket 50-305 NRC-04-01 2 January 30, 2004 Enclosure 4, Page 2 Commitment Description Due Date/Event Sections 8.3.8.4 and 8.4.2.2.1 of attachment 4 state that a cycle-specific heat load calculation will be performed prior to each refueling outage to determine the adequacy of spent fuel pool 5 cooling capability. Reactor engineering refueling procedures have Complete.

already incorporated this confirmatory calculation as a requirement.

The new requirement will administratively control the in core hold time of the fuel after shutdown to ensure the requirements are met.

Instrument and Control System Setpoint changes were summarized in sections 8.3.9.3 and section 8.3.15 of attachment 4. These recommended changes included CST level setpoints, first stage Prior to stretch 6 turbine pressure, and turbine overspeed trip settings. The turbine uprate overspeed trip setting changes were completed during the R26 implementation.

refueling outage. The other recommendations will be reviewed by the plant staff and implemented as appropriate.

Piping and pipe support evaluations concluded in section 8.4.4 that Prior to stretch the systems remain acceptable assuming resolution of open items. uprate Open items remained on the following systems: service water and ip ati component cooling water. These open items will be resolved. implementation.

Modifications to the steam generator level control system were 8 recommended to support the stretch uprate as described in Complete.

Sections 8.6.5 and 8.6.6 of attachment 4. The level control changes were completed during the R26 refueling outage.

Evaluation of the EQ equipment inside containment affected by the 9 higher containment EQ long-term temperature profile will be Complete.

performed per Sections 8.9.3 and 8.9.4 of attachment 4.

The KNPP EQ Plan will be updated, as appropriate, to reflect power Prior to 10 uprate evaluations in accordance with Section 8.8.4.3 of attachment implementation of

4. stretch power uprate.

For those components where the HELB temperatures exceeded the 1 equipment qualification temperature, the EQ equipment required for HELB outside containment will be qualified per Section 8.9.3 and Complete.

8.9.4 of attachment 4.

Plant procedures will be revised as appropriate to accommodate the stretch power uprate. Procedure changes were committed to in Prior to Attachment 1, Section 2.6 and in Attachment 4, Section 8.7.2. Prio o 12 Emergency, abnormal, and operating procedures that are entered implementation of due to a LONE event or have AFW TS requirements will be stretch power changed as appropriate to reflect the new TS requirements for the uprate.

AFW system.

.r ;, -

Docket 50-305 NRC-04-01 2 January 30, 2004 Enclosure 4, Page 3 Commitment Description Due Date/Event New TS requirements, revised procedures, and any control room Prior to 13 changes due to the stretch power uprate will be reviewed by implementation training for determination of being included in the operator training stretch power of program. This was committed to in Attachment 1, Section 2.6. uprate.

Attachment 1, Section 2.6 stated that setpoint changes for reactor Prior to protection and control inputs, alarms, computer constants, and implementation of 14 embedded values, will be updated consistent with operation at stretch power 1772 MWt. Power range nuclear instruments will be recalibrated uprate and checked based on a secondary heat balance.

Sections 9.2.4.1.7 and 9.2.4.2.5 of attachment 4 describe recommended modifications to the high pressure turbine cylinder 15 joint bolting and the low-pressure turbine coupling bolts. These Complete.

modifications were completed during the recent R26 refueling outage.

The response to RAI question #57 to the stretch power upratePrior to submittal stated degraded voltage and thrust calculations for MOV prio o 16 operators outside containment were reviewed for impact of the implementation of uprated post accident temperatures, and will be revised, as strte required, prior to implementation of the 6%Stretch Power Uprate. uprate.

17 NMC will provide NRC with a status update and schedule for Aprl2,2004 resolution of GL 96-06 water hammer issues Note: attachment 4 refers to attachment 4 of reference letter 1 listed on the cover letter.

.V 4w -

ENCLOSURE 5 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 January 30, 2004 Letter from Thomas Coutu (NMC)

To Document Control Desk (NRC)

NMC Responses to NRC Clarification Questions on Responses to Requests for Additional Information Regarding LAR 195 NMC Responses to Letter from John Lamb (NRC) to Thomas Coutu (NMC), "Kewaunee Nuclear Power Plant - Review of License Amendment Request No. 195, Stretch Power Uprate (TAC NO. MB9031)," dated January 26, 2004) 2 Pages Follow

- r Ace Docket 50-305 NRC-04-01 2 January 30, 2004 Enclosure 5, Page 1 In the referenced letter (reference 4) from the NRC, the NRC stated that it still needed additional technical information to complete its review. Below is a listing of the information the staff stated it needed and NMC's response or direction to where NMC's response is located in enclosure 2 of this letter:

(1) an evaluation to address the effects of the stretch power uprate on utilizing the refueling water storage tank to borate the reactor coolant system to cold shutdown, with and without the availability of letdown, NMC Response: See NMC response to NRC SPLB Branch question number 2 found on page 10 of enclosure 2.

(2) the calculations and description of the methodology to eliminate the rod bow penalty, NMC Response: See NMC response to NRC SRXB Branch question number 2 found on page 22 of enclosure 2.

(3) the technical basis for operating the feedwater and condensate motors at service factors of approximately 1.03 at stretch power uprate conditions, NMC Response: See NMC response to NRC EMEB Branch question number 5 found on page 4 of enclosure 2.

(4) the technical basis for operating at the stretch power uprate conditions in light of the water hammer, two-phase flow, and thermal overpressurization concerns that are discussed in Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During a Design-Basis Accident Conditions," dated September 30,1996, NMC Response: See NMC response to NRC SPLB Branch question number 5 found on page 19 of enclosure 2.

(5) a summary of your responses to post-Three Mile Island (TMI) Action Plan Item II.E.LI, that includes a listing of applicable correspondence and a description of the impact that the proposed stretch power uprate will have on your responses to and resolution of this particular post-TMI Action Plan item, NMC Response: See NMC response to NRC SPLB Branch question number 4 found on page 14 of enclosure 2.

(6) how you interpret the proposed auxiliary feedwater (AFW) TS requirements if a steam supply to the turbine-driven AFW pump is inoperable when another AFW pump is inoperable, NMC Response: Based on current KNPP TS 3.4.b.2, when a steam supply to the Turbine Driven AFW (TDAFW) pump is inoperable a 7-day allowed outage time (AOT) limitation is entered (KNPP TS 3.4.b.2.C). As stated in the KNPP TS basis, this condition does not render the TDAFW pump inoperable or render an AFW train inoperable (KNPP TS basis page TS B3.4-1). If one of the AFW pumps is concurrently declared inoperable then one train of AFW is inoperable and a 72-hour AOT is entered (KNPP TS 3.4.b.2.A). The NRC in the staffs' safety evaluation for KNPP license amendment 123 approved this condition.'

b at ;*

Docket 50-305 NRC-04-012 January 30, 2004 Enclosure 5, Page 2 (7) how you plan to coordinate the down power maneuver when two AFW pumps are inoperable to assure decay heat removal requirements will be met in the event of a loss of normal feedwater event, and NMC Response: With the proposed stretch power uprate TS, during 100% power operation (1772 Mwth) when two AFW pumps are inoperable (i.e. two AFW trains are inoperable) two limiting conditions for operation (LCO's) are entered simultaneously. The first is proposed TS 3.4.b.3, "If two of the three AFW trains are inoperable, then within two hours, reduce reactor power to < 1673 MWth. The second is proposed TS 3.4.b.4.B (current TS 3.4.b.2.B), "Two auxiliary feedwater trains may be inoperable for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Thus within two hours of the time of discovery of two AFW trains inoperable the plant power will be reduced to < 1673 MWth, if the condition is not corrected within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the time of discovery the plant will shutdown in accordance with proposed TS 3.4.b.6 with a final end state of the reactor coolant system temperature < 3500 F (8) an example showing that sufficient margin exists between the analytical limit and the setpoint including the tolerance band around the setpoint to illustrate when the channel is considered inoperable.

NMC Response: See NMC response to NRC EEIB Branch question number 1 found on page 24 of enclosure 2.

Enclosure 5

References:

1 Letter from Richard J. Laufer (NRC) to M.L. Marchi (WPSC), "Amendment No. 123 To Facility Operating License No. DPR Kewaunee Nuclear Power Plant (TAC NO. M93977), dated January 3,1996,