ML033570513

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

ML033570513
Person / Time
Site:	Kewaunee
Issue date:	12/15/2003
From:	Coutu T Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC-03-125
Download: ML033570513 (121)

Text

I Committed to Nuclew Excellence Kewaunee Nuclear Power Plant Operated by Nuclear Management Company, LLC NRC-03-125 10 CFR 50.90 December 15, 2003 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

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.

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 54216-9510 Telephone: 920.388.2560

Docket 50-305 NRC-03-1 25 December 15, 2003 Page 2 On November 5, 2003, NMC responded to a request for additional information from the Nuclear Regulatory Commission (NRC) regarding the proposed stretch power uprate (Reference 2). Subsequent to NMC's response the NRC staff has requested clarification on some of the responses submitted in NMC's November 5th letter. This letter, with attachments and enclosures, contains the NMC responses to the NRC request for clarifications. contains the questions the NRC asked NMC for a response. These questions either came by email from the NRC staff or were asked during conference calls with the NRC staff. These questions are listed separately in attachment 1. contains NMC's response to these clarification questions.

Enclosures A and B contain calculations requested by the NRC staff to complete their review of the stretch power uprate submittal.

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. Attachment 1 is the questions the NRC staff asked NMC for clarification to the RAI responses in reference 2. Attachment 2 is NMC's responses to these clarification questions. Enclosures A and B are requested calculations These 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.

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 December 15, 2003.

Thomas Coutu Site Vice-President, Kewaunee Plant BJW Attachments: 1. NRC Clarification Questions on Responses to Requests for Additional Information

2. NMC Responses to NRC Clarification Questions

Docket 50-305 NRC-03-1 25 December 15, 2003 Page 3

Enclosures:

A. Kewaunee Nuclear Power Plant Calculation C1 1473, "Post Accident Operating Time Evaluations for Various Safety-Related Equipment in Support of Power Uprate," Revision 1, dated 11/18/03 B. Kewaunee Nuclear Power Plant Calculation C11543, "Evaluation of the Impact of the Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project,"

Revision 0, dated 12/05/03 cc- US NRC, Region IlIl US NRC Senior Resident Inspector (w/o enclosures)

Electric Division, PSCW (w/o enclosures)

ATTACHMENT 1 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 December 15, 2003 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 5 Pages to Follow

Docket 50-305 NRC-03-125 December 15, 2003 , Page 1 RAI #4: The second bullet in the response indicates that the applicant "will" reevaluate conditions, and this evaluation "will be completed" prior to the uprate. Yet other statements indicate that the evaluation has been completed. I'm not sure which it is.

If the evaluation hasn't been performed prior to the writing of the SER it will be an open item, if the evaluation has been performed then the RAI response should state clearly that it has been performed and available time is adequate to perform manual actions with the higher power uprate power level.

RAI #7: In the answer provided by the licensee to question # 7 regarding the megavolt-amperes reactive (MVARS) supplied by the main generator (MG) the licensee responded the following:

'lf Energy Supply and Control requests an increase in MVARs to support grid voltage, a power reduction would be required to stay within the generator capability curve" The staff recommends that a Plant Procedure be required in in the event that Energy Supply and Control requests an increase in MVARs.

ALSO: There is a situation with the transformers similar to the one that we had with the generator during our conference call. I told Narinder about the problem and I think the licensee should be contacted. The problem is that is not clear from the submittal what's the operating point of these transformers. It says that they will be operated at their rated value but it's not clear what rating they are going to use. I will send an e-mail to Narinder with a draft SE that has a comment regarding the transformers operation. Please see the dratf SE for more information.

RAI #10: The staff needs a clarification for the answer provided to question #10 related to the operation of the RCP and FW pumps. In Table 8.3.14.1 -1 of the WCAP-1 6040-P page 8-73 shows that the Feedwater Pumps will be operated at 5150Hp but is rated at 5000Hp. NEMA MG 1-20.45 Variations from Rated Voltage and Rated Frequency section A states that Motors shall operate successfully, under running conditions, at rated load and frequency with a voltage not more than 10 percent above or below rated voltage but not necessarily in accordance with the standards established for operation at rated voltage. Since this motor is not going to be operated at rated load NEMA MG 1-20.45 standard doesn't seem to be applicable for the operation of this motor.

The staff requires supporting evidence that the overloading of the motors based on their service factors won't affect the operation of these motors. The IEEE standard 666 11.4.2.4 that deals with the operation of General-purpose motors having a service factor greater than one specifically states that:

The operation above rated horsepower will cause higher temperature rise, and the performance may differ from the performance at rated horsepower. Torque capability will be based on rated horsepower. Also, if a motor is operated continuously at its service factor rating, its expected life will be reduced. (According to the 8-120C increase rule, the life of the insulation is reduced by 50% if the temperature is increased by 8-120C.)

Docket 50-305 NRC-03-125 December 15, 2003 , Page 2 NEMA MG 1-20.13 Service factor states that:

In those applications requiring an overload capacity, the use of a higher horsepower rating as given in MG 1-20.10 is recommended to avoid exceeding the temperature rises for the class of insulation used and to provide adequate torque capacity.

RAI #13: Circuit breakers overdutied The licensee stated in the response that:

"If a maximum possible fault current situation actually occurred, the breaker attempting to interrupt the fault could fail catastrophically."

"The justification is that actual test data take for the circuit breakers demonstrated that they could interrupt fault current greater than the maximum available fault current at KNPP."

The staff requires a clarification of the statement of the licensee that the breaker could fail catastrophically in the event the maximum possible fault current situation should occur.

RAI #16: The staff needs supporting evidence for the operation of the Feedwater Pumps and Condensate Pumps at overated horsepower based on the reasons provided in item #3 of this memorandum.

RAI #17: The staff requires that the licensee submits the evaluations that demonstrate the affected equipment is qualified for the EQ long-term temperature.

RAI #18: The staff requires that the licensee submits the evaluations that demonstrate the EQ equipment required for HELB outside containment is qualified for the EQ thermal lag temperatures.

RAI #24: The licensee refers to a commitment they made to reanalyze the Chapter 14 events and update all appropriate documentation for a Framatome fuel core. The staff requested the results of these analyses in the fuel transition LAR, and the licensee stated that this information would be submitted as part of the stretch uprate submittal. The staff needs to see the information requested.

ALSO: Regarding my issue with RAI 24, please have the licensee include a discussion of how they addressed the existing Framatome/ANP fuel rod performance/design criteria under the stretch uprate conditions. The criteria of interest are those addressed in Section 2.4.1 of the Westinghouse report titled, "Technical Design Basis for the Transition to 422V+ Fuel.' This report was Attachment 4 to the KNPP RTSR submittal.

Docket 50-305 NRC-03-125 December 15, 2003 , Page 3 RAI #29: The licensee is performing calculations which are not in accordance with currently approved methodology. In order for the staff to review and approve a new calculational method, significant additional information must be submitted to the staff. Specifically, the RAI response does not discuss the method used to calculate the FdH burndown credit taken to offset rod bow DNBR penalty. However, the staff has some question as to the need to use this methodology, as it appears that the licensee has adequate margin to DNBR limits even if the rod bow penalty is applied.

RAI #40: The existing radiological consequences of a steam generator tube rupture (STGR) accident assumes that the leak flow from the reactor coolant system (RCS) to secondary side of the steam generator (SG) is terminated in 30 minutes following the event initiation. The analysis uses an equilibrium break flow that continues at a constant rate for 30 minutes. The resulting break flow mass transfer is then used to calculate the radiological consequences of the SGTR.

You have considered that this assumption will result in a conservative calculation since the reduction in the break flow rate over the 30 minutes time is ignored. Inherent in this evaluation is the assumption that the operator can terminate the break flow in 30 minutes. Plants with similar design to KNPP have reported to NRC that in simulator exercises, the operators demonstrated that the time to terminate the break flow exceeded the 30 minute assumption.

Should operator significantly exceed the 30 minute termination criteria, this event could lead to an increase in radiological releases from that assumed when the leak flow was terminated within 30 minutes. Please provide the necessary information to confirm that the assumption of termination of the break flow in 30 minutes following a design basis SGTR event is valid for KNPP at the uprated power level of 1772 MWt and will indeed lead to a bounding calculation regarding to the radiological consequences of the event.

RAI #43: Please explain why reduce the assumed initial RCS pressure to 2000 psia will provide conservative results of transient and accident analyses especially for events lead to higher peak RCS pressures.

RAI #44: Please discuss the change of feedwater temperature responses affecting the current NSSS design transients.

RAI #51: The NRC staff has recently become aware, and has discussed with NEI, a problem with use of instrument setpoint methodology if based on "Method 3" of ANSIAISA-S67.04, "Setpoints for Nuclear Safety-Related Instrumentation." Does Kewaunee use this method?

Additional Comments: Discussion items

1. Why was a supplemental table of mass and energy for MSLB (Table 6) added to November 5, 2003 letter?

2. Explain why the accumulators are a heat sink for the main steam line break.

Docket 50-305 NRC-03-125 December 15, 2003 , Page 4

3. WCAP 10325 does not mention specifically the FROTH and EPITOME computer codes. Is there another W document which discusses these codes or an NRC approval of these codes?

4. Describe model for calculating heat transfer to shield building.

5. Do the calculations for high energy line breaks (HELBs) outside containment follow the Standard Review Plan (SRP 6.2.1.2) guidance? In particular, is the entrainment modeled as 100% or is a GOTHIC option used (drop-liquid conversion or equivalent) which results in less than 100 entrainment? The drop-liquid conversion is discussed in Section 8.7 of the GOTHIC 7.0 manual.

6. During a phone call on 12/8/03, John Lamb discussed an additional concern, or clarification the NRC requires:

In our submittal we made the following statements:

Cover Letter The NMC has determined that the information for the proposed amendments does not involve a significant hazards consideration, authorize a significant change in the types or total amounts of effluent release, or result in any significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendments meet the categorical exclusion requirements of 10 CFR 51.22(c)(9) and an environmental impact appraisal need not be prepared.

Attachment 1 page 16 An environmental review summary is contained in attachment 4, report section 8.10, "Environmental Assessment." The conclusion of this review is that the proposed amendments do not involve a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22 (b), an environmental assessment of the proposed change is not required.

In the Licensing Report (WCAP-1 6040-P), Section 8.8, Radiological Assessments, Section 8.8.2.3, Results (pages 8-131 & 132), under both "Liquid Effluents" and "Gaseous Effluents",

the report makes statements about noble gas activity changes that the NRC found confusing.

Under "Liquid Effluents", we talk about increases in Tritium of 17.6%, but that the actual increase is only about 11.4%. We then mention part of the reason is because there is not a change in "mode of operation". We also say that strict adherence to NUREG-0017 would have identified no change in tritium level. The NRC wants clarification on why we did not just

Docket 50-305 NRC-03-1 25 December 15, 2003 , Page 5 stick with the NUREG evaluation, instead of confusing the issue with 17.6% and 11.4%

values. And they were not sure what was meant by "mode of operation".

Under "Gaseous Effluents", similar comments apply. Apparently they do not see a clear logic between the increases we are addressing and the conclusions in 8.8.2.4, Conclusions.

They do not necessarily want to see the analyses, they are confused because the report states that the NUREG method would not have identified a change, but the method we applied does identify increases.

7. TS 3.3.c.1 .A.3.(iii) is being deleted because 2 trains of containment spray (CS) as well as one containment fancoil unit train are required to support the containment integrity analysis.

If this is the case, why are current TSs 3.3.c.1.A.3.(ii) and (iv) still acceptable? Have they been shown to produce acceptable results at the uprated power level?

8. The licensee had not performed the first in-service inspection at the time of the power uprate submittal, therefore stating in it that no loose parts were present in their steam generators (SGs). Kewaunee submitted a license amendment request 199, "Steam Generators Eddy Current Inspection Frequency Extension" where they detected possible loose parts (PLPs) in their SGs. Since this information came up after the submittal, in our review of the uprate we have to acknowledge this PLPs that are present now. Although in their submittal for the power uprate the licensee stated that a generic loose parts was prepared to address undefined PLPs, we will like to know if the PLPs found are still bounded by this evaluation, and if not, whether the licensee has performed another evaluation for leaving the PLPs in service at the new uprate conditions.

9. Please assure that adequate margin exists between the Analytical Limit (AL) and the Allowable Value (AV) that equals or exceeds the value of uncertainties not measured during the channel operations test (COT). Please confirm that this is true for all protection system setpoints. Please provide an example calculation which demonstrates the existence of adequate margin; choose from one of the reactor setpoints being revised, Overtemperature delta T (OTDT) or Overpower delta T (OPDT). Please provide assurance to the staff that TS setting limits will not be revised using method 3 without NRC approval.

ATTACHMENT 2 NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 December 15,2003 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 Responses to NRC Clarification Questions 30 Pages to Follow

Docket 50-305 NRC-03-125 December 15, 2003 , Page 1 RAI #4: The second bullet in the response indicates that the applicant will' reevaluate conditions, and this evaluation "will be completed" prior to the uprate. Yet other statements indicate that the evaluation has been completed. I'm not sure which it is.

If the evaluation hasn't been performed prior to the writing of the SER it will be an open item, if the evaluation has been performed then the RAI response should state clearly that it has been performed and available time is adequate to perform manual actions with the higher power uprate power level.

RAI #4 Staff Concern NMC Response: As part of the Appendix R program evaluation for stretch uprate (SUR) a calculation was performed to determine the steam generator (SG) dry out time due to a loss of main feedwater with the assumption that the initial reactor power level is at a nominal SUR power level of 1772 MWt. This calculation was used to verify that the time to SG dry out is greater than the Appendix R design basis operator response time for initiating AFW flow. The SUR SG dry out calculation was performed consistent with the methodology of the original Appendix R SG dry out calculation, which had been performed with a nominal reactor power assumption of 1720 MWt.

The calculated time to SG dry out, assuming stretch power uprate conditions (i.e. reactor power level is 1772 MWt, plus 0.6% uncertainty, equal 1783 MWt) is 36.6 minutes. Since this time is greater than the Appendix R acceptance criteria for response time for initiating auxiliary feedwater flow, which is 35 minutes, the KNPP fire protection Appendix R design description AFW design basis response time requirement is satisfied at the stretch power uprate conditions.

Therefore the KNPP fire protection Appendix R design description requirement for AFW flow response time is valid and does not need to be revised for stretch power uprate. In addition, no operating procedures or associated training utilized for response to an Appendix R fire require revision for this issue.

RAI #7: In the answer provided by the licensee to question # 7 regarding the megavolt-amperes reactive (MVARS) supplied by the main generator (MG) the licensee responded the following:

"If Energy Supply and Control requests an increase in MVARs to support grid voltage, a power reduction would be required to stay within the generator capability curve" The staff recommends that a Plant Procedure be required in in the event that Energy Supply and Control requests an increase in MVARs.

ALSO: There is a situation with the transformers similar to the one that we had with the generator during our conference call. I told Narinder about the problem and I think the licensee should be contacted. The problem is that is not clear from the submittal what's the operating point of these transformers. It says that they will be operated at their rated value but it's not clear what rating they are going to use. I will send an e-mail to Narinder with a draft SE that has a comment regarding the transformers operation. Please see the draff SE for more information.

RAI #7 Staff Concern NMC Response: At uprated power conditions, KNPP is expected to produce approximately 595.7 MWe gross at a 0.957 pf and 180.3 MVAR. At uprated power

Docket 50-305 NRC-03-125 December 15, 2003 , Page 2 conditions, KNPP will not exceed the design limits of either the main generator (622.389 MVA and 560.15 MWe gross at a 0.90 pf) or the main transformer (649.5 MVA at 65 0C). This is assured by KNPP operating procedures, which require the plant to remain within the main generator capability curve. The capability curve is based on a design limit of 622.389 MVA, and accounts for operation at various power factors, gross MWe outputs, and MVAR outputs.

Operation within the capability curve prevents exceeding main generator design ratings. Since the main transformer power handling capabilities are greater then the power generation capabilities of the main generator, operating within the generator capability curve ensures main transformer ratings are not exceeded.

For the second concern on transformer ratings and operating points, Tables 8.3.14.1-1, 2, & 3 of of our submittal provides the design ratings and expected normal operating loads for the transformers. As stated in Section 8.3.14.1.3.4 of Attachment 4 of our submittal, 'The MAT, RAT, and TAT are capable of supporting station operation at full-power uprate."

RAI #10: The staff needs a clarification for the answer provided to question #10 related to the operation of the RCP and FW pumps. In Table 8.3.14.1-1 of the WCAP-16040-P page 8-73 shows that the Feedwater Pumps will be operated at 5150Hp but is rated at 5000Hp. NEMA MG 1-20.45 Variations from Rated Voltage and Rated Frequency section A states that Motors shall operate successfully, under running conditions, at rated load and frequency with a voltage not more than 10 percent above or below rated voltage but not necessarily in accordance with the standards established for operation at rated voltage. Since this motor is not going to be operated at rated load NEMA MG 1-20.45 standard doesn't seem to be applicable for the operation of this motor.

The staff requires supporting evidence that the overloading of the motors based on their service factors won't affect the operation of these motors. The IEEE standard 666 11.4.2.4 that deals with the operation of General-purpose motors having a service factor greater than one specifically states that:

The operation above rated horsepower will cause higher temperature rise, and the performance may differ from the performance at rated horsepower. Torque capability will be based on rated horsepower. Also, if a motor is operated continuously at its service factor rating, its expected life will be reduced. (According to the 8-120C increase rule, the life of the insulation is reduced by 50% if the temperature is increased by 8-120C.)

NEMA MG 1-20.13 Service factor states that:

In those applications requiring an overload capacity, the use of a higher horsepower rating as given in MG 1-20.10 is recommended to avoid exceeding the temperature rises for the class of insulation used and to provide adequate torque capacity.

RAI #10 Staff Concern NMC Response: These motors will not be operated at overrated conditions based on their service factor (SF). They are rated as 1.15 SF motors, and are rated for operation at any SF up to 1.15, as long as they also stay within their temperature rise limits for their insulation class (in this case, Class B). At uprated power conditions, KNPP does not exceed any of the design limits of these pump motors.

Docket 50-305 NRC-03-125 December 15, 2003 , Page 3 If we were designing a new facility from scratch, all pumps and motors would be selected to have nameplate ratings greater than the expected conditions. However, we are applying to uprate the existing plant, and verify in our application that existing equipment (pump motors) will continue to operate within design specifications. The design specifications are the nameplate ratings +/- the tolerances allowed by the governing standard. For these large motors, the governing standard is NEMA MG 1, Motors and Generators (The NEMA reference for this response is NEMA MG 1-1 993, Rev. 3 of May, 1996).

Per MG 1, Section 1.66, Classification of Insulation Systems, "Insulation systems are divided into classes according to the thermal endurance of the system for temperature rating purposes."

For each class of insulation, "An insulation system which, by experience or accepted test, can be shown to have suitable thermal endurance when operated at the limiting class temperature."

Per MG 1, Section 1.43, Service Factor - AC Motors, service factor is, "a multiplier which, when applied to the rated horsepower, indicates a permissible horsepower loading which may be carried under the conditions specified for the service factor."

Per MG 1, Section 20.14.1, Service Factor of 1.0 - For motors of this service factor, the second paragraph states, "In those applications requiring an overload capacity, the use of a higher rating is recommended to avoid exceeding the adequate torque handling capacity." This applies to motors with a service factor rating of 1.0. Our motors in question have a rating of 1.15, and are therefore, not covered by section 20.14.1.

Per MG 1, Section 20.14.2, Service Factor of 1.15- 'When specified, motors furnished in accordance with this standard will have a service factor of 1.15 and a temperature rise not in excess of that specified in 20.40.2 when operated at the service factor horsepower rating with rated voltage and frequency." (for SF HP rating see Sections 1.43 and 20.14.3 discussed above

& below)

Per MG 1, Section 20.14.3, Application of Motors with a Service Factor of 1.15 - In the "General" sub-section (20.14.3.1) it states, in part, "...the motor may be overloaded up to the horsepower obtained by multiplying the rated horsepower by the service factor shown on the nameplate."

Per MG 1, Section 20.14.3, Application of Motors with a Service Factor of 1.15 - In the "Temperature Rise" sub-section (20.14.3.2) it states, in part, 'When operated at the 1.15 service-factor-load, the motor will have a temperature rise not in excess of that specified in 20.40.2 with rated voltage and frequency maintained.... NOTE - The tables in 20.40.1 and 20.40.2 apply individually to a particular motor rating (that is, a 1.0 or 1.15 service factor), and it is not intended or implied that they be applied as a dual rating to an individual motor....

Operation at the temperature-rise values given in 20.40.2 for a 1.15-service-factor load causes the motor insulation to age thermally at approximately twice the rate that occurs at the temperature-rise values given in 20.40.1 for a motor with a 1.0-service-factor load;...".

Based on the above, and the fact that the KNPP feedwater and condensate pump motors are rated for a 1.15 service factor, operation at any service factor between 1.0 and 1.15, and with a temperature rise of less than 90 C above a reference ambient of 40 C, is within the motor's

Docket 50-305 NRC-03-125 December 15, 2003 , Page 4 design specifications, and does not constitute an overloaded condition. At power uprate conditions the expected temperature rise of the condensate pump motors will increase approximately 2.3 C, and the feedwater pump motors will increase approximately 3.9 0C.

The thermal life of these pump motors is a function of the motor insulation operating temperature, whether the motor is rated at 1.0 service factor or 1.15 service factor. Therefore, the temperature rise of the motor insulation is the key parameter. In general, as a motor's service factor increases, it's motor insulation temperature will increase (not a linear relationship). A rule of thumb for thermal aging states that the thermal life of an insulation is reduced by 50% for every 10 C rise in insulation temperature. However, thermal life has not been a regulatory issue with BOP motors where a projected thermal life has been required to be established. Instead, motor Preventive Maintenance (PM) activities have been established to determine a degrading trend in capability prior to failure.

Under stretch uprate conditions, these motors will continue to be operated within their design limits for Class B motors rated for a 1.15 SF. At KNPP, refueling PMs measure each motor's insulation resistance, and informs engineering if this resistance drops below established values.

These PMs also change air inlet filters and lubricating oil, if so equipped. These motors have temperature monitoring instrumentation which alarms in the control room before the motors exceed their class rise limits for temperature. These are the established programs in place to identify degradation of motor capability.

RAI #13: Circuit breakers overdutied The licensee stated in the response that:

"Ifa maximum possible fault current situation actually occurred, the breaker attempting to interrupt the fault could fail catastrophically."

'The justification is that actual test data take for the circuit breakers demonstrated that they could interrupt fault current greater than the maximum available fault current at KNPP."

The staff requires a clarification of the statement of the licensee that the breaker could fail catastrophically in the event the maximum possible fault current situation should occur.

RAI #13 Staff Concern NMC Response: Calculation C10739,4160 Volt Bus and Switchgear Fault Current Calculation (4/95), was done following a NRC Electrical Distribution System Functional Inspection (EDSFI), and is the most up-to-date fault current calculation for breakers on buses 1 through 6. It's purpose was to determine the worst case (highest) fault current available at buses 5 and 6 (Engineered Safeguard Buses) for use in purchasing vacuum replacement breakers for those buses. It also calculated fault current values for buses 1 through 4 (Non-Safeguard Buses).

In section 5.4 (Results) it states: 'The worst case for buses 1 and 2 is 42.3 kA at 4260 V. The worst case for buses 3 and 4 is 33.5 kA at 4260 V. While these values exceed the existing

Docket 50-305 NRC-03-125 December 15,2003 , Page 5 switchgear's ratings (32.4 kA at 4260 V), the existing gear is adequate for these buses as discussed during the EDSFI (ref. 4.6) based on actual test data on the existing breakers."

The McGraw-Edison Company test report indicates that the switchgear interrupted an average current of 37.7 kA at 5.06 kV. This corresponds to 44.8 kA at 4260 V, approximately 6% above the required 42.3 kA needed for buses 1 & 2, and 34% above the required 33.5 kA needed for buses 3 & 4.

Calculation C1 0739 was primarily focused on buses 5 and 6. The pre-fault voltage of 4260 V used at all buses was intended to be conservatively high for buses 5 and 6. The highest fault current on buses 1 through 4 occurs when they are fed by the Main Auxiliary Transformer (MAT). The MAT is fed directly from the generator, and with voltage drops within the MAT, buses 1 and 2 voltages will remain well below 4160 V (typically between 4000 V and 4050 V).

At 4160 V, the switchgear is rated to interrupt 33.2 kA. The test data indicates it could interrupt 45.9 kA at 4160 V, or approximately 8.5% above the 42.3 kA worst case fault current for buses 1 & 2.

The probability of this maximum fault occurring is also very low. As stated in the EDSFI concern response, American Switchgear (who purchased McGraw-Edison) noted that it is reasonable to assume that a three phase bolted fault can only occur during a maintenance activity, with the bus intentionally grounded for personnel protection, and a breaker is accidentally closed. Such maintenance activities are only performed during plant shutdown, when the available fault current would be much lower since the Feedwater and Reactor Coolant pump motors would not be running. A more credible failure would be a fault due to insulation failure. This type of failure would result in an arcing fault of significantly lower magnitude.

In summary, the probability of the maximum calculated fault occurring is very low. If it did in fact occur, the switchgear has been shown by test to be capable of interrupting a larger current.

RAI #16: The staff needs supporting evidence for the operation of the Feedwater Pumps and Condensate Pumps at overated horsepower based on the reasons provided in item #3 of this memorandum.

RAI #16 Staff Concern NMC Response: See the response to #10 above. These motors are not operated at an overrated HP based on their service factor (SF). They are rated as 1.15 SF motors. Per MG 1, "...the motor may be overloaded up to the horsepower obtained by multiplying the rated horsepower by the service factor shown on the nameplate." Continuous operation in the overloaded region is acceptable as long as they also stay within their temperature rise limits for their class (in this case, Class B). At power uprate conditions, the expected temperature rise of these motors will still be well below their insulation class allowable temperature rise.

RAI #17: The staff requires that the licensee submits the evaluations that demonstrate the affected equipment is qualified for the EQ long-term temperature.

Docket 50-305 NRC-03-125 December 15, 2003 , Page 6 RAI #17 Staff Concern NMC Response: See Enclosure A, Calculation C11473, Rev. 1, Post Accident Operating Time Evaluations for Various Safety-Related Equipment in Support of Power Uprate.

RAI #18: The staff requires that the licensee submits the evaluations that demonstrate the EQ equipment required for HELB outside containment is qualified for the EQ thermal lag temperatures.

RAI #18 Staff Concern NMC Response: See Enclosure B, Calculation C1 1543, Rev. 0, Evaluation of the Impact of the Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project.

RAI #24: The licensee refers to a commitment they made to reanalyze the Chapter 14 events and update all appropriate documentation for a Framatome fuel core. The staff requested the results of these analyses in the fuel transition LAR, and the licensee stated that this information would be submitted as part of the stretch uprate submittal. The staff needs to see the information requested.

ALSO: Regarding my issue with RAI 24, please have the licensee include a discussion of how they addressed the existing Framatome/ANP fuel rod performance/design criteria under the stretch uprate conditions. The criteria of interest are those addressed in Section 2.4.1 of the Westinghouse report titled, "Technical Design Basis for the Transition to 422V+ Fuel." This report was Attachment 4 to the KNPP RTSR submittal.

RAI #24 Staff Concern NMC Response: The information provided below describes the evaluations and results of evaluations performed by Nuclear Management Company (NMC),

Westinghouse Electric Company (WEC), and Framatome ANP (FANP) to demonstrate that the design basis for the Framatome ANP fuel is met for Kewaunee Cycle 26 at the Stretch Uprate conditions. The information is focused upon the fuel system design criteria, including the aspects of thermal and hydraulic design and nuclear design that affect or are related to the fuel system design criteria for FANP fuel.

The table below outlines the evaluations performed and is organized along the criteria given in Reference 1, which has been approved by the NRC as complying with the Standard Review Plan (SRP) in Reference 2. The request for additional information (RAI) contained in RAI #10 of Reference 3 indicates that the criteria of interest are those of the SRP and not specifically those of the Westinghouse Reload Transition Safety Report (RTSR, Ref. 4), the second part of the RAI #24 concern above not withstanding, and is evidence that this approach for presenting the information is acceptable.

Following the table is a summary of the results of the thermal-hydraulic analyses performed by NMC for the FANP fuel DNBR and PCT analyses.

Docket 50-305 NRC-03-1 25 December 15, 2003 , Page 7 Section Criteria Description Criteria SRP Reference Disposition (Organization:

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ D esc ription )

3.2 Fuel Rod Criteria Hydrogen content in FANP: Satisfied since hydrogen components controlled to a content is controlled by 3.2.1 Internal Hydriding minimum level during SRP 4.2 - I.A.2.(a) manufacturing specifications and manufacture to limit verified by QC inspection.

internal hydriding FANP: Satisfied since radial gap Sufficient plenum spring >0.0 inch, plenum spring design defection and cold radial ensures positive downward force, 3.2.2 Cladding Collapse gap to prevent axial gap SRP 4.2 - lI.A.2.(b) and fuel rod internal pressure formation during maintains radial gap, throughout densification densification for all fuel operating in Cycle 26.

95/95 confidence that fuel rods do not experience NMC: Satisfied for normal Overheating of ro ulaeeration opatr and AO0s based on 3.2.3 Cladding boiling (DNB) during SRP 4.2 - I.A.2.(d) TIH (DNBR) analyses for FANP Cladding steady-state or anticipated fuel. R)aalss o FN operational occurrences (AOOs)

WEC: Satisfied criterion for REA.

Overheating of Fuel No centerline melting NMC: Satisfied for normal 3.2.4 Pellets during normal operation SRP 4.2 - I.A.2.(e) operation and other AOOs based and AOOs on T/H (DNBR, non-LOCA PCT) analyses for FANP fuel.

(Stress and Strain Limits) 3.2.5 Pellet/Cladding Cladding strain <1% SRP 4.2 - I.A.2.(g) FANP: Satisfied with the following

Docket 50-305 NRC-03-1 25 December 15, 2003 , Page 8 Section Criteria Description Criteria SRP Reference Disposition (Organization:

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Interaction (PCI) (<0.75% at pellet burnups results:

>60 GWd/MTU) and no Transient (AOO) Strain:

fuel centerline melting U02 Rod = 0.419% Margin Gad Rod = 0.579% Margin Steady-State Strain:

U02 Rod =-0.045%

Gad Rod = -0.249%

See Section 3.2.4 for temperature.

ASME Section 1I1, Appendix Ill, Article Ill FANP: Satisfied since component Cladding Stress (2000), in combination with SRP 42 - A.2.'a' materials maintain margin to the specified 0.2% offset R . _ .. a ASME criteria. Minimum margin is yield strength and ultimate 0.110.

strength of Zircalloy-4 Not underestimated during Loss-of-Coolant AccidentWE:Stsidaprtoth 3.2.6 Cladding Rupture (LOCA) and used in SRP 4.2 - I.A.2.(h) Large Break LOCA evaluation.

determination of 10 CFR LreBekLC vlain 50.46 criteria Fuel Rod Mechanical ASME Section 111, WEC: Satisfied as part of the 3.2.7 Fracturing Appendix F SRP 4.2 - l.A.2.(i) Seismic-LOCA evaluation.

3.2.8 Fuel Densification and See Sections 3.2.2, 3.2.4, FANP: Satisfied with NRC-3..8 Swelling 3.2.5, and 3.3.7 approved fuel performance codes.

3.3 Fuel System Criteria (Stress, strain, and loading limits on assembly 3.3.1 components. See Section 3.3.9 for handling and 3.4 for accident conditions.)

Docket 50-305 NRC-03-1 25 December 15, 2003 , Page 9 Section Criteria Description Criteria SRP Reference Disposition (Organization:

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Spacer Grid SpacerLateral Grid load < load limit SRP 4.2 - I.A.1.(a) ~~~~~~~~~~~~Seismic-LOCA WEC: Satisfied evaluation.

as part of the FANP: Satisfied based on the Limiting load occur during following Seismic-LOCA loads UTPs and LTPs handling and postulated SRP 4.2 - lI.A.1.(a) from WEC:

accidents - 950 lbs on top nozzle

- 3800 lbs on bottom nozzle Cladding See Section 3.2.5 of this SRP 4.2 - lI.A.1.(a) Satisfied, see Section 3.2.5.

table FANP: Satisfied with the following 3.3.2 Fatigue Cumulative (CUEF) <0.67usage factor SRP 4.2 - II.A.1.(b)

SP42-lA..() margins:

U02 Rod CUIF =0.23 Gad Rod CUF = 0.13 No fuel rod failures due to SRP 4.2 Il.A.1 .(c)FANP: Satisfied since spacer/rod 3.3.3 Fretting Wear 3.3.3 Fretting fretting Wear N freng wearrodefaiuresdueto SRP 4.2 - I.A.1.(c) ~~~~~~~~~meet contact condition continues to criteria.

Acceptable maximum oxide thickness. Effects of FANP: Satisfied since peak local oxidation and crud oxide <130 microns (Ref. 5).

3.3.4 Oxidation, Hydriding, included in thermal and SRP 4.2 - .A.1 .(d) Approved fuel rod performance and Crud Buildup mechanical fuel rod code accounts for oxidation and analyses. Stress analysis crud buildup. Metal loss to include metal loss due accounted for in stress analysis.

to oxidation Lateral displacement of the NMC: Satisfied for normal fuel rods shall not be of operation and AOOs based on 3.3.5 Rod Bow sufficient magnitude to SRP 4.2 - I.A.1.(e) T/H (DNBR) analyses for FANP impact thermal margins impact marginsthermal fuel including rod bow penalty.an evaluation of the

Docket 50-305 NRC-03-125 December 15, 2003 , Page 10 Section Criteria Description Criteria SRP Reference Disposition (Organization:

_ _ _ _ _ _ _ _ _ _ _ __________ ___________ ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ D e scriptio n) 3.3.6 Axial Irradiation Growth FANP: Satisfied at fuel assembly design exposure limit with the Clearance remains following margin:

Fuel Rod UTP/LTP at end of l fe SRP 4.2 - I.A.1.(e) Min. EOL clearance = 0.087" UTP/LTPof life at end WEC: Fuel rod design exposure (EOL) limit met for Cycle 26 - maximum rod exposure < 58 GWd/MTU.

The fuel assembly length shall not exceed the FANP: Satisfied since clearance Fuel Assembly upper and lower core SRP 4.2 - I.A.1.(e) exists at EOL under cold plates in the cold condition at EOL FANP Satisfied at fuel assembly design exposure limit with the Acceptable maximum rod following margin:

internal pressure. Gap U02 Rod = 2334.1 psia 3.3.7 Rod Internal Pressure does not open during SRP 4.2 - I.A.1.(f) Gad Rod = 1303.8 psia steady-state or increasing WEC: Fuel rod design exposure power limit meet for Cycle 26 - maximum rod exposure < 58 GWd/MTU.

FANP: Satisfied since no FANP fuel liftoff during pump startup operations. Liftoff evaluation 3.3.8 Assembly Liftoff No liftoff from core support SRP 4.2 - lI.A.1.(g) considers a maximum 3.2% mass flow increased due to mixed fuel core.

WEC: Cycle 26 maximum flow

Docket 50-305 NRC-03-1 25 December 15, 2003 , Page 11 Section Criteria Description Criteria SR P Reference Disposition Description)(Organization:

increase in FANP fuel is 3.17%.

Assembly withstands 21h FANP: Satisfied since assembly 3.3.9 andFuel Assembly withstas sti SRP 4.2 - I.A.1.(a) withstands in excess of 2k times Handling timcestewihasasai SR4.-l.A1() the weight as static force based on testing.

3.4 Fuel Coolability FANP: Satisfied based on the Structural Deformations geometry and ability to SRP 4.2 - I.A.3.(e) deflection: Seismic/LOCA insert control rods as per SRP 4.2 - App. A - Lateral displacement of 1.1 ASME Section III, App. F inches at grid 4 3.4.1 Cladding Embrittlement Include in LOCA analysis SRP 4.2 - I.A.3.(a) WEC: Satisfied in LOCA analysis.

3.4.2 Violent Expulsion of <280 cal/gm energy SRP 4.2 - lI.A.3.(b) WEC: Satisfied in REA analysis.

Fuel deposition 3.4.3 Fuel Ballooning Consider impact of flow SRP 4.2 blockage in LOCA analysisSR4.- - I.A.3.(d) WEC: Satisfied in LOCA analysis.

lA3d)WCSaifeinL Aaays.

4.1 Thermal and Hydraulic Criteria Hydraulic flow resistance WC aife nRla 4.1.1 Hydraulic Compatibility similar to resident FANP SRP 4.4 -11.4 Transition Safety Report (Ref. 4).

fuel assemblies NMC: Satisfied for normal 4.1.2 Thermal Margin 95/95 no DNB SRP 4.4 -11.1 operation and other AOOs based Performance on T/H (DNBR) analyses for FANP fuel.

4.1.3 Fuel Centerline No centerline melting SRP 4.4 -11.1 WEC: Satisfied for REA in RTSR

Docket 50-305 NRC-03-1 25 December 15, 2003 , Page 12 Section Criteria Description Criteria SRP Reference Disposition (Organization:

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Temperature (Ref. 4).

NMC: Satisfied for normal operation and other AQOs based on T/H (DNBR, non-LOCA PCT) analyses for FANP fuel.

NMC: Satisfied for normal operation and other AOOs based 4.1.4 Rod Bow Protect thermal limits SRP 4.4 -11.2 on T/H (DNBR) analyses for FANP fuel including an evaluation of the rod bow penalty.

5.0 Neutronics Criteria 5.1 Power Distribution In accordance with SRP 4.3 -11.1 WEC: Satisfied.

Technical Specifications SP43-1. E:Stsid 5.2 Kinetics Parameters Doppler Reactivity Negative SRP 4.3-11.2 WEC: Satisfied.

Coefficient Power Coefficient Negative relative to HZP SRP 4.3 -11.2 WEC: Satisfied.

Moderator Temperature In accordance with Coefficient Technical Specifications SRP 4.3-11.2 WEC: Satisfied.

5.3 Control Rod Reactivity Technical Specification SRP 4.3-11.3 WEC: Satisfied.

_ _________ ____ ___ ____ ___ margin m antained

Docket 50-305 NRC-03-125 December 15, 2003 , Page 13 The FANP fuel DNBR analyses at SUR conditions were performed by NMC with the VIPRE code as per the methodology described in the KNPP RSE topical report (Refs. 6, 7). For Cycle 26, Westinghouse performed the system and core response conditions for steady-state, anticipated operational occurrences, and transient analyses. The results (statepoints) of the Westinghouse analysis are provided to NMC in the statepoints document, that is part of the reload safety analysis checklist confirmation process. The statepoints are comprised of the limiting sets of system pressure, core inlet temperature, core flow, and core heat flux that bound the set of Chapter 14 DNBR events. All appropriate additional uncertainties (RTDP uncertainties) are accounted for in the statepoint analyses.

Table 2 below describes the disposition of the USAR Chapter 14 events.

Table 2. USAR Non-LOCA Analyses - Fuel-Related Thermal-Hydraulic Analyses USAR Analysis Description ID String Applicability 14.1.1 Uncontrolled RCCA Withdrawal From A Sub-Critical RWFS See Table 3

________ Condition _ _ _ _ _

14.1.2 Uncontrolled RCCA Withdrawal At Power RWAP See Table 3 14.1.3 RCCA Misalignment RMIS See Table 3 14.1.4 Chemical And Volume Control System Malfunction BD N/A(l) 14.1.5 Startup Of An Inactive Reactor Coolant Loop ILS N/A(2) 14.1.6 Excessive Heat Removal Due To Feedwater System FWM See Table 3

____ ____ M alfunctions _ _ _ _ _ _

14.1.7 Excessive Load Increase Incident ELI N/A(')

14.1.8 Loss Of Reactor Coolant Flow LOF See Table 3 14.1.9 Loss Of External Electric Load LOL 14.1.10 Loss Of Normal Feedwater LONF N/A(3) 14.1.11 Anticipated Transients Without SCRAM ATWS N/A(3) 14.1.12 Loss Of AC Power To The Plant Auxiliaries LOOP N/A(4) 14.2.1 Fuel Handling Accidents FHA N/A(3) 14.2.2 Accidental Release - Recycle Of Waste Liquid --- N/A(3) 14.2.3 Accidental Release - Waste Gas --- N/A(3)

Docket 50-305 NRC-03-1 25 December 15, 2003 Attachment 2, Page 14 I Table 2. USAR Non-LOCA Analyses - Fuel-Related Thermal-Hydraulic Analyses Section Analysis Description ID String Applicability 14.2.4 Steam Generator Tube Rupture SGTR N/A(3) 14.2.5 Steam Line Break MSLB See Table 3 14.2.6 Rupture Of A Control Rod Drive Mechanism Housing REA N/A(3)

(RCCA Ejection)

Missile D g o eF o etd_

14.2.7 Turbine Missile Damage To Spent Fuel Pool (Deleted) I --- NA2 N/A - MDNBR results bounded by DNB events listed in RSAC. No other fuel-related thermal-hydraulic analyses required.

(2) N/A - Analysis no longer required (See Ref. 4).

N/A - Not analyzed for MDNBR or other fuel-related thermal-hydraulic acceptance (3)criteria.

(4) N/A - Not analyzed; results bounded by LONF analysis.

Table 3 below lists the RSAC transients that are analyzed for DNBR. The set of RSAC DNBR transients together with the core thermal limits and core axial offset limits provide a bounding set of conditions with which to verify the thermal-hydraulic design basis for the FANP fuel.

Table 3. RSAC Transients DNB Thermal-Transient Hydraulic Event Description in RSAC? Analyses BD Boron Dilution at Power (Mode 1) No N/A BD Boron Dilution at Startup (Mode 2) No N/A BD Boron Dilution at Hot Standby (Mode 3) No N/A BD Boron Dilution at Hot Shutdown (Mode 4) No N/A BD Boron Dilution at Cold Shutdown (Mode 5) No N/A BD Boron Dilution during Refueling (Mode 6) No N/A RWFS Rod Withdrawal from Subcritical - Below Mixing Vane Grid Yes (1)

RWFS Rod Withdrawal from Subcritical - Above Mixing Vane Grid Yes (1)

SRW Single Rod Withdrawal N/A N/A RWAP Rod Withdrawal at Power Yes (1)

Docket 50-305 NRC-03-125 December 15, 2003 Attachment 2, Page 15 Table 3. RSAC Transients DNB Thermal-Transient Hydraulic Event Description in RSAC? Analyses LOF Loss of Flow - Underfrequency Yes (1)

LOF Locked Rotor Rods In DNB Yes (2)

RMIS Rod Out of Position (Static Rod Misalignment) Yes (1)

RMIS Dropped RCCA(s) Yes (3)

REA HZP- BOL No N/A REA HFP-BOL No N/A REA HZP-EOL No N/A REA HFP- EOL No N/A MSLB Credible Break Approach to Criticality After Cooldown with Above No N/A Shutdown Margin No____

MSLB Zero Power Hypothetical Break - 1.4ft2 DER with Offsite Power Yes (1)

Available ____

MSLB Full Power Break N/A N/A FWM Feedwater Malfunction at HZP Yes (4)

(')Departure from Nucleate Boiling Ratio (DNBR) analysis required (2) Peak Clad Temperature (PCT) and % Rods-ln-DNB analysis required (3) Dropped Rod Limit Line (DRLL) analysis required (4) DNBR analysis bounded by HZP MSLB - analysis not required Table 4 presents the actual VIPRE case results performed to verify the DNBR design basis.

Table 4. Cycle 26 MDNBR Results System Inlet Inlet Average Pressure Enthalpy Mass Flux Heat rate Case (psia) (btullbm) (mlbm/hr-ft 2 ) (btu/sec-ft) MDNBR AOL 2374.9 602.0 2.304 5.293 1.633 AOL 2374.9 529.4 2.520 7.809 1.652 AOL 2374.9 659.4 2.109 5.297 1.392 AOL 2374.9 573.6 2.393 7.812 1.502 AOL 2374.9 659.0 2.110 5.277 1.475 l

Docket 50-305 NRC-03-125 December 15,2003 , Page 16 Table 4. Cycle 26 MDNBR Results System Inlet Inlet Average Pressure Enthalpy Mass Flux Heat rate Case (psia) (btu/lbm) (mlbmlhr-ft 2) (btu/sec-ft) MDNBR AOL 2374.9 581.0 2.369 7.784 1.491 AOL 2374.9 679.9 2.046 5.293 1.334 AOL 2374.9 606.1 2.290 7.808 1.342 AOL 2374.9 673.3 2.060 5.294 1.345 AOL 2374.9 600.6 2.308 7.809 1.335 CTL 1649.9 522.0 2.540 7.915 1.455 CTL 1649.9 556.4 2.444 6.596 1.511 CTL 1649.9 573.0 2.395 5.277 1.891 CTL 1749.9 529.9 2.518 7.915 1.472 CTL 1749.9 567.8 2.410 6.596 1.481 CTL 1749.9 584.6 2.358 5.277 1.854 CTL 1949.9 545.5 2.475 7.915 1.494 CTL 1949.9 588.7 2.346 6.596 1.428 CTL 1949.9 607.2 2.286 5.277 1.767 CTL 2199.9 564.3 2.420 7.915 1.501 CTL 2199.9 607.8 2.284 6.596 1.427 CTL 2199.9 635.2 2.193 5.277 1.633 CTL 2374.9 576.7 2.383 7.915 1.500 CTL 2374.9 622.0 2.237 6.596 1.407 CTL 2374.9 655.2 2.123 5.277 1.530 LOF-LR 2226.2 546.9 1.084 0.286* 1.405 LOF-UF 2226.2 546.4 1.689 6.469 1.501 MSLB 701.3 395.5 3.110 1.873 2.433 MSLB 701.3 400.8 3.099 1.875 2.395 RMIS-DRLL 2214.9 545.7 2.475 6.614 2.325 RMIS-DRLL 1649.9 508.4 2.574 7.937 1.552 RMIS-DRLL 1749.9 508.4 2.574 8.136 1.602 RMIS-DRLL 1949.9 508.4 2.574 8.546 1.684

Docket 50-305 NRC-03-125 December 15, 2003 , Page 17 Table 4. Cycle 26 MDNBR Results System Inlet Inlet Average Pressure Enthalpy Mass Flux Heat rate Case (psia) (btu/lbm) (mlbm/hr-ft2) (btu/sec-ft) MDNBR RMIS-DRLL 2199.9 508.4 2.574 9.015 1.754 RMIS-DRLL 2374.9 508.4 2.574 9.253 1.802 RMIS-DRLL 1649.9 539.4 2.492 7.130 1.464 RMIS-DRLL 1749.9 539.4 2.492 7.296 1.515 RMIS-DRLL 1949.9 539.4 2.492 7.659 1.597 RMIS-DRLL 2199.9 539.4 2.492 8.122 1.684 RMIS-DRLL 2374.9 539.4 2.492 8.394 1.740 RMIS-DRLL 1649.9 558.5 2.437 6.654 1.407 RMIS-DRLL 1749.9 558.5 2.437 6.806 1.459 RMIS-DRLL 1949.9 558.5 2.437 7.144 1.546 RMIS-DRLL 2199.9 558.5 2.437 7.587 1.631 RMIS-DRLL 2374.9 558.5 2.437 7.878 1.688 RMIS-SRM 2214.9 545.7 2.475 6.614 1.335 RWAP 2202.6 605.0 2.294 5.417 1.815 RWAP 2363.2 582.7 2.365 7.296 1.571 RWAP 2203.6 602.6 2.302 5.536 1.803 RWAP 2303.4 569.0 2.407 7.540 1.605 RWAP 2254.5 555.0 2.449 7.739 1.676 RWAP 2245.1 555.9 2.445 7.739 1.653 RWFS 2170.0 546.2 1.111 2.569 1.566

• Mbtu/hr-ft 2 Note: ID strings are defined as in Table 1 with the additional definitions:

AOL: Axial Offset Limits CTL: Core Thermal Limits

-LR: Locked Rotor

-UF: Underfrequency Trip

-DRLL: Dropped Rod Limit Lines

-SRM: Static Rod Misalignment

Docket 50-305 NRC-03-125 December 15, 2003 , Page 18 The rod bow penalty calculated by FANP is a linear function of assembly exposure from 0% at 37 GWd/MTU to 17% (a 0.83 multiplier) at 59 GWd/MTU. Since the base analyses are performed at an FDH of 1.38 (i.e., the new FANP fuel COLR limit) and the maximum FDH including uncertainties for the SUR uprate is 1.34 (after 9.5 GWd/MTU core exposure), a conservative FDH burndown credit of 1.38/1.34 can be applied (based on the conservative assumption of a 1 to 1 relation between FDH and DNBR). Therefore, the maximum rod bow penalty that requires consideration is 0.83*(1.38/1.34) = 0.8547. The 95/95 DNBR limit of 1.14 raised to accommodate the maximum rod bow penalty is then 1.14/(0.83*(1.38/1.34)) = 1.3337.

Examination of the DNBR results shows that all cases can accommodate the maximum rod bow penalty of 14.53% (1-0.8547).

The FANP fuel PCT analyses at SUR conditions were performed by NMC with the TOODEE code as per the methodology described in the KNPP RSE topical report (Refs. 6, 7). For Cycle 26, Westinghouse provided the transient normalized power forcing function in addition to the 50% Rods-ln-DNBR Locked Rotor FDH in the statepoints document, that is part of the reload safety analysis checklist confirmation process. All appropriate additional uncertainties are accounted for in the PCT analyses. The PCT is well within the limit of 2700 QF.

Table 5. PCT Results Peak Clad Temperature Case Case (2F) loflr 1 1566

References:

1. Siemens Power Corporation Report, entitled "Generic Mechanical Design Criteria for PWR Fuel Designs", EMF-92-116(P)(A), Revision 0, dated February 1999.

2. U. S. Nuclear Regulatory Commission Letter, from F. Akstulewicz (USNRC) to J. F.

Mallay (SPC), entitled "Acceptance for Referencing of Siemens Power Corporation Topical Report EMF-92-116(P): 'Generic Mechanical Design Criteria for PWR Fuel Designs', (TAC NO. M84245)", dated February 2,1999.

3. Nuclear Management Company Letter from T. Coutu (NMC) to Document Control Desk (NRC), entitled UNMC Responses to NRC Request for Additional Information Concerning License Amendment Request No. 187 to the Kewaunee Nuclear Power Plant Technical Specifications (TAC No. MB571 8)," Docket 50-305, License No. DPR-43, dated February 27, 2003

4. Westinghouse Report, entitled "Reload Transition Safety Report for the Kewaunee Nuclear Power Plant", dated July 2002.

5. Nuclear Management Company Letter from T. Coutu (N MC) to Document Control Desk (NRC), entitled "License Amendment Request No. 187 to the Kewaunee Nuclear Power Plant Technical Specifications (TAC No. MB5718)," Docket 50-305, License No. DPR-43, dated March 19, 2003

Docket 50-305 NRC-03-125 December 15,2003 , Page 19

6. Nuclear Management Company Letter, from K. H. Weinhaue (NMC) to Document Control Desk (NRC), entitled 'Wisconsin Public Service Corporation Reload Safety Evaluation Methods Topical Report", WPSRSEM-NP, Revision 3, dated October 12, 2000.

7. NRC Letter, from J. G. Lamb (NRC) to M. Reddemann (NMC), entitled "Kewaunee Nuclear Power Plant - Review for Kewaunee Reload Safety Methods Topical Report WPSRSEM-NP, Revision 3 (TAC NO. MB0306)", dated September 10, 2001.

RAI #29: The licensee is performing calculations which are not in accordance with currently approved methodology. In order for the staff to review and approve a new calculational method, significant additional information must be submitted to the staff. Specifically, the RAI response does not discuss the method used to calculate the FdH burndown credit taken to offset rod bow DNBR penalty. However, the staff has some question as to the need to use this methodology, as it appears that the licensee has adequate margin to DNBR limits even if the rod bow penalty is applied.

RAI #29 Staff Concern NMC Response: As described in the previous response concerning the rod bow penalty, the methodology used to eliminate the rod bow penalty is essentially identical to the approved methodology of WCAP-8691, Rev. 1. The specific differences with the approved methodology are: (1) the KNPP Cycle 26 reload core is explicitly modeled, whereas in WCAP-8691, Rev. 1 a representative set of reload cores were modeled and (2) the results of the KNPP Cycle 26 specific model afforded the elimination of the 2.6% penalty for all cycle exposures.

The licensee is performing calculations in accordance with the currently approved methodology.

The exceptions noted above do not materially alter this statement. Using this approved method, it was demonstrated that the FdH burndown credit conservatively offsets the rod bow DNBR penalty over the entire KNPP Cycle 26. The rod bow DNBR penalty can therefore be evaluated at BOL and thus be reduced from 2.6 to 0% for this specific cycle. It is acknowledged that according to the current methodology, the rod bow DNBR penalty must be evaluated at a burnup of 24,000 MWD/MTU, the NRC approval being based on a representative set of fuel cycles using Westinghouse fuel. The change is therefore not in calculation method but in burnup limit at which the penalty is applied, the method presented here determined a cycle specific limit based on KNPP Cycle 26 only while the approved method determined a generic limit.

The staff is not being asked to review and approve a new calculational methodology per se. It is more accurate to characterize the request as the staff being asked to review a specific application of the existing methodology, applied with the best available cycle-specific information, for concurrence with the conclusion, which is supported by this specific application, that the 2.6% rod bow penalty can be eliminated for KNPP Cycle 26. Therefore, significant additional information does not need to be submitted.

The previous response concerning the rod bow penalty does discuss the method used to calculate the FdH burndown credit. The previous response states "The FdH burndown credit was based on a conservative one-to-one relation with DNBR". For clarification, the FdH burndown credit at a given burnup is calculated as the ratio of the peak FdH at the beginning of cycle to the peak FdH at the given assembly average burnup. The use of this ratio without alteration implies that a one-to-one relationship between a change in FdH and a change in DNBR was assumed, which is clearly conservative. This is a more conservative methodology

Docket 50-305 NRC-03-125 December 15, 2003 , Page 20 as was used in WCAP-8691, Rev. for the FdH burndown credit where a larger DNBR sensitivity to FdH was used.

The need for this methodology is to avoid close to no DNBR margin (0.3%) for Cycle 26 under Stretch Uprate Rate (SUR) conditions and therefore to provide working analysis margin to address subsequent issues should any arise for Cycle 26 under SUR conditions.

RAI #40: The existing radiological consequences of a steam generator tube rupture (STGR) accident assumes that the leak flow from the reactor coolant system (RCS) to secondary side of the steam generator (SG) is terminated in 30 minutes following the event initiation. The analysis uses an equilibrium break flow that continues at a constant rate for 30 minutes. The resulting break flow mass transfer is then used to calculate the radiological consequences of the SGTR.

You have considered that this assumption will result in a conservative calculation since the reduction in the break flow rate over the 30 minutes time is ignored. Inherent in this evaluation is the assumption that the operator can terminate the break flow in 30 minutes. Plants with similar design to KNPP have reported to NRC that in simulator exercises, the operators demonstrated that the time to terminate the break flow exceeded the 30 minute assumption.

Should operator significantly exceed the 30 minute termination criteria, this event could lead to an increase in radiological releases from that assumed when the leak flow was terminated within 30 minutes. Please provide the necessary information to confirm that the assumption of termination of the break flow in 30 minutes following a design basis SGTR event is valid for KNPP at the uprated power level of 1772 MWt and will indeed lead to a bounding calculation regarding to the radiological consequences of the event.

RAI #40 Staff Concern NMC Response: Although it is considered beyond the licensing basis for Kewaunee, detailed thermal-hydraulic analyses have been performed for the power uprate with the RETRAN code to evaluate the impact on the radiological consequences of the SGTR break flow continuing longer than the 30 minutes considered in the licensing basis and to evaluate the potential for steam generator overfill. The analyses used operator action times supported by Kewaunee simulator exercises.

The analyses modeled the recovery actions leading to break flow termination to determine the primary to secondary break flow, amount of break flow that flashes immediately to steam, and the steam releases to the atmosphere. The analysis used operator action times, supported by Kewaunee simulator exercises, leading to steam generator isolation within 30 minutes and break flow termination at approximately 49 minutes. The mass transfer data was used to calculate offsite and control room doses, which were compared to the licensing basis analysis.

The licensing basis analysis considers a constant break flow for 30 minutes and does not model the expected operator responses. The comparison showed that the licensing basis assumptions are limiting. Even though the operator may not terminate break flow to the ruptured steam generator within 30 minutes, the licensing basis calculations performed using this assumption provide conservatively high offsite and control room dose results.

The potential for steam generator overfill following a steam generator tube rupture was also evaluated for Kewaunee. The results of the analysis indicate that recovery actions can be performed to terminate the primary to secondary break flow before overfill of the ruptured steam generator would occur. The analysis used operator action times, supported by Kewaunee simulator exercises, leading to break flow terminated at approximately 49 minutes. Therefore,

Docket 50-305 NRC-03-1 25 December 15, 2003 , Page 21 water release does not have to be considered in the radiological consequences analyses, even with break flow continuing beyond the 30 minutes assumed in the licensing basis analysis.

RAI #43: Please explain why reduce the assumed initial RCS pressure to 2000 psia will provide conservative results of transient and accident analyses especially for events lead to higher peak RCS pressures.

RAI #43 Staff Concern NMC Response: The table in the previous response to RAI# 43 presented a summary of the initial conditions assumed for the safety analysis operating parameters. The 2000 psia initial RCS pressure is the conservative initial RCS pressure assumption for transients where DNBR is limiting. Use of 2000 psia for RCS pressure may also be a conservative initial assumption for transients where peak RCS pressure results are limiting.

The direction of conservatism for initial condition assumptions depends on the transient being analyzed. With respect to the non-LOCA analyses, Tables 5.1-2 and 5.1-7 of the RTSR (see Attachment B of Letter #NRC-03-01 6 of 2/27/03 to NRC RAIs from LAR 187 submittal for fuel transition) summarize the plant initial conditions that were applied. Initial conditions are chosen to minimize specific transient margin to the limit for the various parameters used in the acceptance criteria. Depending on the specific transient and specific parameter of interest, minimum or maximum initial conditions may be assumed.

With respect to peak reactor coolant system pressure, the bounding transient is the loss of load

/ turbine trip (LOL/TT) transient. This is because the LOL/TT transient results in a limiting power mismatch between the primary and secondary sides. The initial pressurizer pressure assumed in the loss of load analysis is the nominal value minus uncertainties (2250 psia-50 psi=2200 psia) because this conservatively delays the reactor trip on high pressurizer pressure. The LOLITT transient analysis shows the RCS pressure acceptance criteria is satisfied.

RAI #44: Please discuss the change of feedwater temperature responses affecting the current NSSS design transients.

RAI #44 Staff Concern NMC Response: The change in feedwater temperature (Tfeed) is 437.1 F at full power, to 320 F (conservative value) at no-load for the stretch power uprate. This change in Tfeed is slightly greater than the 427.30F at full power, to 32 0F (conservative value) at no-load Tfed change used for the replacement steam generator (RSG) Program.

For the Tfeed transient response, a design transient reanalysis was required for several design transients. This was because the higher full power feedwater temperature resulted in a larger change in feedwater temperature for any design transient that resulted in a power level change.

For the stretch power uprate the feedwater temperature transient response was therefore revised for the following NSSS design transients:

Unit Loading Unit Unloading Large Step Load Decrease Loss of Load Loss of Power Loss of Flow Reactor Trip

Docket 50-305 NRC-03-125 December 15,2003 , Page 22 The revised feedwater temperature design transients were used in the analysis of the various NSSS components as described in Section 5.0 of Attachment 4 of our submittal (WCAP 16040-P). Of the components listed in that section, only the steam generators are affected. As described in Section 5.7.2.4 of WCAP 16040-P, the effect of the revised design transients were included in the steam generator evaluation.

RAI #51: The NRC staff has recently become aware, and has discussed with NEI, a problem with use of instrument setpoint methodology if based on "Method 3" of ANSI/ISA-S67.04, "Setpoints for Nuclear Safety-Related Instrumentation." Does Kewaunee use this method?

RAI #51 Staff Concern NMC Response: Kewaunee (KNPP) uses a variation of Method 3, that in reality, combines the attributes of Methods 1 and 3, as described in ANSI/ISA-S67.04. KNPP first calculates the Nominal Trip Setpoint (NTSP) from the Analytical Limit (AL), and then calculates the Allowable Value (AV) from the NTSP. This has attributes of Method 3 described in the ISA standard. Where KNPP differs, is that in calculating the NTSP, KNPP accounts for channel uncertainties, channel instrument drift, and calibration uncertainties. AV is then calculated from the NTSP by accounting for instrument drift and calibration uncertainties. The end result is that our calculated NTSP is the same value as it would be if Method 1 was utilized.

This results in the difference between the AL and the AV always being 2 the total channel uncertainty.

Additional Staff Discussion Items Item 1.Why was a supplemental table of mass and energy for MSLB (Table 6) added to November 5, 2003 letter?

Item 1 NMC Response: The supplemental MSLB mass and energy release (M&E) table was included because the MSLB M&E calculation and subsequent containment integrity analyses have been revised. The revision was necessary to address NRC SER restrictions on the use of the GOTHIC with MDLM model, which eroded MSLB containment integrity analysis pressure margin. Margin recovery was achieved by crediting increased core shutdown margin in the revised MSLB M&E calculation and by performing a revised containment response calculation using the revised M&E, revised heat sinks, and including a containment shield building model.

Item 2. Explain why the accumulators are a heat sink for the main steam line break.

Item 2 NMC Response: The currently approved containment evaluation model (EM) for main steam line break (MSLB) containment integrity analysis (CIA) credits the accumulators as a passive heat sink. The accumulators are large metal tanks located inside containment that are, for the duration of the MSLB transient, filled with nitrogen and water and remain below the containment atmospheric temperature. Conservatively modeling the accumulators as heat sinks improves the containment temperature and pressure transient response. The crediting of the accumulators is justified since the accumulators meet the basic requirements for classification as a passive heat sink as indicated above.

Docket 50-305 NRC-03-1 25 December 15, 2003 , Page 23 Item 3.WCAP 10325 does not mention specifically the FROTH and EPITOME computer codes.

Is there another W document which discusses these codes or an NRC approval of these codes?

Item 3 NMC Response: The word code" is not used for FROTH, but Section 2.3 on page 2-8 in WCAP-10325-P-A is titled FROTH MODEL". Reference 1 in WCAP-10325-P-A is WCAP-8264-P-A. WCAP-8264 is the base methodology that was revised by WCAP-10325. WCAP-8264-P-A, on pages 11-1-19 through 11-1 -26, called out the FROTH code by name and provides a description of the code.

The EPITOME code is not called by name in either WCAP 10325 or WCAP 8264 but the function of the code is described in the last paragraph on page 3-4 of WCAP-10325-P-A, and from page 11-1-28 through 11-1-34 of WCAP-8264-P-A for the mass and energy balance tables and the intact and broken loop steam generator equilibration calculations. The EPITOME code also performs the decay heat calculations after 3600 seconds with the model that is described in Section 2.4 of WCAP-1 0325-P-A.

Item 4. Describe model for calculating heat transfer to shield building.

Item 4 NMC Response: The shield building is modeled as a separate closed volume initially filled with air at 1 atmosphere and 120F. Natural convection heat transfer is modeled from the outside of the containment shell to the volume representing the shield building. The inside of the containment shell sees natural convection and condensation heat transfer from the containment atmosphere.

Item 5. Do the calculations for high energy line breaks (HELBs) outside containment follow the Standard Review Plan (SRP 6.2.1.2) guidance? In particular, is the entrainment modeled as 100% or is a GOTHIC option used (drop-liquid conversion or equivalent) which results in less than 100 entrainment? The drop-liquid conversion is discussed in Section 8.7 of the GOTHIC 7.0 manual.

Item 5 NMC Response: The temperature response and thermal lag calculations that are performed for the outside containment high energy line breaks with GOTHIC do not follow the guidance in Standard Review Plan 6.2.1.2 for inside containment short term sub-compartment pressurization calculations. The temperature response analysis for HELB's outside containment follows guidance from NUREG-0588 and/or Regulatory Guide 1.89 for equipment qualification calculations.

The droplet option in GOTHIC for the high energy line break flow was not set to entrain any liquid. If the code determined that any liquid would exist in the break flow, that liquid would fall out to the floor. It would not be entrained in the break flow.

Item 6. During a phone call on 12/8/03, John Lamb discussed an additional concern, or clarification the NRC requires:

Docket 50-305 NRC-03-125 December 15,2003 , Page 24 In our submittal we made the following statements:

Cover Letter The NMC has determined that the information for the proposed amendments does not involve a significant hazards consideration, authorize a significant change in the types or total amounts of effluent release, or result in any significant increase in individual or cumulative occupational radiation exposure.

Therefore, the proposed amendments meet the categorical exclusion requirements of 10 CFR 51.22(c)(9) and an environmental impact appraisal need not be prepared.

Attachment 1 page 16 An environmental review summary is contained in attachment 4, report section 8.10, "Environmental Assessment." The conclusion of this review is that the proposed amendments do not involve a significant change in the types of or significant increase in the amounts of any effluents that may be released off site or a significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22 (b), an environmental assessment of the proposed change is not required.

In the Licensing Report (WCAP-16040-P), Section 8.8, Radiological Assessments, Section 8.8.2.3, Results (pages 8-131 & 132), under both "Liquid Effluents" and "Gaseous Effluents",

the report makes statements about noble gas activity changes that the NRC found confusing.

Under "Liquid Effluents", we talk about increases in Tritium of 17.6%, but that the actual increase is only about 11.4%. We then mention part of the reason is because there is not a change in "mode of operation". We also say that strict adherence to NUREG-001 7 would have identified no change in tritium level. The NRC wants clarification on why we did not just stick with the NUREG evaluation, instead of confusing the issue with 17.6% and 11.4%

values. And they were not sure what was meant by "mode of operation".

Under "Gaseous Effluents", similar comments apply. Apparently they do not see a clear logic between the increases we are addressing and the conclusions in 8.8.2.4, Conclusions.

They do not necessarily want to see the analyses, they are confused because the report states that the NUREG method would not have identified a change, but the method we applied does identify increases.

Item 6 NMC Response: Based on an original licensed core power level of 1650 MWL and an uprate core power level of 1772 MWt, it is expected that the radioactive effluents and consequent off-site doses will increase by approximately the percentage increase in core power, i.e. 7.4%.

To estimate the impact of power uprate on radioactive effluents and consequent off-site doses, the uprate analysis utilizes NUREG 0017 equations and assumptions, conservative

Docket 50-305 NRC-03-125 December 15,2003 , Page 25 methodology, and plant operating history relative to radwaste effluents and subsequent doses, by scaling plant operation at 1600.5 MWt (the effective core power level during the 5 year period 1996-2000), to estimate the impact of operating at the uprated core power level of 1782.6 MWt (which includes margin for power uncertainty).

Based on a comparison of plant coolant system parameters (e.g., RCS mass, SG liquid mass, steam flow rate, RCS letdown flow rate, flow rate to cation demineralizer, letdown flow rate for boron control, SG blowdown flow rate, SG moisture carryover, etc.), for both pre-uprate and uprate conditions, the maximum potential increase in coolant activity levels due to uprate, for each chemical group identified in NUREG 0017, was estimated using scaling techniques. The uprate analysis conservatively used the worst-case scaling factor to establish the impact of uprate on plant radwaste effluents and subsequent doses.

Briefly, based on the conservative methodology outlined above, the impact of uprate is as follows:

Reactor Coolant concentrations:

Based on a comparison of pre-uprate vs. uprate input parameters, and the methodology outlined in NUREG 0017, the maximum expected increase in the reactor coolant source between operation at 1600.5 MWt to 1782.6 MWt is approximately 17.6% for noble gases, and 11.4 % for other long half life non-gaseous activity. This is primarily due to differences between the pre and post uprate conditions, i.e., a decrease in RCS mass (-5.5 %), and an increase in the effective core power level (-11.4 % [effective power increase is 1782.6 MW/1 600.5 MW,

{average power level during 1996 - 2000) = 1.114]) The uprate analysis conservatively used the worst case scaling factor of 1.176 for all the isotopes in reactor coolant.

Liquid Effluents There will be a maximum 11.4% increase in activity in liquid effluents based on the pre-uprate case (based on the 5 year period) as activities entering the liquid waste system are based on long half life RCS activity, which is proportional to the effective core uprate percentage increase, and on waste volumes, which are essentially independent of power level within the applicability range of NUREG 0017.

Tritium releases in liquid effluents would also increase approximately 11.4% (corresponding to the effective increase in core power) since the facility is changing its power rating, without changing its operational procedures / philosophy (i.e., mode of operation).

However, for all liquid releases, the uprate analysis conservatively used the worst case scaling factor for all isotopes between the pre-uprate case and the uprate case; i.e.; 1.176 Gaseous Effluents For all noble gases, there will be a maximum 17.6% in activity in gaseous effluents due to the effective increase in core power level. Noble gas effluents have two components, one which is based on RCS inventory and results in a 11.4% increase, and the second which is based on RCS concentration which results in a 17.6% increase between the pre and post uprate conditions. With the exception of the Kr-85 (long half-life), decay is the dominant removal mechanism for noble gases in the RCS.

Docket 50-305 NRC-03-125 December 15, 2003 , Page 26 In actuality, gaseous releases of Kr-85 will increase by approximately the effective percentage increase in power (11.4%). Gaseous isotopes with shorter half-lives will have increases slightly greater than the effective percentage increase in power level up to the bounding value of 17.6%.

For particulates, the methodology of NUREG-0017 specifies the release rate per year per unit per building ventilation system. This is not dependent on power level within the range of applicability. However, particulates due to steam releases will increase by approximately the effective percentage increase in power. The impact of power uprate on particulates can therefore be approximated by the effective increase in core power, i.e., 11.4 %. The impact of power uprate on iodines and tritium will also be limited to the effective increase in core power, i.e., 11.4 %.

However, for all gaseous releases, the uprate analysis conservatively used the worst case scaling factor between the pre-uprate case and the uprate case, i.e.; 1.176 Results & Conclusions The estimated doses following the uprate based on the above methodology are presented in Table 6. It is concluded that the estimated doses due to annual radwaste effluent releases following the uprate will remain a small percentage of the allowable 10CFR50 Appendix I limits.

Docket 50-305 NRC-03-125 December 15,2003 , Page 27 TABLE 6 Estimated Impact of Power Uprate on Annual Effluent Doses 5 Yr Annual Maximum Average Estimated Percentag Doses for Dose Impact e of Appendix I Maximum Following Appendix I Type of Dose Design Individual Uprate for Design Objectives Based on Maximum Objectives Pre-Uprate Individual for Uprate 5-yr Based on Pre- Case Operations* Uprate 5-yr Operations

___Operations**

Liquid Effluents Dose to total body 3 mrem/yr 4.56E-03 5.37E-03 0.18%

from all pathways per unit mrem/r mrem/yr ._ _

Dose to any organ from all 10 mrem/yr 2.78E-02 3.27E-02 0.33%

pathways per unit mrem/yr mrem/yr Gaseous Effluents Gamma Dose in 10 mrad/yr 2.56E-05 3.01 E-05 3.01 E-04%

Air per unit Wyr mrad/yr Beta Dose in Air 20 mrad/yr 1.02E-05 1.19E-05 5.96E-05%

per unit mrad/yr mrad/yr Dose to total body Based on of an individual above airborne 5 mremfdyr per Notdoses, unit Not reported expected to be well within Appendix I guidelines Dose to skin of an Based on individual above airborne 15 mremn/yr Notdoses, per unit Not reported expected to be well within Appendix I guidelines Radioiodines and Particulates Released to the Atmosphere Dose to any organ 15 mrem/yr 5.57E-04 6.55E-04 4.36E-03%

from all pathways per unit mrem/yr mrem/yr I Notes:

• Average Core Power Level for the 5-year operation (base case) is 1600.5 MWt

Docket 50-305 NRC-03-1 25 December 15, 2003 , Page 28

• Core Power Level assumed for uprate analysis is 1782.6 MWt (includes margin for power uncertainty)

Item 7.TS 3.3.c.1.A.3.(iii) is being deleted because 2 trains of containment spray (CS) as well as one containment fancoil unit train are required to support the containment integrity analysis. If this is the case, why are current TSs 3.3.c.1 .A.3.(ii) and (iv) still acceptable?

Have they been shown to produce acceptable results at the uprated power level?

Item 7 NMC Response: The combinations of containment spray trains and containment fan coil unit trains that provide sufficient cooling are listed in KNPP TS Basis page TS B3.3-3. It states that the acceptable combinations are 1) four fan coil units, or 2) two fan coil units and one spray pump (LAR 195 attachment 2 or 3, first paragraph, last sentence). As there are two fan coil units in a train one fan coil unit out-of-service renders the train out-of-service. The only thing that has changed here with the uprate is that both spray trains will no longer remove all the post accident heat.

Current TS's 3.3.c.1 .A.3.ii and 3.3.c.1.A.3.iv are acceptable for power uprate conditions since they are TS LCO's that are consistent with the safety analysis assumptions on containment cooling for the power uprate containment integrity analyses.

Power Uprate safety analysis assumptions for the containment integrity analyses include the loss of a containment cooling safeguard train. Design considerations combine containment cooling equipment into two trains, each train consisting of two containment fan cooling units (CFCU) and one containment spray pump. Each train is fully capable of controlling post-accident containment temperature and pressure, and is electrically and mechanically separated from the opposite train. Each train will remain capable of its design safety function after power uprate.

Therefore the LCO's in TS 3.3.c.1.A.3 ii and iv, which are consistent with the containment integrity analysis containment cooling system configuration assumptions, are acceptable LCO's since the power uprate safety analyses have been shown to have acceptable results satisfying all applicable acceptance criteria.

TS 3.3.c.1.A.3 (ii) One containment spray train may be out of service for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided the opposite containment spray train remains OPERABLE.

(iv) The same containment fan coil unit and containment spray trains may be out of service for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided their opposite containment fan coil unit and containment spray trains remain OPERABLE.

Item 8.The licensee had not performed the first in-service inspection at the time of the power uprate submittal, therefore stating in it that no loose parts were present in their steam generators (SGs). Kewaunee submitted a license amendment request 199, "Steam Generators Eddy Current Inspection Frequency Extension" where they detected possible loose parts (PLPs) in their SGs. Since this information came up after the submittal, in

Docket 50-305 NRC-03-125 December 15,2003 , Page 29 our review of the uprate we have to acknowledge this PLPs that are present now.

Although in their submittal for the power uprate the licensee stated that a generic loose parts was prepared to address undefined PLPs, we will like to know if the PLPs found are still bounded by this evaluation, and if not, whether the licensee has performed another evaluation for leaving the PLPs in service at the new uprate conditions.

Item 8 NMC Response: A generic loose parts analysis was prepared as part of the Power Uprate project to address undefined potential loose parts (PLPs). This generic analysis utilizes parameters applicable to the 7.4% uprated NSSS power level of 1780 MWt with a maximum 10% tube plugging level per steam generator. This generic report provides the methodology to perform steam generator secondary side loose part evaluations to help determine safe operating intervals based on an estimated wear time to reach tube minimum allowable wall thickness.

During the April 2003 refueling outage, PLPs were identified during the inservice examination.

An evaluation was performed to address the size and location of the specific loose parts that were left in service during the 2003 outage. This evaluation was performed assuming uprated conditions utilizing the methodology contained in the generic report. The evaluation was performed assuming that the loose object is located at the tube that exhibits the limiting amplitudes of vibration at uprated conditions, with appropriate values for minimum acceptable tube wall thickness, cross flow fluid velocity and fluid densities at uprated conditions.

Additional conservatisms used in the analysis of the loose parts identified during the Spring 2003 refueling outage include assumptions that the tube has existing 20% though wall degradation, that the object rests upon a sludge pile approximately 6 inches deep, that the object will remain in the same location, and that only the tube (and not the loose part) will experience wear.

Results from the evaluation of the loose part show that the minimum estimated impact-sliding wear time to reach the minimum allowable tube wall thickness at uprated power conditions is 43 months. This result is consistent with the generic evaluation results for a horizontal rectangular wear scar of similar geometry as the loose part identified during the spring 2003 refueling outage.

In summary, an evaluation was performed to address the size and location of the specific loose parts that were left in service during the 2003 outage. Results of the evaluation, under uprated power conditions, demonstrate that the minimum estimated impact-sliding wear time to reach the minimum allowable tube wall thickness is 43 months. A subsequent inservice inspection is required within 40 months by the KNPP Technical Specifications. During this subsequent inservice inspection the affected tube will be re-inspected to determine if tube degradation has occurred and if additional actions (such as plugging) are required.

Item 9. Please assure that adequate margin exists between the Analytical Limit (AL) and the Allowable Value (AV) that equals or exceeds the value of uncertainties not measured during the channel operational test (COT). Please confirm that this is true for all protection system setpoints. Please provide an example calculation which demonstrates the existence of adequate margin; choose from one of the reactor setpoints being revised, Overtemperature delta T (OTDT) or Overpower delta T (OPDT). Please provide

Docket 50-305 NRC-03-125 December 15,2003 , Page 30 assurance to the staff that TS setting limits will not be revised using method 3 without NRC approval.

Item 9, NMC Response: All protection system setpoint calculations performed under the current program have been reviewed. With the exception of three containment pressure actuation setpoints, all associated Technical Specification (TS) Setting Limit values have a margin to their Analytical Limit (AL) greater than the calculated Total Loop Error (which includes both unmeasured uncertainties and measured uncertainties) for that loop.

The three containment pressure actuation loops do not have adequate margin between the TS Setting Limit values and their associated ALs. This was discovered in June of this year while performing setpoint calculations for these loops. A change to the TS Setting Limit values is being processed, and this is being tracked by the KNPP Corrective Action Program. The AL value for these containment pressure loops is not changing for the stretch power uprate.

WCAP-1 5821 -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.

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 AL and the COLR OPDT Trip Setpoint, which is > the total CSA of 0.042.

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.

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

Reference(s):

1. WCAP-15821-P, Rev.0, 'Westinghouse Protection System Setpoint Methodology, Kewaunee Nuclear Plant (Power Uprate to 1757 MWt-NSSS Power with Feedwater Venturis, or 1780 MWt-NSSS Power with Ultrasonic Flow Measurements, and 54F Replacement Steam Generators)"

ENCLOSURE A NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 December 15, 2003 Letter from Thomas Coutu (N MC)

To Document Control Desk (NRC)

Responses to NRC Clarification Questions on Responses to Requests for Additional Information Regarding LAR 195 Kewaunee Nuclear Power Plant Calculation C1 1473, "Post Accident Operating Time Evaluations for Various Safety-Related Equipment in Support of Power Uprate,"

Revision 1, dated 11/18/03 50 Pages to Follow

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 IRev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 Page 1 of 42 Prepared By H t T Reviewed By /on~acrj F. a 1.0 PURPOSE The purpose of this calculation is to demonstrate that the post-accident operability time, of various important-to-safety equipment located inside containment, meets the KNPP requirement using the post-DBE temperatures postulated to exist following the planned power uprate.

2.0 BACKGROUND

The post-accident operability of important-to-safety equipment at KNPP has been previously addressed as part of the K2NPP EQ program to meet the requirements of 10CFR50.49. A proposed change to the plant to increase generation by 7.4% will result in a different accident/post-accident environmental profile inside containment.

The original issue of this calculation was based on thermal analysis of the mass and energy release postulated for line breaks following the 7.4% power uprate. This initial analysis did not include the consequences of the new conditions on long term cooling. Two very minor changes to the qualification temperature envelope developed in the original version of this calculation were necessary and have been incorporated by this revision. Other editorial changes have also been incorporated to improved clarity and make consistent with prcocedure.

3.0 INPUTS AND ASSUMPTIONS 3.1 Due to the power uprate, the accident/post-accident temperature profiles inside containment will change. These profile changes impact all important-to-safety equipment located inside containment.

3.2 Figure 1 of this calculation shows four temperature profiles, the existing qualification profile, the new postulated LOCA and MSLB combined profiles from Reference 5.1, and the new qualification envelope profile.

3.2.1 The existing profile is the basis for the current qualification of all equipment located inside containment. It is shown as Figure E-1 in Reference 5.2.

3.2.2 The new LOCA and MSLB profiles are composites of the maximum temperature for the analyzed break cases (LOCA and MSLB, Reference 5.1) and are shown on Figure I of this calculation.

3.2.3 The qualification envelope curve is generated from the new LOCA and MSLB profiles described above and is shown in Table 3.1. It is a curve consisting of the peak temperature at various times from the various cases analyzed for the power uprate (Reference 5.1). The various curves were analvzed to determine discrete time segments and the maximum temperature for each time segment wvas selected.

WISCONSIN PUBLIC SERVICE CORP. Section No. C1 1473 l Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 2 of 42 l Thus the temperature curve is not the result of a single analyzed case, rather it is a conservative compilation of the maximum values from the analyzed cases.

The peak temperature used for this profile is the saturation temperature of the containment design pressure (293 0F at 46 psig, Reference 5.4). This envelope profile is generated to simplify the evaluation by breaking the decay portion of the new curve into "steps" above and over the new curve.

Table 3.1 Post-accident Plant Conditions (Envelope)

Temperature Duration 293F 1 I00 sec 268'F 0.25 hr I 260F 0.44 hr 235 0F 0.39 hr 219 0F 1.1 hr 213 0F 3.33 hr 198 0F 22.22 hr 165 0F 55.56 hr 153 0F I I11.11 hr 138 0F 55.56 hr 135 0F 27.78 hr 132°F 138.89 hr 128°F 138.89 hr 124°F 138.89 hr 122°F 138.89 hr 120°F 7926.67 hr 3.2.4 The existing profile showed a return to 120°F at 11.6 days. The new analysis profile of the breaks, presented in Reference 5.1 Attachment 3, was performed for a ten million-second (I E7) period (approximately 115 days). Per Reference 5.1 Attachment 3, the temperature at three million seconds (3E6, approximately 34.7 days) is 120.0°F for the worst-case break (double-ended pump suction break, minimum safeguards case). The temperature can be assumed to remain there or below for the remainder of the 365 days.

3.2.5 Comparison of the postulated peaks of the post-SGR profile to the new LOCA/MSLB profile for power uprate reveals the following:

3.2.5.1 The post-SGR profile has a peak temperature of 270.2°F (Reference 5.3, Table 6.4-9), and returns to normal (120 0 F) after 11.6 days.

3.2.5.2 The new LOCARVISLB profile has a peak temperature of 266.6°F, and returns to normal (120°F) after 34.7 days.

3.3 The peak temperature of the new LOCA/MSLB profile (266.6°F, Reference 5.1, Table

7) and the peak temperature of the new qualification envelope profile (293°F, based on the saturation temperature for the peak pressure) are enveloped by the existing envelope peak temperature of 293°F used for inside containment.

....... - .' - .- .. .. - . .. ' - - w%  % -'- . .'.: .. .... : '-" ,.

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 I Rev. I I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 3 of 42 i 3.4 The new qualification envelope profile exceeds the existing envelope profile during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by temperatures ranging from 80 F to 10 F. This includes 16.5 minutes at 268'F (8 0 F above 260'F), 23.3 minutes at 2350 F (4 0 F above 2310 F), 66.7 minutes at 2191F ( 0 F above 218'F) 200 minutes at 2130 F (F above 210 0 F) and 22.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at 1981F (8 0 F above 190'F). However, with the exception of the 16.5 minutes at 2680 F at the beginning of the profile, all of the remaining differences are below the EQ accident profile used prior to the steam generator replacement, which remained above 2280 F for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. As such those differences do not impact the qualification of the equipment. For the 8F difference at 2680 F for the first 16.5 minutes of the accident, the EQER's applicable to equipment inside containment were reviewed and it was verified that the peak accident temperature exceeded 2680 F during the first 16.5 minutes. Table 3.2 shows the EQER number, temperature and duration at that temperature. These are not always the peak temperatures or the entire profile, rather the temperature reported at 16.5 minutes and its duration.

Table 3.2

___________ Initial 15 Minute Temperature and Peak Pressure Evaluation EQER 1 Temperature! Peak Pressure EQER Temperature! Peak Pressure I Duration psig I Duration psig 5.1 5.2 1j 3450 F/8 hr 68 _ _ _ _4__

34.1 l _ _

3450 F13hr

_ _ _ _ h r_

_3 _

112

_ _112 _

5.2 l 346F 8 hr 1 110 34.2 341F / 6 hr 110 7.1 I 318F/75 in 115 I 34.3 345 0 F/3 hr 114 9.1 _ 346°Fi5.6hr j 113 I 34.4 345°F/3 hr 104 9.2 10.1 I

1 390°F/24 min 340°F / 17 min 66 75 I 34.5 34.8 3410 F3 hr 345°F 6 hr T 112 116 12.1 I 320°F/6hr 1 70 35.1 400°F/26rmin 100 12.2 l 300°F / 6 hr 53 3 5.2 400°F /39 rin 72 12.3 l 370°F /20 in 75 35.3 340°F /2.7 hr 77 12.4 325F /20 min o 1 35.4 282°F 62 in 49 12.5 370F /5.6 hr 80 36.1 314°F /32 hr 66 14.1 270°F124hr J 52 I 36.2 350°F/6hr 120 14.2 I 270°F/24hr 52 l 36.3 314°F/32hr 66 14.3 I 270 0 F/24 hr 52 I 37.2 300°F /3 hr 80 14.4 1 270°F/24 hr 52 l 38.1 341°F16hr 114 14.5 I 270°F / 24 hr 52 l 38.2 340°F / 6 hr 105 14.6 i 270°F/24 hr 52 ] 3S.3 340°F / 6 hr 122 20.1 I 320°F/2.8hr 85 I 38.4 343°F/6hr l 120 21.1 1 300°F/2hr 65 1 38.5 22.1 23.1

! 355°F /5 hr l 311 F ! 30 min 70 70 39.3 39.4 l

I 351°F/6hr 320°F /2.8 hr 325°F / 8 hr 132 85 85 25.1 300 0F/ 2 hr 77 39.6 1 330°F / 8 hr 85 27.1 1 346°F i 3 hr 139 I 43.1 346°F / 8 hr 113 27.2 27.3

!1 346°F /3 hr 300°F f 10 hr 139 83.5  !

l 43.2 47.1 312F /7.8 hr 320°F! 12.7 hr 66 113 27.4 340°F 3 hr 95.5 48.2 1 340F /20 min 75

. . .. . . . . . .- ' -... - .. . . . .*.. ... . .. ~ -.-- '

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 4 of 42 Table 3.2 Tnitil 1 Minute Temperature and Peak Pressure Evaluation EQER Temperature! Peak Pressure EQER Temperature! Peak Pressure Duration psDuration psig 29.1 310F / 24 min 70 l 48.3 340'F / 26 min 65 29.2 I 329 0F/I hr 90 49.1 I 340 0F/3 hr 104 31.1 l 350 0F/3 hr 80  ! 52.2 1 340 0F / 3 hr 113 31.2 I 295F/ 40 min [ 64 1 53.1 350F / 3hr 70 33.1 i 340F/3.l hr  ! 80  ! 54.1 350 0 F/3hr 81 33.2 I 340 0 F hr I 90 l 54.3 347 0F1"3 hr 120 33.5 1 3520F1 1 hr 74.5 l 55.1 3400 F! 17 min 78 3.5 The calculations and analysis performed are consistent with that in the appropriate calculation or EQER for that equipment. If the existing evaluations were performed in accordance with the guidance provided in the DOR Guidelines, the evaluation herein also followed that guidance.

3.6 The following EQER's have been excluded from this evaluation as the equipment evaluated is only located outside containment, is required to operate in a radiation harsh environment post-DBE, or is not installed. As such a change to the temperature-pressure inside containment resulting from power uprate will not result in a corresponding change to the post-DBE environment for this equipment.

Table 3.3 Excluded EQER's Rai;at;nn onlv Ou1tsidp Cnntainment or Not Installed EQER Manufacturer l Equipment 02.1 Allis Chalmer Motor 07.2 BIW Tape Splices 11.1 Chromalox Electric Heaters 13.1 Crane Motor 13.2 Crane Motor 15.1 Electroswitch Sw itch 25.2 Barton Pressure Switch 27.5 Kerite Cable 28.1 Labarge Wire 29.3 Limitorgue Actuator 29.4 Limitorgue Actuator 30.1 Magnerrol Level Switch 30.2 Magneol Level Switch 33.4 Namco Limit Switch 33.6 NamcoLimit Switch 33.7 Namco Limit Switch 35.5 Patel Thermoswitch 34.6 Okonite Tape Splices 36.4 Raychem Jacket Repair Tape 37.1 Reliance Motor

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 I Rev. I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 I Page 5 of 42 Table 3.3 Excluded EQER's Radiation Only, Outside Containment or Not Installed EQER i Manufacturer Equipment 37.3 Reliance Motors 38.6 Rockbestos Cable 39.2 Rosemount Pressure Transmitter 39.5 Rosemount Transmitters 43.3 Tarmer Rock Valve 48.1 Westinghouse Motors 50.1 Paul Monroe Actuator 51.1 UEC Temperature Switch 52.1 NVeidmuller Terminal Block 54.2 EGS Grayboot "A" connectors 56.1 Weed RTD 57.1 Phelps Dodge Cable 58.1 Collyer Wire 3.7 The peak pressure postulated for the DBE (45.7 psig, Reference 5.1, Table 7) is essentially the same as than that previously used in the evaluation of equipment located inside containment (45.8 psig, Reference 5.2). The peak test pressure identified in Tab B of EQER was reviewed to verify this value did not exceed the qualification value. These peak pressures are listed in Table 3.2.

3.8 The age-sensitive components for the various pieces of equipment are as identified in their respective EQER's.

3.9 Although normal service temperature will change as a result of the power uprate the analysis of that effect on equipment qualified life has been addressed under separate cover.

3.10 The radiation doses, normal and accident will increase or change slightly as a result of power uprate, however these changes are addressed under separate cover.

4.0 METHODOLOGY AND ACCEPTANCE CRITERIA 4.1 To facilitate the repetitive task of calculating equivalent time, a spreadsheet was prepared using Excel. The formula to used calculate the value teq(hrs) for EQER 5.2 was =SB8*EXP((CS6/0.00008617)*((l/(322.04)-1/((SA8-32)/1.8+273.15)))) where SB8 is the required time at temperature SA8, and CS6 is the activation energy for the appropriate EQER referenced. The value teq(days) was obtained by dividing teq(hrs) by 24. The value teq(acc) was obtained using the formula

=D4*EXP((CS6/0.00008617)*((/11(322.04)- 1/((D6-32)/1.8+273.15))))/24, where D4 is tacc, D6 is Tacc, and CS6 is the activation energy for the appropriate EQER referenced. The output from the spreadsheet is included as Attachment 1.

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 I Page 6 of 42

5.0 REFERENCES

5.1 Containment Integrity and Long Term Cooling Analysisfor 7.4% Pover Uprate, Calculation Number Cl 1546, Revision 0, dated November 6, 2003.

5.2 Kewaunee Nuclear Power PlantEnvironmental QualificationPlan, Rev. 20, dated September 4, 2003.

5.3 Kevaunee Nuclear PowerPlant Steam GeneratorReplacement and Tavg Operating Window Program, dated November, 2000.

5.4 EnironinentalQualifcationof Safety-Related Equipment, USNRC Letter to Eugene Mathews, Wisconsin Public Service Corporation, dated June 18, 1981 (K-8 1-100)

The following is a list of the EQER's reviewed as part of this evaluation. These EQERs qualify important-to-safety equipment located inside containment which are affected as a result of the power uprate.

EOER No. Title 5.1, Rev. 9 ASCO Solenoid Valve Model Types: 206-380, 206-381, 206-832, NP8320 and NP8314 5.2, Rev. 10 ASCO Solenoid Valve Model Types: NP83 16, NP8321, NP8344

7. 1, Rev. 5 Boston Insulated Wire Co. Instrumentation and Control Cable 9.1, Rev. S Brand Rex Fire Retardant Irradiation Cross-Linked Polyethylene XLPE Power and Control Cable 9.2, Rev 4 Brand Rex Fire Retardant and Irradiation Cross-Linked Polyethylene XLPE Instrumentation Cable 10.1, Rev. 1 CKB Industries Thermocouple Connector (LEMO B) 12.1, Rev. 8 Conax Electrical Conductor Seal Assemblies (ECSA) 12.2, Rev. I Conax: RTD Assembly 12.3, Rev. 2 Conax Low Voltage Power and Control Electrical Penetrations 12.4, Rev. 2 Conax Low Voltage Instrumentation Penetrations 12.5, Rev. 2 Conax RTD Models 7P64-10000 and 7P30-10000 14.1, Rev. 2 D. G. O'Brien Medium Voltage Power Electrical Penetrations 14.2, Rev. 3 D. G. O'Brien Low Voltage Power Electrical Penetrations 14.3, Rev. 2 D. G. O'Brien Control Rod Drive Power Electrical Penetrations 14.4, Rev. 2 D. G. O'Brien Nuclear Instrumentation Service Electrical Penetration 14.5, Rev. 2 D. G. O'Brien Radiation Monitoring Electrical Penetration 14.6, Rev. 3 D. G. O'Brien Instrumentation and Control Electrical Penetrations

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 I Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 1 Page 7 of 42 !

20.1, Rev. 13 Foxboro N-ElI and N-E13 Pressure Transmitters 21.1, Rev. 3 Gems Sensors - Liquid Level Transmitters Model No. X.M-54852 and Xi-54853 22.1, Rev. 3 General Atomics RD-23 Radiation Detector

23. 1, Rev. 3 General Electric Terminal Blocks EB-5, EB-25 25.1, Rev. 3 ITT-Barton Differential Pressure Transmitter Model 764 27.1, Rev. 6 Kerite 600 V Low Voltage Power Control Cable, 50 mils FR (HI-70)

Insulation, 65 mils FR (HC-71 1) Jacket 27.2, Rev. 5 Kerite 1 KV Power and 600 V Control Cable HTK (N-98) Insulation/FR (HC-711) Jacket and Taped Splices 27.3, Rev. 4 Kerite 600 V Control Cable, 40 mils FR (HI-70) Insulation, 65 mils FR (HC-71 1) Jacket 27.4, Rev. 2 Kerite Medium Voltage Power Cable 5 KV and 9 KV, 125 mils HTK CN-98 Insulation 80 mils NS (HI-70) Jacket 29.1, Rev. 9 Limitorque Valve Actuators - Inside Containment Type, Limitorque Report No. 600456 29.2, Rev. 7 Limitorque Valve Actuators - Inside Containment Type, Limitorque Report No. 600376A and 600198 plus Addendum I 31.1, Rev. 3 Marathon Terminal Block Assemblies Located Inside and Outside Containment Model 1500 NUC and 142 NUC 31.2, Rev. 1 Marathon Terminal Block Assemblies Located Inside and Outside Containment Series 1500 33.1, Rev. 12 Namco EA1SO Series Limit Switches 33.2, Rev. 9 Namco EA740 Series Limit Switches 33.5, Rev. 4 Requalification of Namco Model EA180 Series Limit Switches 34.1, Rev. 3 Okonite 5KV Power Cable Interlocked Armour, Okoguard Insulated Okolon Jacketed 34.2, Rev. 1 Okonite FMR Insulated Low Voltage Power and Control Cables 34.3, Rev. 4 Okonite Power and Control Cable Okonite Insulated and Okolon Jacketed 34.4, Rev. 3 Okonite Low Voltage Power and Control Cable EPR Insulated, Okoprene or Okoseal Jacketed 34.5, Rev. 4 Okonite Tefzel Insulated Instrumentation & Control Wire 34.8, Rev. 1 Okonite Okozel (Tefzel 280) Low Voltage (600V) Power & Control Cable

- - .- -'. - - . ' . -. - '. . . . .I , - , _:

- -. - - . .. - ..' .. . I.. . t.-Z."W -

.- I ._. 1!

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. I i Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 8 of 42 35.1, Rev. 8 Patel Conduit Seals 35.2, Rev. 1 Patel Thread Sealant Paste 35.3, Rev. 3 Patel Quick Disconnect 1/2 Inch Electrical Connector 35.4, Rev. 0 Patel LOCA Proof Enclosure 36.1, Rev. 4 Raychem Nuclear In-Line Cable Splice Assemblies 36.2, Rev. 6 Raychem Nuclear Plant Kit (NPK) and Nuclear Plant Stub Connection Kit (NPKV) 36.3, Rev. Raychem Breakout Splices 37.2, Rev. 7 Reliance Motors for Containment Fan Coil Units and Containment Dome Fans 3S.1, Rev. 7 Rockbestos Irradiation Cross-Linked Polyethylene Firewall III Cables 38.2, Rev. 6 Rockbestos Chemically Cross-Linked Polyethylene Firewvall III Cables 38.3, Rev. 4 Rockbestos Coaxial Cable 38.4, Rev. 2 Rockbestos RSS-6-161 Coaxial Cable 38.5, Rev. 1 Rockbestos Firewall Silicone Rubber Cable 39.3, Rev. 5 Rosemount, Inc. Pressure Transmitters Model 1153 Series D 39.4, Rev. 1 Rosemount, Inc. Model 1154 Series H Pressure Transmitters 39.6, Rev. 0 Rosemount, Inc. 1154 Series Pressure Transmitters 43.1, Rev. 3 Target Rock Solenoid Operated Globe Valves Model 83QQ-002, Rev. B 43.2, Rev. 4 Target Rock Solenoid Operated Globe Valves Model 80B-001-8 47.1, Rev. 5 Valcor Solenoid Operated Valves 48.2, Rev. 0 Westinghouse Incore Thermocouple Reference Junction Box, Model WX-34 794 48.3, Rev. 0 Westinghouse - Hardline Potting Adaptor/Cable Splice Assemblies 49.1, Rev. 5 3M Co. - 3M Scotch Brand Tapes 52.2, Rev. I Weidmuller Terminal Blocks Located Inside Containment Type SAK4 53.1, Rev. 0 Gamma-Metrics RCS Series Neutron Flux Monitoring System 54.1, Rev. 6 EGS Grayboot Connectors 54.3, Rev. 0 EGS Tape Splices (3M Scotch 130C) 55.1, Rev. 0 RdF Corporation RTD Model 21204

- . .- . __ . - - ..- .. . . , 9..-, . . . -. .- . -..... 7-. _, - -r- - - .. "..  %.i. ., .. " . .... ... .

WISCONSIN PUBLIC SERVICE CORP. Section No. C 1473 IRev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation 11/18/03 l Page 9 of 42 6.0 CALCULATIONSIEVALUATION AND RESULTS Calculations were required when the existing analysis did not cover the post-DBE conditions postulated as a result of the planned power uprate. The EQER's that require evaluation fall into one of three categories.

1. Thermal Degradation - These are cases where the thermal degradation of the equipment resulting from its normal service life and the accident were compared to the thermal properties for its materials of construction.

2. Post-DBE Margin - These are cases where a portion of a thirty day (or longer) accident test was analyzed to address margin

3. Special cases - Instances where the other analyses were performed to either address margin or operability during the post-DBE period.

An example of each type of analysis is provided below followed by a summary of the applicable EQER's, data, and results for that analysis.

6.1 Thermal degradation This analysis was typically performed on equipment that had been qualified in accordance with the DOR Guideline (Enclosure 4 of NRC EB 79-Ol). For this case a combination of test and analysis has been used to demonstrate qualification. In many cases the equipment tested had not been subjected to thermal aging prior to the partial test data used to demonstrate qualification to the accident temperature.

Typically for this equipment, the expected thermal life for the materials of construction was determined using the Arrhenius technique. The thermal equivalent of the accident exposure was determined and compared to the expected life along with the anticipated thermal life of the equipment. This comparison resulted in a life usage factor which, when considered along with the other partial type test data, established the qualification of the equipment.

This is typical of the D.G. O'Brien penetrations. For example, EQER 14.2 documents the exposure of a prototype penetration to two 48-hour steam tests with temperatures from 270'F to 220'F. The exposure was significantly less than the 1-year postulated post-accident duration. Analysis of the thermal equivalence of the steam test exposure did not yield results, which met the 1-year post accident duration. In accordance with the DOR Guidelines, this partial type test data was supplemented with thermal analysis of the materials of construction of the penetrations similar to that shown below.

The Arrhenius equation, in the following form was used:

k = A exp (-(Ea/ KBT)) (1) where, k =

- reaction rate

1% ^

_,-- . --. . Ad_

..........-...ape_.-' . * :A :

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 10 of 42 l A = frequency factor exp = exponent to base e Ea = activation energy KB = Boltzmann's Constant T = Absolute temperature Life is assumed to be inversely proportional to the chemical reaction rate. In terms of life, and converting to Napierian (or natural) base logarithms, Equation (1) becomes:

In (life) = (Ea' KB )(1/T) + Constant (2)

Equation (2) has the algebraic form:

y=mx+b (3)

The constants, m and b, can be estimated by regression analysis of experimental data (Arrhenius plot) for the subject material. For the D.G. O'Brien penetrations, the constants for the most limiting material (Parker S604-70 silicone rubber, Reference WPSC Calculation 14.2.1) are substituted yielding In t= 7343.44 -7.16243 (273.15 +T) where t is the expected material life (hours) at a temperature T( 0 C).

Substituting for the normal service temperature, T = 120'F (48.890 C) yields:

t = 6,202,918 hours0.0106 days <br />0.255 hours <br />0.00152 weeks <br />3.49299e-4 months <br /> Similarly substituting for the plant conditions DBE identified in 3.2.3 yields the following:

Table 6.1 Life Usage Factor for D.G. O'Brien Penetrations Temperature Required Time Qualified Time (hrs) Life Usage Factor (RT) (QT) U = RT/QT Accident Profile 2930F (145.00CC) 100 sec 32,837 0.000085%

268 0F (131.12 0 C) 0.25 hr 55,719 0.000416%

2600 F (126.67 0 C) 0.44 hr 73,467 0.000605%

235 0 F (112.78 0C) 0.39 hr 142,285 0.000273%

219 0 F (103.89 0 C) 1.11 hr 222,834 0.000499%

2130 F (100.56 0C) 3.33 hr 265,112 0.001257%

1980 F (92.23 0C) 22.22 hr 427,921 0.005254%

1. . . ~ ~ I ~~~.

... tl WISCONSIN PUBLIC SERVICE CORP. Section No. Cl 1473 l Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 11 of 42 Table 6.1 Life Usage Factor for D.G. O'Brien Penetrations Temperature Required Time Qualified Time (hrs) Life Usage Factor (RT) (QT) U = RTIQT 165CF (73.890 C) 55.56 hr 1,200,021 0.004630%

153 0F(67.230 C) 111.11 1,881,492 0.006117%

1380 F (58.890 C) 55.56 hr 3,121,345 0.001780%

1350F (57.83CC) 27.78 hr 3,489,771 0.000796%

132CF (5.56 0 C) 138.89 hr 3,906.101 0.0035560%

1280 F (53.340 C) 138.89 hr 4.547,607 0.003054%

124CF (51.12CC) 138.89 hr 5,305.5 15 0.002618%

1220 F (0.00 0 C) 138.89 hr 5,735,154 0.002422%

120 0 F (48.89-C) 7926.67 hr 6,202,918 0.127789%

Normal Operation 120°F 350,640 6,202,918 5.65%

Total 5.81%

The life usage column contains the results of the comparison of the expected life (identified as "Qualified Time) at each temperature and the required plant time at that temperature. These values are summed to yield the portion of the material thermal life used. As can be seen from the above, the life usage is less than 6% of the thermal life, therefore the equipment is qualified to the new postulated accident conditions.

The above analysis applies to the other EQER's for D.G. O'Brien penetrations (EQER's 14.3, 14.4, 14.5, and 14.6).

6.2 Post-DBE margin In most cases, the test profile was analyzed starting after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the temperature had return to 120°F. Since the new profile does not return to 120°F until day 35, the post-DBE margin must be recalculated to show the test levels envelope the plant requirements including margin. After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the postulated plant requirement is shown in Table 6.2. For these EQER's the equivalent time at 120°F for the plant requirement was calculated as well as the equivalent time for the tested profile. These two values were compared to verify the test exposed the equipment to thermal conditions at least as severe as those postulated with margin.

WISCONSIN PUBLIC SERVICE CORP. Section No. C 1473 l Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation 11/18/03 l Page 12 of 42 1 Table 6.2 Post-accident Plant Conditions (after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)

Temperature Duration 0

198 F 3.78 hr 0

165 F 55.56 hr 0

153 F 111.11 138 0F 55.56 hr 1350 F 27.78 hr 0

132 F 138.89 hr 0

128 F 138.89 hr 0

124 F 138.89 hr 122cF 138.89 hr 0

12 F 7926.67 hr Typical of this group was EQER 35.3 where the accident test was 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The equivalent time at 120'F for each plateau of the required profile was calculated using equation 4, and treq = ter and the activation energy for the o-ring material (Ea=1.05) as follows:

I) -I-- I tier = t~ _e KAT- T-) (4)

Where:

tacc = accident (test) time in hours T acc = accident (test) temperature (K)

Tser = service temperature (K) = 322.04K (120'F)

= activation energy K = Boltzmann's Constant = 8.617 x 10 eV/K tser = required time Substituting at (P = 1.05, Tcc = 1980 F (365.38K) and tacr = 3.78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />, yields

( 1.03 y I I trq =3.78 es.6 I7xIO-dA322.04 365.3 8 treq = 336.02 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> / 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day treq = 14.0 days The equivalent time for the remaining portion of the required profile are similarly calculated and shown in Table 6.3. This table also shows the equivalent time calculated for the accident test profile after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> at 250'F).

..I ~ - ~. . - -, - I pI .. . ..

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 l Rev. 1 I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation _Date 11/18/03 l Page 13 of 42 i Table 6.3 Thermal Equivalent of Required ProfilelAccident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 35.3)

Temperature Required Time Equivalent Time (hrs), t l Equivalent Time (days) jE120'F. 0 .S WVJ _ _ _ _ _ _ _ _ _ _ _ _

Accident Profile l l 198CF 3.78 hr 336.02 14.0 1650 F 55.56 hr 848.18 35.3 153cF 111.11 852.72 35.5 138 0F 55.56 hr 173.62 7.2 1350F 27.78 hr 72.14 3.0 1320 F 138.89 hr _ 299.15 12.5 1280 F 138.89 hr 232.44 9.7 124 0 F 138.89 hr 179.98 7.5 1220 F 138.89 hr 158.17 6.6 120 0F 7926.67 hr 7926.67 330.3 Total 461.6 Test Profile Test Time Equivalent Time (hrs), tq

__2rF, CI1O5 eV I 2500 F 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 49,134 2,047 As can be seen above the total thermal exposure experience by the equipment during the accident test (2,047 days) is well in excess of the thermal equivalent of the conditions postulated for post-DBE after power uprate (461.6 days).

Similarly for EQER 5.2, using (D= 0.94 (EQER 5.2, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 624 hours0.00722 days <br />0.173 hours <br />0.00103 weeks <br />2.37432e-4 months <br /> at 2001F, EQER 5.2, Tab B) the results are shown in Table 6.4.

. --:.' "7- -". '- -"' '--i%;

I WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 14 of 42 I Table 6.4 Thermal Equivalent of Required Profile!Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 5.2)

Temperature Required Time l Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198CF 3.78 hr 210.0 8.7 165cF 55.56 hr 637.5 26.6 0

153 F 111.11 688.8 28.7 138 0F 55.56 hr 154.1 6. 4 0

135 F 27.78 hr 65.3 2.7 132 0 F 138.89 hr 276.0 11.5 128 0F 138.89 hr 220.2 9.2 0

124 F 138.89 hr 175.2 7.3 122 0 F 138.89 hr 156.0 6.5 120 0 F 7926.67 hr 7926.67 330.3 Total 437.9 Test Profile Test Time Equivalent Time (hrs) 200°F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 26,272 1,094.7 For EQER 7.1, using tD = 1.23 (EQER 7.1, Tab C), and the last 96.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the accident test profile after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (96.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at 255°F, EQER 7.1 Tab B) the results are shown in Table 6.5.

Table 6.5 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 7.1)

Temperature l Required Time Equivalent Time (hrs) J Equivalent Time (days)

Accident Profile 198°F 3.78 hr 725.2 30.2 165°F 55.56 hr 1353.3 56.4 153°F 111.11 1209.3 50.4 138°F 55.56 hr 211.1 8.8 135-F 27.78 hr 85.0 3.5 132°F7 138.89 hr 341.2 14.2

, ... -- ...i .. 7-7-.-.--- -4.- _w-;- - '-1v: '.'- . . .... - - - --.

WISCONSIN PUBLIC SERVICE CORP. Section No. C 1473 l Rev. I l Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 15 of 42 Table 6.5 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 7.1)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 128 0 F 138.89 hr 253.9 10.6 124CF 138.89 hr 188 7.8 122 0 F 138.89 hr 161.7 6.7 1207F 7926.67 hr 7926.67 330.3 Total 519.0 Test Profile Test Time Equivalent Time (hrs) 2550F 96.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 417,115.5 17,379.8 For EQER 9.2, using <) = 1.07 (EQER 9.2, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 474 hours0.00549 days <br />0.132 hours <br />7.837302e-4 weeks <br />1.80357e-4 months <br /> at 230F, EQER 9.2, Tab B) the results are shown in Table 6.6.

Table 6.6 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 9.2)

Temperature Required Time Equivalent Time (hrs) J Equivalent Time (days)

Accident Profile 1980 F 3.78 hr 366.0 15.3 1650F 55.56 hr 893.4 37.2 1530F 111.11 886.5 36.9 138°F 55.56 hr 177.4 7.4 135 0F 27.78 hr 73.5 3.1 132 0 F 138.89 hr 303.6 12.6 128 0F 138.89 hr 234.7 9.8 1240 F 138.89 hr 180.9 7.5 122 0 F 138.89 hr 158.6 6.6 1200 F 7926.67 hr 7926.67 330.3 Total 466.7 Test Profile Test Time Equivalent Time (hrs) 2300 F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 202,446.4 8,435.3

- . . .. ... .. . I .. - . ...- .. - . "...----- . . - .'. t -, , '.-- -. . f;. '. "- .. . I - .. ...- _- - - ,- . .. -_... ' -

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 16 of 42 i For EQER 10.1, using (D= 0.72 (PS Calculation 10.1.1), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 818.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at 255 0 F, EQER 10.1, Tab B) the results are shown in Table 6.7.

Table 6.7 Thermal Equivalent of Required Profile!Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 10.1)

Temperature l Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile l 1981F 3.78 hr 82.0 3.4 0

165 F 55.56 hr 360.1 15.0 0

153 F 111.11 449.4 1S.7 1387F 55.56 hr 121.4 5.1 0

135 F 27.78 hr 53.4 2.2 132cF 138.89 hr 235.1 9.8 0

128 F 138.89 hr 197.7 8.2 0

124 F 138.89 hr 165.9 6.9 0

122 F 138.89 hr 151.8 6.3 120cF 7926.67 hr 7926.67 330.3 Total 406.0 Test Profile Test Time Equivalent Time (hrs) 255°F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 58,073.6 2,419.7 For EQER 20. 1, using (D= 0.78 (EQER 20. 1, Tab C), and the last 696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> of the accident test profile (696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> at 176.4°F, EQER 20.1, Tab B) the results are shown in Table 6.8. These transmitters have a thirty-day PAOT. As such, only the portion of the accident envelope through 30 days was used in this calculation.

Table 6.8 Thermal Equivalent of Required Profile/Accident Test

__________________ After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 20.1)

Temperature I Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198°F 3.78 hr 106.0 4.4 165°F 55.56 hr 420.8 17.5 0 0492.

153°F _5 21.0 111.11 504.9

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 IRev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 17 of 42 Table 6.8 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 20.1)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 1380 F 55.56 hr 129.5 5.4 1350 F 27.78 hr 56.4 2.4 132 0 F 138.89 hr 245.6 10.2 128 0F 138.89 hr 203.6 8.5 124cF 138.89 hr 168.4 7.0 0

122 F 25.54 hr 28.1 1.2 Total 77.6 Test Profile Test Time Equivalent Time (hrs) 176.4 0F 696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> 8,413.4 350.6 For EQER 21. 1, using (D= 0.78 (EQER 21. 1, Tab C), and the last 696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> of the accident test profile (696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> at 185F, EQER 21. 1, Tab B) the results are shown in Table 6.9.

Table 6.9 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 21.1)

Temperature l Required Time J Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 1980 F 3.78 hr 106.0 4.4 1650 F 55.56 hr 420.8 17.5 1530 F 111.11 504.9 21.0 0

138 F 55.56 hr 129.5 5.4 135F 27.78 hr 56.4 2.4 0

132 F 138.89 hr 245.6 10.2 0

128 F 138.89 hr 203.6 8.5 124 0 F 138.89 hr 168.4 7.0 122 0 F 138.89 hr 153.0 6.4 120 0 F 7926.67 hr 7926.67 330.3 Total 413.1

. -I

- ..I _--

- -. - . ... ..1.. :..

. ---, .. I -__ ' .--.1 . .-. .. -.1r, . - -. . '. . ~~~_._ -- - 1 ~ ~r

- .. 3. 7 . :-

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 IRev. I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation 11/18/03 l Page 18 of 42 Table 6.9 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 21.1)

Temperature l Required Time I Equivalent Time (hrs) l Equivalent Time (days)

Accident Profile I Test Profile Test Time Equivalent Time (hrs) 185 0 F 696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> 11,840.6 493.4

_1- -I For EQER 23.1, using = 0.96 (EQER 23. 1, Tab C), and the last 101 hours0.00117 days <br />0.0281 hours <br />1.669974e-4 weeks <br />3.84305e-5 months <br /> of the accident test profile (101 hours0.00117 days <br />0.0281 hours <br />1.669974e-4 weeks <br />3.84305e-5 months <br /> at 232 0 F, EQER 23.1, Tab B) the results are shown in Table 6.10.

Table 6.10 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 23.1)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile l 198cF 3.78 hr f 228.7 9.5 165 0F 55.56 hr 671.5 28.0 153cF 111.11 716.1 29.8 138 0F 55.56 hr 157.5 6.6 135 0F 27.78 hr 66.5 2.8 132 0F 138.89 hr 280.1 11.7 128 0F 138.89 hr 222.4 9.3 124 0F 138.89 hr 176.0 7.3 122 0F 138.89 hr 156.4 6.5 120 0F 7926.67 hr 7926.67 330.3 Total 441.7 I Test Profile Test Time Equivalent Time (hrs) 232 0F 101 hours0.00117 days <br />0.0281 hours <br />1.669974e-4 weeks <br />3.84305e-5 months <br /> 27,358.4 1,139.9 For EQER 27.1, using $) = 0.924 (EQER 27.1, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> at 212'F, EQER 27.1, Tab B) the results are shown in Table 6.1 1. The equivalent times were evaluated at 600 C (Tscr = 600 C , rather than Tser

= 1201F).

.. - - - - - -- - - .- - .- . . .. .... -. . . -..- ..... . .. I. . .I .

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 I Rev. I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 19 of 42 l Table 6.11 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60'C (EQER 27.1)

Temperature 1 Required Time Equivalent Time (hrs) J Equivalent Time (days)

Accident Profile I .

1987F 3.78 hr 64.6 2.7 1650 F 55.56 hr 201.5 8.4 0

153 F 111.11 220.0 9.2 0

138 F 55.56 hr 49.9 2.1 1350 F 27.78 hr 21.2 0.9 132°F 138.89 hr 89.9 3.7 128 0F 138.89 hr 72.0 3.0 0

124 F 138.89 hr 57.5 2.4 122 0F 138.89 hr 51.3 2.1 120WF 7926.67 hr 2611.0 108.8 Total 143.3 Test Profile Test Time Equivalent Time (hrs) 212°F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 13,611.8 567.2 For EQER 27.3, using O = 0.929 (EQER 27.3, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 624 hours0.00722 days <br />0.173 hours <br />0.00103 weeks <br />2.37432e-4 months <br /> at 210°F, EQER 27.3, Tab B) the results are shown in Table 6.12. The equivalent times were evaluated at 60°C (Ts, =

60°C , rather than Tser = 120°F).

Table 6.12 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60°C (EQER 27.3)

Temperature l Required Time l Equivalent Time (hrs) l Equivalent Time (days)

Accident Profile 198°F 3.78 hr 65.6 2.7 165°F 55.56 hr 202.9 8.5 153°F 111.11 220.8 9.2 138°F 55.56 hr 49.9 2.1 135°F 27.78 hr 21.2 0.9 132'F 138.89 hr 89.7 3.7 12S°F 138.89 hr il.7 3.0

.. .. .. -, - . ,.- - -::X WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation e 11/18/03 l Page 20 of 42 Table 6.12 Thermal Equivalent of Required Profile/Accident Test Temperature Required Time  !

After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60'C (EQER 27.3)

Equivalent Time (hrs) j Equivalent Time (days)

Accident Profile l l 124 0F 138.89 hr 57.2 2.4 122 0F 138.89 hr 51.0 2.1 120 0F 7926.67 hr 2595.3 108.1 Total 142.7 Test Profile Test Time Equivalent Time (hrs) 210 0F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 12,721.8 530.1 For EQER 29.1, using (D= 1.016 (EQER 29.1, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> at 192 0F, EQER 29. 1, Tab B) the results are shown in Table 6.13.

Table 6.13 Thermal Equivalent of Required Profile!Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 29.1)

Temperature l Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198 0F 3.78 hr 290.6 12.1 165 0F 55.56 hr 776.5 32.4 153'F 111.11 798.3 33.3 138 0F 55.56 hr 167.3 7.0 135 0F 27.78 hr 69.9 2.9 132 0F 138.89 hr 291.8 12.2 128 0F 138.89 hr 228.6 9.5 124 0F 138.89 hr 178.5 7.4 122 0F 138.89 hr 157.5 6.6 120 0F 7926.67 hr 7926.67 330.3 Total 453.6 Test Profile Test Time Equivalent Time (hrs) 1927F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 24,672.4 . 1,028.0

. - ..- . . -_  :. . . --.-- . - - '. 1. I ,, 1: 11 I :1 - "' -, . '- !.,- -'. ,  !: - -.' I -': ,, .-' " - . .. I ..- .

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 Page 21 of 42 For EQER 29.2, using c1= 1.016 (EQER 29.2, Tab C), and the last 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> of the accident test profile (144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> at 250'F, EQER 29.2, Tab B) the results are shown in Table 6.14.

Table 6.14 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 29.2)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198 0F 3.78 hr 290.6 12.1 165 0F 55.56 hr 776.5 32.4 153 0F 111.11 798.3 33.3 138 0F 55.56 hr 167.3 7.0 135 0F 27.78 hr 69.9 2.9 132 0F 138.89 hr 291.8 12.2 128 0F 138.89 hr 228.6 9.5 124 0F 138.89 hr 178.5 7.4 122 0F 138.89 hr 157.5 6.6 120°F 7926.67 hr 7926.67 330.3 Total 453.6 Test Profile Test Time Equivalent Time (hrs) 250 0F 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> 117,771.3 4,907.1 For EQER 31. 1, using (1 = 1.59 (EQER 31.1, Tab C), and the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accident test profile (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the 672 hours0.00778 days <br />0.187 hours <br />0.00111 weeks <br />2.55696e-4 months <br /> at 295°F, EQER 31.1, Tab B) the results are shown in Table 6.15.

Table 6.15 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 31.1)

Temperature l Required Time I Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198°F 3.78 hr 3377.8 140.7 165 0F 55.56 hr 3445.5 143.6 153 0F 111.11 2432.1 101.3 138 0F 55.56 hr 312.0 13.0 135eF 27.78 hr 117.8 4.9

.. , .. - . ..- .. .1 -. .. . - '. - -. - - . ., - 't.,  : , ,, ,. : -- - -.. . - --- f .

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 22 of 42 Table 6.15 Thermal Equivalent of Required Profile/Accident Test Temperature l Required Time J After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 31.1)

Equivalent Time (hrs) { Equivalent Time (days)

Accident Profile 132 0F 138.89 hr 443.9 18.5 128 0 F 138.89 hr 302.9 12.6 124 0 F 138.89 hr 205.6 8.6 122 0 F T 138.89 hr 169.1 7.0 120 0F f 7926.67 hr 7926.67 330.3 Total 780.6 Test Profile Test Time Equivalent Time (hrs) 2950 F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 14,139,041.8 589,126.7 For EQER 31.2, using c1 = 0.91 (EQER 31.2, Tab C), and the last 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> of the accident test profile (28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> at 250 0 F, EQ Reference 426, pages 163-164) the results are shown in Table 6.16.

Table 6.16 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 31.2)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198 0 F 3.78 hr 184.7 7.7 165 0F 55.56 hr 589.7 24.6 1530 F 111.11 649.8 27.1 138OF 55.56 hr 149.2 6.2 135 0F 27.78 hr 63.5 2.6 0

132 F 138.89 hr 270.1 11.3 0

128 F 138.89 hr 217.0 9.0 0

124 F 138.89 hr 173.9 7.2 0

122 F 138.89 hr 155.4 6.5 0

120 F 7926.67 hr 7926.67 330.3 Total 432.5 l etTm qiaetTm hs Test Pr_

Tes Profile Test Time Equivalent Time (hrs)

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 23 of 42 Table 6.16 Thermal Equivalent of Required Profile,'Accident Test Temperature  ! After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 31.2)

Required Time I Equivalent Time (hrs) I Equivalent Time (days)

Accident Profile l l l 2500 F 28 hours3.240741e-4 days <br />0.00778 hours <br />4.62963e-5 weeks <br />1.0654e-5 months <br /> 11,375.0 474.0 For EQER 33.1, using = 1.133 (EQER 33.1, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> at 200 0F, EQER 33.1, Tab B) the results are shown in Table 6.17.

Table 6.17 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 33.1)

Temperature l Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198CF 3.78 hr 479.1 20.0 165 0F 55.56 hr 1052.1 43.8 153 0F 111.11 1001.8 41.7 138 0F 55.56 hr 190.0 7.9 135 0F 27.78 hr 77.8 3.2 132 0F 138.89 hr 317.9 13.2 128 0F 138.89 hr 242.1 10.1 124 0F 138.89 hr 183.7 7.7 122 0F 138.89 hr 159.8 6.7 120 0F 7926.67 hr 7926.67 330.3 Total 484.6 Test Profile Test Time. Equivalent Time (hrs) 200 0F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 61,064.9 2,544.4 For EQER 33.2, using (D= 0.999 (EQER 33.2, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 672 hours0.00778 days <br />0.187 hours <br />0.00111 weeks <br />2.55696e-4 months <br /> at 200'F, EQER 33.2, Tab B) the results are shown in Table 6.18.

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calcuilation Date 11/18/03 l Page 24 of 42 I Table 6.18 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 33.2)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile l 0

198 F 3.78 hr 270.2 11.3 0

165 F 55.56 hr 743.0 31.0 0

153 F 111.11 772.4 32.2 0

138 F 55.56 hr 164.3 6.8 135 0F 27.78 hr 68.9 2.9 0

132 F 138.89 hr 288.2 12.0 0

128 F 138.89 hr 226.7 9.4 124cF 138.89 hr 177.7 7.4 0

122 F 138.89 hr 157.2 6.5 0

120 F 7926.67 hr 7926.67 330.3 Total 449.8 Test Profile Test Time Equivalent Time (hrs) 200 0F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 33,999.6 1,416.7 For EQER 33.5, using $d= 1.133 (EQER 33.5, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 624 hours0.00722 days <br />0.173 hours <br />0.00103 weeks <br />2.37432e-4 months <br /> at 207'F, EQER 33.5, Tab B) the results are shown in Table 6.19.

Table 6.19 Thermal Equivalent of Required Profile!Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 33.5)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

.- Accident Profile 198°F 3.78 hr 479.1 20.0 165°F 55.56 hr 1052.1 43.8 153F 111.11 1001.8 41.7 138°F 55.56 hr 190.0 7.9 135°F 27.78 hr 77.8 3.2 132°F 138.89 hr 317.9 13.2 128°F 138.89 hr 242.1 10.1 124°F 138.89 hr 183 .7 7.7

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 25 of 42 l Table 6.19 Thermal Equivalent of Required Profile.'Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 33.5)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 122 0 F 138.89 hr 159.8 6.7 0

120 F 7926.67 hr 7926.67 330.3 Total 484.6 Test Profile Test Time Equivalent Time (hrs) 2070 F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 89,000.9 3,70S.4 For EQER 34.3, using ( = 1.07 (EQER 34.3, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 of the 3,024 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 2127F, EQER 34.3, Tab B) the results are shown in Table 6.20. The equivalent times were evaluated at 60'C (Tser = 60'C, rather than Tser = 120'F).

Table 6.20 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 0C (EQER 34.3)

Temperature Required Time Equivalent Time (hrs) l Equivalent Time (days)

Accident Profile 198 0F 3.78 hr 101.2 4.2 1650 F 55.56 hr 246.9 10.3 1530F 111.11 245.0 10.2 138 0F 55.56 hr 49.0 2.0 1350 F 27.78 hr 20.3 0.8 132 0 F 138.89 hr 83.9 3.5 128 0F 138.89 hr 64.9 2.7 I

124 0F 138.89 hr 50.0 2.1 1220 F 138.89 hr 43.8 1.8 1200 F 7,926.67 hr 2190.8 91.3 Total 129.0 Test Profile Test Time Equivalent Time (hrs) 212 0 F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 23,479.0 978.3

. . .. - -..- ... . ... . ... . . -- - .. - . .1 - .. - -

WISCONSIN PUBLIC SERVICE CORP. Section No. C 1473 l Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation e 11/18/03 i Page 26 of 42 For EQER 34.4, using 1 = 0.883 (Calculation C10974), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 2400 hours0.0278 days <br />0.667 hours <br />0.00397 weeks <br />9.132e-4 months <br /> at 212F, EQER 34.4, Tab B) the results are shown in Table 6.21. The equivalent times were evaluated at 66.6'C (Tser =

66.6 0 C , rather than T, = 120'F).

Table 6.21 Thermal Equivalent of Required Profile/Accident Test Afier 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 66.6 0C (EQER 34.4)

Temperature J Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile l 0

198 F 3.78 hr 31.3 1.3 1650 F 55.56 hr 104.7 4.4 153'F 111.11 117.4 4.9 138 0F 55.56 hr 27.6 1.1 135 0F 27.78 hr 11.8 0.5 0

132 F 138.89 hr 50.4 2.1 128 0F 138.89 hr 40.8 1.7 0

124 F 138.89 hr 32.9 1.4 0

122 F 138.89 hr 29.5 1.2 120F 7,926.67 hr 1509.0 62.9 Total 81.5 Test Profile Test Time Equivalent Time (his) 212F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 6,425.8 267.7 For EQER 35.1, using (t = 2.28 (EQER 35.1, Tab C), and the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accident test profile (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the 216 hours0.0025 days <br />0.06 hours <br />3.571429e-4 weeks <br />8.2188e-5 months <br /> at 250'F, EQER 35.1, Tab B) the results are shown in Table 6.22.

Table 6.22 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 35.1)

Temperature Required Time Equivalent Time (hrs) l Equivalent Time (days)

Accident Profile 1980 F 3.78 hr 64465.4 2686.1 1650 F 55.56 hr 20660.1 860.8 153F 111.11 9280.7 386.7 I3 0 _ 55 hr 65_ 27 138°F 55.56 hr 659.6 27.5

- -- - - - - - . .. --- .. - - . -- . 1. . - - -

WISCONSIN PUBUC SERVICE CORP. Section No. Cl1473 Rev. I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 27 of 42 l Table 6.22 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 35.1)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 135?F 27.78 hr 220.6 9.2 132 0F 138.89 hr 734.9 30.6 1280 F 138.89 hr 424.9 17.7 124 0F 138.89 hr 243.8 10.2 122 0F 138.89 hr 184.2 7.7 120 0F 7926.67 hr 7926.67 330.3 Total 4366.1 Test Profile Test Time Equivalent Time (hrs) 250 0F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 82,515,157.9 3,438,131.6 For EQER 35.3, using 1P = 1.05 (EQER 35.3, Tab C), and the last 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the accident test profile (48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> at 250 0 F, EQER 35.3, Tab B) the results are shown in Table 6.23.

Table 6.23 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 35.3)

Temperature Required Time Equivalent Time (his) Equivalent Time (days)

Accident Profile 1980 F 3.78 hr 336.0 14.0 165F 55.56 hr 848.2 35.3) 153°F 111.11 852.7 35.5 138°F 55.56 hr 173.6 7.2 135°F 27.78 hr 72.1 3.0 132°F 138.89 hr 299.2 12.5 128°F 138.89 hr 232.4 9.7 124°F 138.89 hr 180.0 7.5 122°F 138.89 hr 158.2 6.6 120°F 7926.67 hr 7926.67 330.3

.~~~~~~~~oa 46..6 Total 461.6

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 I Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 28 of 42 Table 6.23 Thermal Equivalent of Required Profile/Accident Test Temperature J After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 35.3)

Required Time l Equivalent Time (hrs) I Equivalent Time (days)

Accident Profile Test Profile Test Time Equivalent Time (hrs) 250F 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 49,134.8 2,047.3 For EQER 38.3, using t = 1.3251 (Calculation C 1206), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 2,496 hours0.00574 days <br />0.138 hours <br />8.201058e-4 weeks <br />1.88728e-4 months <br /> at 227 0 F, EQER 38.3, Tab B) the results are shown in Table 6.24.

Table 6.24 Thermal Equivalent of Required Profile,'Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 38.3)

Temperature Required Time l Equivalent Time (hrs) Equivalent Time (days)

Accident Profile l 198 0F 3.78 hr 1088.8 45.4 165 0F 55.56 hr 1732.3 72.2 0

153 F 111.11 1454.4 60.6 0

138 F 55.56 hr 234.0 9.8 0

135 F 27.78 hr 92.6 3.9 0

132 F 138.89 hr 365.8 15.2 128WF 138.89 hr 266.0 11.1 0

124 F 138.89 hr 192.6 8.0 122 0F 138.89 hr 163.6 6.8 120OF 7926.67 hr 7926.67 330.3 Total 563.2 Test Profile Test Time Equivalent Time (hrs) 227 0F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 736,059.5 30,669.1 For EQER 38.4, using ( = 1.3251 (EQER 38.4, Tab C), and the last 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of the accident test profile (96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> at 212 0F, EQER 38.4, Tab B) the results are shown in Table 6.25.

.. I . . .. ,. .. " . ** ... "- - ,:,~..: ~.. .I - ., , - . . .

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 IRev. 1 I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation e 11/18/03 l Page 29 of 42 Table 6.25 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 38.4)

Temperature i Required Time Equivalent Time (hrs) l Equivalent Time (days)

Accident Profile 198 0F 3.78 hr 1088.8 45.4 165 0 F 55.56 hr 1732.3 72.2 1537F 111.11 1454.4 60.6 138 0 F 55.56 hr 234.0 9.8 135 3F 27.78 hr 92.6 3.9 132 0 F 138.89 hr 365.8 15.2 128 0 F 138.89 hr 266.0 11.1 124 0F 138.89 hr 192.6 8.0 122 0 F 138.89 hr 163.6 6.8 120 0 F 7926.67 hr 7926.67 330.3 Total 563.2 Test Profile Test Time Equivalent Time (hrs) 212 0 F 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 66,486.9 2,770.3 For EQER 38.5, using c = 1.6542 (EQER 38.5, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> at 2340 F, EQER 38.5, Tab B) the results are shown in Table 6.26.

Table 6.26 Thermal Equivalent of Required Profile/Accident Test Temperature After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 38.5)

Required Time Equivalent Time (hrs) T Equivalent Time (days)

Accident Profile 198 0F 3.78 hr 4444.3 185.2 1650 F 55.56 hr 4070.4 169.6 1537F 111.11 2754.8 114.8 1380 F 55.56 hr 334.5 13.9 135 0F 27.78 hr 124.9 5.2 1320 F 138.89 hr 465.2 19.4 128 0 F 138.89 hr 312.6 13.0 124°F 138.89 hr 208.9 8.7

. .. - . .... . - '%z:- -' . -7;. _:.. " ..- ' .:. - :.:, "" -.' : .- .' -_ '.- . , " . ..-. .. A. .- - . .. . .- - . - - .. - ... - . .

WISCONSIN PUBLIC SERVICE CORP. Section No. Cl 1473 I Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 30 of 42 l Table 6.26 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 38.5)

Temperature l Required Time j Equivalent Time (hrs) J Equivalent Time (days)

Accident Profile 1220F 138.89 hr 170.4 7.1 120cF 7926.67 hr 7926.67 330.3 Total 867.2 Test Profile Test Time Equivalent Time (hrs) 234 0 F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 7,762,567.4 323,440.3 For EQER 39.3, using (D= 0.78 (EQER 39.3, Tab C), and the last 696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> of the accident test profile (696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> at 176.40 F, EQER 39.3, Tab B) the results are shown in Table 6.27. These transmitters have a thirty-day PAOT. As such, only the portion of the accident envelope through 30 days was used in this calculation.

Table 6.27 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 393)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 1980 F 3.78 hr 106.0 4.4 1650F 55.56 hr 420.8 17.5 1530F 111.11 504.9 21.0 1387F 55.56 hr 129.5 5.4 1350 F 27.78 hr 56.4 2.4 1320 F 138.89 hr 245.6 10.2 1280 F 138.89 hr 203.6 8.5 124 0F 138.89 hr 168.4 7.0 122°F 25.54 hr 28.1 1.2 Total 77.6 Test Profile Test Time Equivalent Time (hrs) 176.4 0F 696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> 8,413.4 350.6

.I . . . - .. - . . ... . I . .- .... ... -I.... .. . - , '.. - -'. "  : "-

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation 11/18/03 l Page 31 of 42 1 For EQER 39.4, using ($ = 0.78 (EQER 39.4, Tab C), and the last 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of the accident test profile (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 2650 F, EQER 39.4, Tab B) the results are shown in Table 6.28.

Table 6.28 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 39.4)

Temperature Required Time Equivalent Time (hrs) [ Equivalent Time (days)

Accident Profile 198F 3.78 hr 106.0 4.4 165zF 55.56 hr 420.8 17.5 153F 111.11 504.9 21.0 138 0F 55.56 hr 129.5 5.4 135 0F 27.78 hr 56.4 2.4 132 0F 138.89 hr 245.6 10.2 128T 138.89 hr 203.6 8.5 124pF 138.89 hr 168.4 7.0 122 0F 138.89 hr 153.0 6.4 120 0F 7926.67 hr 7926.67 330.3 Total 413.1 Test Profile Test Time Equivalent Time (hrs) 2650F 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> I1,0S0.5 461.7 For EQER 39.6, using (Q= 0.78 (EQER 39.6, Tab C), and the last 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> of the accident test profile (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at 2650 F, EQER 39.6, Tab B) the results are shown in Table 6.29.

Table 6.29 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 39.6)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 1980 F 3.78 hr 106.0 4.4 1650 F 55.56 hr 420.8 17.5 153 0F 111.11 504.9 21.0 138 0 F 55.56 hr 129.5 5.4 13 0 . ,hr 227.78 6. 2.4 135017 56.4 2.4

. .. . . ...- .'. - _; t- .  :. . .M7:'_

- - .. .. . .... -.:%: .':., -- . " _1.:.:_1'!1L:_Z_;t_  %'-'

WISCONSIN PUBLIC SERVICE CORP. Section No. Cl 1473 IRev. I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 32 of 42 l Table 6.29 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 39.6)

Temperature Required Time l Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 132 0F 138.89 hr 245.6 10.2 1280 F 138.89 hr 203.6 8.5 1240 F 138.89 hr 168.4 7.0 122 0 F 138.89 hr 153.0 6.4 1200 F 7926.67 hr 7926.67 330.3 Total 413.1 Test Profile Test Time Equivalent Time (hrs) 2650 F 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> 11,080.5 461.7 For EQER 43.2, using ( = 0.95 (EQER 43.2, Tab C), and the last 294 hours0.0034 days <br />0.0817 hours <br />4.861111e-4 weeks <br />1.11867e-4 months <br /> of the accident test profile (294 hours0.0034 days <br />0.0817 hours <br />4.861111e-4 weeks <br />1.11867e-4 months <br /> at 215'F, EQER 43.2, Tab B) the results are shown in Table 6.30. These valves have a thirty-day PAOT. As such, only the portion of the accident envelope through 30 days was used in this calculation.

Table 6.30 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 43.2)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198 0F 3.78 hr 219.2 9.1 1650 F 55.56 hr 654.3 27.3 153°F 111.11 702.3 29.3 1380 F 55.56 hr 155.8 6.5 I

135°F 27.78 hr 65.9 2.7 132°F 138.89 hr 278.1 11.6 128°F 138.89 hr 221.3 9.2 124°F 138.89 hr 175.6 7.3 122°F 25.54 hr 28.7 1.2 Total 104.2

-Iest roftile I Test Time Equivalent Time (hrs)

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 I Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 33 of 42 Table 6.30 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 43.2)

Temperature Required Time Equivalent Time (hrs)J Equivalent Time (days)

Accident Profile I 0

215 F 294 hours0.0034 days <br />0.0817 hours <br />4.861111e-4 weeks <br />1.11867e-4 months <br /> 36,459.3 1,519.1 1 For EQER 48.2, using (D= 1.0 (WPS Calculation 48.2.1), and the last 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> of the accident test profile (240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> of the 280 hours0.00324 days <br />0.0778 hours <br />4.62963e-4 weeks <br />1.0654e-4 months <br /> at 2300 F, EQER 48.2, Tab B) the results are shown in Table 6.31.

Table 6.31 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 48.2)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198 0F 3.78 hr 271.4 11.3 165'F 55.56 hr 744.9 31.0 153 0F 111.11 773.9 32.2 138 0 F 55.56 hr 164.5 6.9 135 0F 27.78 hr 68.9 2.9 132 0F 138.89 hr 288.4 12.0 1287F 138.89 hr 226.8 9.5 1240 F 138.89 hr 177.8 7.4 122cF 138.89 hr 157.2 6.5 120 0F 7926.67 hr 7926.67 330.3 Total 450.0 Test Profile Test Time Equivalent Time (hrs)

I 2300 F 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> 75,216.0 3,134.0 For EQER 48.3, using (D= 1.0 (EQER 48.3, Tab C), and the last 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the accident test profile (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the 145 hours0.00168 days <br />0.0403 hours <br />2.397487e-4 weeks <br />5.51725e-5 months <br /> at 255 0F, EQER 48.3, Tab B) the results are shown in Table 6.32.

~~~~~~~~~~~~

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WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 IRev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 34 of 42 Table 6.32 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 48.3)

Temperature l Required Time Equivalent Time (hrs) JEquivalent Time (days)

Accident Profile 198 0F 3.78 hr 271.4 11.3 1650 F 55.56 hr 744.9 31.0 153cF 111.11 773.9 32.2 138 0F 55.56 hr 164.5 6.9 135 0 F 27.78 hr 68.9 2.9 132 0 F 138.89 hr 288.4 12.0 128 0F 138.89 hr 226.8 9.5 124 0F 138.89 hr 177.8 7.4 1220 F 138.89 hr 157.2 6.5 120 0F 7926.67 hr 7926.67 330.3 Total 450.0 Test Profile Test Time Equivalent Time (hrs)

I 2550F 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 65,099.2 2,712.5 For EQER 49. 1, using (D= 1.15 (EQER 49. 1, Tab C), and 67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> of the accident test profile (67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> of the 81 hours9.375e-4 days <br />0.0225 hours <br />1.339286e-4 weeks <br />3.08205e-5 months <br /> at 2650 F, EQER 49.1, Tab B) the results are shown in Table 6.33.

Table 6.33 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 49.1)

Temperature Required Time Equivalent Time (rs) Equivalent Time (days)

Accident Profile 198CF 3.78 hr 515.2 21.5 I 165 0F 55.56 hr 1099.6 45.8 153 0F 111.11 1035.4 43.1 138 0F 55.56 hr 193.5 8.1 135 0 F 27.78 hr 79.0 3.3 132 0 F 138.89 hr 321.8 13.4 128F 138.89 hr 244.1 10.2 0

1124°F 13.8 hr 184.A 7.7 138.89 hr 184.5 7.7

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WISCONSIN PUBLIC SERVICE CORP. Section No. Cl 1473 l Rev. I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 _ Page 35 of 42 Table 6.33 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 49.1)

Temperature Required Time l Equivalent Time (hrs) I Equivalent Time (days)

Accident Profile 122 0F 138.89 hr 160.1 6.7 1200 F 7926.67 hr 7926.67 330.3 Total 490.0 Test Profile Test Time Equivalent Time (hrs) 255 0 F 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 267,422.5 11,142.6 For EQER 52.2, using (D= 1.33 (EQER 52.2, Tab C), and the last 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the accident test profile (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at 320'F, EQER 52.2, Tab B) the results are shown in Table 6.34. Since this was only a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test, the time after the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> could not be used for the evaluation. As such the last four hours of the test were used and the last four hours of the accident profile (from time t = 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> to t = 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, which is at 1970 F) were added to the required time (i.e., the Required Time at 197 0 F was increased from 3.78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> to 7.78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br />).

Table 6.34 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 52.2)

Temperature l Required Time l Equivalent Time (hrs) [ Equivalent Time (days)

Accident Profile .

198 0 F 7.78 hr 2288.5 95.4 165 0 F 55.56 hr 1754.5 73.1 1530F 111.11 1468.3 61.2 138 0F 55.56 hr 235.3 9.8 135 0F 27.78 hr 93.0 3.9 I

132°F 138.89 hr 367.1 15.3 128°F 138.89 hr 266.7 11.1 124°F 138.89 hr 192.9 . 8.0 122°F 138.89 hr 163.7 6.8 120°F 7926.67 hr 7926.67 330.3 Total 614.9 Test Profile Test Time Equivalent Time (hrs)

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WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 l Rev. I Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation e11/18/03 i Page 36 of 42 For EQER 53.1, using (P = 1.57 (WPS Calculation WPS.006.0200.53.1.1), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 480 hours0.00556 days <br />0.133 hours <br />7.936508e-4 weeks <br />1.8264e-4 months <br /> at 220'F, EQER

53. 1, Tab B) the results are shown in Table 6.35.

Table 6.35 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 53.1)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198 0F 3.78 hr 3101.1 129.2 165 0F 55.56 hr 3271.2 136.3 153cF 111.11 2339.5 97.5 0

138 F 55.56 hr 305.3 12.7 135 0F 27.78 hr 115.7 4.8 0

132 F 138.89 hr 437.4 18.2 1280 F . 138.89 hr 300.0 12.5 0

124 F 138.89 hr 204.6 8.5 0

122 F 138.89 hr 168.7 7.0 0

120 F 7926.67 hr 7926.67 330.3 Total 757.1 Test Profile Test Time Equivalent Time (hrs) 2200 F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 1,780,362 74,181.8 I For EQER 54. 1, using (D= 0.92 (EQER 54. 1, Tab C), and the last 696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> of the accident test profile (696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> at 180'F, EQER 54.1, Tab B) the results are shown in Table 6.36.

WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 37 of 42 Table 636 Thermal Equivalent of Required Profile'Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (EQER 54.1)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 198 0 F 3.78 hr 192.8 8.0 165 0F 55.56 hr 605.2 25.2 1530 F 111.11 662.6 27.6 1380 F 55.56 hr 150.8 6.3 1350 F 27.78 hr 64.1 2.7 132CF 138.89 hr 272.0 11.3 1280 F 138.89 hr 218.1 9.1 1247F 138.89 hr 174.3 7.3 1220 F 138.89 hr 155.6 6.5 1200 F 7926.67 hr 7926.67 330.3 Total 434.3 Test Profile Test Time Equivalent Time (hrs) 180 0 F 696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> 15,598.6 649.9 For EQER 54.3, using ($ = 1.14 (EQER 54.3, Tab C), and the last 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the accident test profile (432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> of the 696 hours0.00806 days <br />0.193 hours <br />0.00115 weeks <br />2.64828e-4 months <br /> at 227°F, EQER 54.3, Tab B) the results are shown in Table 6.37. . The equivalent times were evaluated at 55.2 0C (Ts..

= 55.2°C , rather than Tser = 120°F).

Table 6.37 Thermal Equivalent of Required Profile/Accident Test Temperature Required Time [

After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 55.2°C (EQER 54.3)

Equivalent Time (hrs) Equivalent Time (days)

Accident Profile I 198°F 3.78 hr 224.2 9.3 165°F 55.56 hr 486.5 20.3 153°F 111.11 461.1 19.2 138°F 55.56 hr 86.9 3.6 135°F 27.78 hr 35.6 1.5 132OF 138.89 hr 145.1 6.0 3.9h 1. .

128eF

_2C 138.89 hr 110.3 4.6

.7.. W.,_%.r, r.. ,, .- -9 .1 WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 Page 38 of 42 Table 6.37 Thermal Equivalent of Required Profile/Accident Test After 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 55.2 0 C (EQER 54.3)

Temperature Required Time Equivalent Time (hrs) Equivalent Time (days)

Accident Profile 124 0 F 138.89 hr 83.6 3.5 122 0F 138.89 hr 72.6 3.0 0

120 F 7926.67 hr 3598.9 150.0 Total 221.0 Test Profile Test Time Equivalent Time (hrs) 180°F 432 hours0.005 days <br />0.12 hours <br />7.142857e-4 weeks <br />1.64376e-4 months <br /> 118,212.5 4,925.5 6.3 Special Cases Solenoid Valves 33809, 33810, 33811, and 33812 The majority of the solenoid valves covered by EQER 5.2, are normally de-energized but cycled periodically post-accident and are covered by the above analysis. Solenoid valve 33809, 33810, 33811 and 33812 are required to be energized for thirty days following the accident. The existing analysis of these valves in the EQER compared the total thermal effect resulting from thirty days energized post accident to the thermal degradation experience by the test valve in the test program. To accomplish this the equivalent time at 1201F was calculated assuming the valve are continuously energized for thirty days. The appropriate heat rise was added to Teq (105 0C heat rise for coils and 41.0 0C for elastomers, See Calculation C10546, Rev. 2, Table 3.3.3) and the equivalent time calculated as follows:

teqv = tacc.e (t (4)

Where:

tacc = accident (test) time in hours Tacc = accident (test) temperature (K) including heat rise Teqv = service temperature (K) = 322.04K (120'F)

= activation energy K = Boltzmann's Constant = 8.617 x 10 5 eV/K teqY = equivalent time For the coil of ASCo solenoid valve 33809 (et.al., EQER 5.2) and the first portion of the accident curve, tacc 100 sec Tacc = 523.15K (2930F+105'C heat rise)

X

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s WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 I Rev. 1 Title Post Accident Operating Time

. Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 39 of 42

• = activation energy (1.0 eV)

Therefore,

( 1.0 y I "

t,, = 1 00 - e~T8617x10'A322.04 23.15) tcqv = 1.04E8 secs / 3600 secs per hour teq, = 28,837 hours0.00969 days <br />0.233 hours <br />0.00138 weeks <br />3.184785e-4 months <br /> /8760 hours per year tcqv = 3.292 years Other results for the coil and for the elastomers (Ea = 0.94) are shown in Table 6.38.

Table 6.38 Equivalent Life Usage Factor for ASCo Solenoid Valves Temperature Required Time Equivalent Time (years) Equivalent Time (years)

(tacc) (coil) (elastomers)

Accident Profile 293F 100 sec 3.29 0.078 2687 0.25 hr 16.18 0.336 2600 F 0.44 hr 23.52 0.462 235F 0.28 hr 7.68 0.131 219 0 F 1.22 hr 21.84 0.333 213CF 3.33 hr 50.37 0.737 1980 F 22.22 hr 218.56 2.866 1650 F 55.56 hr 200.88 2.015 153F 111.11 273.58 2.466 1387F 55.56 hr 83.23 0.651 135CF 27.78 hr 37.59 0.285 132F 138.89 hr 169.66 1.250 128F 138.89 hr 147.82 1.046 124 0 F 138.89 hr 128.61 0.873 122 0 F 2.56 hr 22.06 0.147 Normal Operation (No heat rise) 120 0 F 350,640 40 40 Total 1,444.89 53.707

. . . __.- . . _. ... '. - I 'Z. - '.1 ..-%.

-, , .. ..1. - , . .". - '_ ."-. ".- ..' t-WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation e 11/18/03 l Page 40 of 42 Analysis contained in the calculation associated with EQER (C 10546) evaluated the equivalent time at temperature by comparing similar results to the equivalent thermal degradation seen by the solenoid as a result of thermal aging. In this case, the qualified lives are 90.9 years for the elastomers and 3,261 years for the coils. Comparing the results from this analysis to the values calculated for the thermal life, it can be seen that margin still exists. This is further substantiated by the fact that the accident test performed was not included in the above thermal equivalence, providing additional margin.

6.4 Other EQER's Table 6.39 lists the other EQER's which discuss equipment, located inside containment potentially effected by the change to the postulated accident profile resulting from power uprate. These EQER's have either evaluated the post-DBE period at a higher temperature than postulated, identified insensitivity to the post-DBE environment, or had evaluations which are consistent with the new postulated profile (e.g., started the post-DBE analysis after 35 days). As such no additional calculations or evaluations are required.

Table 6.39 EQER Not effected by postulated change EQER Comments 05.1 30 day PAOT /30 day test at or above 200'F (C10545) 9.1 evaluated at 90 0C (C10520) 12.1 evaluated at 900 C (C 1214) 12.2 contains no temperature sensitive materials (EQER, Tab C) 12.3 evaluated at 901C (Calculation WPS 12.3.1) 12.4 evaluated at 901C (EQER Tab C) 12.5 evaluated at 228 0 F (EQ Reference No.316, Section 6.13.9.2, page 23) 14.1 Contains no temperature sensitive materials (EQER, Tab C) 22.1 Contains no temperature sensitive materials (EQER, Tab C) 25.1 1/ hour PAOT (EQER Tab C) 27.2 evaluated at 90'C (C10554) 27.4 evaluated at 901C (C 1089) 34.1 evaluated at 901C (Cl 11 12) 34.2 evaluated at 901C (C 11 14) 34.5 evaluated at 901C (C1 1051) 34.8 evaluated at 900 C (Cl 1052) 35.2 Contains no temperature sensitive materials (EQER, Tab C) 35.4 Contains no temperature sensitive materials (EQER, Tab C) 36.1 evaluated at 901C (C11070) 36.2 evaluated at 90'C (Cl 1205) 36.3 evaluated at 90'C (Cl 1204) 37.2 not sensitive /below rating (C10744) 38.1 evaluated at 90'C (C10521) 38.2 evaluated at 901C (C105 1)

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WISCONSIN PUBLIC SERVICE CORP. Section No. C 1473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation Date 11/18/03 l Page 41 of 42 Table 6.39 EQER Comments 43.1 30 day PAOT/ 30 day test at or above 200'F (C10710) 47.1 30 dayPAOT/30 day test at orabove 200°F (C10795) 551 contains no temperature sensitive materials (EQER, Tab C)

7.0 CONCLUSION

Based on this evaluation, the equipment evaluated in the listed EQER's and installed inside containment is qualified to the new postulated temperature/pressure profile resulting from the proposed power uprate.

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WISCONSIN PUBLIC SERVICE CORP. Section No. C11473 Rev. 1 Title Post Accident Operating Time Kewaunee Nuclear Power Plant Evaluations for Various Safety-Related Equipment in Support of Power Uprate Calculation e 11/18/03 Page 42 of 42 Profile Comparison 350 300 250 L

0

'D 200 Cal 150 E

100 50-O 1 1.0OE01 1.OE+00 1.OE+01 1.0E+02 1.0E+03 1.OE+04 1.0E+05 1.0E+06 1.OE+07 Time, seconds l newLOCA -new quafification envelope - - existing - newMSLB Figure 1

C1 1473 Rev 1 Attachment I Page 1 of 8 I I I ,

l

• I I _ _ _ I -I EQER tacc -QER tacc EQER ._ ._

EQER Mcc EQER_l tacc 5.2 432 7.1 9.2 EQER tacc

. _. .. 96.4 432 10.1 432 20.1 I,---- 21.1 696 Ea Tacc Ea_ Ea Ea_

Tacc Tace Ea Tacc_ Ea_ Tacc 0.94 200 1.23 255 1.07 230 0.72 255 _0.78

_ Tser_ tser

. 176.4 _t,?8.,S 185

!eR! rsL- teq(days) 7q(.r_ teq(days) teq(irs) _ teq(days) teq(hrs) teq(days) teq(Qrs_

teq(daYs) teq(hrs)_. I "ecqdays2 1 198 3.78 210.0 8.7_ 725.2_ 30.2 366.0 15.3 82.0 3.4 106.0 4.4 _ 106.0 4.4

_165 55.56 637.5 26.6 1353.3 56.4 893.4 37.2 360.1 15.0 40 8 175_ 420.8 886.5 17.5 153 11i.11 688.8 28.7 _1209.3_ 50.4 36.9. 449.4 18.7 504.9 21.0 504.9 21.0 138 55.56 154.1 6.4 211.1 _88 177.4_ 7 .4 121.4 5.4

__ 5.1 129.5 l_129.5_ 5.4 135 27.78 65.3 2.7 85.0 3.5 73.5_ 3.1 2.2 53.4 56.4 2.4 __ 56.4 2.4 132 138.89 276.0 11.5 341.2 14.2 303.6 12.6 235.1 9.8 245.6 10.2 245.6 10.2 128 138.89 220.2 9.2 53.9 10.6 234.7 9.8._ 197.7 _

8.2 203.6 I _ - _

I

_ 8.5 203.6 8.5 1'IA IIQ 2lq 1'75 2 7.31 1RX.2 7.8 180.9 _ 7.5 165.9 6.9 168.4

__ _ ____ _ _ _ _I- I _ 7.0 168.4 7.0 122 138.89 156.0 6.5 161.7 6.7 158.6 a-6.6 Bs! .8,. 6.3 28.1 1.2 153.0 6.4 120_ 7926.67 7926.67 30.3 7926.67 7926.67 7926.67 330.3 7926.67 330.3 330.3 _ 330.3 _

Clim 437.9 519.0_ 466.7 406.0 77.6

_ 4 -s_ . 4- 413.1 teq(acc) days 1094.7 1V3I9.1s 8435.3 2419.7 350.6 493.4 hlours_ 26272.38 417115 i. 2024464 58073.63 30 dayL 8413.4 11840.65

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C11473 Rev 1 Attachment Page 2 of 8

. - - -I I EQER I acc acc .. QER lacc EQER tacc _ EQER tacc EQER 23.1 l......

101 --- 27.1 432 27.3 432 29.1 432 29.2 144 31.1 Ea_ _Tacc_ Ea _Tacc Ea Tacc Ea Tacc Ea Tacc Ea 0.96 232 0.924 _212 __

_0.929. ._

210 1.016 1.016_ 250__ 1.59

_Tser tser _eq(lirs) teq(dayl 6eq(lrs) _eq~ays) teq(lrs)

..- 9 tepq(Ony). _eq(~dys) teq(!lrE teq(days) teq(hrsf 198 3.78 228.7 9.5 64.6 2.7 65.6

. _ _=. ._ _2.7 290.6 12.1 290.6 12.1 3377.8 165 202.9 55.56 671.5 28.0 201.5 8.4.__ 8.5 776.5 32.4_ 7?76.5 _ 32.4 _3445.5 153 111.11 716.1 29.8 220.0 9.2 220.8 .- .9.2 -- _798.3

_- ____ 33.3 798.3 33.3 2432.1 138 55.56 157.5_-- -6.6 49.9 _ 2.1 49.9 2.1 167.3 7.0 167.3 7.0 312.0 135 27.78 66.5 2.8 21.2 0.9_ 21.2 0.9 69.9 2.9 69.9 2.9 117.8 132 138.89 280.1 11.7 89.9 3.7 89.7 3.7 291 .8 12.2 291.8 12.2 443.9 128 138.89 222.4 9.3 72.0 3.0 71.7 228.6 9.5 228.6 9.5 302.9

_ 124_ 138.89 176.0 7.3 57.5 2.4 57.2 _2.4___ 178.5 7.4 178.5

_ . __ . . _. ___. 7.4 205.6 122 138.89 _156.4_ 6.5.._ 51.3 51.0 157.5 6.6_

.__..2.1 157.5 6.6 169.1 120 7926.67 7926.67 330.3 2611.0 108.8 25i~9-5.3i 108.1 92f66 330.3.7926.67 - 330.3 7926.67 14 .

AA1 '7 142.7

-- I - - - - -I :, ilit 143.3

- -~I

• 453.6 _ 453.6 ___

teq(acc) _1139.9 567. 1028.0_ _ 4907.1_

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ENCLOSURE B NUCLEAR MANAGEMENT COMPANY, LLC KEWAUNEE NUCLEAR PLANT DOCKET 50-305 December 15, 2003 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 Kewaunee Nuclear Power Plant Calculation C11543, "Evaluation of the Impact of the Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project," Revision 0, dated 12/05/03 29 Pages to Follow

WISCONSIN PUBLIC SERVICE CORP. Calc. No. Cl 1543 l Rev. 0 Title Evaluation of the Impact of the Auxiliary Kewaunee Nuclear Power Plant Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project Calculation Date 12/05/03 l Page I of 6 Prepared By , e 9-.._ Reviewed By /;lmi , Fl. pa",

1.0 PURPOSE The purpose of this evaluation is to address the impact on the environmental qualification of equipment of higher Main Steam Line Break (MSLB) temperatures resulting from the 7.4% Power Uprate Project.

2.0 BACKGROUND

The environmental qualification program for electrical equipment important-to-safety must consider the time dependent temperature and pressure at the location of the equipment for the most severe design accident.

The methodology, for computing mass and energy releases for postulated MSLB accidents for Westinghouse reactors,.was described in WCAP-8821 and WCAP-8822. A review of WCAP-8822 noted that the steam generator blowdown model did not account for the heat transfer from the uncovered portion of the steam generator tube bundle to the escaping steam. It was determined that certain break sizes would produce superheated blowdown and greater energy releases than had been previously calculated. The NRC issued JE Information Notice 84-90: Main Steam line Break Effect on EnvironmentalQualificationof Equipment (Reference 5.1).

Wisconsin Public Service Corporation's (WPSC) actual plant response and subsequent mass and energy released from the breaks were determined by the University of Wisconsin-Madison (Reference 5.2). Both Fluor Engineers Inc. and ABB Impell Corporation used this data to develop compartment temperatures. The compartment temperatures were calculated by Fluor using the computer model CONTEMPT/MOD3. Impell calculated the area temperatures using the computer model RELAP5/MOD2 (Reference 5.3). The two computer models were then evaluated and it was determined that Impell's analysis was to be used for determining High Energy Line Break (HELB) temperatures outside containment (Reference 5.4). These analyses were repeated using the GOTHIC computer code to support the Steam Generator replacement (Reference 5.5).

The Steam Generator Replacement project included a new main steam line break analysis using the, GOTHIC code. The first part of that analysis builds on the previous RELAP models and develops Auxiliary Building models using the GOTHIC computer code. The GOTHIC models were then run using the same inputs as the original RELAP models in order to develop a comparative benchmark between the RELAP and GOTHIC models. In the second half of the analysis the GOTHIC model was revised to include compartment walkdown data collected in support of the new analysis. The revised models were run with the new blowdown data developed in support of the Steam Generator replacement design change and the results are documented in Calculation Cl 1074, Auxiliary Building Northwest Quadrantand EastSide GOTHIC Mfodels for use in Equipment Qualification (Reference 5.5).

WISCONSIN PUBLIC SERVICE CORP. Caic. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project Calculation Date 12/05/03 l Page 2 of 6 Calculation C10951, Revision 0 was generated as an evaluation of superheated main steam line break temperatures on EQ equipment located outside containment. The calculation identified equipment located in the areas effected by the MSLB accident transients that was required to be operable and provided a brief evaluation of the qualification of equipment for those environments.

The calculation was revised to address the changes to the break analyses associated with the steam generator replacement.

The Power Uprate project included a new main steam line break analysis using the GOTHIC code.

The GOTHIC analysis is composed of two parts. In the first part of the analysis, GOTHIC models were then using the same inputs as the GOTHIC models used in the Steam Generator replacement change in order to develop a comparative benchmark between the GOTHIC versions and executing platforms. In the second half of the analysis the GOTHIC model was revised to include the new releases generate by Westinghouse at the uprated conditions. These results are documented in Calculation Note CN-CRA-02-64, Kewaunee 7.4% UprateProject:MISLB Outside Containment TemperatureAnalysis (Reference 5.6).

3.0 INPUTS AND ASSUMPTIONS 3.1 The limiting peak temperature along with its applicable break case for each compartment is listed in Table 3.1. The table also shows a worst case pressure for each compartment (Reference 5.6, Tables 6.5-3 through 6.5-6):

Table 3.1 Peak Comnartment Termnerature and Presvirte Compartment Location Peak Break Case Peak Break Case Temperature Pressure, psia (if harsh)

Al Basketball Court 495.4 0F Case 5bAI-1 19.8 NW Case Ia BI A MSIV Area 457.5 0F NWCase2a 20.1 NW Case la BI A MSIV Area 410.4 0F NN Case 5b 17.7 NW Case 5b Cl A FW Area 213.8 0 F NW Case la 20.1 NW Case la Cl A FWV Area 176.60 F NW Case b 17.8 Case 5bAI-2 CH B FWVArea 319.7 0F EastCase5b 15.2 East Case Sb DII BMSIVArea 374.70 F EastCase7b 15.2 East Case 5b Dla SGBlowdown 333.60 F EastCase7b 15.2 EastCaseSb Fla PotatoBinsArea 169.8 0F NWCaselb 20.1 NWCaselb Fla Potato BinsArea 147.9 0 F NWCase6b 17.8 NWCase9b IL ases I tnrougn 4 are large oreaks and case through IU are small breaKs.

3.2 The equipment identified in Appendix II of C10951, Revision 0, Equipment and Subcomponent Listing by EQER, is the same equipment which requires evaluation to the new conditions resulting from power uprate.

WISCONSIN PUBLIC SERVICE CORP. Caic. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project Calculation Date 12105/03 l Page 3 of 6 3.3 As documented in calculation C10951, Revision 0, the equipment located in Compartment AI is only required for a LOCA. Therefore qualification is not required for the MSLB temperatures. Compartment AI has been included in the above table only for completeness.

3.4 The MSLB analysis resulted in the identification of harsh temperatures in Compartment Fla.

Table I indicates that the peak temperature in Compartment Fla is 169.8 0 F. The only EQ components/subcomponents installed in this compartment are cables and splices (RAY-3, ROC-25 and BIW-1) as described in Calculation C10951, Revision 0. These components are associated with the circuits for the Steam Generator Steam Pressure transmitter that located in Compartment BI. Qualification to the peak temperatures is demonstrated by the same equipment types installed in Compartments with higher peak temperatures.

3.5 Appendix I of this calculation, Superheat Evaluation by EQER, includes the superheat evaluation for each EQER listed in Appendix II of calculation C10951, Revision 1. This list provides the current peak qualification temperature, the applicable EQER file, test report, EQ Reference No., and the required compartment temperature or thermal lag peak to which the equipment must be qualified. Note that Appendix I also includes all of the compartments which are applicable to a particular EQER as shown in Calculation C10951, Revision 0, along with the associated peak temperatures for those compartments.

3.6 Although it is not believed that the cables associated with EQERs 27.3 and 38.2 are routed through the cable spreading area (Compartment DII), qualification will still be documented and the EQERs included in Appendix I.

3.7 EGS performed MSLB testing on ASCo NP8136 solenoid valves, taken from KNPP stock.

The temperature profile was a composite of the different break cases for Compartment BI. As such, an extended plateau (dwell) at the peak temperature was included as the test requirement. During the testing, the solenoid valves demonstrated operability at the beginning and end of the high temperature exposure (peak reading 501'F, average temperature approximately 475 0F for - 10 minutes, Reference 5.7 Figure 1), as well as at the completion of the test.

3.8 The peak surface and/or internal air temperature for the various components are shown in Table 3.2 below. These values are taken from Reference 5.8, Tables 7-1, 7-2 and 7-3. The components were modeled using GOTHIC and the various post power uprate break cases were run. The table shows the peak temperature for Compartments BI and DII and the break case that resulted in the highest temperature.

WISCONSIN PUBLIC SERVICE CORP. Calc. No. C11543 IRev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project Calculation Date 12/05/03 l Page4of 6 Table 3.2 Component Peak Surface / Internal ir Temperature Component BI Case DII Case 1.5" Rigid Al conduit(X) 427.3 lb 327.0 7b 3/4" Rigid Galvanized steel conduit(0) 412.9 lb 319.0 7b 3/" Flex Galvanized steel conduits) ,Sealtight 420.2 lb 324.8 7b 10" x 10" x 6" Galvanized steel junction box(2) 401.1 2b 313.3 7b 3/" Conduit Seal enclosure"s 363.2 lb 295.0 9b Rockbestos cable/3) NA NA 354.4 7b Okonite cable(3) NA NA 357.6 7b BIW cable(3) NA NA 325.5 7b (1) Internal Enclosure Wall Temperature (2) Internal Enclosure Air Temperature (3) Cable Outside Wall Surface Temperature 3.9 Most of the equipment exposed to the MSLB conditions outside containment has already been demonstrated to be qualified to the LOCA conditions inside containment. Those conditions, although not reaching the same peak temperatures, are much more sever than the MSLB due to the overall length of the exposure in the tests and the dwell times at the various peak temperatures. For that equipment already qualified for the in-containment conditions, demonstrating its ability to operate at the MSLB peak temperature or its equivalent to the actual equipment by thermal lag analysis is sufficient to assure the equipment is qualified for the application.

3.10 The junction box modeled in the thermal lag analysis is 10" x 10" x 6" Galvanized steel. This is the smallest box used in Compartment BI. Using the smallest box is conservative as it has, the smallest mass and the distance to the internal components and equipment is minimized.

3.11 The BIW cable modeled in the thermal lag analysis is that shown on BIW sketch 9104-H-002. This cable is the smallest of the BIW cables installed. All other constructions have more conductors, thicker insulation and jackets and thus greater mass. Therefore, the modeled cable conservatively represents all BIW cables at KNPP.

3.12 Cable in the cable spreading area (Compartment DII) were modeled in open cable trays.

WISCONSIN PUBLIC SERVICE CORP. CaIc. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Pover Uprate Project Calculation Date 12/05/03 l Page 5 of 6 3.13 Condulets were not specifically modeled in the thermal lag analysis. Due to the small size and significant wall thickness of the condulet, the results for the junction boxes have been assumed for these components.

3.14 The conduits modeled are typical for those found in Compartment BI and DII. The results for the 1.5" Al conduit will be conservatively used to represent all the conduit in these areas since it had the hichest result from the thermal lag analysis.

4.0 METHODOLGY AND ACCEPTANCE CRITERIA 4.1 A review of available test reports was performed to identify additional testing and/or to verify the peak temperature referenced in the EQER enveloped the accident conditions for all of the EQER's listed in Appendix I.

4.2 The Evaluation discussions included in Appendix I have been included where the applicable EQER does not demonstrate qualification to the new analysis or additional information is necessary to explain why the existing qualification is adequate.

4.3 The results of the initial evaluation identified a number of components where additional information or analysis were necessary to satisfy the new temperature requirements. As such a thermal lag analysis was performed to facilitate the review of the existing and new test data.

4.4 The results of the thermal lag analysis further identified components (ASCO solenoid valves) for which no information was available to support qualification to the new temperature peaks. For these components a test program was undertaken to demonstrate operability for this equipment at the new temperature peaks.

5.0 REFERENCES

5.1 IE Information Notice No. 84-90: Main Steam Line Break Effect on Environmental Qualificationof Equipment, dated December 7, 1984.

5.2 High Energy Line Break Analysis for the Kewaunee PressurizedWater Reactor, prepared by University of Wisconsin-Madison, dated July, 1986 (EQ Reference No. 452).

5.3 Auxiliary Building Temperature Response to SuperheatedMain Steam Line Breaks Outside Containment, prepared by ABB Impell Corporation, dated October, 1990 (EQ Reference No.

446).

5.4 Evaluation of Compartment TemperatureAnalyses for HELB-Superheat Project, prepared by KA Hoops, dated 09/18/96 (EQ Reference No. 446).

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WISCONSIN PUBLIC SERVICE CORP. Caic. No. C 1543 lRev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project Calculation IDate 12105/03 l Page 6 of 6 5.5 Kewaunee Nuclear Power Plant Calculation, Cl 1074, AlLiliary Building Northwest Quadrantand East Side GOTHIC Models, Revision 0, dated May 2001 (EQ Reference No.

500).

5.6 Kewaunee 7.4% Uprate Project:MSLB Outside Containment Temperature Analysis, Westinghouse Electric Company, Calculation Note Number CN-CRA-02-64, dated January 25,2003 (EQ Reference No. 513).

5.7 MSLB Test Report for ASCO Solenoid Valves Model NP8316E34E & NPL8316B76E, EGS Report Number EGS-TR-23050-0251-02, Revision 0, dated November 17, 2003 (EQ Reference No. 515).

5.8 KNPP HELB Thermal Lag Analysis, SCENTECH Report Number 17550-M-001, dated November 9, 2003 (EQ Reference No. 516).

5.9 Kewaunee Nuclear Power Plant Calculation, C10951, AuLiliary Building Northwest Quadrantand East Side GOTHIC Models, Revision 0, dated August 21, 1998 and Revision 1, dated November 26, 2001.

6.0 CALCULATIONS AND RESULTS The results of this evaluation are contained in Appendix I.

7.0 CONCLUSION

AND RECOMMENDATIONS Based on the results of the attached evaluation and documentation, all EQ equipment located in areas subject to superheated steam from a MSLB outside containment that is required to operate to mitigate those accidents is qualified.

WISCONSIN PUBLIC SERVICE CORP. Calc. No. Cl 1543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 1 of 23 EQER 5.1 Rev: 9

Title:

ASCO Solenoid Valve, Series 206-380, 206-381, 206-832, NP8320, NP8314 and NP8316 Manufacturer: Automatic Switch Company Model/Series 206-380, 206-381, 206-832, NP8320, NP8314, NP8316 Category Solenoid Valve EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 180 AOR-67368 Rev.1 450 100.7 Figure 4.2, page 26 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, see psia

Reference:

B! Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 BI Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 CIH Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 & -6 DII Design Basis Crack 374.8 1202 -, 15.2 513 Tables 6.5-4 & -6 Super Heat Qualification EQ Ref Report ID Comments:

515 EGS-TR-23050-0251-02 501 72.4 Figure 1, page x Evaluation: The existing qualification did not envelope the new peak temperature. A MSLB test program was conducted on solenoid valves taken from KNPP stock. The valves were aged to the same levels as in the existing qualification documentation to an equivalent life of 40 years. They were then subjected to superheated steam to simulate the conditions associated with a main steam line break in compartment BI. The test profile was a composite of the various postulated breaks in the compartment and as such much more severe that any single case. The test profile Included 15@F margin in the required profile peak (4580 F + 15'F = 473'F). The valves successfully operated at the peak temperature (both the beginning and the end of the peak plateau) and at the completion of the test.

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than tha postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Additional testing on aged solenoid valves demonstrates qualification to the peak temperature and operability time in superheated steam conditions In the above compartments

WISCONSIN PUBLIC SERVICE CORP. Calc. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 L Page 2 of 23 EQER 7.1 Rev: 4

Title:

Boston Insulated Wire Co. Instrumentation & Control Cable Manufacturer: Boston Insulated Wire, BIW Model/Series See EQER Tab A Category Cable EQ Type: H2 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 391 IPS-348, Rev. 0 345 124.7 12 6901 318 104.7 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 Cl1 Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 & -6 DIl Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Dlla Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 & -6 Thermal Lag Comp. Component

Reference:

D1I BIW Cable surface 325.5 516 Tables 7.3 Super Heat Qualificatlon EQ Ref Report ID Comments:

391 IPS-348, Rev. 0 345 Peak surface temperature In Dil below qualification temperature Evaluation: Results of the thermal lag analysis show the peak surface temperature of the cable to be 325.52F in Compartment DIl. The peak temperature of the qualification profile (LOCA) Is 3450F and Is maintained for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This is very conservative compared to the MSLB where the peak is maintained for less than 10 minutes. Thus the existing qualification demonstrates operability at conditions, which are thermally more severe than the postulated MSLB outside containment

Conclusions:

Based on current testing and performance, BIW Bostrad-7 cable is fully qualified and will perform its safety function in superheated steam conditions In the above compartments post accident.

WISCONSIN PUBLIC SERVICE CORP. Calc. No. C1 1543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation b EQER Calculation Date 12/05/03 l Page 3 of 23 EQER 9.1 Rev: 5

Title:

Brand Rex Fire Retardant Irradiation Cross-Linked Polyethylene XLPE Power and Control Cable Manufacturer: Brand Rex Model/Series See EQER Tab A Category Cable EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 286 F-C5120-1 415 80.7 page 5-5 and 5-6 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

8I Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 B' Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 CI Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 Cl Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 & -6 DII Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Dila Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 & -6 Thermal Lag Comp. Component

Reference:

BI 1.5" rigid conduit 427.3 516 Tables 7.1 Super Heat Qualification EQ Ref Report ID Comments:

483 SANDS0-2629 503 Based on MOkD criteria Table 4, page 18 Evaluation In testing performed by Sandia National Laboratories, Brand Rex cable was exposed to high temperature steam. The test specimen was a 3/C #12 AWG XLPE insulated cable similar to those Installed at KNPP. In this testing the lowest failure temperature for a 10kQ acceptance criteria for a 100-metercable was 5030F (conductor 1). This converts to 3289kf)-ft (100 meterx 3.289 feet/meter x 10 k 0-100m). For power and control circuits the DC grounds analysis (EQ Reference 466) 525k0-ft was the minimum value used in circuit analysis. This temperature/failure criteria combination adequately meets the installed equipment requirements.

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation of the qualification to the peak temperature is adequate to show acceptability of the equipment Based on current testing and performance, EQER 9.1 cables are fully qualified and will perform their safety function In superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBUC SERVICE CORP. Calc. No. C11543 lIRev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 4 of 23 EQER 12.1 Rev: 8

Title:

Conax Electrical Conductor Seal Assemblies (ECSA)

Manufacturer: Conax Model/Series See EQER Tab A Category ECSA EQ Type: HI Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 148 IPS-409. IPS-409.1 430 84.7 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 Super Heat Qualification EO Ref Report ID Comments:

148 IPS-409, PS-409.1 430 84.7 Evaluation

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation of the qualification to the peak temperature is adequate to show acceptability of the equipment Based on current testing and performance, EQER 12.1 ECSA's are fully qualified and will perform their safety function in superheated steam conditions in the above compartments post-accident.

..... . ..- . . .. . - . - . .. - . . - ... z:_

WISCONSIN PUBLIC SERVICE CORP. Calc. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 5 of 23 EQER 23.1 Rev: 3

Title:

General Electric Terminal Blocks EB-5, EB-25 Manufacturer: General Electric ModelSeries EB-5, EB-25 Category Terminal Block EQ Type: H2 Temperature Profiles Current Qualification Peak Pressure, ED Ref. Report Temp, F psia Comments 313 80119 312 84.7 181 110-11004 470 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

81 Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 Bi Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 CHi Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 & -6 Dil Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Thermal Lag Comp. Component

Reference:

SI junction box 401.1 516 Tables 7.1 Super Heat Qualification EQ Ret Report ID Comments:

181 110-11004 470 74.0 Evaluation: These terminal blocks located nside junction boxes in Compartment Si where the peak internal temperature was shown to be 401.1 F by the thermal lag analysis.

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation the of the qualification to the peak temperature Is adequate to show acceptability of the equipment Based on testing and performance, General Electric EB-5 terminal blocks are fully qualified and will perform their safety function in superheated steam conditions In the above compartments post-accident.

.. .1.. I .. .1. .--. -.. 1... . .. . - -- .. ...- - . .. - . -- -- ... .... - - .- - -. --- --. -...

WISCONSIN PUBLIC SERVICE CORP. Caic. No. C1 1543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 6 of 23 EQER 27.1 Rev: 6

Title:

Kerite 600 VLow Voltage Power Control Cable, 50 mils FR (HI-70) Insulation, 65 mils FR (HC-71 1) Jacket Manufacturer: Kerite ModelSeries See EQER Tab A Category Cable EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 205 F-C4020-1 346 139 FIRL Technical Report Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

el Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 BI Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -;

Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 C1l Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 & -6 Dll Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 D1la Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 & -6 Thermal Lag Comp. Component

Reference:

BI 1.5' rigid conduit 427.3 516 Tables 7.1 Super Heat Qualification EQ Ref Report ID Comments:

484 20255.1 450 Internal box temperature. peak surface Figure G.1 temperature external 540'F Evaluation: The 450'F temperature was from readings of thermocouples inside the junction box, near the surface of the cable/terminal block assemblies. Thermocouples located outside the box showed the peak test temperature to be 540'F(EQ Reference No 484). These cables are located inside junction boxes or conduit In Compartment BI where the peak Internal temperature was shown tot be 427.3'F by the thermal lag analysis.

Conclusions:

EQER 27.5 Kerite 600V Control Cable, 40 Mils FR(HI-70) Insulation, 65 Mils FR(HC-71 1) Jacket (Superheated HELB - Supplement to ECER 27.1 and 27.3) evaluates this cable using the Farwell & Hendricks report, which envelop the accident conditions in compartment BI. Based on the testing and performance documented in EQER 27.5 the Kerite FR/FR cable is fully qualified and will perform its safety function in the above compartments

WISCONSIN PUBLIC SERVICE CORP. Calc. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calcilation Date 12/05/03 l Page 7 of 23 EQER 27.3 Rev: 5

Title:

Kerite 600V Control Cable, 40 mils FR (HI-70) Insulation, 65 mils FR (HC-711) Jacket Manufacturer: Kerite Model/Series See EQER Tab A Category Cable EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EO Ref. Report Temp, F psla Comments 378 F-C4158 320 98.2 FIRL Technical Report Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

Oil Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 &-6 Super Heat Qualification EQ Ref Report ID Comments:

484 20255.1 450 Internal box temperature, peak surface Figure G.1 temperature box external 540'F Evaluation: The 450 0F temperature was from readings of thermocouples inside the junction box, near the surface of the cablelterminal block assemblies. Thermocouples located outside the box showed the peak test temperature to be 540'F.

Conclusions:

EQER 27.5, Kente 600V Control Cable, 40 Mils FR(HI-70) Insulation, 65 Mils FR(HC-71 1) Jacket (Superheated HELB - Supplement to ECER 2Z71 and 27.3) evaluates this cable using the Farwell & Hendricks report (EQ Reference 484) which envelop the accident conditions in compartment Dil. Based on the testing and performance documented In EqER 27.sthe Kerite FR/FR cable is fully qualified and will perform its safety function in the above compartments

WISCONSIN PUBUC SERVICE CORP. Calc. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 i Page 8 of23 EGER 29.1 Rev: 9

Title:

Limitorque Valve Acutuators-lnside/Outside Containment Type Limitorque Report No. 600456 and B0027 Manufacturer: Limitorque Model/Series SMB-2, SMB-00, SMB-000, SB-00, HOBC-SMB-000 Category Valve Actuator EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 100. Limitorque Qual Rpt 600456 310 84.7 100 Repor? B0027 492 Accident data Peak Time Into Pressure Camp. MSLB Accident Temp, F Event, sec psIa

Reference:

Dii Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Super Heat Qualification EQ Ref Report ID Comments:

100 Umitorque Qual. Rpt B0027 492 80.7 Evaluation

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) nside containment, which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation of the qualification to the peak temperature Isadequate to show acceptability of the equipment Based on current testing and performance, EQER 29.1 Lmitorque valve actuators are fully qualified and will perform their safety function in superheated steam conditions in the above compartment post accident.

.... -_ 1.- - .- . '_ I -1 ..... .-- - -... - - - -I WISCONSIN PUBLIC SERVICE CORP. Calc. No. C11543 Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 9 of 23 EQER 29.3 Rev: 8

Title:

Limitorque Valve Actuators - Outside Containment Type Limitorque Report No. B0003 Manufacturer: Limitorque Model/Series SMB-O0, SMB-1, SMB-2 Category Valve Actuator EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EO Ref. Report Temp, F psia Comments 100 80003 250 39.7 Accident data Peak Time into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 Super Heat Qualification EO Ref Report ID Comments:

100 B0003 250 39.7 Evaluation:

Conclusions:

This equipment has been qualified for outside containment, to a test profile, which is thermally more severe than the postulated MSLB outside containment. Based on current testing and performance. EQER 29.3 actuators are fully qualified and will perform their safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. Caic. No. C1 1543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendrx I - Evaluation by EQER Calculation Date 12/05/03 l Page 10 of 23 EQER 29.4 Rev: 1

Title:

Limitorque Valve Actuators - Outside Containment Type Limitorque Report No. 80009 Manufacturer: Limitorque ModeVSerles SMB-000 Category Valve Actuator EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EO Ref. Report Temp, F psia Comments 100 B0009 340 143.7 Accident data Peak Time into Pressure Comp. MSLB Accident Temp, F Event, see psia

Reference:

Dila Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 &-6 Super Heat Qualification EQ Ref Report ID Comments:

100 Limitorque Qual. Rpt B0027 492 80.7 Evaluation:

Conclusions:

Based on current testing and performance, EQER 29.4 actuators are fully qualified and wmiperform their safety function in superheated steam conditions in the above compartments post-accident. Discussion regarding the thermal lag experienced by the actuator internals In 80027 (EQ Reference 100) confirms the adequacy of the existing testing.

WISCONSIN PUBLIC SERVICE CORP. Calc. No. Cl 1543 _ l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 11 of 23 EQER 31.1 Rev: 2

Title:

Marathon Terminal Block Assemblies Located Inside and Outside Containment Model 1500 NUC and 142 NUC Manufacturer: Marathon Model/Series 1500 NUC, 142 NUC Category Terminal Block EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 244 45603-1 380 94.7 Accident data Peak Time nto Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

DlI Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Thermal Lag Comp. Component

Reference:

DII junction box 313.3 516 Tables 7.1 Super Heat Qualification EQ Ref Report ID Comments:

244 45603-1 380 94.7 Evaluation: These terminal blocks are located in junction boxes in compartment D. Peak internal temperature from the thermal lag analysis is 313.30F.

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, than the the qualification to the peak temperature is adequate to show acceptability of the equipment Based on testing evaluation of and performance.

ECER 31.1 terminal blocks are fully qualified and will perform their safety function In superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. CaIc. No. C11543 Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 Page 12 of 23 EQER 31.2 Rev: 1

Title:

Marathon Terminal Block Assemblies Located Inside and Outside Containment Series 1500 Manufacturer: Marathon Model/Series 1508, 1512 Category Terminal Block EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 426 60538 303 80.7 484 20255.1 450 26.7 Accident data Peak Time into Pressure Camp. MSLB Accident Temp, F Event, sec psia

Reference:

81 Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 Si Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Thermal Lag Camp. Component

Reference:

81 junction box 401.1 516 Tables 7.1 Super Heat Qualification EQ Ref Report ID Comments:

484 20255.1 450 Internal box temperature. peak surface Figure G.1 temperature box external 540'F Evaluation: The 4500F temperature was from readings of thermocouple inside the junction box, near the surface of the cable/terminal block assemblies. Thermocouples located outside the box showed the peak test temperature to be 5400F. These terminal blocks located inside junction boxes in Compartment B where the peak internal temperature was shown to be 401.1*F by the thermal lag analysis.

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which Is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation of the qualification to the peak temperature is adequate to show acceptability of the equipment Based on current testing and performance, EQER 31.2 terminal blocks are fully qualified and will perform their safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. Caic. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 13 of 23 EQER 33.1 Rev: 12

Title:

Namco EA1 80 Series Limit Switches (BOM Revision Level H through N (Code 4086))

Manufacturer: Namco Controls Model/Series EA180 Series Category Limit Switch EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 175 QTR-105, Rev. 4 340 94.7 Namco test report Accident data Peak Time into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 CHi Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 & -6 Super Heat Qualification EO Ref Report ID Comments:

175 QTR-105, Rev. 4 340 94.7 Evaluation:

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation of the qualification to the peak temperature is adequate to show acceptability of the equipment Based on current testing and performance, Namco EA180 series limit switches are fully qualified and will perform their safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. Caic. No. C11543 lRev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 14 of 23 EQER 33.5 Rev: 4

Title:

Requalification of NAMCO EA1 80 Series Limit Switches (BOM Revision Level Later Than N (Code 4086))

Manufacturer: Namco Controls Model/Series EA1 80 Series Category Limit Switch EQ Type: HI Temperature Profiles Current Qualification Peak Pressure, EQ Ret. Report Temp, F psia Comments 422 QTR-155 357 89.3 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

SI Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 BI Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 CII Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 &-6 Dil Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Super Heat Qualification EC Ref Report ID Comments:

484 20255.1 600 56.7 (See EQER 33.7)

Figure F.1 Evaluation:

Conclusions:

EQER 33.7, NAMCO EA 180 Series Umit Switches (MSLB Qualification) evaluates these limit switches using the Farwell &

Hendrick's report (EQ Reference 484) which envelops accident conditions in compartment Bl. Based on the testing and performance documented in EQER 33.7, the Namco limit switches are fully qualified and will perform their safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. CaIc. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 15 of 23 EQER 34.4 Rev: 3

Title:

Okonite - Low Voltage Power and Control Cable EPR Insulated, Okoprene or Okoseal Jacketed Manufacturer: Okonite Cable Company Model/Series See EQER Tab A Category Cable EQ Type: H2 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 333 Okonite Eng. Report No. 141 345 118.7 483 SAND9O-2629 750 91.2 Testing was performed in groups at several elevated temperature levels up to 400 C (>750 F) 11 Okonite Eng Report No 110E 307 74.7 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 DIl Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Super Heat Qualification EQ Ref Report ID Comments:

483 SAND90-2629 750 Table 4, Page 18 Evaluation

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment. which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation of the qualification to the peak temperature is adequate to show acceptability of the equipment. Based on current testing and performance, EQER 34.4 cable is fully qualified and will perform its safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. Caic. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Iine Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 16 of 23 EQER 35.1 Rev: 8

Title:

Patel Conduit Seals Manufacturer: Patel Engineers/EGS Corp. International Model/Series 8412-06 Category Conduit Seal EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 387 PEI-TR-841203-12 415 114.7 Appendix VII, Figure 3 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

BI Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 81 Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 CI Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 CHI Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 & -6 Dil Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Dlla Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 & -6 Thermal Lag Comp. Component

Reference:

SI Conduit Seal internal surface 363.2 516 Tables 7.1 Super Heat Qualification EQ Ref Report ID Comments:

387 PEI-TR-841203-12 415 114.7 Appendix V1I, Figure 3 Evaluation: The test exposed the conduit seal to a peak of 415@F and temperatures above 400'F for a total of 34 minutes, followed by approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at or above 340 0F. The results of the thermal lag show for the peak temperature of 458'F and total time above 340°F of less than 12 minutes, the internal surface of the conduit seal only reaches 363.2 0F. The high pressure end of the conduit seal was open to the test chamber environment (no conduit was installed on the Field side). As such the testing exposed the internals to actual temperatures higher than those anticipated for the as-installed configuration.

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation of the qualification to the peak temperature is adequate to show acceptability of the equipment. Based on current testing and performance, EQER 35.1 conduit seals are fully qualified and will perform their safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. CaIc. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12105/03 l Page 17 of 23 EQER 35.2 Rev: 1

Title:

Patel Thread Sealant Paste Manufacturer: Patel Engineers ModeUSeries P-1 Category Thread Sealant EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 420 PEI-TR-841209-04 412 86.7 Accident data Peak Time into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

El Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 Bi Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 Cl' Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 & -6 DII Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 DIla Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 & -6 Super Heat Qualification EQ Ref Report ID Comments:

420 PEI-TR-841209-04 412 86.7 Evaluation: The thread sealant paste consists of a graphite flake In a linseed oil carrier. The carrier is volatile and essentially gone after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, leaving the graphite to effect the seal. The graphite is chemically Inert and thermally stable to temperatures in excess of 8001F.

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation of the qualification to the peak temperature is adequate to show acceptability of the equipment. Based on current testing and performance, EQER 35.2, thread sealant is fully qualified and will perform its safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBUC SERVICE CORP. Calc. No. C11543 Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12105/03 l Page 18 of 23 EQER 36.1 Rev: 4

Title:

Raychem Nuclear In-ULne Cable Splice Assemblies WCSF-N Manufacturer: Raychem ModeVSeries WCSF-N Category Cable Splice EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 269 EDR-5019 400 146.7 Raychem Report Accident data Peak Time into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 Dil Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Dila Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 & -6 Super Heat Qualification EQ Ref Report ID Comments:

269 EDR-5019 400 162.7 Raychem Report Evaluation:

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation of the qualification to the peak temperature is adequate to show acceptability of the equipment Based on current testing and performance, EQER 36.1 cable splices are fully qualified and will perform their safety function in superheated steam conditions in the above compartments post-accident.

I-,,

WISCONSIN PUBLIC SERVICE CORP. Calc. No. C1 1543 Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 19 of 23 EQER 36.2 Rev: 6

Title:

Raychem Nuclear Plant Kit (NPK) and Nuclear Plant Stub Connection Kit (NPKV)

Manufacturer: Raychem Model/Series NPKC,KPKP,NPKX,NPKV,NPKS Category Cable Splice EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psla Comments 269 58722-1 420 132.3 Accident data Peak Time into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

81 Large Break 457.5 163 20.1 513 Tables 6.5-3 -5 BI Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 CI Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 DlI Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Dila Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 & -6 Thermal Lag Comp. Component

Reference:

8I junction box 401.1 516 Tables 7.1 Super Heat Qualification EQ Ref Report ID Comments:

269 58722-1 420 134.7 Evaluation: These splices are installed inside junction boxes or condulets (See section 3.13) at KNPP. The results of the thermal lag analysis show the internal air temperature of a typical unction box to be 401.1OF for the worst case accident in Compartment B. This is below the existing qualification temperature of 420 0F.

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, evaluation f 9

the qualification to the peak temperature is adequate to show acceptability of the equipment Based on current testing and performance, EQER 36.2 splices are fully qualified and will perform their safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBUC SERVICE CORP. Calc. No. C1 1543 lRev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12105/03 l Page 20 of 23 EQER 38.1 Rev: 7

Title:

Rockbestos Irradiation Cross-Linked Polyethylene Firewall III Cable Manufacturer: Rockbestos Company ModeVSeries See EQER Tab A Category Cable EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 323 OR-5805 341 129.1 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

BI Large Break 457.5 163 20.1 513 Tables 6.5-3 &-5 81 Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 &-5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 &-5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 Dil Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Dila Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 &-6 Thermal Lag Comp. Component

Reference:

B 1.5' Aluminum conduit 427.3 516 Tables 7.1 Super Heat Qualification EQ Ret Report ID Comments:

445 TR-880707-04 435 87.7 (See EQER 38.6) 483 SAND90-2629 515 Based on 10k criteria Table 4, page 18 Evaluation: These cables are installed inside conduit In Compartment Bl. The results of the thermal lag analysis show the internal air temperature of the conduit to be 427.30F for the worst-case accident In Compartment 81. This is below the existing qualification temperature of 435'F. In testing performed by Sandia National Laboratories, Rockbestos cable was exposed to high temperature steam. The test specimen was a 3/C #12 AWG XLPE insulated cable similar to those installed at KNPP. In this testing the lowest failure temperature for a 100k acceptance criteria for a 100-meter cable was 5151F (conductor 14). This converts to 3289kD-ft. For power and control circuits the DC grounds analysis (EQ Reference 466) 525kQ-ft was the minimum value used in circuit analysis.

This temperature/failure criteria combination adequately meets the installed equipment requirements.

Conclusions:

EQER 38.6, Rockbestos Cable (MSLB Superheat Supplement to EOER 38.1, 38.2 and 38.5), evaluates these cables using the EGS report, which envelops the accident conditions in compartment B. Based on the testing and performance documented in EQER 38.6 the Rockbestos cable is fully qualified and will perform its safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. Caic. No. C 1543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12105103 l Page 21 of 23 EQER 38.2 Rev: 6

Title:

Rockbestos Chemically Cross-Linked Polyethylene Firewall I Cables Manufacturer: Rockbestos Company ModeVSeries See EQER Tab A Category Cable EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, Et Ret. Report Temp, F psia Comments 325 OR-5804 341 121.7 Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, see psia

Reference:

DII Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Super Heat Qualification Ea Ret Report ID Comments:

445 TR-880707-04 435 87.7 (See ECER 38.6)

Evaluation:

Conclusions:

EQER 38.6. Rockbestos Cable (MSLB Superheat Supplement to EQER 38. ,38.2 and 38.5), evaluates these cables using the EGS report which envelops the accident conditions in compartment BI. Based on the testing and performance documented in EQER 38.6 the Rockbestos cable is fully qualified and will perform its safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. Caic. No. C11543 l Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Buildin, Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 22 of 23 EQER 39.3 Rev: 4

Title:

Rosemount Inc. Pressure Transmitters Model 1153 Series D Manufacturer: Rosemount, Inc.

ModeVSerles 1153 Series D Category Pressure Transmitter EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psla Comments 301 D8300040 350 99.7 Rosemount test report Accident data Peak Time Into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

81 Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 Bl Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 DII Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 Dlla Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 & -6 Super Heat Qualification EQ Ref Report ID Comments:

485 20255.2 495 26.7 (See EQER 39.5)

Figure H.1 and H.3 Evaluation:

Conclusions:

EQER 39.5, Rosemount Inc. Pressure Transmitters Model 1153 Series 0 (Superheated HELB - Supplement to EQER 39.3),

evaluates these transmitters using the Farwell & Hendrick's report which envelops the accident conditions in compartment Based on testing and performance documented in EQER 39.5. the Rosemount transmitters are fully qualified and will performSI.

their safety function in superheated steam conditions in the above compartments post-accident.

WISCONSIN PUBLIC SERVICE CORP. Calc. No. C11543 I-Rev. 0 Title Evaluation of the Impact of the Kewaunee Nuclear Power Plant Auxiliary Building Main Steam Line Break Analysis on EQ Equipment in Support of the Power Uprate Project -

Appendix I - Evaluation by EQER Calculation Date 12/05/03 l Page 23 of 23 EQER 54.1 Rev: 6

Title:

EGS Grayboot Connector Manufacturer: EGS Corporation Model/Series GB-1, GE-2, GB-3 Series, GB-S-2, GB-S-3 Category Connector EQ Type: H1 Temperature Profiles Current Qualification Peak Pressure, EQ Ref. Report Temp, F psia Comments 445 TR-880707-04, Rev. D 450 87.7 Accident data Peak Time into Pressure Comp. MSLB Accident Temp, F Event, sec psia

Reference:

81 Large Break 457.5 163 20.1 513 Tables 6.5-3 & -5 81 Design Basis Crack 410.4 1225 17.7 513 Tables 6.5-3 & -5 Cl Large Break 213.8 27 20.1 513 Tables 6.5-3 & -5 Cl Design Basis Crack 176.6 66 17.8 513 Tables 6.5-3 & -5 CII Design Basis Crack 319.7 1202 15.2 513 Tables 6.5-4 & -6 DI1 Design Basis Crack 374.8 1202 15.2 513 Tables 6.5-4 & -6 DIla Design Basis Crack 333.6 1202 15.2 513 Tables 6.5-4 & -6 Thermal Lag Comp. Component

Reference:

SI unction box 401.1 516 Tables 7.1 Super Heat Qualification EQ Ref Report ID Comments:

445 TR-880707-04 450 87.7 Evaluation: These connectors are install inside junction boxes or condulets. The results of the thermal lag analysis show the temperature of a typical junction box to be 401.1 OF for the worst-case accident In Compartment B1. This is below ntemal air the existing qualification temperature of 4350F.

Conclusions:

This equipment has been qualified to the qualification profile (LOCA) inside containment, which is thermally more severe than the postulated MSLB outside containment due to the duration and extended periods at elevated temperature. Therefore, the qualification to the peak temperature is adequate to show acceptability of the equipment. Based on current evaluation 9f performance, EQER 54.1 connectors are fully qualified and will perform their safety function in superheated testing and steam conditions in the above compartments post-accident.