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#REDIRECT [[ET 10-0013, License Amendment Request to the Approved Fire Protection Program to Remove High/Low Pressure Interface Designation from the Pressurizer Power Operated Relief Valves and Their Associated Block Valves]]
| number = ML101100479
| issue date = 04/13/2010
| title = License Amendment Request to the Approved Fire Protection Program to Remove High/Low Pressure Interface Designation from the Pressurizer Power Operated Relief Valves and Their Associated Block Valves
| author name = Garrett T J
| author affiliation = Wolf Creek Nuclear Operating Corp
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000482
| license number = NPF-042
| contact person =
| case reference number = ET 10-0013
| document type = Letter
| page count = 16
| project =
| stage = Request
}}
 
=Text=
{{#Wiki_filter:WOLF CREEK'NUCLEAR OPERATING CORPORATION Terry J. Garrett Vice President Engineering April 13, 2010 ET 10-0013 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
 
==Subject:==
Docket No. 50-482: License Amendment Request (LAR) to the Approved Fire Protection Program to Remove the High/Low Pressure Interface Designation from the Pressurizer Power Operated Relief Valves (PORVs) and their Associated Block Valves Gentlemen:
Pursuant to 10 CFR 50.90, Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests an amendment to the Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). This license amendment request (LAR) is seeking approval by the Commission, pursuant to License Condition 2.C(5), to make changes to the approved fire protection program as described in the WCGS Updated Safety Analysis Report (USAR). Specifically, a revision to the response to Nuclear Regulatory Commission (NRC)Question 280.5, documented on page 280-5 of the NRC Questions Section at the end of the USAR is requested regarding the removal of the high/low pressure interface designation from the pressurizer Power Operated Relief Valves (PORVs) and their associated block valves.Attachment I provides the evaluation and justification for the proposed license amendment.
Attachment II provides a markup of Page 280-5 of the NRC Question Section of the USAR.Attachment III provides a list of regulatory commitments.
P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET ET 10-0013 Page 2 of 3 It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92. The amendment application was reviewed by the WCNOC Plant Safety Review Committee.
In accordance with 10 CFR 50.91, a copy of this application is being provided to the designated Kansas State official.This LAR supports an element of the corrective action associated with Licensee Event Report (LER) 2008-009-00, "Inadequate Compensatory Actions for a Fire Area." The primary corrective action for LER 2008-009-00 is the implementation of Design Change 012944 to ensure the ability to close a fire induced, spuriously open PORV from the Control Room. As identified in LER 2008-009-00, this change will be implemented prior to startup following Refueling Outage 18. The PORV modification scope has been developed under the supposition that this LAR will be approved.
A compensatory measure fire watch and mitigating action procedure guidance, documented in WCNOC procedure OFN KC-016, "Fire Response," will not be lifted until LAR approval is received and Change Package 012944 is implemented.
If LAR approval is not received, further modification will be required for the PORV circuits to address the more stringent circuit failure modes associated with a high/low pressure interface component.
WCNOC requests approval of this proposed amendment by March 1, 2011 to support the PORV modification in Refueling Outage 18. Refueling Outage 18 is scheduled for Spring 2011.Once approved, the amendment will be implemented within 90 days of receipt and the USAR will be revised in accordance with 10 CFR 50.71(e).If you have any questions concerning this matter, please contact me at (620) 364-4084, or Mr.Richard D. Flannigan at (620) 364-4117.Sincerely, Terry J. Garrett TJG/rlt Attachments:
I Evaluation of Proposed Change II Markup of USAR Pages III List of Regulatory Commitments cc: E. E. Collins (NRC), w/a T. A. Conley (KDHE), w/a G. B. Miller (NRC), w/a B. K. Singal (NRC), w/a Senior Resident Inspector (NRC), w/a ET 10-0013 Page 3 of 3 STATE OF KANSAS )SS COUNTY OF COFFEY )Terry J. Garrett, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.By ,/4vV; , Terry J. arrett Vice President Engineering SUBSCRIBED and sworn to before me this I3+/- day of APrI 2010.Notary(_ublic
'GAYLE SHEPHEARD1 M Notary Public -State of Kansas. E p t[My Appt. Expires 7/1-/4:01/I Expiration Date /.:- 201 Attachment I to ET 10-0013 Page 1 of 9 EVALUATION OF PROPOSED CHANGE
 
==Subject:==
License Amendment Request (LAR) to the Approved Fire Protection Program to Remove the High/Low Pressure Interface Designation from the Pressurizer Power Operated Relief Valves (PORVs) and their Associated Block Valves 1.
 
==SUMMARY==
DESCRIPTION
: 2. DETAILED DESCRIPTION
: 3. TECHNICAL EVALUATION
: 4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Significant Hazards Consideration 4.3 Conclusions
: 5. ENVIRONMENTAL CONSIDERATION
: 6. REFERENCES Attachment I to ET 10-0013 Page 2 of 9 1.
 
==SUMMARY==
DESCRIPTION This evaluation supports a request to'amend Renewed Facility Operating License NPF-42 for the Wolf Creek Generating Station (WCGS).The proposed amendment would revise the Renewed Facility Operating License to deviate from certain WCGS Fire Protection Program requirements.
In the WCGS response to Nuclear Regulatory Commission (NRC) Question 280.5, documented on page 280-5 of the NRC Questions Section at the end of the WCGS Updated Safety Analysis Report (USAR), the Pressurizer Power Operated Relief Valves (PORVs) and their associated block (isolation) valves were classified as high/low pressure interfaces.
Wolf Creek Nuclear Operating Corporation (WCNOC) does not consider this classification to be correct. The reason for this amendment is to remove the high/low pressure interface designation from the PORVs and their associated block valves by revising the response to NRC Question 280.5, documented on page 280-5 of the NRC Questions Section at the end of the WCGS USAR.The predominant basis for the change is NRC Regulatory Guide (RG) 1.189, "Fire Protection for Nuclear Power Plants," Revision 2, issued in October 2009, which states that the approach outlined in Appendix C to NEI 00-01 provides an acceptable methodology for the determination of high/low pressure interface components.
Appendix C to NEI 00-01, "Guidance for Post-Fire Safe-Shutdown Circuit Analysis," Rev. 2, dated May 2009, specifically states that the PORV/PORV Block Valve pathway is not required to be treated as a high/low pressure interface because it does not meet the criteria of an interfacing loss of coolant accident (LOCA) outside of primary containment.
However, since the initial WCGS response to Question 280.5 included the pressurizer PORVs and associated block valves as high/low pressure interface components, it was determined that the removal of the PORVs and block valves as high/low pressure interface components is a reduction in the post-fire safe shutdown (PFSSD) analysis methodology contained in E-1 F991 0, Rev. 6, "Post-Fire Safe Shutdown Fire Area Analysis," when applying License Condition 2.C(5)(b).
The referenced license condition allows WCNOC to make changes to the approved fire protection program without prior NRC approval only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Considering the above discussion, WCNOC has conservatively determined that NRC review and approval of the proposed change is warranted.
: 2. DETAILED DESCRIPTION NRC Question 280.5, documented on pages 280-4 and 280-5 of the USAR, states the following: "The residual heat removal system is generally a low pressure system that interfaces with the high pressure primary coolant system. To preclude a LOCA through this interface, we require compliance with the recommendations of Branch Technical Position RSB 5-1. Thus, this interface most likely consists of two redundant and independent motor operated valves with diverse interlocks in accordance with Branch Technical Position ICSB 3. These two motor operated valves and their Attachment I to ET 10-0013 Page 3 of 9 associated cable may be subject to a single fire hazard. It is our concern that this single fire could cause the two valves to open resulting in a fire-initiated LOCA through the subject high-low pressure system interface.
To assure that this interface and other high-low pressure interfaces are adequately protected from the effects of a single fire, we require the following information:
a) Identify each high-low pressure interface that uses redundant electrically controlled devices (such as two series motor operated valves) to isolate or preclude rupture of any primary coolant boundary.b) Identify each device's essential cabling (power and control) and describe the cable routing (by fire area) from source to termination.
c) Identify each location where the identified cables are separated by less than a wall having a three-hour fire rating from cables for the redundant device.d) For the areas identified in item 5c above (if any), provide the bases and justification as to the acceptability of the existing design or any proposed modifications." The WCGS response to Question 280.5 was as follows: "The reactor coolant system high-low pressure interfaces that rely on redundant electrically controlled devices for isolation include the RHR letdown isolation valves and the pressurizer power-operated relief valves and associated isolation valves.The fire hazards analysis, Appendix 9.5B, demonstrates that no single credible fire could cause the spurious opening of these valves in a manner that would breach the primary coolant boundary." The reason for this amendment is to remove the high/low pressure interface designation from the pressurizer PORVs and their associated block valves by revising the response to NRC Question 280.5, documented on page 280-5 of the NRC Questions Section at the end of the WCGS USAR.As identified in the NRC response to Question 5.3.1 in Generic Letter 86-10, "Implementation of Fire Protection Requirements," high/low pressure interface equipment is subject to more stringent PFSSD analysis criteria than non-high/low pressure interfaces, when considering fire induced spurious operation.
Section 5.3.2 of NRC Regulatory Guide 1.189, Rev. 2, dated October 2009, states that the approach outlined in Appendix C to NEI 00-01 provides an acceptable methodology for the determination of components as high/low pressure interface components, when applied in conjunction with the regulatory guide. Appendix C to NEI 00-01, Rev. 2, dated May 2009, specifically states that the PORV/PORV Block Valve pathway is not required to be treated as a high/low pressure interface because it does not meet the criteria of an interfacing LOCA outside Attachment I to ET 10-0013 Page 4 of 9 of primary containment.
This is discussed in detail below in the Technical Evaluation (Section 3). Therefore, WCNOC is requesting removal of the high/low pressure interface designation from the pressurizer PORVs and their associated block valves.,',-
I This license amendment request (LAR) supports an element of the corrective action associated with Licensee Event Report (LER) 2008-009-00, "Inadequate Compensatory Actions for a Fire Area." The primary corrective action for LER 2008-009-00 is the implementation of Design Change 012944 to ensure the ability to close a fire induced, spuriously open pressurizer PORV from the Control Room. As identified in LER 2008-009-00, this change will be implemented prior to startup following Refueling Outage 18. The PORV modification scope has been developed under the supposition that this LAR will be approved.
A compensatory measure fire watch and mitigating action procedure guidance, documented in WCNOC procedure OFN KC-016, "Fire Response," will not be lifted until LAR approval is received and Change Package 012944 is implemented.
If LAR approval is not received, further modification will be required for the PORV circuits to address the more stringent circuit failure modes associated with a high/low pressure interface component.
The USAR will be revised in accordance with 10 CFR 50.71(e).3. TECHNICAL EVALUATION Pressurizer PORV System Description For PFSSD purposes, the pressurizer PORVs and their associated block valves have been classified as high/low pressure interfaces at WCGS. This increases the circuit fault scenarios that can result in a fire induced open PORV.The PORVs have electrical solenoid actuators.
They are operated automatically based on Reactor Cololant System (RCS) pressure or by remote manual control. The PORVs are designed to limit pressurizer pressure to a value below the fixed high pressure reactor trip setpoint.
They are designed to fail to the closed position on loss of power. The PORVs also assist administrative controls to prevent violation of pressure limits during low temperature operation and provide the safety related means for RCS depressurization to achieve cold shutdown.
Discharged steam from the PORVs is piped to the pressurizer relief tank (inside containment) where it is condensed and cooled by mixing with water.Regulatory Guidance Document Evaluation The Reactor Safety Study, WASH-1400, also known as NUREG 75/014, first identified a safety risk from an intersystem LOCA. The report identified that a significant risk of radiation release to the public exists in piping systems that connect to the RCS and also penetrate the containment.
The concern is stated in paragraph 5.3.2.5 of the NUREG as "...a break in the system will lead into a safeguards building outside the containment so there will be a direct path for radioactive release to the atmosphere..." All systems that connect to the RCS were investigated by the study and most were dismissed as interfacing systems LOCA risks due to one or more reasons. One reason for dismissal listed in Appendix I, paragraph 4.1.6 of the study is that "Failure of the barriers would involve a LOCA into the containment..."
Attachment I to ET 10-0013 Page 5 of 9 Generic Letter 80-14 references the WASH-1400 report in regards to a LOCA that bypasses containment and requests licensees to determine if an event V configuration exists on the residual heat removal (RHR) lines. .NRC Regulatory Guide (RG) 1.189, "Fire Protection for Nuclear Power Plants," Revision 2, issued in October 2009, provides a comprehensive fire protection guidance document, and identifies the scope and depth of fire protection that the staff would consider acceptable for nuclear power plants. Section 5.3.2 of RG 1.189, Revision 2, states that the licensee should evaluate the circuits associated with high/low pressure interfaces for the potential to adversely affect safe shutdown.
As identified in the NRC response to Question 5.3.1 in Generic Letter 86-10, the NRC has taken the position that high/low pressure interface equipment must be evaluated to more stringent requirements than non-high/low pressure interfaces when considering fire induced spurious operations.
The purpose of the requirements is to ensure that a fire-induced LOCA, with a direct path for radioactive release to the atmosphere, does not occur. RG 1.189 further states that the approach outlined in Appendix C to NEI 00-01 provides an acceptable methodology for the determination of components as high/low pressure interface components, when applied in conjunction with the regulatory guide.Appendix C, "High/Low Pressure Interfaces," to NEI 00-01, Revision 2, dated May 2009 states that the PORV/PORV Block Valve pathway is not required to be treated as a high/low pressure interface because it does not meet the criteria of an interfacing LOCA outside of primary containment.
See excerpts from NEI 00-01 Appendix C below: Section C.2, "Introduction," states in part: "10 CFR 50 Appendix R analyses must evaluate the potential for spurious operations that may adversely affect the ability to achieve and maintain safe shutdown.
A subset of components considered for spurious operation involves reactor coolant pressure boundary (RCPB) components whose spurious operation can lead to an unacceptable loss of reactor pressure vessel/Reactor Coolant System (RPV/RCS) inventory via an interfacing system loss of coolant accident (ISLOCA).
Because an ISLOCA is a significant transient, it may be beyond the capability of a given safe shutdown path to mitigate.
As a result of this concern, selected RCPB valves are defined as high/low pressure interface valve components requiring special consideration and criteria." Section C.3, "Identifying High/Low Pressure Interface Components," states in part -"...the following criterion is established to determine if a RCPB valve is considered a high/low pressure interface valve component:
A valve whose spurious opening could result in a loss of RPV/RCS inventory and, due to the lower pressure rating on the downstream piping, an interfacing LOCA outside of Primary Containment (i.e., pipe rupture in the low pressure piping).Although spurious relief valve operations do result in a loss of RPV/RCS inventory, the down stream piping is designed for the discharge pressures.
As a result, spurious relief valve operation will not result in an interfacing system LOCA and spurious relief valve operation is not considered to be a high low pressure interface condition.
Similarly, a PORV/PORV Block Valve pathway that could result in a similar fluid path loss concern is not considered to be a high/low pressure interface condition.
Even Attachment I to ET 10-0013 Page 6 of 9 though this PORV/PORV Block Valve pathway could result in the eventual rupture of the pressurizer relief tank rupture disk, this is not a piping system failure and, as such, does not require the PORV/PORV Block Valve pathway to be treated as a high/low pressure interface." The spurious opening of the pressurizer PORVs and their associated block valves at WCGS would not result in an interfacing LOCA outside of primary containment.
Additionally, the above excerpts from Appendix C to NEI 00-01 clearly state that the PORVs and their associated block valves are not high/low pressure interface components.
Based on the above discussion, it is apparent that high/low pressure interface concerns only apply to systems that connect to the RCS and penetrate the containment.
The discharge piping of the PORVs and associated block valves at WCGS do not penetrate containment, and therefore can be reclassified as non-high/low pressure interface components.
WCGS License Condition 2.C.(5)(b) allows the Station to make changes to the fire protection program provided PFSSD is not adversely affected.
Removal of the high/low pressure interface designation for the pressurizer PORVs and associated block valves and invoking less stringent PFSSD analysis criteria is considered a reduction in the PFSSD analysis methodology contained in E-1F9910, Rev. 6, "Post-Fire Safe Shutdown Fire Area Analysis," when applying License Condition 2.C(5)(b).
The referenced license condition allows WCNOC to make changes to the approved fire protection program without prior NRC approval only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Therefore, even though regulatory guidance clearly reflects that the PORVs and their associated block valves are not high/low pressure interface components, considering the above discussion, WCNOC has conservatively chosen to request NRC review and approval of the proposed change to remove the high/low pressure interface designation from the pressurizer PORVs and associated block valves.4. REGULATORY EVALUATION 4.1 Applicable Regulatory Reguirements/Criteria 10 CFR 50, Section 48, Fire Protection, which states in paragraph (a)(1) that "Each holder of an operating license issued under this part or a combined license issued under part 52 of this chapter must have a fire protection plan that satisfies Criterion 3 of appendix A to this part." 10 CFR 50, Appendix A, General Design Criterion 3, Fire Protection, requires that structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions.
10 CFR 50, Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979", applies to licensed nuclear power electric generating stations that were operating prior to January 1, 1979. Since WCGS was licensed after January 1, 1979, WCNOC is not obligated to meet the requirements of 10 CFR 50, Appendix R. However, USAR Appendix 9.5E, provides a comparison of the WCGS design to Appendix R.
Attachment I to ET 10-0013 Page 7 of 9 In the "APPENDIX R QUESTIONS AND ANSWERS" Section of NRC Generic Letter 86-10,"Implementation of Fire Protection Requirements," Section 5.3.1, "Circuit Failure Modes," asks what circuit failure modes must be :considered in identifying circuits associated by spurious actuation.
The NRC response states in part: "Sections III.G.2 and III.L.7 of Appendix R define the circuit failure modes as hot shorts, open circuits, and shorts to ground....
For three-phase AC circuits, the probability of getting a hot short on all three phases in the proper sequence to cause spurious operation of a motor is considered sufficiently low as to not require evaluation except for any cases involving Hi/Lo pressure interfaces.
For ungrounded DC circuits, if it can be shown that only two hot shorts of the proper polarity without grounding could cause spurious operation, no further evaluation is necessary except for any cases involving Hi/Lo pressure interfaces." 4.2 Significant Hazards Consideration The proposed amendment would revise the Renewed Facility Operating License to deviate from certain Wolf Creek Generating Station (WCGS) Fire Protection Program requirements.
In the WCGS response to Nuclear Regulatory Commission (NRC) Question 280.5, documented on page 280-5 of the NRC Questions Section at the end of the WCGS Updated Safety Analysis Report (USAR), the pressurizer Power Operated Relief Valves (PORVs) and their associated block valves were classified as high/low pressure interfaces.
The reason for this amendment is to remove the high/low pressure interface designation from the PORVs and their associated block valves by revising the response to NRC Question 280.5, documented on page 280-5 of the NRC Questions Section at the end of the WCGS USAR.Wolf Creek Nuclear Operating Corporation (WCNOC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:
No.The design function of structures, systems and components are not impacted by the proposed change. This amendment classifies the pressurizer PORVs and their associated block valves based on the guidance in Regulatory Guide 1.189, "Fire Protection for Nuclear Power Plants," Revision 2, and Nuclear Energy Institue (NEI) 00-01, "Guidance for Post-Fire Safe-Shutdown Circuit Analysis," Revision 2, Appendix C. The classification change only affects the post fire safe shutdown (PFSSD) analysis methodology for the PORVs and block valves. Reclassification of the PORVs and block valves will not impact the use of the valves to depressurize the Reactor Coolant System (RCS) to recover from certain transients if normal pressurizer spray is not available.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Attachment I to ET 10-0013 Page 8 of 9 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evalu'ated?
-Response:
No.There are no changes in the method by which any safety related plant system performs its safety function and the normal manner of plant operation is unaffected.
No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this change. There will be no adverse effect or challenges imposed on any safety related system as a result of this change. The classification change only affects the PFSSD analysis methodology for the PORVs and block valves.Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Does the proposed amendment involve a significant reduction in a margin of safety?Response:
No.There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions.
There will be no impact on departure from nuclear boiling ration (DNBR) limits, heat flux hot channel factor (FQ(Z)) limits, nuclear enthalpy rise hot channel factor (FXH) limits, peak centerline temperature (PCT) limits, peak local power density or any other margin of safety.Therefore, the proposed change does not involve a significant reduction in a margin of safety.Based on the above, WCNOC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
4.3 Conclusions In conclusion, based on the considerations discussed above, (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.5. ENVIRONMENTAL CONSIDERATION WCNOC has evaluated the proposed changes and determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant Attachment I to ET 10-0013 Page 9 of 9 increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
: 6. REFERENCES
: 1. WASH-1400 (NUREG 75/014), Reactor Safety Study, An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants, dated October 1975 2. Generic Letter 80-14, LWR Primary Coolant System Pressure Isolation Valves, dated February 23, 1980 3. Generic Letter 86-10, Implementation of Fire Protection Requirements, dated April 24, 1986 4. NEI 00-01,Guidance for Post-Fire Safe-Shutdown Circuit Analysis, Rev. 2, dated May 2009 5. Regulatory Guide 1.189, Fire Protection Program for Nuclear Power Plants, Rev. 2, dated October 2009 6. WCNOC Letter WO 08-0028, Docket No. 50-482: Licensee Event Report 2008-009-00, Inadequate Compensatory Actions for a Fire Area, dated December 19, 2008 7. OFN KC-016, Fire Response, Rev. 23 8. E-1 F991 0, Post-Fire Safe Shutdown Fire Area Analysis, Rev. 6 Attachment II to ET 10-0013 Page 1 of 3 Markup of USAR Pages Attachment II to ET 10-0013 Page 2 of 3 WOLF CREEK No change made to this USAR page: containing the primary shutdown equipment.
For each cable so identified provide the results of an analysis that demonstrates that failure (open, ground, or hot short) of the associated cable will not adversely affect the alternate, dedicated, or remote method of shutdown.RESPONSE A discussion of safe shutdown and a list of systems necessary for safe shutdown are in Section 7.4. Section 7.4 also describes the capability of the auxiliary shutdown panel for safe shutdown from outside the control room.The final fire hazards analysis, USAR Appendix 9.5B, considers primary, alternate, and associated circuits and demonstrates that any single fire will not prevent the safe shutdown of the plant.Q280.5 The residual heat removal system is generally a low pressure system that interfaces with the high pressure primary coolant system. To preclude a LOCA through this interface, we require compliance with the recommendations of Branch Technical Position RSB 5-1. Thus, this interface most likely consists of two redundant and independent motor operated valves with diverse interlocks in accordance with Branch Technical Position ICSB 3. These two motor operated valves and their associated cable may be subject to a single fire hazard. It is our concern that this single fire could cause the two valves to open resulting in a fire-initiated LOCA through the subject high-low pressure system interface.
To assure that this interface and other high-low pressure interfaces are adequately protected from the effects of a single fire, we require the following information:
a) Identify each high-low pressure interface that uses redundant electrically controlled devices (such as two series motor operated valves) to isolate or preclude rupture of any primary coolant boundary.b) Identify each device's essential cabling (power and control) and describe the cable routing (by fire area) from source to termination.
280-4 Rev. 14 Attachment II to ET 10-0013 Page 3 of 3 WOLF CREEK c) Identify each location where the identified cables are separated by less than a wall having a three-hour fire rating from cables for the redundant device.d) For the areas identified in item 5c above (if any), provide the bases and justification as to the acceptability of the existing design or any proposed modifications.
RESPONSE The reactor coolant system high-low pressure interfaces that rely on redundant-electrically controlled devices for isolation include the RHR letdown isolation valves ard tvHp pr~ssuri ,ýgr p~wer- np~rat-pd rc~lj -L-~ -4ii~ -ncI ýoran The fire hazards analysis, Appendix 9.5B, demonstrates that no single credible fire could cause the spurious opening of these valves in a manner that would breach the primary coolant boundary.Q280.6 Notification of Appendix R to 10 CFR Part 50 as a Licensing Requirement.
Appendix R to 10 CFR Part 50 will also be used as guidance for our review of your fire protection program. Your compliance with the requirement set forth in Appendix R as modified by accepted exceptions will be made a license condition.
Identify any exceptions your program takes to the requirements of Appendix R as well as BTP ASB 9.5-1, and describe your alternative for providing an equivalent level of fire protection.
RESPONSE Table 9.5E-1 provides the requested comparisons and identifies the exceptions of the Wolf Creek Generating Station to 10 CFR 50 Appendix R. Table 9.5B-1 provides the WCGS Fire Protection comparisons to APCSB 9.5-1 Appendix A.280-5 Rev. 13 Attachment III to ET 10-0013 Page 1 of 1 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by WCNOC in this document.
Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.
Please direct questions regarding these commitments to Mr.Richard Flannigan at (620) 364-4117.REGULATORY COMMITMENTS Requlatory commitment Due Implement License Amendment Once approved, the amendment will be implemented within 90 days of receipt.The compensatory measure fire watch and OFN KC-016 mitigating Prior to startup from action for LER 2008-009-00 will not be lifted until LAR approval is Refueling Outage 18 received and design change package 012944 is implemented.
Revise the Updated Safety Analysis Report response to NRC In accordance with Question 280.5. 10 CFR 50.71(e).}}

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