ML080650219

License Amendment 176 Replacement of Main Stream and Feedwater Isolation Valves

ML080650219
Person / Time
Site:	Wolf Creek
Issue date:	03/21/2008
From:	Donohew J NRC/NRR/ADRO/DORL/LPLIV
To:	Muench R Wolf Creek
Donohew J N, NRR/DLPM,415-1307
Shared Package
ML080650198	List: ML080650219
References
TAC MD4840
Download: ML080650219 (46)

Text

March 21, 2008 Mr. Rick A. Muench President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839

SUBJECT:

WOLF CREEK GENERATING STATION - ISSUANCE OF AMENDMENT RE:

REPLACEMENT OF MAIN STEAM AND MAIN FEEDWATER ISOLATION VALVES (TAC NO. MD4840)

Dear Mr. Muench:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 176 to Facility Operating License No. NPF-42 for the Wolf Creek Generating Station. The amendment consists of changes to the license in response to your application dated March 14, 2007, as supplemented by letters dated September 12, October 16, and December 14 (two letters), 2007, and January 18, 2008 (ET 07-0004, ET 07-0040, ET 07-0050, ET 07-0060, ET 07-0062, and ET 08-0005, respectively).

The amendment authorizes (1) the replacement of the main steam isolation valves (MSIVs) and main feedwater isolation valves (MFIVs) and (2) the use of Figures B 3.7.2-1 (MSIVs) and 3.7.3-1 (MFIVs) as the limiting closure times for these valves to demonstrate that these valves meet the limiting conditions for operation with respect to the valve closure time. The remaining amendment requests in the application that have not yet been addressed by the NRC are the proposed (1) addition of main feedwater regulating valves and bypass valves to TS 3.7.3, "Main Feedwater Isolation Valves," and (2) modification of the main steam and feedwater isolation system controls. These requests will be addressed in future letters.

A copy of our related Safety Evaluation is enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Sincerely,

/RA/

Jack N. Donohew, Senior Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482

Enclosures:

1. Amendment No. 176 to NPF-42

2. Safety Evaluation cc w/encls: See next page

ML080650219

(*) See previous concurrence OFFICE NRR/LPL4/PM NRR/LPL4/LA IOLB/BC SRXB/BC AADB/BC(A)

SCVB/BC SNPB/BC OGC - NLO w/edits NRR/LPL4/BC NAME JDonohew (*)

JBurkhardt NSalgado (*)

GCranston (*)

MHart (*)

RDennig (*)

AMendiola (*)

MSmith (*)

THiltz Section All All 3.4, 4.7 3.3, 4.5 3.1, 4.3 3.2, 4.4 3.3, 4.6 All All DATE 3/19/08 3/21/08 02/07/08 01/24/08 10/26/07 10/24/07 03/06/08 03/18/08 3/20/08

Wolf Creek Generating Station (2/2006) cc:

Jay Silberg, Esq.

Pillsbury Winthrop Shaw Pittman LLP 2300 N Street, NW Washington, D.C. 20037 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 Senior Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 311 Burlington, KS 66839 Chief Engineer, Utilities Division Kansas Corporation Commission 1500 SW Arrowhead Road Topeka, KS 66604-4027 Office of the Governor State of Kansas Topeka, KS 66612 Attorney General 120 S.W. 10th Avenue, 2nd Floor Topeka, KS 66612-1597 County Clerk Coffey County Courthouse 110 South 6th Street Burlington, KS 66839 Chief, Radiation and Asbestos Control Section Kansas Department of Health and Environment Bureau of Air and Radiation 1000 SW Jackson, Suite 310 Topeka, KS 66612-1366 Vice President Operations/Plant Manager Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 Supervisor Licensing Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 U.S. Nuclear Regulatory Commission Resident Inspectors Office/Callaway Plant 8201 NRC Road Steedman, MO 65077-1032

WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 176 License No. NPF-42

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Wolf Creek Generating Station (the facility)

Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated March 14, 2007, as supplemented by letters dated September 12, October 16, and December 14 (two letters), 2007, and January 18, 2008, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended to authorize the replacement of the main steam isolation valves (MSIVs) and main feedwater isolation valves (MFIVs) and the use of Figures B 3.7.2-1 (MSIVs) and 3.7.3-1 (MFIVs) as the limiting closure times for these valves to demonstrate that these valves meet the limiting conditions for operation with respect to their limiting closure time, as described in the licensee's application dated March 14, 2007, as supplemented by letters dated September 12, October 16, and December 14 (two letters), 2007, and January 18, 2008 (ET 07-0004, ET 07-0040, ET 07-0050, ET 07-0060, ET 07-0062, and ET 08-0005, respectively).

3.

The license amendment is effective as of its date of issuance and shall be implemented before entry into Mode 3 in the restart from Refueling Outage 16, which is to be conducted in the spring of 2008.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Thomas G. Hiltz, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Date of Issuance: March 21, 2008

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 176 TO FACILITY OPERATING LICENSE NO. NPF-42 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482

1.0 INTRODUCTION

By application dated March 14, 2007, as supplemented by letters dated September 12, October 16, and December 14 (two letters), 2007, and January 18, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML070800193, ML072620128, ML072970684, ML073550325, ML080230167, and ML080250061, respectively)

(References 1 through 4), Wolf Creek Nuclear Operating Corporation (the licensee) proposed to replace the existing main steam isolation valves (MSIVs) and main feedwater isolation valves (MFIVs) with new valve actuators to perform the safety function of isolating and closing the main steam lines and main feedwater lines during accidents. In the upcoming spring 2008 refueling outage, the licensee has proposed to replace the MSIVs and MFIVs with new valves.

In the application dated March 14, 2007, the licensee requested the following changes to the Technical Specifications (TSs, Appendix A to Facility Operating License No. NPF-42) for the Wolf Creek Generating Station (WCGS):

1.

Changes to TS 3.3.2, Engineered Safety Features Actuation System (ESFAS)

Instrumentation, to separate the main steam and feedwater isolation system (MSFIS) from the solid state protection system (SSPS) in ESFAS Functions 4.b and 5.a of Table 3.3.2-1.

2.

Changes to TS 3.7.3, Main Feedwater Isolation Valves (MFIVs), for the addition of the main feedwater regulating valves (MFRVs), and associated MFRV bypass valves to TS 3.7.3.

3.

Changes to Surveillance Requirements (SRs) 3.7.2.1 and 3.7.3.1 to adopt the Nuclear Regulatory Commission (NRC)-approved Technical Specification Task Force Traveler 491, Revision 2 (TSTF-491R2), Removal of Main Steam and Main Feedwater Isolation Times.

4.

Changes to the TS Table of Contents to change the title of TS 3.7.3 (see No. 2 above) and to re-number the TS pages for TS 3.7.4 through TS 3.7.18.

None of the above changes to the TSs are being addressed in this amendment. The above TS change Nos. 2 and 4 have not been addressed in an amendment and will be the subject of a future letter before the spring 2008 refueling outage is completed. The above TS change Nos. 1 and 3 were addressed in the following amendments to the WCGS operating license:

A.

The changes in No. 1 above were issued in Amendment No. 175 by letter dated March 3, 2008, except for the changes to footnote j and addition of footnote k in TS Table 3.3.2-1.

B.

The changes in No. 3 above were issued in Amendment No. 174 by letter dated August 28, 2007.

In the application, the licensee also proposed to modify the MSFIS controls. Therefore, the remaining proposed changes in the application, which have not been addressed by the NRC staff, are the changes to add the MFRV and MFRV bypass valves to TS 3.7.3 and revise the TS Table of Contents, and the changes to modify the MSFIS controls. These two sets of changes will be addressed in future letters to the licensee.

In Attachments IV and V to the application, the licensee identified (1) changes to the TS Bases for the proposed amendment and (2) the list of regulatory commitments. The identified changes to the TS Bases come under TS 5.5.14, Technical Specification (TS) Bases Control Program, which requires that [l]icensees may make changes to the [TS] Bases without prior NRC approval provided the changes do no require either of the following: 1. a change to the TS incorporated in the license; or 2. a change to the updated FSAR [Final Safety Analysis Report]

that requires approval pursuant to 10 CFR [Title 10 of the Code of Federal Regulations] 50.59.

In the identified changes to the TS 3.7 Bases are the two figures B 3.7.2-1, "MSIV Isolation Time vs. Steam Generator Pressure," and B 3.7.3-1, "MFIV Isolation Time vs. Steam Generator Pressure." Although these figures were present for information only, the licensee in effect is proposing to replace the current valve closure limit of 5 seconds by the proposed figures of valve closure time as a function of steam generator (SG) pressure. These figures give the limiting MSIV/MFIV closure time as a function of the SG pressure since the actuators to close these valves are system-medium actuators and the closure time decreases as the system pressure increases. In replacing the current MSIVs and MFIVs with valves having system-medium actuators, the licensee will be testing these new valves at the SG pressure existing at the time of the test and using the proposed Figures B 3.7.2-1 (MSIVs) and B 3.7.3-1 (MFIVs) to show that the valves meet the requirement of verification that the valves meet their safety isolation closure limit in SRs 3.7.2.1 and 3.7.3.1. The SRs do not have to be conducted at the SG pressure in the accidents and the maximum 15-second valve closure time demonstrated. Also, in approving this amendment, the licensee, in accordance with 10 CFR 50.71(e), is required to update its final safety analysis report to reflect all changes to its facility in support of approved license amendments. In the case of this amendment, the licensee would be required to update the Updated Safety Analysis Report (USAR) for WCGS to describe the new MSIVs and MFIVs and the method by which the valves are tested to demonstrate valve operability, which in this case is the use of TS Bases Figure B 3.7.2-1 and Figure B 3.7.3-1. These changes to the MSIV and MFIV closure time limit are of the type of change to the license that, pursuant to 10 CFR 50.59, requires NRC review and approval, and, thus, NRC approval of these changes is required.

The supplemental letters dated September 12, October 16, and December 14 (two letters),

2007, and January 18, 2008, provided additional information that clarified the proposed changes in the application. The letters did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination published in the Federal Register on June 19, 2007 (72 FR 33785).

In the application and supplemental letters, the licensee submitted proprietary topical reports addressing the MSIV/MFIV valve replacement in the letters dated October 16, 2007, and January 18, 2008. These proprietary reports addressed the small-break loss-of-coolant accident (SBLOCA) analysis presented in Section 4.6 of Attachment II to the application. The withholding of the proprietary information in these two topical reports from the public was approved pursuant to 10 CFR 2.390(b)(5) and Section 103(b) of the Atomic Energy Act of 1954, as amended, in the two letters from NRC to the licensee dated February 29, 2008 (ADAMS Accession Nos. ML080450004 and ML080450010, respectively).

2.0 BACKGROUND

The licensee described the replacement of the MSIVs and MFIVs in Attachment I to its application. There are four MSIVs with one installed in each of the four main steam lines outside containment and downstream of the main steam safety valve for that line. There are four MFIVs with one installed in each of the four main feedwater lines outside containment.

The licensee stated in its application that the hydraulic actuators for the existing MSIVs and MFIVs have a poor maintenance history, being very complex and containing numerous O-rings under high pressure, of leaks that have resulted in loss-of-power-generation capacity, delays in starting up the plant from refueling outages, and increased exposure to personnel of hazardous material. The replacement of the existing valve actuators with system-medium actuators will result in a system pressure dependent valve closure time of between 6 seconds to 33 seconds for a steam pressure ranging from 1100 pounds per square inch gauge (psig) to 100 psig. The replacement of the actuators and the increase in the MSIV/MFIV closure times has been evaluated by the licensee with respect to the impact of the valve closure times on the germane accident analyses in Section 4.0 of Attachment I to the application dated March 14, 2007. In replacing the valve actuators, the entire valve is being replaced.

In Section 4.0 of Attachment I to the application, the licensee stated that the following design-basis accidents (DBAs) at WCGS either include the MSIVs or the MFIVs or both to ensure that the applicable acceptance criteria are met. The licensee stated that the impact of the change to the MSIVs and MFIVs were evaluated for the following DBAs and addressed in Section 4.0 of the application:

1.

Feedwater system malfunctions that result in an increase in feedwater flow (USAR Section 15.1.2) for the MFIVs

2.

Loss of non-emergency alternating current (AC) power to the station auxiliaries (USAR Section 15.2.6) for the MFIVs

3.

Loss of normal main feedwater flow (USAR Section 15.2.7) for the MFIVs

4.

Steam system piping failure, or main steam line break (MSLB) (USAR Section 15.1.5) for the MSIVs

5.

Feedwater system pipe break, or main feedwater line break (MFLB) (USAR Section 15.2.8) for the MFIVs

6.

Steam generator tube rupture (SGTR) (USAR Section 15.6.3) for both the MSIVs and MFIVs

7.

LOCAs resulting from postulated piping breaks within the reactor coolant pressure boundary (RCPB) (USAR Section 15.6.5) for both the MSIVs and MFIVs

8.

Mass and energy release analysis for postulated secondary pipe rupture inside containment (USAR Section 6.2.1.4)

9.

MSLB mass and energy release analysis outside containment (Chapter 3, Appendix 3B)

For the first three DBAs listed above, the licensee stated that only the MFIV affected the accidents and the effect was expected to not be significant so that a qualitative evaluation was performed. For the DBAs above Nos. 3 to 7, both the MSIV and MFIV affected the accidents and the effect was expected to be significant so that a reanalysis was performed and presented by the licensee. The licensee addressed the first seven DBAs in Sections 4.1 through 4.6, respectively, in Attachment I to its application. Section 4.2 addresses both the (1) loss of non-emergency AC power to the station auxiliaries (USAR Section 15.2.6), and (2) loss of normal main feedwater flow (USAR Section 15.2.7) for the MFIVs.

For each of the above DBAs, the licensee assessed the effect of the valve replacements by assessing the effect of extending the valve closure time of 5 seconds (the current licensing basis valve closure time) to 15 seconds for the new valves. Although the two proposed valve closure time figures, as a function of SG pressure, for the MSIVs (TS Bases Figure 3.7.2-1) and the MFIVs (TS Bases Figure 3.7.3-1) show a range of closure times from about 8 to 40 seconds for SG pressures of about 50 to 1200 psig, the licensee stated that a valve closure time of 15 seconds was assumed for the MSIVs and MFIVs for the above accident analyses involving these valves because the 15 seconds closure time is conservative and bounding for all the accidents.

The proposed figures for the TS 3.7 Bases are valve closure time as a function of SG pressure.

Because the valve actuators are system-medium actuators where the system pressure (steam line or feedwater line pressure) assists the valve closure, the closure time will be different depending on the system pressure, which can be shown to be a function of the SG pressure.

The figures will be used to demonstrate, per SRs 3.7.2.1 and 3.7.3.1, that the valves are operable with respect to the valve closure time at a lower SG pressure instead of having to conduct the test at a SG pressure consistent with an accident. The WCGS accident analyses are based on a 15-second valve closure time.

The basis for the licensee's statement that a valve closure of 15 seconds is conservative is that the SG pressures expected for all the accidents will be above 400 psig when the valves are approaching the closed position. A valve closure time of 15 seconds by the proposed figures

are for SG pressure of above 240 psig for MSIVs and 260 psig for MFIVs. The 400 psig SG pressure gives valve closure times of less than 13 seconds for the MSIVs and MFIVs. In the licensee's response letter dated December 14, 2007 (page 16 of 32), to the NRCs request for additional information (RAI), Question 2, the licensee stated that the system pressures are significantly higher than 400 pounds per square inch absolute (psia) at the times the MSIVs and MFIVs are required to close; therefore, the closure times for these valves by the proposed figures will be lower than the assumed 15 seconds.

For the above DBAs, the licensee described the effect of the extending the valve closure time from the current licensing basis time of 5.0 seconds to the new 15 seconds for the new valves, and concluded as follows in Sections 4.1 through 4.12 in Attachment I to the application:

1.

In Section 4.1 of application, for the feedwater system malfunctions that result in an increase in feedwater flow, USAR Section 15.1.2: Based on the reanalysis of the accident, the impact on the minimum departure from nucleate boiling ratio (DNBR) for this accident is negligible due to the longer valve closure time, and remains above the current licensing basis limit.

2.

In Section 4.2. for the loss of non-emergency AC power to the station auxiliaries and loss of normal feedwater flow, USAR Sections 15.2.6 and 15.2.7: The current licensing basis of the 392 second delay time, which accounts for the 60-second delay for diesel generator and auxiliary feedwater (AFW) pump start, and the filling of the associated feedwater piping, accommodates the increased MFIV closure of 15 seconds. Thus, there is no effect on the results of the two analyses and the unchanged acceptance criteria continues to be met.

3.

In Section 4.3, for the steam system line failure, USAR Section 15.1.5: A departure from nucleate boiling (DNB) analysis was performed for this DBA and the analysis shows that the minimum DNBR remains above the current licensing basis limit and, thus, the DNB design basis is met for this accident.

4.

In Section 4.4, for the feedwater system pipe break, USAR Section 15.2.8: The results of the analysis of the feedwater line rupture demonstrate that all current licensing basis criteria, which are not being changed, are satisfied for the extended valve closure time.

5.

In Section 4.5, for the SGTR, USAR Section 15.6.3: Although the analysis of the SGTR with a postulated failure of the AFW discharge flow control valve indicated an overfilling of the SG, the accident dose analysis show that the doses remain well within the limiting valves in 10 CFR Part 100, and a SGTR will cause no subsequent to the reactor coolant system (RCS) or the reactor core.

6.

In Section 4.6, for the SBLOCA, USAR Section 15.6.5: The results of the analysis show that the current licensing basis acceptance criteria, which is not being changed, will be met with the extended valve closure time. The peak cladding temperature, maximum cladding oxidation, maximum hydrogen generation, and maintaining a coolable geometry and long-term cooling will be met.

7.

In Sections 4.7 and 4.8, for the MSLB and energy releases to containment, USAR Section 6.2.1.4: Based on the examination of the integrated mass and energy releases, the total MSLB blowdown mass and energy increases within containment because of the longer closure time of the valves and, thus, a greater amount of water is allowed into the faulted SG and into containment before valve closure. This would increase the calculated peak containment pressure or temperature; however, conservative assumptions concerning the degradation of the fan cooler performance and operator response time to terminate the AFW flow to the faulted SG were revised. With these reductions in conservative assumptions, the plant stays within its current licensing basis, and the existing temperature profiles for equipment qualification remain bounding.

8.

In Sections 4.9 and 4.10, for the MSLB mass and energy releases outside containment, USAR Chapter 3, Appendix 3B: The mass and energy releases from MSLB cases outside containment have been analyzed to assure that the existing environmental qualification envelope is maintained with the extended valve closure times.

9.

In Section 4.11, for the radiological consequences of a postulated SGTR, USAR Section 15.6.3: The calculated radiological consequences do not exceed (1) the exposure guidelines set forth in 10 CFR Part 100 for the SGTR with an assumed pre-accident radioiodine spike and (2) 10 percent of these exposure guidelines for the SGTR with an equilibrium in combination with an assumed accident-initiated radioiodine spike.

10.

In Section 4.12, for the full-power core response to steam line rupture: This accident is not analyzed in the USAR. The licensee stated that it analyzed this event to determine the impact on the plant. The licensee stated that the results were that the DNBR safety limit is met and, therefore, there is no melting at the fuel centerline. The licensee pointed out that the MSIV closure time has no impact on this event.

11.

In Section 4.12, for the MSLB at power coincident with rod withdrawal: The licensee stated that this event has been analyzed in response the NRC Information Notice No. 79-22, "Qualification of Control System." The licensee stated that the increase in MSIV and MFIV closure time would not perceptibly affect the calculated limiting DNBR as the actuations of the steam line and feedwater line isolation occur after the limiting DNBR is reached during rod motion from the reactor trip.

3.0 REGULATORY EVALUATION

In Section 50.36 of 10 CFR, the Commission established its regulatory requirements related to the content of the TSs. Pursuant to 10 CFR 50.36, TSs are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs);

(3) SRs; (4) design features; and (5) administrative controls. The rule does not specify the particular requirements to be included in a plants TSs.

As stated in 10 CFR 50.36(d)(2)(i), LCOs "are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications... The remedial actions in the TSs are specified in terms of LCO conditions, required actions, and completion times (CTs), or allowed outage times, to complete the required actions. The conditions and required actions specified in the TSs must be acceptable remedial actions for the LCO not being met, and the CTs must be a reasonable time for completing the required actions while maintaining the safe operation of the plant.

As required by 10 CFR 50.36(d)(3), SRs are the requirements related to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

As explained in Section 4.2 of this safety evaluation (SE), the technical areas that were reviewed by the NRC staff in evaluating the MSIV/MFIV valve replacement and the longer valve closure time for the new valves are the following: accident consequences analysis, containment integrity analysis, non-LOCA transient analysis, LOCA transient analysis, and human factors analysis.

The following are the regulatory requirements within these five technical area reviews:

3.1 Accident Consequences Review This SE addresses the impact of the proposed changes on previously analyzed DBA radiological consequences and the acceptability of the revised analysis results. The NRC staff used the regulatory guidance provided in NUREG-0800, Standard Review Plan (SRP) for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 15.6.3, "Radiological Consequences of Steam Generator Tube Failure (PWR)," Revision 2, to evaluate the licensees SGTR accident analysis. In addition, the staff used the guidance from Regulatory Guide (RG) 1.195, Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light Water Nuclear Power Reactors, dated May 2003.

The applicable dose criteria, for the evaluation of DBAs, depends on the source term incorporated in the dose consequence analyses. The licensee uses the source term defined in Technical Information Document (TID) 14844, AEC, 1962, Calculation of Distance Factors for Power and Test Reactors Sites, for dose consequence analyses. The dose acceptance guidelines are specified in 10 CFR Part 100 in terms of the maximum dose to the whole body and the thyroid that an individual at the exclusion area boundary (EAB) can receive for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an accident, and at the low population zone (LPZ) outer boundary for the duration of the radiological release. As specified in 10 CFR 100.11, these guidelines are 25 roentgen equivalent man (rem) total whole body dose and 300 rem thyroid dose from iodine exposure. The accident dose guideline criteria in 10 CFR 100.11 are supplemented by accident-specific dose acceptance criteria in Table 4 of RG 1.195.

For control room dose consequence analyses that use the TID 14844 source term, the regulatory requirement for which the staff bases its acceptance is General Design Criterion (GDC) 19 of Appendix A to 10 CFR Part 50, Control Room. GDC 19 requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. NUREG-0800

SRP Section 6.4, Control Room Habitability System, Revision 2, July 1981, provides guidelines defining the dose equivalency of 5 rem whole body as 30 rem for both the thyroid and skin dose.

For licensees adopting the guidance from RG 1.196, Control Room Habitability at Light Water Nuclear Power Reactors, May 2003, Section C.4.5 of RG 1.195, May 2003, states that in lieu of the dose equivalency guidelines from Section 6.4 of NUREG-0800, the 10 CFR 20.1201 annual organ dose limit of 50 rem can be used for both the thyroid and skin dose equivalent of 5 rem whole body. The current licensing basis control room dose limits at WCGS are 5 rem whole body and 30 rem thyroid.

3.2 Containment Integrity Review The regulatory requirements and the guidance which the staff used to review the containment functional design are based on the GDC in Appendix A, 10 CFR Part 50, and are as follows:

1.

GDC 16, "Containment Design," requires that the containment and associated systems be designed to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment, and to assure that the containment design conditions important to safety are not exceeded as long as postulated accident conditions require. The design pressure of the WCGS containment is 60 psig and a design temperature is 320 degrees Fahrenheit (°F)

(USAR Table 6.2.1-2). For WCGS, the associated systems referred to in GDC 16 are the containment spray system and the containment fan coolers.

2.

GDC 38, "Containment heat removal," requires that these systems rapidly reduce the containment pressure and temperature following any LOCA and maintain them at acceptably low levels. This is verified by the accident analyses.

3.

GDC 50, "Containment design basis," requires that the containment be designed so that the containment structure can accommodate, without exceeding its design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA.

The NRC staff also used the following sections of NUREG-0800, Standard Review Plan, for this review: Section 6.2.1, Containment Functional Design, Section 6.2.1.1.A, PWR

[Pressurized-Water Reactor] Dry Containments, Including Subatmospheric Containments, and Section 6.2.2, Containment Heat Removal Systems.

3.3 Non-LOCA and LOCA Transient Review General Design Criterion (GDC) 10 in Appendix A to 10 CFR Part 50 specifies that specified acceptable fuel design limits are not exceeded during any anticipated operational occurrences (AOOs).

GDC 15 specifies that the design conditions of the RCPB are not exceeded during any AOOs.

The regulation 10 CFR 100.11 specifies dose acceptance guidelines in the mitigation of the radiological consequences of an accident.

The regulation 10 CFR 50.46 specifies that the performance of an emergency core cooling system (ECCS) shall be calculated in accordance with an acceptable evaluation model for a limiting LOCA to meet the following acceptance criteria: the peak cladding temperature does not exceed 2200 °F; the maximum cladding oxidation does not exceed 17 percent of the total cladding thickness; the maximum metal-water reaction does not exceed 1 percent of the total amount of metal in the core; the core geometry remains amenable to allow long-term cooling of the core; and decay heat can be removed for an extended period of time with clad temperatures remaining at acceptable values.

The NRC staffs evaluation of the design-basis-event analyses is to assure continued compliance of WCGS with the requirements of GDCs 10 and 15, 10 CFR 50.46 and 10 CFR 100.11.

3.4 Human Factors Review The NRC staffs review covered changes to operator actions, human-system interfaces, procedures, and training needed for the proposed TS modifications. The NRCs acceptance criteria for human factors are based on the following documents: GDC-19, 10 CFR 50.120, 10 CFR Part 55, NUREG-1764, Guidance for the Review of Changes to Human Actions, and the guidance in Generic Letter (GL) 82-33. Specific review criteria are contained in SRP Sections 13.2.1, 13.2.2, and 13.5.2.1, and Chapter 18.0.

4.0 TECHNICAL EVALUATION

4.1 Licensee-Identified Changes to the WCGS License In its application, the licensee has proposed to replace the current valve closure limit of 5 seconds for the MSIVs and MFIVs by the proposed Figure B 3.7.2-1 and Figure B 3.7.3-1, where the valve closure time is a function of SG pressure. These figures provide curves to show that the limiting valve closure time decreases from (1) about 40 seconds at 50 psig to 10 seconds at 1200 psig for the MSIVs and (2) about 35 seconds at 50 psig to 8 seconds at 1200 psig for the MFIVs. The results of accident analyses presented in Attachment I to the application address the longer time for the MSIVs and MFIVs to close and isolate the steam lines and feedwater lines, respectively.

As addressed in Section 2.0, "Background," the licensee stated that a maximum valve closure time of 15 seconds was assumed for the MSIVs and MFIVs for the accident analyses involving these valves because the 15 seconds closure time is conservative and bounding for all the accidents. The licensee stated that the basis for this statement is that a valve closure of 15 seconds is conservative in that the SG pressures and, therefore, the system pressure at the actuators for these valves expected for all the accidents will be above 400 psig when the valves are approaching the closed position. The proposed Figure B 3.7.2-1 (MSIVs) and Figure 3.7.3-1 (MFIVs) show the valve closure time as a function of SG pressure for the valves to demonstrate the valves meet the requirement in TS 3.7.2 and 3.7.3 to operability in terms of the valve closure time.

4.2 Areas of Review by NRC Staff The technical areas that were reviewed in evaluating the MSIV/MFIV valve replacement and the longer valve closure time for the new valves are the following:

1.

Accident consequences analysis

2.

Containment integrity analysis

3.

Non-LOCA transient analysis

4.

LOCA transient analysis

5.

Human factors analysis These technical areas are addressed in the following Sections 4.3 through 4.7, respectively, of this SE.

4.3 Accident Consequences Analysis 4.3.1 Radiological Consequences of Design Basis Accidents Based on its application and supplemental letters, the licensee used the radiological consequence computer code, ARADTRAD: Simplified Model for RADionuclide Transport and Removal and Dose Estimation,@ Version 3.03, to evaluate the integrated whole body and thyroid dose at the EAB for first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following the onset of the accident and the integrated whole body and thyroid doses at the outer boundary of the LPZ for the duration of the accident. The NRC sponsored Sandia National Laboratories to develop the RADTRAD code. RADTRAD, Version 3.03, as described in NUREG/CR-6604, is currently used by the staff for the performance of confirmatory dose consequence analyses. The code estimates transport and removal of radionuclides and radiological consequence doses at selected receptors. The NRC staff performs independent confirmatory dose evaluations using the RADTRAD computer code.

The results of the evaluations performed by the licensee, as well as the applicable dose acceptance criteria from RG 1.195, Table 4, are shown in Table 1, which is attached to this SE.

The licensee used dose conversion factors (DCFs) from Federal Guidance Reports (FGR) 11 and 12 to determine the dose consequences of the SGTR accident. The use of DCFs from FGR-11 and FGR-12 is in accordance with RG 1.195 guidance and is, therefore, acceptable to the NRC staff.

4.3.1.1 SGTR As described in its application, the licensee re-analyzed the radiological consequences of an SGTR accident to assess the impact of the change to the valve closure characteristics of the MSIVs and MFIVs, resulting from the replacement of the MSIVs, MFIVs, and their associated actuators. This is representing the limiting valve closure times as a function of SG pressure in Figures B 3.7.2-1 (MSIVs) and 3.7.3-1 (MFIVs). In addition, the licensee incorporated revisions to the assumed critical operator action times based upon current operating procedures that

reflect operator performance from simulator exercises. The operator actions are addressed in Section 4.7 of this SE.

The SGTR accident is evaluated based on the assumption of an instantaneous and complete severance of a single SG tube. The postulated break allows primary coolant liquid to leak to the secondary side of the ruptured SG. Integrity of the barrier between the RCS and the main steam system is significant from a radiological release standpoint. The radioactivity from the leaking SG tube mixes with the shell-side water in the ruptured SG. The resulting sharp increase in radioactivity in the secondary system will be detected by radiation monitors that will automatically terminate SG blowdown. A reactor trip occurs automatically as a result of low pressurizer pressure and will automatically trip the turbine. For the SGTR DBA radiological consequence analysis, a loss-of-offsite power (LOOP) is assumed to occur shortly after the trip signal. The assumed coincident LOOP will cause closure of the steam dump valves to protect the condenser. Following a reactor trip and turbine trip, the SG pressure will increase rapidly, resulting in steam and activity releases through the SG safety and relief valves. Because the LOOP renders the main condenser unavailable, the plant is cooled down by releasing steam to the environment. Venting from the ruptured SG will continue until the secondary system pressure is below the SG safety valve setpoint. At this time, the ruptured SG is effectively isolated ending the activity release from the ruptured SG. Core decay heat is removed by venting steam through the atmospheric relief valves (ARVs) of the remaining unaffected SGs until the controlled cooldown is terminated.

Based on its letters, the licensee evaluated two SGTR scenarios in order to ensure that operators can respond to the accident in a timely manner so as to minimize the resulting offsite releases and prevent overfilling the affected steam line. For the overfill SGTR scenario, the licensee assumed a failure in the AFW system requiring operator action to terminate feedwater in order to prevent the affected steam line from filling with water. The specific failure involves the discharge flow control valve on the discharge side of the motor-driven AFW pump feeding the ruptured SG. To maximize flow to the ruptured SG, the licensee assumed that this valve fails in the wide-open position. Failure of this valve coupled with the contribution from the turbine-driven AFW pump has the potential for overfilling the ruptured SG and subsequently, releasing water via a safety valve. The radioactive releases are maximized by assuming the safety valve is stuck-open following water relief, with an effective flow area equal to 5 percent of the total safety valve flow area. The increase in MFIV stroke time, associated with the MSIV/MFIV replacement, affects the analysis for the SGTR accident by introducing additional feedwater into the ruptured SG. As a result, it will take less time to overfill the ruptured SG and the total release to atmosphere will increase. This scenario credits a manual operator action to terminate the AFW from the motor-driven AFW pump to the ruptured SG, by locally closing the failed discharge flow control valve. This manual operator action is assumed to occur 18 minutes after initiation of the safety-injection (SI) signal.

The licensee analyzed a second SGTR scenario with the assumption of a stuck-open ARV.

During the SGTR accident sequence, as pressure rises on the secondary side, the SG ARVs open to release excess secondary pressure. For the stuck-open ARV scenario, ARVs for the unaffected SGs are assumed to close within 7 minutes. The licensee conservatively assumed that the ARV for the ruptured SG remains open for 20 minutes releasing steam and radioactivity, until an operator manually closes the associated block valve. Failure of the ruptured SG's ARV to close maximizes the offsite release by assuming a direct path from the ruptured SG to the atmosphere.

4.3.1.1.1 Source Term If a licensee demonstrates that no or minimal fuel damage is postulated for the limiting event, the activity released should be the maximum coolant activity allowed by TS. Two radioiodine spiking cases are considered. The first case is referred to as a pre-accident iodine spike and assumes that a reactor transient has occurred prior to the postulated SGTR that has raised the primary coolant iodine concentration to the maximum value permitted by the TS for a spiking condition.

For WCGS, the maximum iodine concentration allowed by the TSs as a result of an iodine spike is 60 microcuries per gram (Ci/gm) dose equivalent iodine 131 (DEI).

The second case assumes that the primary system transient associated with the SGTR causes an iodine spike in the primary system. This case is referred to as a concurrent iodine spike.

Initially, the plant is assumed to be operating with the RCS iodine activity at the TS limit for normal operation. For WCGS, the RCS TS limit for normal operation is 1 Ci/gm DEI. The increase in primary coolant iodine concentration for the concurrent iodine spike case is estimated using a spiking model that assumes that as a result of the accident, iodine is released from the fuel rods to the primary coolant at a rate that is 335 times greater than the iodine equilibrium release rate corresponding to the iodine concentration at the TS limit for normal operation. The iodine release rate at equilibrium is equal to the rate at which iodine is lost due to radioactive decay, RCS purification, and RCS leakage. The iodine release rate from the fuel is also referred to as the iodine appearance rate. The concurrent iodine spike is assumed to persist for a period of eight hours.

The licensee=s evaluation indicates that no fuel damage is predicted as a result of an SGTR accident. Therefore, consistent with the current licensing analysis basis and regulatory guidance, the licensee performed the SGTR accident analyses for the pre-accident iodine spike case and the concurrent accident iodine spike case. In accordance with regulatory guidance, the licensee assumed that the activity released from the iodine spiking mixes instantaneously and homogeneously throughout the primary coolant system.

The licensee evaluated the radiological dose contribution from the release of secondary coolant iodine activity at the TS limit of 0.1 Ci/gm DEI.

4.3.1.1.2 Transport RG 1.195, Appendix E, Regulatory Position 2.2, states that A[t]he density used in converting volumetric leak rates (e.g., gpm [gallons per minute] to mass leak rates (e.g., lbm/hr [pounds mass per hour]) should be consistent with the basis of the surveillance tests used to show compliance with leak rate technical specifications. These tests are typically based on cooled liquid. Facility instrumentation used to determine leakage is typically located on lines containing room temperature liquids. In most cases, the density should be assumed to be 1.0 gm/cc

[grams per cubic centimeter] (62.4 lbm/ft3 [pounds mass per cubic foot]).@ The licensee=s leak rate testing results are adjusted so that the allowable leakage corresponds to a density of 62.4 lbm/ft3 and accordingly this density was used to convert the volumetric leak rate to a total mass flow rate due to SG tube leakage in the SGTR dose consequence analysis.

RG 1.195, Appendix E, Regulatory Position 2.3, states that A[t]he primary to secondary leakage should be assumed to continue until the primary system pressure is less than the secondary

system pressure, or until the temperature of the leakage is less than 100 °C [degrees Celsius]

(212 °F). The release of radioactivity from the unaffected steam generators should be assumed to continue until shutdown cooling is in operation and releases from the steam generators have been terminated.@ The licensee assumed a total primary-to-secondary leak rate equal to 1 gpm.

This assumption is conservative with respect to the TS limit of 150 gallons per day which is equivalent to 0.104 gpm. The licensee assumed that all of the 1 gpm primary-to-secondary leakage enters the shell side of the three intact SGs until the residual heat removal (RHR) entry conditions are reached at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. At this point in the accident sequence, steaming is no longer required for cool down and releases from the intact SGs are terminated.

The licensee assumed that the source term resulting from the radionuclides in the primary system coolant, including the contribution from iodine spiking, is transported to the ruptured SG by the break flow. A portion of the break flow is assumed to flash to steam because of the higher enthalpy in the RCS relative to the secondary system. The licensee assumed that the flashed portion of the break flow will ascend through bulk water of the SG, enter the steam space of the ruptured SG, and be immediately available for release to the environment with no credit taken for scrubbing.

The radioactivity in the non-flashed portion of the break flow is assumed to mix uniformly with the SG liquid mass and be released to the environment in direct proportion to the steaming rate and in inverse proportion to the applicable partition coefficient (PC). The licensee=s evaluation of the releases from the steaming of the liquid mass in the SGs credits a PC of 100.

Also, following the applicable NRC regulatory guidance, the licensee assumed that all noble gas radionuclides released from the primary system are released to the environment without reduction or mitigation. Assumptions and data used by the licensee and found acceptable by the NRC staff for use in the reanalysis of the SGTR accident are shown in Table 3, which is attached to this SE.

4.3.1.1.3 Control Room Habitability The current licensing basis for WCGS only includes radiation doses to a control room operator due to a postulated LOCA. This licensing basis is predicated on the results of a study of the radiological consequences in the control room due to various postulated accidents indicating that the LOCA is the limiting case.

In a letter dated September 12, 2007 (Reference 2), the licensee stated that the calculated offsite LPZ and EAB radiological consequences, for both the pre-accident and concurrent iodine spike cases reanalyzed for this license amendment, are exceeded by or similar to the radiological consequences of the current limiting SGTR with forced overfill analysis of record.

The licensee further stated that the integrated break-flow changes due to the MSIV and MFIV valve and actuator replacement were minimal. The licensee provided the staff with a table showing the integrated break flows for the updated analysis compared to the flows used in the analysis of record to substantiate the conclusion that the flow changes due to the MSIV and MFIV valve and actuator replacement are minimal. This comparison table is included as part of the SGTR data and assumptions shown in Table 3, which is attached to this SE.

The steam release to the atmosphere from the intact SGs, until shutdown cooling is in operation and all releases are secured, is the only change that results in significantly higher steam

releases for the revised analysis. In the revised analysis, the steam releases were assumed to continue from the intact SGs for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> until long-term cooling operation was achieved. This is a departure from the current analysis of record in which long-term cooling operation was achieved within four hours. However, since releases from the intact SG are afforded a considerable reduction due to the steaming rate and assumed PC, the resulting contribution to dose from the steaming of the intact SGs is low compared to the contribution from the ruptured SG.

The licensee stated that changes in the use of a bounding flashing fraction in the reanalysis, compared with a variable flashing fraction based upon actual primary and secondary conditions in the analysis of record, tended to increase the calculated radiological consequences. These increases were mitigated by the adoption of dose conversion factors from FGR 11 and 12. The use of DCFs from FGR 11 and 12 is consistent with the current NRC guidance, as stated in RG 1.195, therefore, the NRC staff concludes that these DCFs are acceptable. In addition, the updated analysis incorporates an accident-initiated iodine spiking factor of 335 while the analysis of record used a spiking factor of 500. The use of an accident-initiated spiking factor of 335 is consistent with current NRC guidance, as stated in RG 1.195; therefore, the NRC staff concludes that this spiking factor is acceptable.

The licensee asserts, and the NRC staff agrees, that the calculated offsite radiological consequences in the revised SGTR analysis provide assurance that the control room habitability radiological consequences due to the proposed change would remain bounded by the current large-break LOCA radiological consequence analysis of record.

4.3.2 Atmospheric Dispersion The licensee used atmospheric dispersion factors from Table 15A-2 of the WCGS USAR for the SGTR dose consequence analysis evaluated in this SE. Since no changes were made to the previously approved atmospheric dispersion factors, as shown in Table 2 (which is attached to this SE), the NRC staff finds that the use of these factors in the evaluation of the SGTR accident for this license amendment is acceptable 4.3.3 Conclusion As discussed in Section 4.3 of this SE above, the NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of the SGTR at WCGS. Based on its review, the NRC staff finds that the licensee used analysis methods and assumptions consistent with the conservative regulatory requirements and guidance identified in Section 3.1 of this SE above. The NRC staff also finds with reasonable assurance that the licensee=s estimates of the EAB and LPZ doses will comply with these criteria. The NRC staff further finds reasonable assurance that WCGS as modified by this license amendment, will continue to provide sufficient safety margins with adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameters. Therefore, the staff finds that the proposed license amendment is acceptable with respect to the radiological consequences of the SGTR DBA and, therefore, acceptable for the proposed replacement of the MSIVs and MFIVs.

4.4 Containment Integrity Analysis 4.4.1 Containment Technical Findings The replacement of MSIV and MFIV and their actuators will result in an increase in the valve closing time from the current 5 seconds to 15 seconds, as discussed in Section 2.0 of this SE.

During an MSLB accident inside containment, the increase in the valve closing times results in additional feedwater added to the SGs, resulting in an increased mass and energy released inside the containment. In order to verify these changes will not result in exceeding the containment design pressure and temperature, the licensee has evaluated the effect of the proposed changes on the MSLB mass and energy release analyses and reanalyzed the containment response to a postulated design basis MSLB inside containment.

4.4.2 MSLB Mass and Energy Release Analysis The licensee analyzed mass and energy release for the MSLB accident inside the containment using NRC-approved methodology documented in Reference 10. The analysis was done for the same 16 cases listed in USAR Table 6.2.1-56, but with power level percentage scaled to the re-rated power of 3579 megaWatts thermal (MWt). The licensed power level for WCGS is 3565 MWt. The analysis was done using the LOFTRAN code, which was also used for the current licensing-basis MSLB mass and energy release analysis inside the containment. The licensee listed 15 assumptions out of which 14 were conservative from the standpoint of maximizing mass and energy release in the containment following an MSLB accident, and acceptable to the NRC staff. The remaining assumption made by the licensee is the 15-second maximum closing for the MSIVs and MFIVs at a SG pressure of 400 psig, which is the change in the valves being replaced in this amendment. The NRC staff had an RAI regarding this assumption from the licensee. By letter dated September 12, 2007, the licensee provided responses that are addressed below.

(a)

The NRC staff requested design information for the MSIVs and MFIVs showing their maximum closing time is 15 seconds at a steam generator pressure of 400 psig. The NRC staff also requested a description of the method of validating the closing time. The licensee stated that the design specifications for MSIVs and MFIVs require the closing time for these valves to be no greater than 5 seconds from the receipt of the signal under the most adverse flow conditions, which bounds the closing time of 15 seconds at 400 psig used in accident analyses. The licensee also stated that the closing time requirement for each actuator was validated by its vendor by performing functional testing, showing that the actuators are capable of closing the valve within the required time.

(b)

The NRC staff requested the licensee to provide the bases for expecting the SG pressure to be higher than 400 psig during any accident condition when the closing of MSIVs and MFIVs are required. The licensee responded that the transient analyses of all affected DBAs showed the SG pressure to be higher than 400 psig when the MSIVs and MFIVs approached their closed positions.

The licensee provided results of SG pressure at the isolation times of feedwater and steam lines for all 16 cases analyzed. This is also addressed in Section 2.0 of this SE.

(c)

The NRC staff requested the licensee to discuss the assurance that the replacement MSIVs and MFIVs closing time would not degrade below that needed for accident analyses in the time interval between the times to test the valves within the intervals specified within the in-service test program. There is no proposed change to the current surveillance test interval for testing the MSIV/MFIV valve closure time in SRs 3.7.2.1 (MSIVs) and 3.7.3.1 (MFIVs) in this amendment. The licensee stated that, for the replacement MSIVs and MFIVs, the valve type is the same as the existing valves and there are fewer active components in the new actuators than in the existing ones. The licensee stated that any unidentified degradation of valve performance due to its body is unlikely, and the degradation of the possibility actuator performance is reduced because it has fewer active components. The current test interval should still be adequate for the new valves.

(d)

The NRC staff requested the licensee to provide results of an evaluation confirming that in a faulted SG loop, the mass and energy release inside the containment with the MFIV closing bounds the mass and energy release in case the MFIV fails to close and the feedwater isolation is achieved by the MFRV and MFRV bypass valves, which the licensee has proposed to be a backup to the MFIV in its application. In its response, the licensee stated that this case will result in higher mass and energy input in the containment and was calculated to be about 4.5 percent higher than the case in which the MFIV does not fail to close. The higher mass and energy is attributed to additional volume of water in the feedwater line between the MFRV and the MFIV. The licensee included the effects of additional volume of feedwater trapped between the SG and the MFRV in the analysis. The results presented in Section 4.8 of Attachment I to its application, Table 4.8-2, show that the resulting containment temperature and pressure do not challenge the containment design limits. As stated in Section 1.0 of this SE, the proposed changes to add the MFRV and MFRV bypass valves to TS 3.7.3 will be addressed in a future letter to the licensee.

(e)

The NRC staff requested the licensee to provide closing time data for the MFRV and MFRV bypass valves to address the assurance that the valve closing time would not degrade below that needed for accident analyses in the time interval between the times to test the valves. In its response, the licensee stated the MFRV and its bypass valve are not now included in the in-service testing program; however, the licensee has proposed in its TS changes to TS 3.7.3 on the MFIVs to add the MFRVs and MFRV bypass valves to its in-service testing program. During the refueling outage of fall of 1996, after replacing valve positioners, closing time was measured to be 3.5 to 3.8 seconds. The licensee stated that with good reliability control components and the valve live-load stem packing adjusted to the manufacturers specified torque, it is expected that the performance of MFIV and its bypass valve will not degrade during the operating cycle and will perform its intended function.

The NRC staff reviewed the above responses by the licensee to the NRC staff's questions and concludes that the licensee's responses are acceptable.

The results of integrated mass and energy release from the licensee show higher values than the current licensing basis analysis results given in the USAR. For the current containment temperature limiting case which is double-ended break at 50 percent power, the mass and energy release in USAR Table 6.2.1-62 is 550 x 103 lbm and 669.9 x 106 Btu (British thermal unit), respectively, whereas the revised analysis for the same case gives mass and energy release values of 614.4 x 103 lbm and 731.3 x 106 Btu, respectively. The increase is mainly attributed to the longer closing times of the MSIVs and MFIVs (because of the replacement valves) used in the analyses, and a higher initial SG fluid mass.

4.4.3 Containment Response to a Postulated MSLB Accident The current containment analysis model for the MSLB is based on the CONTEMPT code. For the revised analyses, the licensee used Generation of Thermal-Hydraulic Information for Containment (GOTHIC) code version 7.2(a) to model the containment for evaluating its pressure and temperature response. The NRC staff has approved the use of this version of GOTHIC code for containment analysis. The licensee made assumptions which are consistent with those in the current licensing basis except for the fan cooler heat-removal capacity degradation, for which the licensee assumed its heat-removal capacity degraded by 20 percent, whereas the current licensing basis assumed a degradation ranging from 32 percent to 95 percent. In an RAI, the NRC staff requested the basis for assuming 20 percent degradation of the fan cooler heat-removal capability, the value of heat-removal capability used to determine that the fan cooler is operable, and the uncertainty in determining the heat-removal capability.

By a response letter dated September 12, 2007, the licensee stated that during the last surveillance test based on Generic Letter (GL) 89-13, Service Water System Problems Affecting Safety-Related Equipment, the fan cooler heat-removal capability was determined to be 77,569,693 Btu per hour. After adjusting for a test uncertainty of 14,857,539 Btu per hour calculated per the surveillance procedure, the net heat-removal rate was 62,712,154 Btu per hour. The minimum heat-removal capability needed at accident conditions after allowing for 20 percent degradation is 15,701.33 Btu per second (56,524,788 Btu per hour) which is bounded by the surveillance test result value of 62,712,154 Btu per hour. Although the fan cooler is not required to be tested by the TSs, the licensee also stated that the fan coolers successfully passed the previous 18-month frequency surveillance test per GL 89-13. The NRC staff considers the licensees response acceptable.

The licensees presented containment analysis results for a postulated MSLB accident based on the GOTHIC model for all 16 cases. The limiting case for peak pressure was 0.40-square-foot split with MSIV failure at zero percent power, which gave the highest peak containment pressure of 51.5 psig at 1202 seconds from the beginning of the transient. The limiting case for peak temperature was for a full double-ended break at 102 percent power (i.e., re-rated power 3579 MWt), which gave the highest peak containment atmosphere temperature of 360 °F at 18 seconds from the beginning of the transient. Although the containment design temperature is stated to be 320 °F, the licensee stated that this design temperature is the temperature of the safety-related equipment and instrumentation inside containment and not the maximum temperature allowed for the containment atmosphere. The calculated maximum containment atmosphere temperature may exceed this containment design temperature for a short time, which is the case for this MSLB accident. However, the licensee pointed out that this containment temperature is less than the 386.5 °F for the maximum containment atmosphere temperature in the current analysis of record (AOR) for the accident. Because there is sufficient

margin with respect to the containment design pressure of 60 psig and the design temperature of 320 °F for the safety-related equipment and instrumentation inside containment, the NRC staff concludes that the results of the containment analysis are acceptable.

4.4.4 MSLB Mass and Energy Release Outside Containment In its letters, the licensee states that the current licensing basis mass and energy release for the MSLB outside the containment used a generic analysis performed in response to Notice No. 84-90 which used non-conservative input assumptions for the WCGS-specific MSLB mass and energy release. The licensees concern was that the non-conservative assumptions could have resulted in lower compartment temperature thus leading to a non-conservative equipment qualification (EQ) temperature. The licensee identified two of these non-conservative assumptions in the generic analysis as (a) the AFW flow rate used was higher than the WCGS-specific value, and (b) the shutdown margin used was higher than WCGS-specific value.

The licensee recalculated mass and energy release outside the containment using WCGS-specific input values in addition to the higher value of MSIV closing time due to the proposed valve and actuator replacement, using the same methodology as for the current licensing basis.

The analysis was performed by the licensee for a subset of current licensing basis cases having break sizes of 4.6 square feet, 1.0 square feet, 0.7 square feet, and 0.5 square feet at 102 percent power. The results of mass and energy release in this analysis were used as input for calculating the EQ temperature for equipment in the main steam tunnel. The licensee used GOTHIC code to perform the main steam tunnel compartment temperature analysis using the mass and energy input of the four cases of MSLB described in Section 4.2 above. The maximum peak temperature was calculated to be 436 °F at 84 seconds from the initiation of the event, which occurred for break size of 4.6 square feet. This temperature is bounded by the maximum temperature of 469 °F in the EQ temperature profiles given in USAR Figures 3B-6a through 3B-6d. Therefore, the proposed valve replacement does not change EQ temperature profiles for the plant.

4.4.5 Conclusions The licensee proposes to (a) replace MSIVs and MFIVs and their actuators with similar valves but with system-medium actuators with increased stroke time which impacts the mass and energy release both inside and outside the containment in the MSLB accident. Based on its evaluation in Sections 4.4.2 through 4.4.4 of this SE above, the NRC staff determined that the proposed valve changes meet the requirements of (a) GDC 16 because the licensee showed that the primary containment design conditions important to safety are not exceeded from the values currently reported in the USAR following a postulated DBA, (b) GDC 38 because the licensee showed that the containment heat-removal system would maintain its pressure and temperature below their design limits following DBA, and (c) GDC 50 because the licensee showed that the primary containment heat-removal system is designed so that the primary containment structure and its internal compartments can accommodate without exceeding the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from the design-basis accident. Based on this, the NRC staff concludes that the proposed valve replacement is acceptable with respect to the effect on containment integrity.

4.5 Non-LOCA Transient Analysis In Attachment I to its application, the licensee evaluated the results of the analyses for transients and accidents in Chapter 15 of the USAR and provided a rationale for those events that were not affected by the proposed valves characteristics of the replacement MSIVs and MFIVs. For those events that were affected by the valve closure characteristics, the licensee analyzed the events and provided the results of the analyses for the NRC staff to review. The following evaluation by the NRC staff addresses (1) the transients and accidents that were not affected by the valve characteristics in Section 4.3.1 of this SE, (2) the transients and accidents that were affected by the valve characteristics, but bounded by the AOR in Section 3.3.2, and (3) reanalyzed transients and accidents that were affected by the valve characteristics in Section 4.3.3 of this SE.

4.5.1 Transients and Accidents That Were Not Affected by the Isolation Valve Characteristics For those events that did not model or credit the MSIV and MFIV closure characteristics, the NRC staff agrees with the licensee that the AOR for those events remains unchanged. The events are the following:

1.

Feedwater System Malfunctions that Result in a Decrease in Feedwater Temperature

2.

Excessive Increase in Secondary Steam Flow

3.

Loss of External Electrical Load

4.

Turbine Trip

5.

Loss of Condenser Vacuum and Other Events Resulting in Turbine Trip

6.

Partial Loss of Forced Reactor Coolant Flow

7.

Complete Loss of Forced Reactor Coolant Flow

8.

Reactor Coolant Pump Shaft Seizure (Locked Rotor)

9.

Reactor Coolant Pump Shaft Break

10.

Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal from a Subcritical or Low Power Startup Condition

11.

Uncontrolled RCCA Bank Withdrawal at Power

12.

Rod Cluster Control Assembly Miss-operation

13.

Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature

14.

Chemical and Volume Control System (CVCS) Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant

15.

Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position

16.

Spectrum of Rod Cluster Assembly Ejection Accidents

17.

Inadvertent Operation of the ECCS during Power Operation

18.

Chemical and Volume Control System (CVCS) Malfunction That Increases Reactor Coolant Inventory

19.

Inadvertent Opening of a Pressurizer Safety or Relief Valve

20.

Break in Instrument Line or Other Lines from the Reactor Coolant Pressure Boundary that Penetrate the Containment

21.

Radioactive Release from a Subsystem or Component 4.5.2 Transients That Were Affected by the Valve Closure Characteristics, But Bounded by the AOR The licensee identified the following events that it had modeled MFIVs and provided qualitative evaluation of those events with respect to the extend valve closure time for the NRC staff to review.

4.5.2.1 Feedwater System Malfunctions That Result in an Increase in Feedwater Flow In the AOR, the feedwater malfunction cases with the full-power and zero-power conditions were analyzed. Feedwater isolation was assumed to occur from a high-high SG water level signal.

The results for both cases showed very little variation on the core conditions, including core inlet enthalpy and core exit pressure at the time of peak core average heat flux. The small difference in the core conditions would result in a small difference in the calculated minimum departure from nucleate boiling ratio (DNBR). Therefore, the NRC staff agrees with the licensee that the minimum DNBR with greater than 6.9 and 10 percent margin to the DNBR limit documented in the AOR for the respective full-power and zero-power cases would be negligible impacted, due to the reanalysis modeling with an MFIV closure time increased from the current 5 seconds to 15 seconds.

4.5.2.2 Loss of Non-Emergency AC Power to the Station Auxiliaries/Loss of Normal Feedwater Flow In the AOR for the loss of non-emergency AC power (LOAC) and loss of normal feedwater flow (LONF) cases, the loss of main feedwater resulted in a reactor trip on a low-low SG water level trip signal. The signal initiated AFW flow to the SGs. Based on the plant design, the feedwater check valves are installed downstream of the AFW injection point. Thus, before AFW can be delivered to the SG, the MFIVs must be fully closed to enable the AFW to fill the piping volume from the MFIVs to the SG. The LOAC/LONF analyses in the AOR assumed that the total delay to the start of AFW from the low-low SG water level signal events was 394 seconds, which accounted for a 2-second delay from the low-low SG water level signal to reactor trip, the 60-second delay for diesel generator, and AFW pump start and allowed for the filling of the

associated feedwater piping. The 394-second delay time sufficiently accommodated the MFIV closure time increased from 5 to 15 seconds. Therefore, the NRC staff agrees with the licensee that the conclusions for the LOAC/LONF events in the AOR remain valid.

4.5.3 Accidents That Were Affected by the MSIV and MFIV Closure Characteristics The licensee identified the following accidents that modeled the MSIVs and/or MFIVs, and provided quantitative results of the reanalysis for those events for the NRC staff to review. The accidents are the MSLB, MFLB, SGTR, and SBLOCA accidents. The NRC staffs review of the reanalysis of accidents is addressed for the MSLB, MFLB, and SGTR accidents in Section 4.5.3, and for the SBLOCA in Section 4.6 of this SE.

4.5.3.1 Computer and Methods Used for the non-LOCA Transient Reanalyses The licensee used the following computer codes to perform reanalyses for the non-LOCA transients:

The ANC computer code is an NRC-approved code (Reference 5) that is used to predict nuclear reactor core reactivity and assembly and rod distributions for normal and off-normal conditions. The code allows for the treatment of enthalpy, xenon, and Doppler feedback. A high degree of automation has been incorporated into the code to address fuel depletion, reactivity coefficients, control rod worth for non-uniform inlet temperature distribution.

The RETRAN-3D computer code used in the RETRAN-02 mode computes pertinent plant transient information including core power level, RCS pressure and temperature (Reference 3). Because the RETRAN-3D in RETRAN-02 mode was previously approved for the WCGS by the NRC (Reference 6), the use of the computer model for the reanalysis of the applicable transients in USAR Chapter 15 is acceptable.

The VIPRE-01 computer code is an NRC-approved open channel thermal-hydraulic code (Reference 7) designed to evaluate DNBR and coolant state for steady-state operations and transients using subchannel analysis.

4.5.3.2 Assumptions That Were Different from the Existing Analyses In response to the RAI, the licensee identified the following assumptions that were different from those used in the AOR.

4.5.3.2.1 Effects of MSIV and MFIV Closure Characteristics on System Response The licensee indicated that for the MFLB and SGTR reanalyses, the respective MSIV and MFIV closure models assuming the isolation valves ramping closed linearly over a specific closure time period remained unchanged from that assumed in the AOR. For the MSLB reanalysis, the AOR MSIV closure model of an instantaneous closure after a specific time delay was changed to ramp closed linearly over a specific closure time period. For all three events, the closure times for the valves were changed to 15 seconds.

In response to the RAI (Reference 4), the licensee performed sensitivity studies to identify the effects of the valve closure patterns and MSIV/MFIV closure times on the reanalysis for the MFLB, SGTR, and MSLB events.

4.5.3.2.1.1 MSLB Analysis The MSLB event was analyzed by the licensee for the four different MSIV and MFIV closure patterns: (1) a step closure of MSIVs and MFIVs with time delay; (2) a step closure of MSIVs with time delay and linear closure of MFIV; (3) a linear closure of MSIVs and step closure of the MFIVs with time delay; and (4) a linear closure of MSIVs and MFIVs. The results showed that the assumed valve closure pattern combination (case 1) corresponding to the greater cooldown of the RCS primary system resulted in the most rapid power increase and the most limiting minimum DNBR.

As discussed in Section 4.5.3.2.1.4 below, the NRC staff concludes that the assumed linear valve closure pattern for MSIVs and MFIVs with a 15-second closure time used in the analysis of case 4 is acceptable. Since the SLB sensitivity study (Reference 4) showed that the calculated minimum DNBRs increased for case 1 through 4 as the cooldown effects of each case decreased, the NRC staff concludes that the use of case 3 as the new AOR is conservative, resulting in a higher DNBR as compared to case 4 that used the acceptable valve closure pattern and time and, therefore, is acceptable.

4.5.3.2.1.2 MFLB Analysis The MFLB analysis was performed for the following five different MSIV conditions: (1) a linear MSIV closure; (2) an MSIV step closure after time delay; (3) an MSIV immediate step closure; (4) a 15-second linear MSIV closure; and (5) a 5-second MSIV linear closure. The results showed that the plant system response to the MFLB event was similar for cases with different valve closure patterns and closure times. The key plant parameters such as the maximum hot-leg temperature, core inlet pressure and pressurizer water volume occurred before the MSIVs were actuated. Therefore, the NRC staff agrees with the licensee that the model used for the valve closure patterns and closure times will have no effect on the margin to hot-leg saturation, RCS pressurization, and pressurizer overfilling following an MFLB event.

4.5.3.2.1.3 SGTR Analysis The SGTR with forced overfill case was selected for sensitivity study since it was identified in the AOR as the limiting SGTR case in terms of the dose release consequence. The case was analyzed for two different MFIV closure patterns: (1) a linear MFIV closure and (2) an MFIV step closure after time delay. The results showed that the MFIV closure patterns would result in slight changes to the SGTR system response. For example, the faulted SG filled sooner if the MFIV closure pattern of an MFIV step closure after time delay was assumed. Consequently, water relief through the safety valves began earlier when the faulted SG and associated steam line up to the MSIV were filled with water. However, a comparison of the integrated break flow, the main source of the radiological release, showed that the comparative results were essentially identical. Therefore, the NRC staff agrees with the licensee that the assumption of the MFIV closure patterns has negligible effect on the STGR results in meeting the acceptance criterion of the radiological release limits.

4.5.3.2.1.4 MSIV and MFIV Isolation Times and Closure Patterns In the RAI response (Reference 4), the licensee presented the single train and dual train valve performance test information for the replacement MSIVs and MFIVs. The valve test information showed that the valve closure time decreased as the SG system pressure increased, while the analyses for the limiting MFLB, SGTR, and MSLB events showed that the SG pressures were significantly higher than 400 psia at the times the MSIVs and MFIVs were required to close. At the SG system pressure of greater than 400 psia, the valve test information showed that the valve closure time was significantly less than 15 seconds. Therefore, the NRC staff concludes that the use of the high value of 15 seconds is conservative, resulting in a greater cooldown effect and lower DNBR, and, therefore, is acceptable for the analysis of the MFLB, SGTR, and MSLB events.

In support of the valve closure pattern used in the accident analysis, the licensee presented the closure patterns of a replacement MSIV/ MFIV in Figure 2-30 of Reference 4. The figure showed that the replacement valves provided a quick closing characteristic. For example, based on a linear valve closure pattern assumed in the reanalyses, a 50 percent stroke time would result in a 50 percent flow area reduction, compared to the recorded flow area reduction of approximately 90 percent. Therefore, the NRC staff concludes that the use of the linear valve closure pattern is conservative, resulting in a greater lower cooldown effect and lower DNBR, and, therefore, is acceptable.

4.5.3.2.2 Changes in the RETRAN Computer Code Base Deck In the RETRAN base deck for the reanalyses of all non-LOCA transients, the licensee stated that changes (Reference 3) were made to revise the pressurizer safety valve (PSV) nominal setpoint from 2485 psig to 2460 psig, the re-closure pressure from 2375 psia to 2351 psia, and the full-open (3 percent accumulation) pressure from 2574 psia to 2548 psia. The NRC staff concludes that the change of the PSV nominal setpoint is acceptable because it is consistent with the value in TS 3.4.10, which was previously approved by the NRC staff on the bases discussed in Reference 8. The changed values for the PSV full-open and closure pressures are also acceptable because they bounded the range of the PSV setpoint (2460 psig + 2 percent) in TS 3.4.10.

In the RETRAN base deck used in the reanalysis of the SGTR with stuck-open ARV, the licensee stated that the value of the thermal design flow (TDF) was changed (Reference 3) to be consistent with the value specified in TS Table 3.2.1. The change of the TDF is acceptable because it is consistent with TS Table 3.2.1 that was previously approved by the NRC on the bases documented in Reference 9.

4.5.3.3 Results of the Reanalyses 4.5.3.3.1 MSLB Reanalysis The steam release during an MSLB event causes a decrease in the RCS temperature and SG pressure. In the presence of a negative moderator temperature coefficient, the RCS temperature decrease results in an addition of positive reactivity. If the positive reactivity addition is not offset by the negative reactivity resulting from the reactor control rod insertion after reactor trip and boron addition from the SI system, the reactor core may return to criticality,

which could result in a low DNBR and fuel failure. The licensee analyzed the MSLB event using the following NRC-approved codes: the RETRAN-3D code in the RETRAN-02 mode for the system response calculation; the ANC code for a detailed core analysis to determine if the RETRAN-predicted reactivity feedback model was conservative, resulting in a greater positive reactivity addition; and the VIPRE code for the core thermal-hydraulic analysis to determine if DNB occurred.

The licensee analyzed the hot standby MSLB case with a postulated limiting break size and an available offsite power, which is the limiting case identified in the AOR. The following conditions assumed in the current licensing basis AOR were used in maximizing the MSLB effects on the margin to the DNBR limits:

1.

SI is actuated on either low-pressurizer pressure or low-compensated steam line pressure, with a 2-second time delay.

2.

A limiting end-of-life shutdown margin of 1.3 percent k was assumed.

3.

A limiting negative end-of-life moderator coefficient was assumed.

4.

The core power distribution was assumed uniform and the two channels had equal reactivity coefficient weighting.

5.

Minimum boron injection capability was assumed corresponding to the most restrictive single failure in the SI system.

6.

SI flow was assumed to be delivered to the RCS with a delay of 52 seconds after the initial signal.

7.

The Doppler feedback curve was based on USAR Figure 15.1-14.

8.

A single failure of one SI system train was assumed.

In the RAI response (Reference 3), the licensee identified the following four assumptions that were different from those used in the current AOR:

1.

RCS Initial Conditions The analysis used the thermal design flow (TDF) of 361,296 gpm instead of the design flow value of 374,000 gpm used in the current AOR. The licensee indicated that an increase in the peak heat flux, associate with the assumption of a flow rate higher than the thermal design flow rate, would be offset by crediting the higher flow rate in the DNBR calculation.

The analysis also used the nominal RCS pressure as the initial RCS pressure.

This assumption represented a change from the -30 psia pressure uncertainty assumed in the current AOR. The licensee indicated that for a large MSLB, the depressurization was determined by the break flow and was not significantly affected by the initial conditions.

During the course of the review, the NRC staff requested the licensee to provide the results of a plant-specific sensitivity analysis for the limiting MSLB event to demonstrate that the changes of the initial RCS flow and pressure would be conservative, resulting in a minimum DNBR lower than that of the AOR. In response, the licensee provided the results of the SLB sensitivity study (Reference 4) and showed that the MSLB analysis with a nominal initial pressure and TDF resulted in a slightly lower minimum DNBR value, as compared with analysis with AOR initial conditions (RCS flow and pressure). Based on this, the NRC staff concludes that the changes of the initial RCS flow and pressure were conservative and, therefore, acceptable.

2.

AFW Flow The AFW delivered to the faulted SG would increase the cooldown of the RCS primary system during an MSLB event. The analysis used the AFW rate of 1360 gpm to the faulted SG and 700 gpm to the intact SGs, for a total 2060 gpm.

The assumed AFW rates were different from the assumed AFW flow of 2020 gpm total to the faulted SG in the AOR. In the RAI response to support the assumed AFW flow rates (Reference 4), the licensee indicated that the piping network model for the AFW system used in the SLB analysis was previously used in calculations to determine AFW flow for safety analyses. The model accounted for the as-built piping configuration including various hydraulic components such as pumps, valves, and pipe fittings. The AFW calculations applied the following key assumptions:

Failure of the AFW control device was assumed in order to maximize the AFW flow to the faulted SG, resulting in an excess cooldown of the RCS primary system during an MSLB event. Specifically, the discharge flow-control device installed on the AFW header common to both the motor-driven and turbine-driven AFW pumps was assumed to fail to perform its safety function. This device was designed to regulate the AFW flow from the motor-driven AFW pump to the faulted SG such that a minimum flow would be delivered to the intact SG from both the motor-driven and turbine-driven AFW pumps. Without the flow-control device, all the flow from the motor-driven AFW pump would be injected to the faulted SG and only the flow from the turbine-driven AFW pump would be delivered to the intact SGs.

The AFW flow was assumed to take suction from the condensate storage tank.

The mini-recirculation flow rate of 75 gpm for each motor-driven AFW pump and 145 gpm for the turbine-driven AFW pump were modeled as a constant external demand at the respective pump discharge junction node.

The certified AFW pump characteristic curves were used in the calculation.

AFW flow rates were calculated for various pressure conditions in the faulted and intact SGs. The total flow of 2060 gpm, consisting of 1360 gpm to the faulted SG and 700 gpm to the intact SGs, used in the MSLB analysis, was based on the faulted SG pressure of 14.7 psia and the intact SG pressure of 1200 psia. As shown in the pressure transient resulting from a limiting MSLB event, the pressures of the intact SGs and the faulted SG remained above 600 psia and 200 psia, respectively. Since the AFW flow rate delivered to the SGs increased as the SG pressure decreased, the AFW flow to the faulted SG based on the lowest SG pressure of 14.7 psia would be greater than the actual AFW flow delivered to the affected SG based on a SG pressure of 200 psia that was the lowest pressure predicted during the limiting MSLB event.

The NRC staff found that the use of high AFW flow rate to the faulted SG was conservative, resulting in an increase in cooldown of the RCS primary system and a decrease in the calculated minimum DNBR during an MSLB event. Therefore, the NRC staff determined that the AFW flow rates assumed in the MSLB analysis were acceptable.

3.

MSIV Closure Characteristics MSIV closure was assumed to occur beginning 2 seconds after a low steam line pressure signal was received. The MSIV closed in a 15-second linear ramp.

This assumption differed from the AOR, which assumed the MSIV closed instantly at 10 seconds (Reference 3). The changes of the MSIV were acceptable on the basis discussed in Sections 4.5.3.2.1.1 and 4.5.3.2.1.4 of this SE.

4.

MFIV Closure Characteristics In the AOR, the feedwater was assumed to be isolated when safety injection was initiated, with a 2-second time delay. The new MSLB analysis assumed that the feedwater flow ramped instantaneously to zero at the end of the 15-second time delay. The assumed feedwater model was acceptable on the basis discussed in Sections 4.5.3.2.1.3 and 4.5.3.2.1.4 of this SE.

The results of the MSLB reanalysis showed that the minimum DNBR remained above the acceptable value of 1.50 for the DNBR safety limit used in the AOR, and demonstrated full compliance with the same acceptance criteria used in the AOR: (1) pressure in the RCS and SGs shall be maintained below 110 percent of the design pressures; (2) any fuel damage that may occur during the transient should be of a sufficiently limited extent so that the core coolability can be maintained; and (3) any activity released must be such that the calculated doses at the site boundary are a small fraction of the 10 CFR 100.11 guideline criteria. Based on this, the NRC staff concluded that the MSLB reanalysis was acceptable.

4.5.3.3.2 MFLB Reanalysis An MFLB is defined as a break in a feedwater line large enough to prevent the addition of sufficient feedwater to maintain shell-side water inventory in the SG. The MFLB may reduce the ability to remove heat generated by the core from the RCS because fluid in the SG is discharged

through the break, and the break may be large enough to prevent the addition of main feedwater after the reactor trip. During the event, the AFW will function to prevent substantial overpressurization of the RCS and maintain sufficient liquid in the RCS to provide core coolability. A reactor trip may be actuated on signals of the SG low-low water level, high-pressurizer pressure, high pressurizer level, overtemperature delta-T (OT -T), and high-containment pressure.

The licensee reanalyzed the full-power MFLB case with a double-ended rupture of the largest feedwater line, which was identified in the AOR as the limiting MFLB case. The event was analyzed using the RETRAN-3D code in the RETRAN-02 mode, which was previously approved by the NRC for non-LOCA transient analyses. The following conditions assumed in the current AOR were used in maximizing the MFLB effects on the margin to the RCS pressure limits:

1.

The initial plant power was assumed to be 102 percent of the rated core power.

2.

No credit was taken for the pressurizer pressure control (i.e., pressurizer power-operated relief valves (PORV) or pressurizer spray).

3.

Initial pressurizer level was at the nominal programmed value, plus 5 percent of span error.

4.

Main feedwater was assumed to be lost to all SGs at the event initiation due to malfunction in the feedwater control system under adverse environment. The MFLB was assumed to occur when the SG inventory reached to 0 percent narrow range span (i.e., reactor trip on low-low SG water level of 23.5 percent, minus environmental allowance and uncertainties). The combination of errors described above yielded the most severe MFLB transient, with control and protection interaction considered.

5.

The AFW system was actuated by the low-low SG water level signal. A 60-second time delay was assumed following the low-low SG water level signal to allow time for startup of the standby diesel generators and AFW pumps. An additional 314 seconds was assumed before the feedwater lines were purged and relatively cold AFW entered the intact SGs.

6.

Failure of one protection train was taken as the worst single failure so that the second motor-driven AFW pump was assumed inoperable. The total AFW flow delivered to the three intact SGs was assumed to be 563 gpm. Credit was taken for the discharge flow control device installed on the AFW header common to both the motor-driven and turbine-driven AFW pumps. Due to this discharged flow control device, the intact SG receiving AFW from both the motor-driven and turbine-driven AFW pumps was assumed to receive no more than 250 gpm. The remaining two intact SGs received AFW flow from only the turbine-driven AFW pump and these SGs were assumed to receive approximately 157 gpm each.

In response to the RAI (Reference 3), the licensee indicated that the initial SG water level was assumed to be 55.7 percent instead of the nominal value of 50 percent assumed in the AOR.

The 5.7 percent increase from the nominal value was used to include the water level measurement uncertainty. With an increase in the initial SG water level assumed in the MFLB

analysis, the reactor trip initiated by a SG low-low level signal would be delayed. The delayed reactor trip prolonged the RCS heat-up and pressurization, as well as the moderator feedback induced power increase as the temperature coefficient for the beginning of life fuel cycle assumed in the MFLB analysis was positive. The net effect of an increase in the initial SG water level would increase the peak RCS pressure during the MFLB event. Therefore, the NRC staff concluded that the use of 55.7 percent for the initial SG water level was adequate and acceptable.

In the MFLB analysis, the MSIVs and MFIVs were assumed to close following a 2-second time delay, and ramped closed in 15 seconds compared to 5 seconds (Reference 1) assumed in the AOR. Since the feedwater was terminated as part of the initiating sequence for the MFLB analysis, the MFIV closure pattern and time would not affect the results of the analysis. The assumed MSIV characteristics were found acceptable on the basis discussed in Sections 4.5.3.2.1.2 and 4.5.3.2.1.4 of this SE.

The licensee analyzed two MFLB cases with and without offsite power, which were the limiting cases identified in the AOR. The results of the MFLB analysis demonstrated full compliance with the same acceptance criteria used in the AOR: (1) pressure in the RCS and SGs shall be maintained below 110 percent of the design pressures; (2) any fuel damage that may occur during the transient should be of a sufficiently limited extent so that the core coolability can be maintained; and (3) any activity released must be such that the calculated doses at the site boundary are a small fraction of the 10 CFR 100.11 guideline criteria. Therefore, the NRC staff concludes that the MFLB reanalysis is acceptable.

4.5.3.3.3 SGTR Reanalysis The licensee analyzed two limiting SGTR scenarios in the AOR: (1) an SGTR with postulated failure of the faulted SG AFW flow control valve (SG overfill scenario); and (2) SGTR with postulated stuck-open ARV for the faulted SG (stuck-open ARV scenario). The RETRAN-3D in the RETRAN-02 mode was used to calculate the transient response of both the primary and secondary systems for the two SGTR scenarios analyzed. Operator actions in response to an STGR event were assumed to follow emergency operating procedure EMG E-3, Steam Generator Tube Rupture. The SGTR operator actions included: (1) to identify that an SGTR has occurred, then to identify and isolate the faulted SG; (2) to prepare the RCS cooldown in order to maintain subcooling after subsequent depressurization; (3) to cool and then depressurize the RCS to reduce the primary-to-secondary leakage; (4) to terminate safety injection in order to prevent repressurization of the RCS; and (5) to take the plant to cold shutdown conditions in order to establish cooling by the RHR system.

The analysis applied the assumptions used in the current AOR in maximizing the amount of radioactive release. These AOR assumptions are discussed below.

In the SG overfill scenario, the AFW flow to the faulted SG was maximized by assuming a wide-open position for the control valve on the discharge side of the motor-driven AFW pump feeding the faulted SG in order to increase probability for faulted SG overfill and subsequent water relief from its safety valve. The radioactive releases were maximized by assuming that the safety valve was stuck open following water relief with an effective flow area equal to 5 percent of the total safety valve flow area. The following assumptions were also used:

1.

Credit of a low pressurizer or OT T signal for reactor trip;

2.

LOOP initiated at reactor trip;

3.

AFW flow varied with the change in the faulted SG pressure;

4.

RCS cooldown initiated at 30 minutes following the initiation of the SI signal;

5.

RCS depressurization using a pressurizer PORV; and

6.

Delayed SI termination to ensure sufficient liquid entered the ruptured SG steam line to cause the safety valves to open and result in water relief.

In the stuck-open ARV scenario, the discharge of contaminated secondary fluid was maximized by assuming the faulted SG ARV stuck open for a specific period of time (consistent with the observed simulator exercises) until the ARV block valve was manually closed. The following assumptions were also used in the analysis:

1.

Credit of a low pressurizer or OT T signal for reactor trip;

2.

LOOP initiated at reactor trip;

3.

An AFW flow of 250 gpm delivered to each SG to maintain the SG narrow range (NR) level below 15 percent;

4.

RCS cooldown initiated at the SG NR level of greater than 10 percent and the SG pressure of greater than 630 psia until RCS temperature reduced to 50 °F less than the faulted SG saturation;

5.

RCS depressurization initiated within the time frame (consistent with observed exercises) after completion of cooldown; and

6.

SI termination delayed by a specific period of time (consistent with observed simulator exercises) after primary side depressurization completed and SI termination criteria met.

7.

In addition, the analysis assumed that following SI termination, the operators equalized pressure in the RCS and faulted SG within a specific time frame consistent with observed simulator exercises. During this time, break flow in the faulted SG continued.

8.

After pressures were equalized, it assumed that the transition to cold shutdown was made utilizing steam release to the atmosphere from the intact SGs.

The results of both SGTR cases showed that the radiological consequences resulting from a SGTR event remained well within the limits specified in 10 CFR 100.11 and Section 15.6.3 of the NRC SRP. The analysis demonstrated full compliance with the same acceptance criteria used in the AOR: (1) pressure in the RCS and SGs shall be maintained below 110 percent of the design pressures; (2) any fuel damage that may occur during the transient should be of a sufficiently limited extent so that the core coolability can be maintained; and (3) any activity released must be such that the calculated doses at the site boundary are a small fraction of the 10 CFR 100.11 guideline criteria (see Section 3.0 of this SE). Based on this, the NRC staff concludes that the SGTR event is acceptable.

4.5.4 Conclusion Based on its review of the licensee's application and its supplemental letters referenced above, the NRC staff concludes the following: the non-LOCA transient analyses were performed using the NRC-approved methods; the assumptions used in the analyses were consistent with those used in the AOR; and the results of the analyses met the applicable acceptance criteria used for the AOR, which satisfied the requirements of GDCs 10 and 15, and 10 CFR 100.11. Based on this, the NRC staff has determined that the licensee's analyses for this amendment are acceptable. The NRC staff also finds that the proposed valve closure times and patterns were adequately considered in the acceptable analyses and, therefore, concludes that the valve closure times and patterns were acceptable for the replacement MSIVs and MFIVs at WCGS.

Based on this, the NRC staff concludes that the proposed replacement of the MSIVs and MFIVs is acceptable.

4.6 LOCA Transient Analysis In Section 4.6 of Attachment I to its application, the licensee presented its SBLOCA analysis in support of the MSIV/MFIV valve replacement. It is only the SBLOCA of the LOCA transients that is affected by the longer valve closure time for the new MSIVs and MFIVs. The licensee's reanalysis of the DBA was presented in the application to demonstrate the conformance of the plant to the 10 CFR 50.46 requirements, using approved SBLOCA methods, with the proposed use of the new MSIV/MFIV valves.

The licensee's SBLOCA reanalysis employed the NOTRUMP code and evaluated the 2-, 3-, 4-,

6-, and 8.75-inch diameter breaks in the cold leg at the reactor coolant discharge leg. The worst break was found to be the 4-inch break with a peak clad temperature (PCT) of 935.5 °F. Due to the low clad temperatures, the NRC staff does not require further evaluation to identify a more limiting break between these integer pipe sizes. At these low temperatures, clad oxidation is negligible and was calculated to be less than 0.01 percent. The low PCTs are due to the high capacity of the ECCS, low refueling water storage tank (RWST) temperature, and the use of the NRC-approved COSI condensation model in the NOTRUMP code. The NRC staff notes that the licensee properly evaluated the consequences of a severed ECCS injection line and found that this assumption was less limiting than the limiting 4-inch diameter break.

In an RAI, the NRC staff questioned the licensee about the low RWST temperature of 100 °F assumed in the SBLOCA analysis in that typical RWST temperatures assumed in such analyses

are typically 120 °F. The licensee responded that the 100 °F RWST maximum temperature is controlled by the TS 3.5.4, "Refueling Water Storage Tank (RWST)," in which SR 3.5.4.1 requires verification to be performed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> that the RWST water temperature is within limits (i.e., 37 °F and 100 °F). Therefore, the RWST water temperature assumed in the SBLOCA analysis is the maximum RWST of 100 °F allowed by TS 3.5.4. The NRC staff agrees with the licensee that this is the correct RWST water temperature for the SBLOCA analysis.

In another RAI, the NRC staff requested if there were new model changes included in the SBLOCA reanalyses submitted in the licensee's application. The licensee responded that it had used NOTRUMP Code Version 39.0, which is the same code version used for recent SBLOCA analyses (since July 2003) including those supporting the extended power uprates of Beaver Valley Power Station, Units 1 and 2, and R.E. Ginna Nuclear Power Plant. No changes have been made to the NOTRUMP EM codes other than (1) those for error corrections, which have been reported through the 10 CFR 50.46 process and (2) the implementation of the COSI condensation model in NRC-approved Westinghouse topical report WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection into the Broken Loop and COSI Condensation Model," dated July 1997.

Because the bottom elevation of the loop seal in the suction leg piping is 6.3 feet below the top elevation of the core, additional RAI questions were issued by the NRC staff requesting the response to breaks located on the top of the cold leg at the reactor coolant discharge legs. Due to the low elevation of the suction leg piping, breaks on the top of the discharge leg will eventually result in refilling of the loop-seal region with ECCS injection. Therefore, during the long term following the largest small break that does not allow the RCS to completely refill with ECCS, the loop seals will remain full of liquid which creates a large resistance to steam flow to the break from the core. As such, the pressure differential between the core and broken cold leg will increase sufficiently resulting in the depression of the two-phase level into the core for an extended period of time. Core uncovery will ensue and clad temperatures could increase above that for the limiting break size located on the bottom of the discharge leg piping. Because the WCGS plant has flow nozzles penetrating the upper portion of the core barrel into the downcomer, a steam venting path to the break is available which will preclude the potential for long-term core uncovery. The licensee stated that the upper barrel flow nozzle total area is approximately 27 square inches. This vent area will relieve the pressure buildup in the upper plenum should the loop seals refill with ECCS during an SBLOCA. Based on this, the NRC staff concluders that long-term core uncovery is not an issue for WCGS due to the presence of the upper core barrel flow nozzles.

Also, when questioned, the licensee noted that the hot-leg nozzle gaps and upper core barrel leakage paths into the upper head were not modeled in the NOTRUMP nodalization model.

This prevents further venting of core generated vapor from bypassing the external loops to exit through the break which will tend to maximize PCT. Based on this, the NRC staff concludes that not modeling the hot-leg nozzle gaps and upper core barrel leakage paths into the upper head is conservative with respect to the SBLOCA analysis, and, therefore, acceptable.

Based on the review of the NOTRUMP SBLOCA analyses performed at the full power level of 3565 MWt plus uncertainty and total core peaking factor of 2.5, described above, the NRC staff concludes that acceptable ECCS performance within the limits set forth in 10 CFR 50.46 has

been demonstrated for the MSIV/MFIV valve replacement and, therefore, the MSIV/MFIV valve replacement is acceptable.

In its review, the NRC staff questioned the validation of the COSI condensation model. Although this model was previously approved by the NRC staff for use in the NOTRUMP SBLOCA code, the NRC staff has recently come to the conclusion that there appears to be inadequate validation of the new method against integral SBLOCA experimental data. The NRC staff considers this a generic issue not related to the licensee's current valve replacement amendment and concludes that the valve replacement amendment is not expected to adversely impact plant operation at WCGS with the application of the NRC-approved COSI model for the SBLOCA analysis. With this amendment, the licensee remains within its current licensing basis for the SBLOCA analysis. However, the NRC staff will be addressing its concerns about the COSI condensation model generically with Westinghouse Electric Company, the owner of topical report WCAP-10054-P-A, Addendum 2, Revision 1.

4.7 Human Factors Analysis 4.7.1 Operator Action Response Times Analyses for Accidents Based on its review of the accidents addressed in Section 4.0 of Attachment I to the licensee's application, the NRC staff concluded that the only accidents that needed to be addressed with respect to operator response times are the following, which are addressed below:

SGTR with Overfill from AFW Flow, SGTR with ARV Stuck Open, and Termination of AFW to faulted SG during MSLB 4.7.2 SGTR Scenario 4.7.2.1 SGTR with Overfill from AFW Flow The licensee provided the time sequence of assumed operator response times used in the SGTR with overfill from AFW flow accident scenario in Tables 4.5-1 and 4.5-2, in Attachment I to the licensee's application. This table appears to differ from the operator responses times for this accident scenario in USAR Table 15.6-1. The licensee stated in its response to the NRC staffs RAI, which is in its letter dated December 14, 2007, that the assumed response times in the reanalysis are the same as those assumed in the current AOR for this accident scenario in USAR Table 15.6-1. The tables in the licensee's application and in the USAR list the operator actions and response times differently so that these tables can not be compared. Also, accounting for the replacement valves with the longer closure time changes the analysis in the AOR in the USAR and these changes result in an apparent differences in Tables 4.5-1 and 4.6-2 with respect to Table 15.6-1; however, the licensee states that the actual operator response times for this accident scenario are not changing. Therefore, the actual operator response times used in the analysis for this SGTR scenario remain unchanged from the times given in the USAR. Based on this, the NRC staff finds that the operator response times for this accident scenario are acceptable because these times are the same as those times used in the current AOR and are justified on the same basis as those in the current AOR.

4.7.2.2 SGTR with ARV Stuck Open As stated by the licensee in its application, the SGTR with ARV stuck-open scenario analysis revises critical operator action times. In its response to an RAI question, in its supplemental letter dated December 14, 2007, the licensee stated that, in the reanalysis of this scenario, it made the following changes to the analysis: (1) the licensee maximized the discharge of contaminated fluid by assuming the ARV was stuck open for 21.53 minutes (compared to the AOR assumption of 20 minutes), (2) the licensee initiated RCS depressurization 5.6 minutes after cooldown is complete (compared to the AOR assumption of 3 minutes), and (3) the licensee terminated SI after a 9.8-minute delay (compared to the assumption of 3 minutes).

These extensions in the reanalysis allow more time for operator actions and the assumed operation action times are based upon a combination of simulator demonstrations and conservative extrapolation developed with operator subjective judgment. Therefore, the proposed license amendment increases the amount of time available for operators to complete the actions as compared to the current analysis of this accident. Based on this, the staff finds the change to the accident scenario analysis acceptable because the reanalysis allows more response time for operator actions than the current safety analysis and the response times are based on a combination of simulator demonstrations and conservative extrapolation based on operator judgment.

4.7.3 Termination of AFW to Faulted SG during MSLB As stated by the licensee in its application, the MSLB analysis for the valve replacement decreases the operator action time for re-aligning the AFW to the faulted SG from the current 30 minutes to 20 minutes. The current AOR uses the more conservative operator response time of 30 minutes; however, the replacement of the MSIVs and MFIVs with valves that have longer closure times increases the associated containment pressure and temperature. This increase is the cause of the change to the operator reaction times used in this accident to regain margin in the new analysis and, therefore, stay within the margins that exist in the current AOR for the accident.

The licensee stated that the actual time for the operator to complete their tasks is stated in USAR Section 6.1.2.3.3 as between 6 and 7 minutes. The licensee added that this actual operator action response time remains unchanged. This is to state that the valve replacement in this amendment request does not change the operator action response time. The licensee has stated that it has run simulator scenario measurements for six crews to verify the operator response time.

Based on the above, the NRC staff finds that the change to this accident scenario in reducing the operator action response time is acceptable. The NRC staff concludes that the margin between the actual CT for the operator action and the 20 minutes assumed in the reanalysis for the operator to complete this action is acceptable. The time available for the completion of the operator actions is less in the reanalysis of the termination of the AFW flow to the faulted SG during the MSLB for the proposed amendment; however, the actual completion time for the operator action remains unchanged from the current AOR of this event.

4.7.4 Conclusion The NRC staff has reviewed (1) the human factors aspects submitted by the licensee in its proposed replacement of the MSIVs and MFIVs with new valves and actuators that extend the closure time for these valves, and (2) the effect of this replacement on operator action response times in accidents. Based on the evaluation given above, the NRC staff finds that the licensee has acceptably addressed the operator action response times for the accidents address in its amendment request and these response times continue to meet the regulations. Based on this, the NRC staff further concludes that the proposed amendment to replace the MSIVs and MFIVs is acceptable with respect to the operator action response times assumed in the analyses for the accidents involving the new replacement MSIVs and MFIVs.

4.8 Amendment Conclusions The NRC staff has reviewed the licensee's amendment request application dated March 14, 2007, and its supplemental letters dated September 12, October 16, and December 14 (two letters), 2007, and January 18, 2008, to replace the MSIVs and MFIVs, and their actuators. The new valves have a longer closure time in response to accidents in that the current closure time of 5 seconds would be extended to 15 seconds. In Sections 4.3 through 4.7 of this SE, the NRC staff has addressed the following technical areas: accident consequences analysis, containment integrity analysis, non-LOCA transient analysis, LOCA transient analysis, and human factors analysis. Based on the conclusions in these sections, the NRC staff further concludes that the amendment meets the appropriate regulatory requirements and, therefore, is acceptable.

In accordance with 10 CFR 50.36, the LCO for a system must have SRs to demonstrate that the system is operable. In the case of the new MSIVs and MFIVs, the licensee has proposed the use of Figures B 3.7.2-1 (MSIVs) and 3.7.3-1 (MFIVs) as the valve closure times to demonstrate the new valves are operable with respect to their closure time. SRs 3.7.2.1 (MSIVs) and 3.7.3.1 (MFIVs) require that the isolation times of these valves must be verified to be within "limits." The "limits" is the valve closure time for the DBAs and, in accordance with Amendment No. 174 issued in letter dated August 28, 2007, the valve closure limits are given in the TS 3.3.2 Bases, which are identified in Attachment IV to the licensee's application. The limits for the MSIVs and the MFIVs in Attachment IV are the two proposed figures. However, the licensee stated in its application that the valve closure time used in DBAs addressed in the application is 15 seconds.

This use of the proposed Figures B 3.7.2-1 and 3.7.3-1, and the valve closure limit of 15 seconds used in the DBAs, are addressed in Section 2.0 of this SE. Based on that evaluation, the NRC staff concludes that the use of the two proposed figures to demonstrate the operability of the MSIVs and MFIVs as to valve closure time meets 10 CFR 50.36, and is, therefore, acceptable.

Based on the above, the NRC staff concludes that the proposed replacement of the MSIVs and MFIVs and their actuators and the use of proposed Figures 3.7.2-1 and 3.7.3-1 are acceptable.

Therefore, the proposed amendment is acceptable.

In Attachment IV to its application, the licensee identified changes to the following TS Bases that are associated with the MSIV/MFIV valve replacement and the results of the reanalysis of DBAs with the longer valve closure time: TS 3.6.4 on containment pressure, TS 3.6.5 on containment temperature, TS 3.6.6 on containment spray and cooling systems, TS 3.7.2 on the MSIVs, and

TS 3.7.3 on the MFIVs. The identified changes to the TS 3.7.3 Bases to account for adding the MFRVs and MFRV bypass valves to TS 3.7.3 are not being addressed at this time. The NRC staff reviewed the identified changes to the TS bases for the MSIV/MFIV valve replacement and the reanalysis of the DBAs and has no disagreement with the changes the licensee intends to make to the TS Bases; however, the licensee does not explain in the TS Bases changes how, in the MSIVs and MFIVs meeting the valve closure time in Figures B 3.7.2-1 (MSIVs) and B 3.7.3-1 (MFIVs), it is demonstrated that the valves meet the valve closure times used in DBAs. The licensee committed to make the changes the TS Bases as part of its implementation of this amendment (see the next paragraph) and has to make these changes to allow its current licensing basis for the MSIVs and MFIVs to be changed to the next valves and the longer accident closure times. Without these changes to the TS Bases, the MSIVs and MFIVs would have to meet the current valve closure time of no more than 5 seconds to be considered operable.

In Attachment V to its application, the licensee made a regulatory commitment that this license amendment will be implemented prior to startup from refueling outage 16 and that the final TS Bases changes will be implemented pursuant to TS 5.5.14 at the time the amendment is implemented. With the authorization of the use of Figures B 3.7.2-1 (MSIVs) and B 3.7.3-1 (MFIVs) of valve closure times, these figures may be added to the TS 3.7 Bases as shown in Attachment IV to the application.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Kansas State official was notified of the proposed issuance of the amendment. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding published in the Federal Register on June 19, 2007 (72 FR 33785). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

8.0 REFERENCES

1.

Letter from T. J. Garrett, (WCNOC) to NRC, Docket No. 50-482: Revision to Technical Specification (TS) 3.3.2, Engineering Safety Feature Actuation System (ESFAS)

Instrumentation, TS 3.7.2, Main Steam Isolation Valves (MSIVs), and TS 3.7.3, Main Feedwater Isolation Valves (MFIVs), dated March 14, 2007.

2.

Letter from T. J. Garrett, (WCNOC) to NRC, Docket No. 50-482: Response to Request for Additional Information Relating to Replacement of the Main Steam and Feedwater Isolation Valves and Controls, dated September 12, 2007.

3.

Letter from T. J. Garrett, (WCNOC) to NRC, Docket No. 50-482: Response to Request for Additional Information Relating to Replacement of the Main Steam and Feedwater Isolation Valves and Controls, dated October 16, 2007.

4.

Letter from T. J. Garrett, (WCNOC) to NRC, Docket No. 50-482: Response to Request for Additional Information Relating to Replacement of the Main Steam and Feedwater Isolation Valves and Controls, dated December 14, 2007.

5.

WCAP-10965, "ANC - A Westinghouse Advanced Nodal Computer Code," 1985.

6.

Letter from J. Donohew (NRC) to R. Muench (WCNOC), Wolf Creek Generating Station - Issuance of Amendment Re: Addition of Surveillance requirement 3.3.1.16 to Technical Specification Table 3.3.1-1 (TAC No. MD0027), dated August 29, 2006.

7.

Electric Power Research Institute (EPRI) NP-2511-CCM-A, "VIPRE-01: A Thermal-Hydraulic Code for Reactor Core," dated August 1987.

8.

Letter from J. Donohew (NRC) O. Maynard (WCNOC), Wolf Creek Generating Station -

Issuance Of Amendment Re: Pressurizer Safety Valves (TAC No. MA6969), dated March 23, 2000.

9.

Letter from J. C. Stone (NRC) N. S. Carns (WCNOC), Wolf Creek Generating Station-Amendment 99 to Facility Operating License No. NPF-42 (TAC No. M94882),

dated April 4, 1996.

10.

WCAP-8822, Mass and Energy Release Following a Steam Line Rupture, R.E. Land, September 1976.

Attachments: Table 1. WCGS SGTR Radiological Consequences (rem)

Table 2. WCGS Offsite Atmospheric Dispersion Factors (sec/m3)

Table 3. WCGS Data and Assumptions for the SGTR Accident

Principal Contributors: Kamishan Martin John Parillo Ahsan Sallman Summer Sun Leonard Ward Date: March 21, 2008

ATTACHMENTS TABLES 1, 2, AND 3

Table 1 WCGS SGTR Radiological Consequences (rem)

Pre-Accident Iodine Spike Location Dose Type SG Overfill Stuck-Open ARV Limit EAB Thyroid 5.2E+01 4.9E+01 3.0E+02 Whole-Body 2.3E!01 1.5E!01 2.5E+01 LPZ Thyroid 7.2E+00 6.6E+00 3.0E+02 Whole-Body 3.4E!02 2.1E!02 2.5E+01 Concurrent Iodine Spike Location Dose Type SG Overfill Stuck-Open ARV Limit EAB Thyroid 2.3E+01 1.6E+01 3.0E+01 Whole-Body 1.3E!01 8.3E!02 2.5E+00 LPZ Thyroid 4.8E+00 2.6E+00 3.0E+01 Whole-Body 2.5E!02 1.3E!02 2.5E+00 Note: Licensee results are expressed to two significant figures

Table 2 WCGS Offsite Atmospheric Dispersion Factors (sec/m3)

Receptor / Duration

/Q (sec/m3)

EAB 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.5E!04 LPZ 0 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2.0E!05 8 - 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.3E!05 24 - 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> 5.4E!06 96 - 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.5E!06

Table 3 WCGS Data and Assumptions for the SGTR Accident Power level for radiological source term 3636 MWt RCS and secondary coolant system density 8.33 lbm/gallon RCS mass 2.30E+08 gm Initial RCS equilibrium activity, 1% failed fuel 1.0 µCi/gm DEI Maximum pre-accident spike iodine concentration 60 µCi/gm DEI-131 Initial secondary side equilibrium DE I-131 activity 0.1 µCi/gm Primary to secondary leakage rate at an assumed density of 62.4 lbm/ft3 Intact SGs (ISGs) 1.0 gpm Ruptured SG (RSG) 0 gpm Credited operator action times Overfill scenario Terminate AFW to RSG by closing the failed discharge flow control valve T= 18 minutes post SGTR Stuck-open ADV scenario Close block valve on RSG T= 20 minutes post SGTR Comparison of integrated break flows and releases for Analysis of Record and Reanalysis, lbm Analysis of Record Reanalysis Integrated break flow to the end of the transient 196,930 195,371 RSG steam release to atmosphere, 0-2 hrs 145,650 140,406 131,760 120,960 RSG steam release to atmosphere, until shutdown cooling is in operation and SG release is terminated ISGs steam release to atmosphere, 0-2 hours 298,783 309,069 63,360 299,520 ISGs steam release to atmosphere, until shutdown cooling is in operation and SG release is terminated