IR 05000482/2002008
| ML023250239 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 11/21/2002 |
| From: | Marschall C NRC/RGN-IV/DRS/EMB |
| To: | Maynard O Wolf Creek |
| References | |
| FOIA/PA-2003-0358 IR-02-008 | |
| Download: ML023250239 (17) | |
Text
ber 21, 2002
SUBJECT:
WOLF CREEK GENERATING STATION - NRC TRIENNIAL FIRE PROTECTION INSPECTION REPORT 50-482/02-08
Dear Mr. Maynard:
On October 11, 2002, the NRC completed the onsite portion of the subject inspection at your Wolf Creek Generating Station. Review of additional documentation provided to the NRC subsequent to the onsite inspection was performed in the Region IV office October 14 -25, 2002. The enclosed report documents the inspection findings which were discussed with Mr. W. B. Norton, Director, Engineering, and other members of your management and staff on November 4, 2002.
This inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
Within these areas, the inspection consisted of selected examination of procedures and representative records, observations of activities, and interviews with personnel.
Based on the results of this inspection, the NRC has identified one issue that was evaluated under the risk significance determination process as having very low safety significance (green). The NRC has also determined that a violation was associated with this issue. This violation is being treated as a noncited violation, consistent with Section VI.A of the Enforcement Policy. The noncited violation is described in the subject inspection report. If you contest the violation or significance of the noncited violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Wolf Creek Generating Station facility.
Wolf Creek Nuclear Operating Corporation -2-In accordance with 10 CFR 2.790 of the NRCs "Rules of Practice," a copy of this letter, its enclosure, and your response will be made available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Charles S. Marschall, Chief Engineering and Maintenance Branch Division of Reactor Safety Docket: 50-482 License: NPF-42
Enclosure:
NRC Inspection Report 50-482/02-08
REGION IV==
Docket: 50-482 License: NPF-42 Report No.: 50-482/02-08 Licensee: Wolf Creek Nuclear Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane, NE Burlington, Kansas Dates: September 23 - 27 and October 7 - 11, 2002 Team Leader R. L. Nease, Senior Reactor Inspector Engineering and Maintenance Branch Inspectors: R. P. Mullikin, Senior Reactor Inspector Engineering and Maintenance Branch M. F. Runyan, Senior Reactor Inspector Engineering and Maintenance Branch F. L. Brush, Senior Resident Inspector Project Branch B J. Cruz, Resident Inspector Project Branch B Accompanying B. Henderson, Reactor Inspector Personnel: Engineering and Maintenance Branch Approved By: Charles S. Marschall, Chief Engineering and Maintenance Branch Division of Reactor Safety
-2-SUMMARY OF FINDINGS IR05000482-02-08; on 09/23-27/2002 and 10/7-11/2002; Wolf Creek Nuclear Operating Corporation; Wolf Creek Generating Station; Triennial Fire Protection Inspection.
The inspection was conducted by a team of three regional inspectors, one senior resident inspector, and one accompanying NRC Region IV employee. The inspection identified one green finding, which was a violation of NRC regulatory requirements. The significance of most findings is indicated by their color (green, white, yellow, red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be "green" or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," revision dated July 2000.
Cornerstone: Mitigating Systems C Green. The team identified a noncited violation of Technical Specification 5.4.1 for the failure to provide an adequate procedure for ensuring the safe shutdown of the reactor in the event of a fire in the control room that requires control room evacuation.
Procedure OFN RP-17, "Control Room Evacuation, Revision 17, was inadequate because certain operator actions specified in Attachment C to the procedure could not be performed within the required time. The licensee entered this finding into their corrective action program as Performance Improvement Request 2002-2393.
This finding was of greater than minor significance because it impacted the mitigating systems cornerstone. This resulted from the issue's potential to affect the licensee's capability to safely shutdown the reactor in response to a fire in the control room requiring control room evacuation and remote shutdown. For fire protection findings, the Phase 1 screening worksheet in Manual Chapter 0609, Appendix A, refers fire protection findings to Manual Chapter 0609, Appendix F, for significance evaluation.
Using the significance determination process described in Appendix F, this finding was determined to be of very low safety significance, due to the licensee's demonstration that operators would have performed the most time-critical step (to isolate the power-operated relief valves) in time to prevent core damage (Section 1R05.4).
Report Details 1. REACTOR SAFETY 1R05 Fire Protection The purpose of this inspection was to review the Wolf Creek Generating Station fire protection program for selected risk significant fire areas. Emphasis was placed on verification of the licensees post-fire safe shutdown capability. The inspection was performed in accordance with the new Nuclear Regulatory Commission (NRC) reactor oversight process using a risk-informed approach for selecting the fire areas and attributes to be inspected. The team used the "Wolf Creek Generating Station Individual Plant Examination of External Events," dated July 7, 1995, to choose several risk-significant areas for detailed inspection and review. The fire areas chosen for review during this inspection were:
- A-16, general area, 2026' elevation of the auxiliary building
- A-18, north penetration room, 2026' elevation of the auxiliary building
- C-9, north switchgear room, control building
- C-10, south switchgear room, control building
- C-35, corridor, 2015' elevation, control building For each of the selected fire areas, the team focused the inspection on the fire protection features and on the systems and equipment necessary for the licensee to achieve and maintain safe shutdown conditions in the event of a fire in those fire areas.
.1 Systems Required to Achieve and Maintain Post-Fire Safe Shutdown a. Inspection Scope The team reviewed piping and instrumentation diagrams and the list of safe shutdown equipment documented in the licensee's post-fire safe shutdown analysis to verify whether their shutdown methodology had properly identified the components and systems necessary to achieve and maintain safe shutdown conditions for equipment in the fire areas selected for review. The team focused on the following functions that must be ensured to achieve and maintain post-fire safe shutdown conditions.
- Reactivity control capable of achieving and maintaining cold shutdown reactivity conditions
- Reactor coolant makeup capable of maintaining the reactor coolant inventory
- Reactor heat removal capable of achieving and maintaining decay heat removal
- Supporting systems capable of providing all other services necessary to permit extended operation of equipment necessary to achieve and maintain hot shutdown conditions
-2-A review was also conducted to ensure that all required electrical components in the selected systems were included in the licensees safe shutdown analysis. The team identified the systems required for each of the primary safety functions necessary to shut down the reactor. These systems were then evaluated to identify the systems that interfaced with the fire areas inspected and were the most risk significant for reaching both hot and cold shutdown. The following systems were selected for review.
- Reactor coolant pump seal injection
- Component cooling water system
- Essential service water system
- High pressure coolant injection system Documents reviewed by the team are listed in the attachment.
b. Findings No findings of significance were identified.
.2 Fire Protection of Safe Shutdown Capability a. Inspection Scope The team reviewed licensee documentation to verify that at least one post-fire safe shutdown success path was free of fire damage in the event of a fire in the selected fire areas. Specifically, the team examined the separation of safe shutdown cables, equipment, and components within the same fire areas. The team reviewed the licensee's methodology for meeting the requirements of 10 CFR 50.48, and the bases for the NRC's acceptance of this methodology as documented in NRC safety evaluation reports. In addition, the team reviewed license documentation, such as the Wolf Creek Updated Final Safety Evaluation Report, submittals made to the NRC by the licensee in support of the NRC's review of their fire protection program, and deviations from NRC regulations to verify that the licensee met license commitments. Documents reviewed by the team are listed in the attachment.
b. Findings No findings of significance were identified.
.3 Post-fire Safe Shutdown Circuit Analysis a. Inspection Scope On a sample basis, the team verified that cables of equipment required to achieve and maintain hot shutdown conditions in the event of fire in selected fire areas had been properly identified and either adequately protected from the potentially adverse effects of fire damage or analyzed to show that fire-induced faults (e.g., hot shorts, open circuits, and shorts to ground) would not prevent safe shutdown. During the inspection, a sample of redundant components associated with systems required to achieve and
-3-maintain hot shutdown conditions were selected for review. The sample included components associated with the auxiliary feedwater system, component cooling water system, and the essential service water system. From this list of components, the team reviewed cable routing data depicting the routing of power and control cables associated with each of the selected components. Additionally, on a sample basis the team verified the adequacy of electrical protective device coordination (e.g., circuit breaker, fuse, relay), for cables of equipment required to achieve and maintain safe shutdown conditions. Documents reviewed by the team are listed in the Attachment.
b. Findings No findings of significance were identified.
.4 Alternative Safe Shutdown Capability and Implementation a. Inspection Scope The team reviewed the systems required to achieve alternative safe shutdown to determine if the licensee had properly identified the components and systems necessary to achieve and maintain safe shutdown conditions from remote shutdown panel and alternate shutdown locations. The team also focused on the adequacy of the systems to perform reactor pressure control, reactor makeup, decay heat removal, process monitoring, and support system functions. The team reviewed Procedure OFN RP-17, "Control Room Evacuation, Revision 17, which would be used by operators to shut down the reactor in the event of a control room fire that required evacuation of the control room. The team also timed licensed and non-licensed personnel as they stepped through the procedure, to determine its adequacy to direct safe shutdown from remote shutdown locations. Documents reviewed by the team are listed in the Attachment.
b. Findings Introduction: The team identified a noncited violation of Technical Specification 5.4.1 for failure to provide an adequate procedure for operators to use in shutting down the plant from outside the control room. Procedure OFN RP-17, "Control Room Evacuation, was inadequate because operators could not complete Attachment C of the procedure within the times specified by Letter SLNRC 84-0109, dated August 23, 1984. Failure to complete the actions within the times specified could adversely affect the ability to achieve and maintain safe shutdown conditions. The team determined that the violation was of very low safety significance (green).
Description: In a field walkdown of the licensee's alternative shutdown Procedure OFN RP-17, "Control Room Evacuation, the team found that operators could not perform certain actions within the times specified in Letter SLNRC 84-0109, dated August 23, 1984. Specifically, Attachment C required the reactor operator to perform the first 18 steps within the Phase A time frame. Phase A, as defined in Letter
-4-SLNRC 84-0109, is 5 minutes. During the walkdown of Procedure OFN RP-17 with a reactor operator, the team noted that it took him approximately 14 minutes to complete these 18 steps. Upon learning of the teams timing concern, the licensee reperformed these steps several times using other operators. The shortest time the operators performed the steps was approximately 7 1/2 minutes.
Risk Analysis: The team determined that the issue was of greater than minor significance because it affects the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of the system that responds to the event to prevent undesirable consequences. The issue had the potential to affect the ability of the licensee to safely shut down the reactor in response to a control room fire requiring control room evacuation and remote shutdown.
For fire protection findings, the Phase 1 Screening Worksheet in Manual Chapter 0609, Appendix A, refers fire protection findings to Manual Chapter 0609, Appendix F, for significance evaluation. The Region IV senior reactor analyst and the team leader evaluated the risk using the NRC's Significance Determination Process described in Manual Chapter 0609, Appendix F. The most time-critical actions specified in Attachment C of Procedure OFN RP-17 were to verify that the power-operated relief valves (PORVs) were closed, and to isolate power to ensure they remained closed.
These actions were required to ensure the pressurizer level did not go below the indicating band, thus, ensuring the core remained covered. Although the licensee was not successful in performing the first 18 steps of Attachment C within 5 minutes, they did demonstrate that operators would have isolated the PORVs in less than 5 minutes. In subsequent discussions with the licensee, the Region IV senior risk analyst learned that with both PORVs open, uncovering the core would begin in 6-9 minutes. Core damage would commence at a 40 percent void fraction, which would occur in approximately 12 minutes. The team leader and the Region IV senior reactor analyst determined that the risk of this finding was very low (green) because the actions to ensure both PORVs were closed could have been performed in time to prevent core damage.
Enforcement: Technical Specification 5.4.1 requires, in part, that the licensee establish procedures for implementing activities covering fire protection program implementation.
The licensee failed to provide an adequate procedure for ensuring the safe shutdown of the reactor in the event of a fire in the control room that requires control room evacuation. Specifically, Attachment C of Procedure OFN RP-17, "Control Room Evacuation, did not ensure that operators could perform time-critical actions within the times required. This is a violation of Technical Specification 5.4.1. This violation is being treated as a noncited violation consistent with Section VI.A of the NRC Enforcement Policy (NCV 50-482/0208-01).
Upon identification of this finding, the licensee implemented interim actions to make an additional operator available to perform some of the actions in Attachment C of Procedure OFN RP-17. With the additional operator, the first 18 steps in Attachment C (including isolation of the PORVs) could be performed in less than 5 minutes. As
-5-long-term corrective actions, the licensee planed to evaluate the basis for the time requirement, and to revise the procedure to ensure time-critical actions are performed in order to meet those time requirements. The team considered these corrective measures acceptable. The licensee entered this finding into their corrective action program as Performance Improvement Request 2002-2393.
.5 Emergency Communications a. Inspection Scope The team reviewed the adequacy of the communication system to support plant personnel in the performance of alternative safe shutdown functions and fire brigade duties. The team verified that adequate communication equipment was available consistent with the licensing basis. The team performed a review of the electrical power supplies and cable routing for the radio repeater system, phone system, and plant paging system. Documents reviewed by the team are listed in the attachment.
b. Findings No findings of significance were identified.
.6 Emergency Lighting a. Inspection Scope The team reviewed the emergency lighting system required for safe shutdown activities to verify that it was adequate for supporting the performance of manual actions required to achieve and maintain hot shutdown conditions, and for illuminating access and egress routes to the areas where manual actions are required.
b. Findings No findings of significance were identified.
.7 Cold Shutdown Repairs a. Inspection Scope The team reviewed licensee Procedure OFN RP-014, Hot Standby to Cold Shutdown From Outside the Control Room, Revision 5, to determine whether repairs were required to achieve cold shutdown and to verify that the required material was available.
The team verified that the licensee had pre-staged the necessary equipment as required by procedure.
b. Findings No findings of significance were identified.
-6-
.8 Fire Protection Systems, Features, and Equipment a. For the selected fire areas, the team evaluated the adequacy of fire protection features, such as fire suppression and detection systems, fire area barriers, penetration seals, and fire doors. To do this, the team observed the material condition and configuration of the installed fire detection and suppression systems, fire barriers, and construction details and supporting fire tests for the installed fire barriers. In addition, the team reviewed license documentation, such as NRC safety evaluation reports and deviations from NRC regulations and the National Fire Protection Association code to verify that fire protection features met license commitments. Documents reviewed by the team are listed in the attachment.
b. Findings No findings of significance were identified.
.9 Compensatory Measures a. Inspection Scope The team verified, by sampling, that adequate compensatory measures were put in place by the licensee for out-of-service, degraded, or inoperable fire protection and post-fire safe shutdown equipment, systems or features (e.g., detection and suppression systems, or passive fire barrier features). Documents reviewed by the team are listed in the attachment.
b. Findings No findings of significance were identified.
4OA6 Meetings, including Exit On, October 11, 2002, at the conclusion of the teams onsite inspection, the team leader debriefed Mr. O. L. Maynard, Chief Executive Officer, Mr. G. B. Fader, Vice President, Technical Services, and other licensee management and staff members on the triennial fire protection inspection results.
On November 4, 2002, the team leader conducted a telephone exit meeting with Mr. W. B. Norton, Director, Engineering, and other licensee management and staff members, during which the results of the triennial fire protection inspection were characterized.
The licensee was asked whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
ATTACHMENT KEY POINTS OF CONTACT Licensee D. Dixon, Design Engineering S. Fellers, Engineer, Licensing P. Herrman, Fire Protection Engineer B. Ketchum, Probabilistic Risk/Safety Analyst B. Masters, Supervisor, Electrical/I&C Engineering B. Selby, Post-fire Safe Shutdown Engineer L. Ratzlaff, Supervisor, Fire Protection J. Suter, Fire Protection Engineer J. Yunk, Supervisor, Licensing NRC D. Loveless, Senior Reactor Analyst T. Pruett, Senior Reactor Analyst P. Qualls, Office of Nuclear Reactor Regulation ITEM OPENED AND CLOSED Opened and Closed 50-482/0208-01 NCV Inadequate alternative shutdown procedure (Section 1R05.4)
DOCUMENTS REVIEWED The following documents were selected and reviewed by the inspectors to accomplish the objectives and scope of the inspection and to support any findings:
CALCULATIONS Number Title Revision/Date H-09-004-CN001 The Unit Auxiliary Transformer (XMA02) has been 3/17/01 rewired due to damage.
-2-DRAWINGS Drawing Number Description/Title Revision 10466-A-1801 Architectural Fire Delineation Floor Plan, EL. 1974-0" 7 10466-A-1802 Architectural Fire Delineation Floor Plan, EL. 2000-0" 10 10466-A-1803 Architectural Fire Delineation Floor Plan, EL. 2026-0" 3 10466-A-1804 Architectural Fire Delineation Floor Plan, EL. 2047-6" 6 E-1991 Single Line Diagram - Site Area Loads M E-01024(Q) Time-Current Characteristic Curves 0 Sheets 4 - 15 E-1F1301 Fire Detection/Protection System - Auxiliary Building EL. 4 2000-0" E-11NB01 Lower Medium Voltage Sys. Class 1E 4.16 KV Single Line 0 Meter and Relay Diagram E-11NB02 Lower Medium Voltage Sys. Class 1E 4.16 KV Single Line 0 Meter and Relay Diagram E-13QF02 Schematic Diagram - Radio Communications Antenna 4 System E-13QF03 Schematic Diagram - Radio Communications Antenna 2 System KD-7496 One Line Diagram 25 M-1G020 Equipment Location - Reactor and Auxiliary Bldgs. Plan - 9 Basement EL. 1974-0" M-1G021 Equipment Location - Auxiliary Building Partial Plan 2 EL. 1988-0" & EL. 2013-6" M-1G022 Equipment Locations - Reactor and Auxiliary Buildings 11 Ground Floor EL. 2000-0" M-1G023 Equipment Locations - Reactor and Auxiliary Buildings 7 Plan EL. 2026-0" M-1G024 Equipment Location - Reactor and Auxiliary Buildings Plan 14 Operating Floor EL. 2047-6" M-1G025 Equipment Locations - Reactor and Auxiliary Buildings 4 Plan EL. 2068-8" M-1G051 Equipment Locations - Control & Diesel Gens. Bldgs. & 8 Comm. Corridor Plan EL. 2000-0" & EL. 2016-0"
-3-Drawing Number Description/Title Revision M-1G052 Equipment Locations - Control & Diesel Gens. Bldgs. & 4 Comm. Corridor Plan EL. 2032-0" & EL. 2047-6" M-1Y1902C Penetration Closure Schedule Aux. Bldg 0 M-663-00017 Penetration Seal Detail E-2A - A Three Hour Fire Rated W17 Foam Seal for a Cable Tray Through Concrete or Concrete Block Floor/Wall M-12BG01 Chemical and Volume Control System 11 M-12BG02 Chemical and Volume Control System 13 M-12BG03 Chemical and Volume Control System 34 M-12BG04 Chemical and Volume Control System 06 M-K2EF01 Essential Service Water System 42 M-12EF01 Essential Service Water System 17 M-12EF02 Essential Service Water System 21 M-12EG01 Component Cooling Water System 13 M-12EG02 Component Cooling Water System 16 M-12EG03 Component Cooling Water System 07 M-12EM01 High Pressure Coolant Injection System 26 M-K2GD01 Essential Service Water Pump House HVAC 07 M-12AL01 Auxiliary Feedwater System 09 M-12FC02 Auxiliary Turbines- Auxiliary Feedwater System 19 M-12AP01 Condensate Storage and Transfer System 06 M-12EJ01 Residual Heat Removal System 28 M-12BN01 Borated Refueling Water Storage Tank 9 PERFORMANCE INPROVEMENT REQUESTS 2002-1670 2002-1781 2002-2393 2002-2290 2002-2502 2002-2507
-4-PROCEDURES Number Title Revision OFN KC-016 Fire Response 8 STN GP-009 Emergency Radio and Equipment Check and Inventory 33 OFN RP-014 Hot Standby to Cold Shutdown From Outside the Control Room 5 ALP KC-888 Fire Protection Panel KC-008 Alarm Response 8 STS AB-2010 Atmospheric Relief Valve Inservice Valve Test 12 OFN RP-017 Control Room Evacuation 17 MISCELLANEOUS Letter SLNRC 84-0037, dated February 24, 1984, addressed to Mr. Harold R. Denton, Director, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission from Nicholas A.
Petrick, Executive Director, SNUPPS Standardized Nuclear Unit Power Plant System Letter SLNRC 84-0014, dated February 1, 1984, addressed to Mr. Harold R. Denton, Director, Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission from Nicholas A.
Petrick, Executive Director, SNUPPS Standardized Nuclear Unit Power Plant System Letter SLNRC 84-0109, dated August 23, 1984, addressed to Mr. Harold R. Denton, Director, Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission from Nicholas A.
Petrick, Executive Director, SNUPPS Standardized Nuclear Unit Power Plant System Cable Routing Data for various components and fire areas E-11024, "Relay Setting Tabulation & Coordination Curves Systems NG/PG," Revision 2 TE 40650 (90-25), Telephone Memorandum dated October 16, 1992 Various sections of the Wolf Creek Generating Station Updated Safety Analysis Report Change Package 06513, "Thermo-Lag Removed for A-16 Fire Stop," Revision 0 Change Package 06605, "Install Manual Controller for ARVs in Fire Area A-23 in Auxiliary Building," Revision 9 Technical Specifications, Wolf Creek Generating Station, Unit 1, Docket STN 50-482, Appendix A to License NPF-42, Amendment 145 Facility Operating License, Wolf Creek Generating Station, Unit 1, Docket STN 50-482, License NPF-42, Amendment 144
-5-NUREG-0881, "Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit 1," dated April 1982 NUREG-0881, Supplement 3, "Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit 1," dated August 1983 NUREG-0881, Supplement 5, "Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit 1," dated March, 1985