ML110530183
| ML110530183 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 03/09/2011 |
| From: | Balwant Singal Plant Licensing Branch IV |
| To: | Matthew Sunseri Wolf Creek |
| Gibson, Lauren, NRR/DORL/LPL4, 415-1056 | |
| References | |
| TAC ME3766 | |
| Download: ML110530183 (19) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 9, 2011 Mr. Matthew W. Sunseri 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:
REMOVING HIGH/LOW PRESSURE DESIGNATION FROM THE PRESSURIZER POWER-OPERATED RELIEF VALVES (TAC NO. ME3766)
Dear Mr. Sunseri:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 193 to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). The amendment consists of changes to the WCGS Updated Safety Analysis Report in response to your application dated April 13, 2010, as supplemented by letters dated June 1,2010, and February 17, 2011.
The amendment revises the approved fire protection program, as described in the response to Question Q280.5 of the WCGS Updated Safety Analysis Report, by removing the high/low pressure interface designation of the pressurizer power-operated relief valves and their associated block valves. The amendment also revises license condition 2.C.(5)(a) to include the change approved by this amendment request.
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, b . . ,\.,~-\ t :S'V\.'J~
Balwant K. Singal, 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. 193 to NPF-42
- 2. Safety Evaluation cc w/encls: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 193 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)
Renewed Facility Operating License No. NPF-42 filed by the Wolf Creek Nuclear Operating Corporation (the Corporation), dated April 13, 2010, as supplemented by letters dated June 1, 2010, and February 17, 2011, 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.
Enclosure 1
-2
- 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraphs 2.C.(2) and 2.C.(5)(a) of Renewed Facility Operating License No. NPF-42 is hereby amended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 193, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(5) Fire Protection (Section 9.5.1, SER Section 9.5.1.8, SSER #5)
(a) The Operating Corporation shall maintain in effect all provisions of the approved fire protection program as described in the SNUPPs Final Safety Analysis Report for the facility through Revision 17, the Wolf Creek site addendum through Revision 15, as approved in the SER through Supplement 5, Amendment No. 191, and Amendment No. 193 subject to provisions band c below.
- 3. The license amendment is effective as of its date of issuance and shall be implemented within 90 days of the date of issuance. Consistent with the requirements in 10 CFR 50.71 (e), implementation shall include revision to the Wolf Creek Generating Station Updated Safety Analysis Report, Response to Question Q280.5, regarding the removal of the high/low pressure interface designation of the pressurizer power-operated relief valves and their associated block valves.
FOR THE NUCLEAR REGULATORY COMMISSION Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: l'vlarch 9, 2011
ATTACHMENT TO LICENSE AMENDMENT NO. 193 RENEWED FACILITY OPERATING LICENSE NO. NPF-42 DOCKET NO. 50-482 Replace the following pages of the Renewed Facility Operating License No. NPF-42 and Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.
The corresponding overleaf pages are provided to maintain document completeness.
Renewed Facility Operating License REMOVE INSERT 4 4 5 5 Technical Specifications REMOVE INSERT None None
4 (5) The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and (6) The Operating Corporation, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility, C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission, now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) Maximum Power Level The Operating Corporation is authorized to operate the facility at reactor core power levels not in excess of 3565 megawatts thermal (100% power) in accordance with the conditions specified herein.
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 193, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated in the license. The Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
(3) Antitrust Conditions Kansas Gas & Electric Company and Kansas City Power & Light Company shall comply with the antitrust conditions delineated in Appendix C to this license.
(4) Environmental Qualification (Section 3.11, SSER #4, Section 3.11, SSER #5)*
Deleted per Amendment No. 141.
- The parenthetical notation following the title of many license conditions denotes the section of the supporting Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
Renewed License No. NPF-42 Amendment No. 193
5 (5) Fire Protection (Section 9.5.1, SER, Section 9.5.1.8, SSER #5)
(a) The Operating Corporation shall maintain in effect all provisions of the approved fire protection program as described in the SNUPPs Final Safety Analysis Report for the facility through Revision 17, the Wolf Creek site addendum through Revision 15, as approved in the SER through Supplement 5, Amendment No. 191, and Amendment No. 193 subject to provisions band c below.
(b) The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
(c) Deleted.
(6) Qualification of Personnel (Section 13.1.2, SSER #5, Section 18, SSER tlll Deleted per Amendment No. 141.
(7) NUREG-0737 Supplement 1 Conditions (Section 22, SER)
Deleted per Amendment No. 141.
(8) Post-Fuel-Loading Initial Test Program (Section 14, SER Section 14, SSER #5)
Deleted per Amendment No. 141.
(9) Inservice Inspection Program (Sections 5.2.4 and 6.6, SER)
Deleted per Amendment No. 141.
(10) Emergency Planning Deleted per Amendment No. 141.
(11) Steam Generator Tube Rupture (Section 15.4.4, SSER #5)
Deleted per Amendment No. 141.
(12) LOCA Reanalysis (Section 15.3.7, SSER #5)
Deleted per Amendment No. 141.
Renewed License No. NPF-42 Amendment No. 193
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555"()001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 193 TO RENEWED 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 April 13, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession No, ML101100479), as supplemented by letters dated June 1,2010, and February 17, 2011 (ADAMS Accession Nos. ML101590671 and ML110550140, respectively),
Wolf Creek Nuclear Operating Corporation (WCNOC, the licensee) submitted a license amendment request (LAR) in which it requested changes to the approved fire protection program (FPP) for the Wolf Creek Generating Station (WCGS),
The proposed changes to the WCGS FPP would allow removal of the high/low pressure interface designation from the pressurizer power-operated relief valves (PORVs) and their associated block valves. The current WCGS Updated Safety Analysis Report (USAR),
Response to Question Q280.5, defines the PORVs and their associated downstream block valves as highllow pressure interfaces, The supplemental letter dated February 17, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U,S, Nuclear Regulatory Commission (NRC) staffs original proposed no significant hazards consideration determination as published in the Federal Register on June 29, 2010 (75 FR 37477). The supplemental letter dated June 1, 2010, was included in the original notice, 2,0 REGULATORY EVALUATION Regulatory requirements for fire protection are contained in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.48, "Fire protection," Appendix A to 10 CFR Part 50, General Design Criterion 3, "Fire protection," and Appendix R, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." to 10 CFR Part 50, The types of circuit failures to be addressed in this license amendment are specified in Section IILG.2 of Appendix R to 10 CFR Part 50 (1ILG.2).
Enclosure 2
- 2 The regulations in Appendix R to 10 CFR Part 50 apply to licensed nuclear power electric generating stations that were operating prior to January 1, 1979. Since WCGS was licensed after January 1, 1979, the licensee is not required to meet the requirements of III,G.2. However, the licensee committed to meeting the requirements of III.G.2 in its approved FPP, per License Condition 2.C.(5) and Appendix 9.5E of the WCGS USAR. Therefore, the WCGS FPP must provide the established level of protection as intended by III,G.2.
This LAR is seeking approval by the Commission, pursuant to License Condition 2.C.(5), to make changes to the approved FPP as described in the WCGS USAR.
3.0 TECHNICAL EVALUATION
The WCGS USAR classifies the pressurizer PORVs and their associated block (isolation) valves as high/low pressure interfaces. The technical evaluation was performed to determine if the high/low pressure interface designation were removed, the defense-in-depth measures would still provide reasonable assurance that a fire that does occur will be limited in severity and that there would be reasonable assurance that safe shutdown could be achieved.
3.1 Area Description The licensee's Fire Hazard Analysis 1 (FHA) provided the information for Tables 1, 2, and 3 below. All plant fire areas described in the following tables have interior wall and structural components that are constructed of steel, reinforced concrete, and other noncombustible materials. Interior wall surfaces are generally painted concrete masonry units or concrete.
Floor areas have also been coated. These paints and coatings have been considered in the fire hazards analysis.
Table 1 ~ Fire Area Description Fire Area Description A~8 Fire Area A-8 is the Auxiliary Building and includes the following:
corridors, filter, valve and demineralizer compartments, Sampling Room, Boron Meter and R.e. Activity Monitor Room, Volume I Control Tank Room, Containment Spray Additive Tank Area, Seal Water Heat Exchange Room, and the Exit Vestibule.
A~11 Fire Area A-11 is a single room cable chase.
A-16 Fire Area A-16 is the general corridor area in the Auxiliary Building.
South A 20-ft. combustible control zone separation area is located between columns A7 and A8 from the containment wall to the east wall of RHR Heat Exchanger Room 1309.
A-17 Fire Area A-17 is the South Electrical Penetration Room in the Auxiliary Building.
Wolf Creek Nuclear Operating Corporation document, "E-1F9905, Revision 3, Fire Hazard Analysis,"
dated June 19, 2009 (Not publicly available).
- 3 Fire Area Description A-18 Fire Area A-18 is the North Electrical Penetration Room in the Auxiliary Building.
A-27 Fire area A-27 is the Reactor Trip Switchgear Room in the Auxiliary Building.
C-18 Fire Area C-18 is a single room cable chase located in the Control I Building.
C-21 Fire Area C-21 is the Lower Cable Spreading Room in the Control I Building.
C-22 Fire Area C-22 is the Upper Cable Spreading Room in the Control Building.
C-23 Fire Area C-23 is a single room (3505) cable chase located in the Control Building.
C-24 Fire Area C-24 is a single room (3504) cable chase located in the I Control Building. i C-30 Fire Area C-30 is a single room (3617) cable chase located in elevation of the Control Building.
C-33 Fire Area C-33 is a single room (3804) cable chase located in the Control Building.
RB Fire Area RB consists of Fire Areas RB.1 through RB.11, which are separate fire areas and encompasses all elevations of the Reactor Building. However, for the purposes of the PFSSD [post-fire safe shutdown] analysis[2], the reactor building is treated as one fire area, which is designated as Fire Area RB.
As indicated in the WCNOC FHA, all of the fire areas in Table 1 are separated from all other fire areas by fire barrier construction and penetration closure assemblies (fire door, fire dampers, and penetration seals) that satisfy at least one of the following criteria:
- They have been directly qualified by 3-hour fire endurance testing in accordance with criteria of the applicable controlling standard (American Society for Testing and Materials (ASTM) E-119, "Standard Test Methods for Fire Tests of Building Construction and Materials" (walls, ceilings, floors); ASTM E-152, "Standard Methods of Fire Tests of Door Assemblies," Underwriters Laboratories (UL)-108, "Standard for Safety Fire Tests of Door Assemblies," and UL-555, "Standard for Fire Dampers" (fire dampers), or Institute of Electrical and Electronics Engineers (lEEE)-634, "Standard for Cable Penetration Fire Stop Qualification Test" (cable penetration fire stop test) for the respective barrier assembly protective feature.
- They have been evaluated as providing an equivalent level of protection.
2 WolfCreek Nuclear Operating Corporation document, E-1F9910, Revision 6, "Post-Fire Safe Shutdown Fire Area Analysis," July 31, 2009 (not publicly available).
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- They have been evaluated as providing protection commensurate with the fire hazards present.
3.2 Defense-In-Depth Review The regulations in 10 CFR 50, Appendix R, Section II.A, "Fire protection program," require that licensees ensure that the ability to achieve and maintain safe shutdown is preserved during and following a fire event by extending the concept of defense-in-depth:
(1) To prevent fires from starting; (2) To detect rapidly, control, and extinguish promptly those fires that do occur; (3) To provide protection for structures, systems, and components important to safety so that a fire that is not promptly extinguished by the fire suppression activities will not prevent the safe shutdown of the plant.
3.2.1 Fire Prevention In its supplemental letter dated June 1, 2010, the licensee stated that, Combustible materials in the areas where the PORV and block valve cables are routed are administratively controlled so as not to allow large quantities of unattended combustibles in the plant. No unattended combustible materials are allowed in containment during power operation.
Additionally, all cables in the plant that are routed in raceways with PORV cables are IEEE-383, "Standard for Qualifying Class 1E Electric Cables and Field Splices for Nuclear Power Generating Stations," rated thermoset cables.
The WCNOC FHA states that Safety-related cable in the general plant area is qualified to IEEE-383-1974. All single conductors inside control panels meet the flame resistance requirements of [Insulated Power Cable Engineers Association (lPCEA)] S-19-81 or S-61-402.
The FHA states that:
Lighting, fire protection, communication and specialty cables which are flame retardant, but not qualified to IEEE-383-1974; and other communication and specialty cable, such as cords and computer ribbon cable, are limited in use in the following manner:
- 1. Covered with a flame retardant coating per the requirements of [Branch Technical Position (BTP)] APCSB 9.5-1, Appendix A; or
- 2. Installed in a totally enclosed metal conduit system; or
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- 3. Consist of short lengths of exposed cable between the end of a totally enclosed metal conduit system routed to a component and the connection to the component (e.g. at light fixtures and public address devices); or
- 4. Located in non-safety-related areas which are separated from safety related areas by fire rated boundaries; or
- 5. Evaluated on a case-by-case basis for adverse impact on the fire protection program.
Table 2 - Additional Fire Prevention Information by Fire Area Fire Area Discussion A-8 The cumulative combustible loading classification for Fire Area A-8 is Low. Administrative controls ensure that only new cartridges for the Chemical and Volume Control System (CVCS) filters in quantities required for immediate use will be brought into this area and the containers are removed after they are emptied.
A-11 No credible in situ ignition sources are present within cable chases and introduction of a transient ignition source is administratively controlled.
A-16 South The cumulative combustible loading classification for Fire Area A-16 is Low. A 20-ft. combustible control zone separation area is administratively maintained to strictly control transient combustibles in the limiting separation area between redundant PFSSD circuits. In situ combustibles within the combustible control zone do not pose a fire propagation path that would disable both trains of redundant PFSSD equipment.
A-17 The cumulative combustible loading classification for Fire Area A-17 is Low. All cable in panels RJ160NB/CID meets the vertical flame requirements of Insulated Power Cable Engineers Association (IPCEA) S-19-81 and/or IEEE 383 for flame resistance. The panels are provided with key locks to control access to the panel interiors.
A-18 The cumulative combustible loading classification for Fire Area A-18 is Low. All cable in panels RJ159NB/C/D meets the vertical flame requirements of IPCEA S-19-81 and/or IEEE-383 for flame resistance. The panels are provided with key locks to control access to the panel interiors.
A-27 The cumulative combustible loading classification for Fire Area I A-27 is Low.
C-18 No credible ignition sources are present within cable chases and introduction of a transient ignition source is administratively controlled.
-6 Fire Area Discussion C-21 The cumulative combustible loading classification for Fire Area I C-21 is Low. The C-21 cable spreading room contains predominantly cable.
C-22 The cumulative combustible loading classification for Fire Area !
C-22 is Low. The C-22 cable spreading room contains predominantly cable.
C-23, C-24, No credible in situ ignition sources are present within cable C-30, C-33 chases and introduction of a transient ignition source is administratively controlled.
RB The cumulative combustible loading classification for Fire Area RB is Low. A system is provided to collect and contain lubricating oil for each reactor coolant pump (RCP).
3.2.2 Detection, Control and Extinguishment In its supplemental letter dated June 1, 2010, the licensee indicates that automatic fire suppression and detection is provided in all areas, outside containment, where PORV cables are routed. The cable trays of concern within containment have wire-type heat detection installed.
The WCGS USAR, Section 9.5.1.2.2, states that the powerblock fire protection system is designed in substantial compliance with the requirements of the American Nuclear Insurers (ANI) and the National Fire Codes of the National Fire Protection Association (NFPA). Codes and standards considered in the design of the fire protection system are listed in USAR Table 9.5.1-1. The key applicable NFPA detection and suppression codes from the table are NFPA 720-1975, "Proprietary Protective Signaling Systems for Guard, Fire Alarm and Supervisory Service," for fire detection, NFPA 12A-1973, "Halon 1301 Fire Extinguishing Systems," for Halon systems, and NFPA 13-1973, -1975, -1976, and -1991, "Standard for the Installation of Sprinkler Systems," for sprinkler systems.
USAR Table 9.5E-1 states that the fire brigade compliment, training, medical qualification, and personal protective equipment complies with 10 CFR 50 Appendix R, Section H, "Fire brigade."
The WCGS USAR Section 9.5.1.7.5.1.2.2, "Fire Brigade," also states, in part, that, The WCGS Fire Brigade is composed of a minimum of five persons from the on duty work force. The Brigade does not include any of the plant physical security personnel. ..
-7 Per the WCNOC FHA, all of the fire areas listed in Table 3 below have fire extinguishers and hose stations provided in the general area for manual fire fighting with the exception of fire areas A-17 and A-18 which have hose stations only.
Table 3 - Additional Detection, Control and Extinguishment Information Fire Area Detection, Control and Extinguishment Capability A-8 An automatic smoke detection system is installed throughout the fire area except in several rooms with low combustible loading and no safe shutdown circuits or equipment. An automatic preaction sprinkler system is installed over cable tray concentrations in Rooms 1301,1314, and 1320.
A-11 Total coverage automatic fire detection and suppression are installed in this area. A wet pipe sprinkler system is provided for the area.
A-16 South Total coverage automatic smoke detection is provided in this area. A pre-action sprinkler system is located in areas with high concentrations of cable trays. In addition to the ceiling level sprinkler system, the east corridor, from column line A 1 to A4, is provided with an intermediate level sprinkler system, located below the lowest cable tray in the corridor.
A-17, A-18 These fire areas have automatic fire detection and a fixed and A-27 automatic Halon fire suppression system installed. The detection and suppression systems are both total coverage systems.
C-18, C-21, All fire areas have total coverage automatic fire detection and C-22, C-23, suppression installed with the exception of C-21 (Lower Cable C-24, C-30, Spreading Room) and C-22 (Upper Cable Spreading Room) and C-33 having partial automatic suppression.
Both cable spreading rooms have an automatic preaction sprinkler system installed at the ceiling. The location of the closed sprinkler heads considers cable tray sizing and arrangement and cables are designed to allow wetting down with suppression water without electrical faulting in accordance with USAR Table 9.5A-1 comparison to APCSB 9.5-1, Appendix A, Section F.3 (Cable Spreading Room).
- 8 Fire Area Detection, Control and Extinguishment Capability RB Linear heat detection is installed above each reactor coolant pump and in areas where cable trays are concentrated. Duct smoke detection is provided for each containment cooler. The Containment atmospheric control filter adsorber units are provided with a thermistor type continuous detector. The above fire detectors are alarmed in the control room and the alarms are zoned for quick identification of the area in alarm. Manual-pull fire alarm stations are located near hose stations and alarm locally and in the Control Room.
A fixed, manually charged, closed head sprinkler system is provided over the cable trays in Zones RB-3 and RB-4. Portable extinguishers and manual hose stations are also installed. The hose stations are spaced at no more than 100-foot intervals. The hose station locations are such that all accessible areas of the Reactor Building are adequately covered by at least one hose stream.
3.2.3 Preservation of Safe Shutdown Capability In its letter dated June 1, 2010, the licensee states that combustible materials in the areas where the PORVs and block valves cables are routed are administratively controlled. The licensee also states that, In all of the fire areas outside containment where the PORV cables are routed, the cables are routed in raceway that is protected by automatic fire suppression and detection. Within containment, the cable trays of concern have protectowire
[wire-type] heat detection installed.
With the existing fire prevention controls, the detection and suppression in the areas where PORV cables are routed, a fire damaging PORV cables is unlikely, and if a damaging fire were to occur it would be rapidly annunciated to the control room and the fire brigade dispatched.
All cables in raceway with PORV cables are IEEE Standard-383 rated thermoset (TS) cables and thermoset cables are not considered vulnerable to self-ignited cable fires. Section 9.1.1 of NUREG/CR-6931, "Cable Response to Live Fire (CAROLFIRE) Volume 1: Test Descriptions and Analysis of Circuit Response Data," April 2008 (not publicly available), states that, Inter-cable shorting between two TS-insulated cables that could cause hot shorts and the spurious actuation of plant equipment was found to be a plausible failure mode, although the likelihood of this failure mode is low in comparison to intra cable short circuits leading to spurious operation. While no detailed statistical analysis has been performed, it appears that the conditional probability (given cable failure) of spurious actuations arising from this specific failure mode is .
small in comparison to that previously estimated for spurious actuations from intra-cable shorting.
- 9 In response to the NRC staff's request for information dated May 20, 2010 (ADAMS Accession No. ML101370091), the licensee stated in its letter dated June 1, 2010, that:
The design change that WCNOC will implement during Refueling Outage 18 is based on the assumption that the license amendment request will be approved.
This modification will re-wire the control room hand switch so that both the negative and positive sides of the auxiliary relay (AR) are de-energized when the operator places the hand switch in the Closed position. De-energizing both polarities ensures a single proper polarity hot short will not prevent operators from de-energizing the AR using the hand switch in the control room. The operation of the hand switch will remain un-changed and the hand switch itself will not be changed.
The PORVs are designed to close on a loss of electrical power. Therefore, if the cables with the hot shorts lose power due to fire damage or the fire-induced opening of breakers or fuses, the PORVs should close, as designed. Therefore, only a limited number of hot short failure modes can cause a loss of control of the PORVs.
The only fire-induced failure that could circumvent the proposed plant modifications is two simultaneous independent proper polarity (positive-to-positive and negative-to-negative) inter cable hot shorts. As stated above, the only reason for WCGS to assume these inter-cable hot shorts is the high-low pressure designation of the PORVs. If the designation were to remain, WCGS would be required to postulate two simultaneous cable failures modes, which would necessitate operator manual actions from outside the control room to remove power from the PORVs and other plant equipment that have 125 Volts direct current cables in the same raceway with PORV cables. To preemptively remove power from other plant equipment that may not be damaged by the fire could reduce the equipment and systems available to the operators for post-fire safe shutdown.
In the unlikely event of a fire damaging PORV control cables and the inter-cable positive-to positive and negative-to-negative combination of hot shorts occur, a PORV may open and remain open until the hot shorts clear. The discharged steam from the PORV is piped to the pressurizer relief tank (PRT), inside containment, where it is cooled and condensed by mixing with water. The piping downstream of the pressurizer PORVs and block valves will not rupture upon opening of the PORVs because it is designed for the expected pressure. Unless manual actions are taken to close the PORVs or the hot shorts clear, the steam will continue to be contained within the PRT until the pressure reaches a level that the PRT rupture disk ruptures, as designed. This will not lead to an outside containment loss of coolant accident.
3.3 NRC Staff Conclusion
s The proposed LAR would allow WCGS to remove the highllow pressure interface designation from the PORVs and their associated block valves. Based on the discussion above, the following defense-in-depth measures are in place to prevent damage to the PORV control cables: 1) combustible materials in the areas where the PORVs and related block valves cables are routed are administratively controlled; 2) automatic fire suppression and detection is provided in all of the areas outside containment, where PORV cables raceways are routed;
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- 3) within containment, the cable concentrations have wire-type heat detection installed and no unattended combustible materials are allowed in containment during power operation.
The proposed modifications would allow operating a control room switch to prevent or mitigate a fire-induced spurious opening of the PORVs. The only scenario that would remain that could cause the PORVs to open would involve two simultaneous inter-cable proper polarity cable failures. To mitigate this scenario, the licensee would have to remove power from the PORVs and other plant equipment. The operator manual action to remove power from other plant equipment to prevent or mitigate the results of the analyzed two simultaneous cable failures could have an adverse affect on PFSSD. Changing the high-low pressure commitments will allow the licensee to analyze PORV cables for a single proper polarity hot short between two cables (inter-cable) in combination with any number of internal cable (intra-cable) hot shorts.
Research has shown that the likelihood of inter-cable hot shorts that can cause spurious equipment operation is low in comparison to intra-cable shorts that can cause spurious equipment operation. The proposed modifications and control room operator actions provide an adequate level of protection as described in Section III.G.2 of Appendix R to 10 CFR Part 50.
Based on the above, the proposed change would revise the Renewed Facility Operating License to deviate from certain WCGS FPP commitments. The WCGS USAR classifies the pressurizer PORVs and their associated block (isolation) valves as highllow pressure interfaces.
The removal of the PORVs and block valves as highllow pressure interface components is a reduction in the PFSSD analysis methodology contained in WCNOC's PFSSD analysis.
However, the defense-in-depth measures provide reasonable assurance that a fire that does occur will be limited in severity and that there is reasonable assurance that safe shutdown can be achieved.
3.4 Changes to License Condition In its letter dated June 1, 2010, the licensee states:
The Final Safety Analysis Report (FSAR), Appendix 9.5B, contained the original fire hazards analysis for the station. Each fire area analysis contained a summary of the safe shutdown capability for a fire in that area. The safe shutdown summary was based on the original PFSSDA [post-fire safe shutdown analysis] prepared during plant design. The following is stated in each of the fire areas where one or both PORVs could spuriously open:
.... However, should the PORV fail open and should the block valve fail as-is in the open position, the RCS [reactor coolant system] would blow down to the PRT. In this case, the control room operator would place the PORV in manual and close the PORV from the control room.
The licensee confirmed the proposed action to re-classify the pressurizer PORV and implement the LAR modifications would effectively make the safe shutdown summary statement in each fire area in the original Final Safety Analysis Report true. The operation of the PORVs will not change. The 'changes will not affect the safety-related function of the PORVs.
- 11 In its letter dated February 17, 2011, the licensee proposed the following revised license condition to include deviation approved by this amendment request:
(5) Fire Protection (Section 9.5.1, SER, Section 9.5.1.8, SSER #5)
(a) The Operating Corporation shall maintain in effect all provisions of the approved fire protection program as described in the SNUPPs Final Safety Analysis Report for the facility through Revision 17, the Wolf Creek site addendum through Revision 15, as approved in the SER through Supplement 5, Amendment No. 191, and Amendment No. XXX [193J subject to provisions band c below.
The NRC staff reviewed the revised license condition and concluded that the change is acceptable.
4.0 REGULATORY COMMITMENTS In its letter dated April 13, 2010, the licensee made the following regulatory commitments:
The compensatory measure fire watch and OFN KC-016 mitigating action for LER 2008-009-00 will not be lifted until LAR approval is received and design change package 012944 is implemented; to be completed prior to the startup from Refueling Outage 18.
Revise the Updated Safety Analysis Report response to NRC Question 280.5; in accordance with 10 CFR 50.71(e).
The NRC staff considers the above to be regulatory commitments and acceptable.
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 29, 2010 (75 FR 37477). 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.
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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.
Principal Contributors: G. Cooper D. Frumkin Date: March 9, 2011
March 9, 2011 Mr. Matthew W. Sunseri 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:
REMOVING HIGH/LOW PRESSURE DESIGNATION FROM THE PRESSURIZER POWER-OPERATED RELIEF VALVES (TAC NO. ME3766)
Dear Mr. Sunseri:
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 193 to Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). The amendment consists of changes to the WCGS Updated Safety Analysis Report in response to your application dated April 13, 2010, as supplemented by letters dated June 1, 2010, and February 17, 2011.
The amendment revises the approved fire protection program, as described in the response to Question Q280.5 of the WCGS Updated Safety Analysis Report, by removing the high/low pressure interface designation of the pressurizer power-operated relief valves and their associated block valves. The amendment also revises license condition 2.C.(5)(a) to include the change approved by this amendment request.
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/
Balwant K. Singal, 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. 193 to NPF-42
- 2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:
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