IR 05000482/2011003

IR 05000482-11-003 & Notice of Violation, on 04/1/2011 6/30/2011, Wolf Creek Generating Station, Integrated Resident Report, Adverse Weather Protection, Equipment Alignments, Inservice Inspection Activities, Post Maintenance Testing, Event

ML112240223
Person / Time
Site:	Wolf Creek
Issue date:	08/12/2011
From:	Geoffrey Miller NRC/RGN-IV/DRP/RPB-B
To:	Matthew Sunseri Wolf Creek
References
EA-11-149 IR-11-003
Preceding documents:	ML112620676 (19 September 2011)
Download: ML112240223 (103)

Text

August 12, 2011 EA-11-149 Matthew Sunseri, President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839 Subject: WOLF CREEK GENERATING STATION

- NRC INTEGRATED INSPECTION REPORT AND NOTICE OF VIOLATION 05000482/201100

Dear Mr. Sunseri:

On June 30, 2011, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Wolf Creek Generating Station. The enclosed integrated inspection report documents the inspection findings, which were discussed on July 13, 2011 , with Mr. Stephen Hedges, Site Vice President, and other members of your staff.

The inspections examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.

Based on the results of this inspection, one violation is cited in the enclosed Notice of Violation (Notice) and the circumstances surrounding this violation are described in detail in the enclosed report. The violation involved the failure to implement procedures for opening of main steam isolation valves without causing safety system actuations (EA-1 1-149). Although determined to be of very low safety significance (Green), this violation is being cited in the Notice because Wolf Creek failed to restore compliance within a reasonable time after the violation was identified in NRC Inspection Report 05000 482/20 10 004, per Section 2.3.2 of the NRC Enforcement Policy. The current Enforcement Policy is included on the NRC's Web site at http://www.nrc.gov/about-nrc/regulatory/enforcement/enforce-pol.html. Please note you are required to respond to this letter and should follow the instructions specified in the enclosed Notice. If you have additional information that you believe the NRC should consider, you may provide it in your response to the Notice.

The NRC will use your response, in part, to determine whether further enforcement action is necessary to ensure compliance with regulatory requirements.

This report also documents nine additional NRC-identified and self-revealing issues that were evaluated under the risk significance determination process as having very low safety U N I T E D S T A T E S N U C L E A R R E G U L A T O R Y C O M M I S S I O N R E G I O N I V 6 12 EAST LAMAR BLVD

, S U I T E 4 0 0 A R L I N G T O N , T E X A S 7 6 0 1 1-4125 Wolf Creek Nuclear Operating Corporation

- 2 - EA-11-149 significance (Green). The NRC determined that violations are associated with eight of these issues. Additionally, two licensee-identified violations, which were determined to be of very low safety significance, are listed in this report. However, because of the very low safety significance and because they were entered into your corrective action program, the NRC is treating these findings as noncited violations, consistent with Section 2.3.2 of the NRC Enforcement Policy.

If you contest the violation or the significance of the noncited violations, 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, D.C.

20555-0001, with copies to the Regional Administrator, U.S.

Nuclear Regulatory Commission, Region IV, 612 E. Lamar Blvd, Suite 400, Arlington, Texas, 7601 1-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the facility. In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region IV, and the NRC Resident Inspector at the facility

.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response, if you choose to provide one for cases where a response is not required, will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal privacy or proprietary, information so that it can be made available to the Public without redaction.

Sincerely,

/RA/ Geoffrey B. Miller, Chief Project Branch B Division of Reactor Projects Docket No. 50-482 License No. NPF-42

Enclosure:

NRC Inspection Report and Notice of Violation 05000482/2011003 w/Attachment: Supplemental Information

REGION IV Docket: 05000 482 License: NPF-42 Report: 05000482/201100 3 Licensee: Wolf Creek Nuclear Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane NE Burlington, Kansas Dates: April 1 to June 30, 201 1 Inspectors:

C. Long, Senior Resident Inspector C. Peabody, Resident Inspector D. Reinert, Acting Resident Inspector J. Drake, Senior Reactor Inspector A. Fairbanks, Reactor Inspector G. Guerra, CHP, Emergency Preparedness Inspector G. Pick, Senior Reactor Inspector D. Strickland, Operations Engineer Approved By:

G. Miller, Chief, Project Branch B Division of Reactor Projects

- 4 - Enclosure

SUMMARY OF FINDINGS

IR 05000482/2011003, 4/1

- 6/3 0/2011; Wolf Creek Generating Station, Integrated Resident Report , Adverse Weather Protection

, Equipment Alignment s , Inservice Inspection Activities, Pos tmaintenance Testing

, Event Follow-up, and Other Activities. The report covered a 3

-month period of inspection by resident inspectors and announced baseline inspections by region-based inspectors. One Green cited violation, eight Green noncited violations, and one finding of significance were identified. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using Inspection Manual Chapter 0609, "Significance Determination Process." The cross-cutting aspect is determined using Inspection Manual Chapter 0310, "Components Within the Cross-Cutting Areas." Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG

-1649, "Reactor Oversight Process," Revision 4, dated December 2006.

Cornerstone: Initiating Events

Green.

The inspectors identified a noncited violation of Technical Specification 5.4.1.a, "Administrative Procedures," for having no procedure to address onsite debris impacting plant equipment during severe weather. The inspectors walked down external areas of the plant on June and June 9, 2011, prior to the onset of predicted severe thunderstorms and tornadoes. The inspectors found loose debris each time and brought it to the attention of the licensee who secured the materials. The inspectors walked down the transformer yard and tank yard during a thunderstorm on June 16 and found loose debris such as plywood, trash, wood planks, and fiberglass planks. The inspectors brought this to the attention of Wolf Creek and the materials were removed or secured. Wolf Creek initiated several condition reports but they only addressed immediate cleanup. Wolf Creek procedures had no steps for securing potential wind-driven projectiles prior to severe weather. After June 16, Wolf Creek wrote Condition Report 40573 which started a weekly maintenance activity to remove loose materials and added procedure steps to have operations walk down external areas prior to severe weather.

This finding was more than minor because it impacted the protecti on against external factors attribute of the Initiating Events Cornerstone

, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Inspection Manua l Chapter 0609.04, and determined that it was of very low safety significance (Green) for June 16, 2011, because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment would be unavailable since the reactor was shutdown.

Inspectors used Manual Chapter 0609 Appendix G, Checklist for the other occurrences because Wolf Creek was in

Modes 4 or 5. The finding again screened to Green because it did not increase the likelihood of a loss of inventory, did not cause the loss of reactor coolant system instrumentation, did not degrade the ability of the licensee to terminate a leak path or add inventory when needed, or degrade the ability to recover residual heat removal if it was lost. This finding has a cross-cutting aspect in the area of problem identification and resolution, specifically the corrective action program attribute because licensee's short-term corrective actions failed to ensure debris was secured or removed prior to severe weather P.1(d)(Secti on 1R01).

Green.

The inspectors documented a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion IX, "Control of Special Processes." Specifically, in October 2009, welders failed to ensure the fillet weld between the train B charging header and the half coupling used to attach two vent valves met the specified weld requirements. This weld failed in January 2011, rendering the train B charging system inoperable. The licensee's extent of condition review identified 12 vent line welds which did not meet ASME code weld size requirements and/or procedural requirements for 2:1 weld taper configuration. Additionally, quality assurance inspectors failed to identify that the 2:1 taper weld requirements specified by procedure, and ASME minimum weld size requirements, were not met in multiple vent line welds. The weld was repaired and built up to the correct 2:1 aspect ratio. This issue was entered into the licensee's corrective action program as Condition Reports 32648, 33686, 33689, and 36438. The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. T he inspectors performed a Phase screening in accordance with Inspection Manual Chapter 0609.04, "Phase 1 - Initial Screening and Characterization of Findings," and determined that the finding was of very low safety significance (Green) because the issue did not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. This finding had a cross-cutting aspect in the area of human performance, resources, because the licensee failed to ensure that personnel, specifically welders and quality assurance inspectors, were adequately trained in the procedural requirements and methods for measuring weld dimensions to assure nuclear safety H.2(b)(Section 1R08).

Green.

The inspectors identified a noncited violation of 10 CFR Part 50 involving the failure of the licensee to ensure that weld preparation was protected from deleterious contamination in that drawers (located in the hot tool room) containing files, grinding wheels, flapper wheels, and cutting wheels, used for the purpose of weld preparation, contained a mixture of both stainless steel tools and carbon steel tools. The failure to separate tools used for stainless steel weld preparation from tools used for carbon steel preparation could result in the contamination of stainless steel welds by carbon steel and affect the material integrity and corrosion resistance. The licensee immediately removed the tools and replaced them with new tools stored separately for use on specific types of metal. This issue was entered into the licensee's corrective action program as Condition Report 36444. The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations

, and if left uncorrected the finding would become a more significant safety concern. The inspectors performed a Phase 1 screening in accordance with Inspection Manual Chapter 0609.0 4 , "Phase 1 - Initial Screening and Characterization of Findings," and determined that the finding was of very low safety significance (Green)because the issue did not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. This finding had a cross-cutting aspect in the area of human performance, resources, because the licensee did not provide complete, accurate, and up-to-date procedures for the preparation of stainless steel and carbon steel welds H.2(c)(Section 1R08).

Green.

The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," involving the failure of the licensee to review the suitability of installing brass fittings and leaving test fittings on pressure, differential pressure, and flow transmitter equalizing block valve drain ports instead of the design specified stainless steel manifold plugs. During a boric acid walkdown, the inspectors identified that drain ports on the equalizing block of two separate reactor coolant system flow transmitters had brass fittings installed instead of the design specified stainless steel fittings. In response to inspector concerns about the brass fittings, the licensee subsequently discovered that a design configuration nonconformance existed by leaving the test fittings on the drain port during plant operation. License e Drawing J-17D22 specifies that manifold plugs be installed in the drain ports during plant operation. The licensee immediately replaced the brass caps with stainless steel fittings. This issue was entered into the licensee's corrective action program as Condition Report 36439. The finding was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. T he inspectors performed a Phase screening in accordance with Inspection Manual 0609.0 4, "Phase 1 - Initial Screening and Characterization of Findings," and determined that the finding was of very low safety significance (Green) because the issue would not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. The inspectors also determined that the finding had a cross-cutting aspect in the area of human performance, resources, because the licensee did not provide adequate training of personnel so that the inappropriately installed fittings could be identified during system walkdowns H.2(b)

(Section 1R08).

Green.

The inspectors identified a cited violation of Technical Specification 5.4.1.a, "Administrative Procedures," involving Wolf Creek's failure to correct Procedure SYS AB-120 for main steam isolation valve operation. Specifically, between March 3, 2010 , and March 19, 2011, Wolf Creek experienced repeat cases of safety-system actuations due to Procedure SYS AB-120 containing inadequate steps to establish conditions necessary to open a main steam isolation valve. Corrective actions were previously limited to steam header pressures below 300 psi. Wolf Creek commenced a root cause evaluation of the March 19, 2011 , safety injection under Condition Report 34964. Due to Wolf Creek's failure to restore compliance from previous NCV 05000482/2010004

-01 within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with the Enforcement Policy

. Failure to correct deficiencies in Procedure SYS AB-120 for steam pressures above 300 psi was a performance deficiency. The inspectors determined that this finding was more than minor because it impacted the equipment performance attribute for the Initiating Events Cornerstone

, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, this issue relates to the configuration control attribute for shut down equipment alignment. The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609.04. Assuming worst case degradation, the finding resulted in exceeding the technical specification limit for reactor coolant system leakage due to the pressurizer power-operated relief valve cycling. Therefore, the inspectors screened the finding to a Phase 2 review by the senior reactor analyst.

The senior reactor analyst used the Wolf Creek SPAR model and concluded that the incremental core damage probability was 3.7E-7 (Green). The inspectors found that the cause of the finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program. Specifically, several evaluations failed to have an adequate extent of condition review and did not find that procedures were inadequate for opening a main steam isolation valve above 300 psi P.1(c)(Section 4OA3.1).

Green.

The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1.a, "Administrative Procedures," for failure to follow procedural requirements to maintain reactor coolant system pressure below 350 psig. Control room operators increased charging flow at too great a rate with the reactor coolant system water-solid which caused the pressurizer power-operated relief valve to cycle three times over several minutes until adjustments to letdown could be made to reduce reactor coolant system pressure. Also, the letdown pressure controller was left in manual when automatic control would have lessened the pressure increase. Wolf Creek wrote Condition Report 35244 to correct the deficiency by changing several procedures for water-solid plant operations

. The failure to maintain pressure below the power-operated relief valve setpoint was a performance deficiency. The performance deficiency was more than minor because it impacted the Initiating Events Cornerstone objective of configuration control to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The significance of the finding was determined using Inspection Manual Chapter 0609, Significance Determination Process, Appendix G, Checklist 2, and determined to be of very low safety significance (Green), because it did not cause the loss of mitigating capability of core heat removal, inventory control, power availability, containment control, or reactivity control. Additionally, the finding also did not cause any low temperature overpressure technical specifications to be exceeded. The inspectors found that the cause of the finding had a cross-cutting aspect in the area of human performance. Specifically, operators had to rely on skill of the craft when procedures should have supplied more instruction for manipulating charging and letdown with a water-solid plant H.2.c](Section 4OA3.2)

.

Green.

The inspectors reviewed a self-revealing noncited violation of License Condition 2.C.5 for failure to implement adequate fire watches which affected both trains of vital ac and dc switchgear. The inadequate fire watches occurred during an actual fire which negated the Halon system discharge because internal fire doors were not shut, as required, by the fire watch. The inspectors found problems with fire impairments and watches from 2008 that had not been corrected.

Subsequent to the fire, Wolf Creek again briefed and trained its personnel on the requirements for fire watches. This issue is captured in the corrective action program as Condition Report 36719. Failure to implement adequate fire impairments such that the fire watches ensured the success of the Halon system was a performance deficiency. The performance deficiency was more than minor because it impacted the Initiating Events Cornerstone and its objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.

Specifically, the protection against external factors attribute was impacted by the fire impairment. To determine significance, the inspectors used Inspection Manual Chapter 0609.04 to screen the find ing to Inspection Manual Chapter 0609, Appendix F, because the fire protection defense-in-depth strategies involving automatic suppression, fire barriers, and administrative controls were degraded. The senior reactor analyst conducted a Phase 3 review of this finding and concluded that the incremental core damage frequency was 1.6 E-8 per year, or very low safety significance (Green). The inspectors found that the cause of the finding had a cross-cutting aspect in the area of problem identification and resolution. Specifically, corrective actions from ineffective fire watches in 2008 did not prevent recurrence of the inadequate fire watch on April 5, 2011 P.1.d](Section 4OA3.3)

.

Cornerstone: Mitigating Systems

Green.

The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1a, "Administrative Procedures," for a loss of component cooling water train B inventory caused by inadequate clearance order verification. Valve HBV110 was stuck in position and was partially open. When the clearance order was implemented, the operators concluded the valve was already closed. Subsequently, the valve created a leakage path which exceeded the surge tank makeup flow capacity and required manual isolation by the control room operators to protect safety-related components. Wolf Creek has taken corrective actions to include communication of expected as-found equipment positions in pre-job briefings and the clearance order template. This issue is captured in the corrective action program as Condition Reports 34505 and 40219.

Failure to properly establish clearance order boundary isolation was a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance and human performance attributes of the Mitigating Systems Cornerstone and impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, the finding was determined to be of very low safety significance because the finding did not result in the loss of operability or functionality of the component cooling water train or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event

. The inspectors found that the finding had a cross-cutting aspect of work practices in the area of human performance associated with the communication of human error prevention techniques, such as holding pre-job briefings, self- and peer-checking, and proper documentation of activities H.4(a)(Section 1R04).

Green.

The inspectors identified a finding involving the failure to follow t he requirements of Procedure AP 16E-002, "Post Maintenance Testing Development," for the startup feedwater pump. On November 4-6, 2010, Wolf Creek workers disassembled the startup feedwater pump for numerous preventive and corrective activities including removing the rotating element. On November 17, 2010, Wolf Creek conducted surveillance Procedure STN AE-007, "Startup Main Feedwater Pump Operational Test," following reassembly. The only acceptance criteria listed in this procedure is that the motor-driven feedwater pump starts from the control room with no local operator action. The inspectors found this contrary to Procedure AP 16E-002, which requires acceptance criteria for a pump flow capacity test, vibration, bearing and lubrication temperatures, motor current, external leakage, and lubrication level be found satisfactory. This issue is captured in the corrective action program as Condition Report 39494. Wolf Creek issued a new work package to conduct a single-point pump capacity test and complete the required postmaintenance testing. Wolf Creek found, pending final review, that initial calculations show that the pump design is capable of enough flow to provide a heat sink in emergency operating procedures

.

Failure to follow Procedure AP 16E-002 for developing test criteria for plant equipment after the completion of maintenance activities is a performance deficiency. The finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of equipment performance and it adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, the inspectors determined that the finding had very low safety significance (Green) because it did not result in a loss of system safety function, an actual loss of safety function of a single train for greater than its technical specification allowed outage time, or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution. Specifically, Wolf Creek created a testing procedure in response to a root cause evaluation, but did not consider acceptance criteria to ensure that the pump performs acceptably P.1(d)(Section 1R19). GreenFailure to implement design control measures to analyze whether containment spray piping remained full of water was a performance deficiency. This finding was more than minor because it affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of the containment spray system to respond to initiating events and prevent undesirable consequences. Specifically, the inspectors had reasonable doubt on the capability of the containment spray system to properly inject because of vortexing in the containment spray additive tank. The inspectors performed the significance determination using Inspection Manual Chapter 0609.04. The finding was determined to be of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Although the failure to have this calculation had existed since original construction, the inspectors determined this finding reflected current performance since the licensee was required to evaluate likelihood of tanks allowing gas intrusion into the emergency core cooling systems in response to Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems." Consequently, this finding had problem identification and resolution cross-cutting aspects associated with the corrective action program in that the licensee did not thoroughly evaluate the potential for gas intrusion from all possible tanks P.1(c)(Section 4OA5). . The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the failure to translate the design basis into instructions, procedures, and drawings. The inspectors found that the licensee failed to assess whether vortexing occurred in the containment spray additive tank in the event of a design-basis accident. Wolf Creek entered this issue in the corrective action program as Condition Report 38715.

B. Licensee-Identified Violations

Violations of very low safety significance, which were identified by the licensee, have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensee's corrective action program. These violations and condition report numbers are listed in Section 4OA7.

REPORT DETAILS

Summary of Plant Status

Wolf Creek began the quarter shut down for Refueling Outage 18. Wolf Creek restarted on June 22, 2011. Reactor operators manually tripped the reactor from 82 percent power on June 26 due to a trip of main feedwater pump B. Wolf Creek restarted on June 29 and ended the quarter holding at 55 percent power to complete troubleshooting and repairs on main feedwater pump B.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, and Emergency Preparedness

1R01 Adverse Weather Protection

.1 Summer Readiness for Offsite and Alternate-ac Power

a. The inspectors performed a review of preparations for summer weather for selected systems, including conditions that could lead to loss

-of-offsite power and conditions that could result from high temperatures. The inspectors reviewed the procedures and communications protocols between the transmission system operator and the plant to verify that the appropriate information was being exchanged when issues arose that could affect the offsite power reliability. Examples of aspects considered in the inspectors' review included:

Inspection Scope The coordination between the transmission system operator and the control room personnel during off

-normal or emergency events

The explanations for the events

The estimates of when the offsite power system would be returned to a normal state The notifications from the transmission system operator to the plant when the offsite power system was returned to normal

During the inspection, the inspectors focused on plant

-specific design features and the procedures used by plant personnel to mitigate or respond to adverse weather conditions. Additionally, the inspectors reviewed the Updated Safety Analysis Report (USAR) and performance requirements for selected systems, and verified that operator actions were appropriate per station procedures. Specific documents reviewed during this inspection are listed in the attachment. The inspectors also reviewed corrective acti on documents to verify that the licensee was identifying adverse weather issues at an appropriate threshold and entering them into their corrective action program. These activities constitute completion of one readiness for summer weather affect on offsite and alternate ac power sample as defined in Inspection Procedure 71111.01-05.

b. Findings

No findings were identified.

.2 Readiness for Impending Adverse Weather Conditions

a. When thunderstorms

, tornados , and high winds were forecast in th e site vicinity on June 1, 9, and 16 , 2011, the inspectors reviewed the plant preparations for the expected weather conditions. On June 1, 9, and 16, the inspector s walked down the offsite power system , refueling water storage tank

, and reactor makeup water storage tank because their safety function s could be affected by high wind-generated missiles or a loss of offsite power. The inspectors evaluated the se preparations against the site procedures and determined that actions by the plant staff were adequate. During the inspection, the inspectors focused on plant

-specific design features and the station procedures used to respond to specified adverse weather conditions. The inspectors also toured outdoor areas of the plant to look for any loose debris that could become a wind-driven projectile. The inspectors evaluated operator staffing and accessibility of instrumentation and controls for systems required to operate the plant. Additionally, the inspectors reviewed the USAR and performance requirements for the selected systems and verified that operator actions were appropriate per station procedures. The inspectors also reviewed a sample of corrective action documents to verify that the licensee

-identified adverse weather issues at an appropriate threshold and entered them into the corrective action program. Specific documents reviewed during this inspection are listed in the attachment.

Inspection Scope These activities constitute completion of two readiness for impending adverse weather condition sample s as defined in Inspection Procedure 71111.01-05. b. Introduction

. The inspectors identified a Green noncited violation of Technical Specification 5.4.1.a for having no procedure to address onsite debris impacting plant equipment during severe weather.

Findings Description

. On June 1, 2011, a severe thunderstorm watch was announced by the national weather service. The inspectors walked down the transformer yard at 6 p.m.

, with the storms forecast to arrive later that night. The inspectors found numerous pieces of unsecured plywood and 2"x4" and 2"x8" planks. The inspectors brought this to the licensee's attention , and Wolf Creek personnel secured the materials. The inspectors reviewed station Procedure AI 14-006, "Severe Weather

," Revision 9A. The procedure directed public address system announcements for national weather service severe weather declarations and instructions on personnel sheltering, but included no steps on equipment protection from onsite debris.

The inspectors reviewed Procedure OFN SG-003, "Natural Events

," Revision 20A, but it did not direct entry until a tornado is sighted or a tornado warning is issued.

The national weather service issued a tornado warning for the site area on June 9, at 3:20 p.m. The inspectors walked down the transformer yard at 5 p.m. The inspectors again found unsecured debris in the transformer and tank areas. The inspectors reported the debris to the control room and outage control center who sent personnel to secure the material. On June 10, a severe thunderstorm watch was issued at 5 p.m., and the inspectors walked down the transformer and tank yards at 6 p.m. to verify the corrective action from the previous day had been implemented for the pending storms. The inspectors found that some material was removed or secured, but also found numerous unsecured sections of scaffolding

, wood, palettes, diamond plate, and debris. The inspectors discussed this with the outage control center. Condition Report 40351 was written with immediate actions to secure the loose materials. The extent of condition description included any area where inclement weather has the potential of creating airborne objects that could challenge plant equipment. On June 16, the inspectors walked down the transformer yard and tank areas during a thunderstorm. The inspectors found numerous unsecured pieces and stacks of wood and other debris that could impact plant equipment if winds were more severe. Wolf Creek responded by securing or removing the debris and writing Condition Report 40573 which implemented a weekly preventive maintenance activity to clean up outside areas and changed Procedure AI 14-006 to perform walkdowns of outside areas prior to severe weather. The inspectors found previously written condition reports on lack of adverse weather preparations for outdoor areas prior to the inspection.

Analysis.

Failure to remove potential wind

-driven debris from the transformer and tank areas before severe weather is a performance deficiency. This finding was more than minor because it impacted the protecti on against external factors attribute of the Initiating Events Cornerstone

, and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated this finding using Inspection Manual Chapter 0609.04, and determined that it was of very low safety significance (Green)for June 16 because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment would be unavailable since the reactor was shutdown.

Inspectors used Manual Chapter 0609, Appendix G, Checklist 4 , for the other occurrences because Wolf Creek was in Modes 4 or 5. The finding again screened to Green because it did not increase the likelihood of a loss of inventory, did not cause the loss of reactor coolant system instrumentation, did not degrade the ability of the licensee to terminate a leak path or add inventory when needed, or degrade the ability to recover residual heat removal if it was lost. This finding has a cross

-cutting aspect in the area of problem identification and resolution, specifically the corrective action program attribute because licensee sho rt-term corrective actions failed to ensure debris was secured or removed prior to severe weather P.1(d)

. Enforcement

. Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, 1978. Regulatory Guide 1.33, Appendix A, Section 6.w, requires, i n part, written procedures for acts of nature (e.g., tornado, flood, dam failure, earthquakes). Contrary to the above, prior to June 16, 2011, Wolf Creek had not established written procedure s for acts of nature associated with tornados. Specifically, there were no procedural directions that addressed how the licensee was to protect from wind-driven projectiles, associated with tornados, in the protected area. Because this violation was of very low safety significance and was entered into the licensee's corrective action program as Condition Report 40573, this violation is being treated as a noncited violation, consistent with Section 2.3.2 of the NRC Enforcement Policy:

NCV 05000485/2011003

-0 1 , "No Procedure for Debris in Transformer and Tank Yards Prior to Severe Weather

."

1R04 Equipment Alignments

.1 Partial Walkdown

a. The inspectors performed a partial system walkdown of the following risk

-significant system: Inspection Scope March 8, 2011, Component cooling water The inspectors selected this system based on its risk significance relative to the Reactor Safety Cornerstone at the time it was inspected. The inspectors attempted to identify any discrepancies that could affect the function of the system, and, therefore, potentially increase risk. The inspectors reviewed applicable operating procedures, system diagrams, USAR, technical specification requirements, administrative technical specifications, outstanding work orders, condition reports, and the impact of ongoing work activities on redundant trains of equipment in order to identify conditions that could have rendered the systems incapable of performing their intended functions. The inspectors also inspected accessible portions of the system to verify system components and support equipment were aligned correctly and operable. The inspectors examined the material condition of the components and observed operating parameters of equipment to verify that there were no obvious deficiencies. The inspectors also verified that the licensee had properly identified and resolved equipment alignment problems that could cause initiating events or impact the capability of mitigating systems or barriers and entered them into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one partial system walkdown sample as defined in Inspection Procedure 71111.04-05.

b. Introduction

. The inspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.4.1a, "Administrative Procedures," for an inadequate clearance order verification which caused a loss of component cooling water B inventory.

Findings Description

. On March 8, 2011, Wolf Creek was preparing to implement Temporary Modification Order (TMO) 10-017-EG-00 to install temporary equipment to cool the radwaste system heat loads.

These preparation s included hanging clearance order D-HB-N-029 which required station operators to verify closed manual valves EGV0079 and HBV0110 and open and uncap the associated piping header hose connection valves HBV0088 and HBV0111. Until TMO 10-017 is implemented, component cooling water must be periodically aligned to the radwaste building to cool its associated nonsafety

-related heat loads. This nonsafety component cooling water function adds seismic vulnerabilities that render the aligned train inoperable (NCV 05000482/2010007

-01). At 9:30 a.m., station operators attempted to move valve HBV0110 in the close d direction and found that the valve would not turn. The operators compared the stem position relative to that of an identical model valve. The operators successfully manipulated travel of valve EGV0079 in the previous step from the fully open to fully closed position. This apparent position verification was made using the naked eye, and was the basis for assuming that the valve was firmly on its seat and signed the clearance order verifications accordingly.

At 2:37 p.m., the control room operators performed a planned routine alignment of component cooling water train B to radwaste. This action immediately resulted in a 200 gpm component cooling water leak through valve HBV0110 and out of the hose connection piping penetrations. The rapidly decreasing component cooling water B surge tank level caused an auto start of the demineralized water makeup to the component cooling water B surge tank and simultaneously sent an alarm to the control room operators. However, the demineralized water makeup capacity is only 60 gpm, resulting in a component cooling water B inventory loss of 140 gpm and a decreasing surge volume.

Without prompt manual actions, the 5000 gallon component cooling water train B surge tank volume would have been exhausted in 25 minutes , at which point component cooling water train B would void and fail.

For the duration of the leak , component cooling water train B was unavailable because it was unable to meet its accident mission time.

Operators identi fied the cause and isolate d the component cooling water supply from the radwaste building. The leak was determined to be approximately 500 gallons total

, or 10 percent of the normal component cooling water surge tank inventory.

The leak revealed that valve HBV0110 was not fully closed but was stuck in a throttled position. Station operators were directed by the control room to attempt to move the valve in the open direction, which they did with an approved torque assist device. When the operators subsequently moved the valve in the closed direction, it moved beyond its previous position and was properly seated. Later that evening, when component cooling water was once again aligned to radwaste, no leakage occurred. Wolf Creek entered the event into their corrective action as Condition Report 34505.

The inspectors reviewed the history for valve HBV0110. All four of the subject valves had minimal manipulation since the waste evaporator package th ey were originally associated with had been abandoned in place in the early 1990s. Also, periodic maintenance to inspect and lubricate the valve internals has not been performed during this time. The last position verification made on valve HBV0110 was conducted April 21, 2006, and indicated that the valve was throttled partially open. The valve was also listed on drawing M-12HB02 as normally throttled.

The clearance order paperwork specified to leave the valve 1.4 turns open upon removal of the clearance order.

The inspectors determined that the operator s failed to meet the requirements of station Procedure AP 21E-001, step 6.4.2.1, to properly position the equipment/components in the sequence specified on the clearance order tag hang list, as well as step 6.4.3.1, the independent verification of that component or equipment condition. The inspectors' interviews with operators and station management indicated that the cause of the leakage was a lack of information communicated to the operators performing the tagout. Wolf Creek tagout practices d id not provide expected, as

-found component positions for taggers and verifiers in the clearance order tag "Hang List" nor is this information communicated during pre-job briefings. Wolf Creek initiated Condition Report 40219 which directed oral communication of the expected initial component positions during pre-job briefings and on the clearance order paperwork template.

Analysis.

Failure to properly establish clearance order boundary isolation is a performance deficiency. The performance deficiency is more than minor because it impacted the equipment performance and human performance attributes of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, the finding was determined to be of very low safety significance (Green) because the finding is not a design or qualification deficiency confirmed not to result in loss of operability or functionality; the finding does not represent a loss of system safety function; the finding does not represent actual loss of safety function of a single train for more than its technical specification allowed outage time; the finding does not represent an actual loss of safety function of one or more nontechnical specification trains of equipment designated as risk significant per 10 CFR 50.65 for more than 24 h ours; and the finding does not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors found that the finding had a cross

-cutting aspect of work practices in the area of human performance. The licensee communicates human error prevention techniques, such as holding pre

-job briefings, self

- and peer-checking, and proper documentation of activities. Specifically

, Wolf Creek did not communicate the expected as-found condition of valve HBV0110 to the taggers and verifiers of the clearance order

H.4(a). Enforcement

. Wolf Creek Technical Specification 5.4.1a requires that procedures be established, implemented, and maintained covering the activities described in Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33, Revision 2, Appendix A, Section 1(c) requires , in part , procedures governing equipment control, including locking and tagging. Licensee Procedure AP 21E-001 "Clearance Orders,"

step s 6.4.2.1 and 6.4.3.1 , specifies that equipment and components be positioned and verified in the sequence specified on the clearance order tag list. Contrary to the above, on March 8, 2011, the licensee failed to ensure the component was positioned and verified in the sequence specified on the clearance order tag list. Specifically, while performing clearance order D-HB-N-029, valve HBV0110 was not properly positioned and verified as specified on the clearance order tag list. These actions directly resulted in a loss of component cooling water train availability. Because this finding is of very low safety significance and was entered into the licensee corrective action program as Condition Report s 34505 and 40219, this violation is being treated as a noncited violation in accordance with Section 2.3.2 of the Enforcement Policy:

NCV 05000482/2011003

-02, "Failure to Properly Establish Clearance Order Boundary Isolation Resulting in Loss of Component Cooling Water Inventory."

.2 Complete Walkdown and System Walkdown

Associated with Temporary Instruction (TI) 2515/177 a. On April 27, 2011, the inspectors performed a complete system alignment inspection of the containment spray system to verify the functional capability of the system. The inspectors selected this system because it was considered both safety significant and risk significant in the licensee's probabilistic risk assessment. The inspectors inspected the system to review mechanical and electrical equipment lineups, electrical power availability, system pressure and temperature indications, as appropriate, component labeling, component lubrication, component and equipment cooling, hangers and supports, operability of support systems, and to ensure that ancillary equipment or debris did not interfere with equipment operation. The inspectors reviewed a sample of past and outstanding work orders to determine whether any deficiencies significantly affected the system function. In addition, the inspectors reviewed the corrective action program database to ensure that system equipment

-alignment problems were being identified and appropriately resolved. Specific documents reviewed during this inspection are listed in the attachment.

Inspection Scope The inspectors conducted a walkdown of the containment spray system in sufficient detail to reasonably assure the acceptability of the licensee's walkdowns (TI 2515/177, Section 04.02.d). The inspectors also verified that the information obtained during the licensee's walkdown was consistent with the items identified during the inspector's independent walkdown (TI 2515/177, Section 04.02.c.3).

In addition, the inspectors verified that the licensee had isometric drawings that describe the containment spray system configurations and had acceptably confirmed the accuracy of the drawings (TI 2515/177, Section 04.02.a). The inspectors verified the following related to the isometric drawings.

High point vents were identified Other areas where gas can accumulate and potentially impact subject system operability, such as at orifices in horizontal pipes, isolated branch lines, heat exchangers, improperly sloped piping, and under closed valves were acceptably referenced in documentation Horizontal pipe centerline elevation deviations and pipe slopes in nominally horizontal lines that exceed specified criteria were identified All pipes and fittings were clearly shown The drawings were up

-to-date with respect to recent hardware changes and that any discrepancies between as

-built configurations and the drawings were documented and entered into the corrective action program for resolution The inspectors verified that piping and instrumentation diagrams accurately described the subject systems; that they were up

-to-date with respect to recent hardware changes; and any discrepancies between as

-built configurations, the isometric drawings, and the piping and instrumentation diagrams were documented and entered into the corrective action program for resolution (TI 2515/177, Section 04.02.b). Documents reviewed are listed in the attachment to this report.

These activities constitute completion of one complete system walkdown sample as defined in Inspection Procedure 71111.04-05. Also, this inspection effort counts toward the completion of TI 2515/177. See Section

4OA5 for additional information.

b. No findings were identified.

Findings

1R05 Fire Protection

.1 Quarterly Fire Inspection Tours

a. The inspectors conducted fire protection walkdowns that were focused on availability, accessibility, and the condition of firefighting equipment in the following risk

-significant plant areas:

Inspection Scope March 19, 2011, Safety injection pump room A March 19, 2011, Control room ventilation equipment room B March 20, 2011, Auxiliary building 1988' pipe chase April 6, 2011, Containment building The inspectors reviewed these areas to assess if licensee personnel had implemented a fire protection program that adequately controlled combustibles and ignition sources within the plant; effectively maintained fire detection and suppression capability; maintained passive fire protection features in good material condition; and had implemented adequate compensatory measures for out of service, degraded or inoperable fire protection equipment, systems, or features in accordance with the licensee's fire plan. The inspectors selected fire areas based on their overall contribution to internal fire risk as documented in the plant's Individual Plant Examination of External Events with later additional insights, their potential to affect equipment that could initiate or mitigate a plant transient, or their impact on the plant's ability to respond to a security event. Using the documents listed in the attachment, the inspectors verified that fire hoses and extinguishers were in their designated locations and available for immediate use; that fire detectors and sprinklers were unobstructed

that transient material loading was within the analyzed limits; and fire doors, dampers, and penetration seals appeared to be in satisfactory condition. The inspectors also verified that minor issues identified during the inspection were entered into the licensee's corrective action program. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of four quarterly fire

-protection inspection sample s as defined in Inspection Procedure 71111.05-05. b. No findings were identified.

Findings

1R08 Inservice Inspection Activities

.1 Inspection Activities Other Than Steam Generator Tube Inspection, Pressurized Water Reactor Vessel Upper Head Penetration Inspections, Boric Acid Corrosion Control (71111.08-02.01)

a. The inspection procedure required review of two or three types of nondestructive examination activities and, if performed, one to three welds on the reactor coolant system pressure boundary. It also required review of one or two examinations with relevant indications (if any were found)that had been accepted by the licensee for continued service.

Inspection Scope The inspectors directly observed the following nondestructive examinations:

SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Pressurizer TBB03-CIRCUM-1-W Ultrasonic Examination Pressurizer TBB03-SEAM-4W Ultrasonic Examination Pressurizer TBB03-10-B-W Ultrasonic Examination Pressurizer TBB03-10-C-W Ultrasonic Examination Pressurizer TBB03-10-B-IR Ultrasonic Examination Pressurizer TBB03-10-C-IR Ultrasonic Examination Steam Generator EBB01A-SEAM-5-W Ultrasonic Examination SYSTEM WELD IDENTIFICATION EXAMINATION TYPE Steam Generator EBB01A-SEAM-8-W Ultrasonic Examination RV Closure Head Studs and Nuts CH-STUD-19 through 36 Ultrasonic Examination Main Steam Piping Support AB-01-R001 Visual Examination 3 Main Steam Piping Support AB-01-R003 Visual Examination 3 Feedwater Piping Support AE05-R028 Visual Examination 3 Feedwater Piping Support AE-04-R019 Visual Examination 3 Feedwater Piping Support AE05-C001 Visual Examination 3 Main Steam Integral Attachment AB-01-R010 Magnetic Examination Feedwater Integral Attachment AE-05-R028 Magnetic Examination During the review and observation of each examination, the inspectors verified that activities were performed in accordance with ASME Boiler and Pressure Vessel Code requirements and applicable procedures. Indications were compared with previous examinations and dispositioned in accordance with ASME code and approved procedures. The qualifications of all nondestructive examination technicians performing the inspections were verified to be current.

Only the visual examination of AE05-R028, "Piping Support," identified any relevant indications. Repairs were made to AE05-R028 and it was reexamined and was satisfactory. Wolf Creek personnel stated that no relevant indications were accepted by the licensee for continued service.

The inspectors directly observed a portion of the following welding activities

SYSTEM WELD IDENTIFICATION WELD TYPE Reactor Coolant System 13BG22-W-33 Shield Metal Arc Welding Reactor Coolant System 13BG22-W-34 Shield Metal Arc Welding The inspectors verified, by review, that the welding procedure specifications and the welders had been properly qualified in accordance with ASME Code,Section IX, requirements. The inspectors also verified through record review that essential variables for the welding process were identified, recorded in the procedure qualification record, and formed the bases for qualification of the welding procedure specifications.

Specific documents reviewed during this inspection are listed in the attachment.

.1 Failure to Ensure Fillet Weld Met Size Requirements on Train

B Charging Header Vent Line

Introduction.

The inspectors documented a self

-revealing Green noncited violation of 10 CFR Part 50, Appendix B, Criterion IX, "Control of Special Processes," after the licensee failed to ensure that that the fillet weld between the train B charging header and the half-coupling used to attach two vent valves met 2:1 taper weld requirements. The undersized weld subsequently resulted in a 300 drop-per-minute leak in January 2011.

Description.

On January 3, 2011, the licensee identified a 300 drop-per-minute pinhole leak at the weld joint between the train B charging header and/or the half coupling used to attach vent valves BGV0845 and BGV0846. The licensee measured the subject weld and concluded that the weld was undersized and the required 2:1 aspect ratio was not obtained. The weld was performed in the October/November 2009 timeframe during the installation of vent valves in the chemical and volume control system, the residual heat removal system, and the high pressure coolant injection system. Also, quality assurance inspectors inappropriately accepted the undersized weld.

Wolf Creek's extent-of-condition review concluded that 12 additional welds either did not meet the procedurally required 2:1 aspect ratio or did not meet ASME minimum weld size requirements. No other undersized welds developed leaks. After the leak was identified, the train B charging system was declared inoperable and the weld was repaired and built up to the correct 2:1 aspect ratio.

Wolf Creek performed a hardware failure analysis on the failed weld and concluded that although the main characteristics of the weld fracture were consistent with stress corrosion cracking, the crescent shape of the fracture indicated cyclic crack growth. The licensee also concluded that the configuration of the vent line with no lateral support could have created a cantilever effect on the line and in combination with a notch created by the lack of fusion in the weld root served as a stress concentrator. This issue was entered into the licensee's corrective action program as Condition Report 36438.

Analysis.

Failure to meet ASME code weld size requirements is a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone. The finding adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The inspectors performed a Phase 1 screening in accordance with Inspection Manual Chapter 0609.04 and determined that the finding was of very low safety significance (Green) because the issue did not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. This finding had a cross

-cutting resources aspect in the area of human performance, because the licensee failed to ensure that welders and quality assurance inspectors were adequately trained in the procedural requirements and methods for measuring weld dimensions to assure nuclear safety H.2(b)

.

Enforcement

. Title 10 CFR Part 50, Appendix B, Criterion IX, "Control of Special Processes," requires in part, that measures be established to ensure that special processes, including welding are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements.

Contrary to the above, in October 2009, the licensee failed to ensure that special processes, including welding

, were controlled and accomplished using qualified procedures. Specifically, welders failed to ensure that the fillet weld between the train B charging header and the half

-coupling used to attach two vent valves met 2:1 taper weld requirements, which subsequently resulted in a 300 drop-per-minute leak in January 2011. This issue was entered into the licensee's corrective action program as Condition Report 36438. Because this finding was determined to be of very low safety significance and was entered into the license's corrective action program, this violation is being treated as a noncited violation consistent with Section 2.3.2 of the NRC Enforcement Policy:

NCV 05000482/2011003

-03, "Failure to Assure Fillet Weld Met Size Requirements on Train B Charging Header Vent Line."

.2 Failure to

Ensure Separation of Stainless Steel and Carbon Steel Grinding and Cutting Tools Introduction

. The inspectors identified a Green noncited violation of 10 CFR 50.55a , "Codes and Standards," after the licensee failed to ensure that stainless steel and carbon steel grinding wheels, flapper wheels, cutting wheels, and files were stored separately and used only for the weld preparation of the designated steel.

Description

. During inspection of the tool issue room in the radiologically controlled area , the inspectors identified that tools designated for either stainless steel or carbon steel weld preparation were not stored separately. The inspectors noted that although stainless steel grinding wheels, flapper wheels, and cutting wheels were marked for stainless steel use, they were stored with carbon steel grinding wheels, flapper wheels, and cutting wheels. Additionally, the inspectors identified that although stainless steel files and carbon steel files were stored in separate drawers, there were files in the stainless steel drawer that appeared to have been used on carbon steel, and there was a file marked for use on stainless steel in the carbon steel drawer. There was also no procedure or written guidance pertaining to proper storage and control of the equipment

. The failure to separate tools used for stainless steel weld preparation from tools used for carbon steel preparation could result in the contamination of stainless steel welds by carbon steel and affect the material integrity and corrosion resistance.

The licensee immediately removed the tools and replaced them with new tools stored separately for use on specific types of metals. This issue was entered into the licensee's corrective action program as Condition Report 3644.

Analysis.

Failure to protect stainless steel welds from deleterious contamination is a performance deficiency. The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events Cornerstone. The finding adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations and if left uncorrected, the finding would become a more significant safety concern. The inspectors performed a Phase 1 screening in accordance with Inspection Manual Chapter 0609.04 and determined that the finding was of very low safety significance (Green) because the issue did not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. This finding had a resources cross

-cutting aspect in the area of human performance, because the licensee did not provide adequate procedures for the preparation of stainless steel and carbon steel welds H.2(c)

. Enforcement

. Title 10 CFR 50.55a , states in part, that "Each operating license for a boiling or pressurized water

-cooled nuclear power facility is subject to the conditions in paragraphs

(f) and (g)." Title 10 CFR 50.55a(g)(4), requires, in part, that component s classified as ASME Code Class 1, Class 2, and Class 3 meet the requirements set forth in Section XI of the ASME Boiler and Pressure Vessel Code and Addenda.Section XI of the ASME Code, Part IWA

-4221(b)(2)

, states that "

When adding a new component to an existing system, the Owner shall specify a Construction Code."

The licensee specified Section III of the subject code when adding a new component to an existing system.Section III, Part NG4412, states that "The work [weld preparation] shall be protected from deleterious contamination." Contrary to the above, prior to June 2011, the licensee did not ensure that weld preparation was protected from deleterious contaminatio n. Specifically , the licensee failed to ensure weld preparation was protected, in that tools located in the hot tool room drawers containing files, grinding wheels, flapper wheels, and cutting wheels that were used for the purpose of weld preparation, were found to contain a mixture of both stainless steel tools and carbon steel tools. This issue was entered into the licensee's corrective action program as Condition Report 36444. Because this finding was determined to be of very low safety significance and was entered into the licensee's corrective action program, this violation is being treated as a noncited violation consistent with Section 2.3.2 of the NRC Enforcement Policy:

NCV 05000482/2011003

-04, "Failure to Assure Separation of Stainless Steel an d Carbon Steel Grinding and Cutting Equipment."

.2 Vessel Upper Head Penetration Inspection Activities (71111.08

-02.02) a. The inspectors witnessed portions of the licensee's performance of the required visual inspection (VT

-2) of the reactor head and pressure

-retaining components above the reactor pressure vessel head in accordance with requirement of ASME Code Case N-729-1 as mandated by 10 CFR 50.55a. Implementation required ASME Code IWA-2212 VT-2 under the mirror insulation on top of the reactor head through multiple access points. The inspectors reviewed the results of this inspection for evidence of leaks or boron deposits at reactor pressure boundaries and related insulation above the head. Specific documents reviewed during this inspection are listed in the attachment.

Inspection Scope

These actions constitute completion of the requirements for Section 02.02 of Inspection Procedure PI 71111.08. b. No findings were identified.

Findings

.3 Boric Acid Corrosion Control Inspection Activities (71111.08-02.03)

a. Inspection Scope

The inspectors evaluated the implementation of the licensee's boric acid corrosion control program for monitoring degradation of those systems that could be adversely affected by boric acid corrosion. The inspectors reviewed the documentation associated with the licensee's boric acid corrosion control walkdown as specified in Procedure STN PE-040D, "RCS Pressure Boundary Integrity Walkdown," Revision 3, and Procedure AP 16F-001, "Boric Acid Corrosion Control Program," Revision 6A. The inspectors also reviewed the visual records of the components and equipment. The inspectors verified that the visual inspections emphasized locations where boric acid leaks could cause degradation of safety

-significant components. The inspectors also verified that the engineering evaluations for those components where boric acid was identified gave assurance that the ASME code wall thickness limits were properly maintained.

The inspectors confirmed that the corrective actions performed for evidence of boric acid leaks were consistent with requirements of the ASME c ode. Specific documents reviewed during this inspection are listed in the attachment.

These actions constitute completion of the requirements for Section 71111.08-02.03. b. Failure to Assure Configuration Control of Safety

-Related Systems Findings

Introduction.

The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," involving the failure of the licensee to review the suitability of replacing the design specified stainless steel manifold plugs with test fittings and brass caps on various flow transmitter equalizing block valve drain ports.

Description

. During a boric acid walkdown, the inspectors identified that drain ports on the equalizing block of two separate reactor coolant system flow transmitters had brass fittings installed instead of stainless steel fittings. The inspectors brought this condition to Wolf Creek's attention. The licensee determined that a design configuration nonconformance existed in that licensee D rawing J-17D22 specified that stainless steel manifold plugs be installed in the drain ports during plant operation. The licensee failed to review the suitability of installing brass fittings and leaving test fittings on flow transmitter equalizing block valve drain ports instead of the design specified stainless steel manifold plugs. Wolf Creek immediately replaced the brass caps with stainless steel fittings. This issue was entered into the licensee's corrective action program as Condition Report 36439.

Analysis.

Plugging instrument lines with test fittings of a different material is a performance deficiency. The finding was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone. The finding affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The inspectors screened the finding per Inspection Manual Chapter 0609.04 and determined that the finding was of very low safety significance (Green) because the issue would not result in exceeding the technical specification limit for identified reactor coolant system leakage or affect other mitigating systems resulting in a total loss of their safety function. The inspectors also determined that the finding had a resources cross

-cutting aspect in the area of human performance, because the licensee did not provide adequate training of personnel so that the inappropriately installed fittings could be identified during system walkdowns H.2(b)

. Enforcement

. Title 10 CFR Part 50, Appendix B, Criterion III, "Design Control," requires, in part, that measures be established for the selection and review of suitability of application of materials, parts, equipment, and processes that are essential to the safety

-related functions of the structures, systems, and components. Contrary to the above, the licensee failed to establish measures for the selection and review for suitability of parts that are essential to the safety

-related functions of systems and components. Specifically, the licensee failed to review the suitability of replacing the design specified stainless steel manifold plugs with brass caps and test fittings on various equalizing block valve drain ports for pressure, differential pressure, and flow transmitters. This issue was entered into the licensee's corrective action program as Condition Report 36439. Because this finding was determined to be of very low safety significance and was entered into the licensee's corrective action program, this violation is being treated as a noncited violation consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000482/201100 3-05 , "Failure to Assure Configuration Control of Safety

-Related Systems

."

.4 Steam Generator Tube Inspection Activities (71111.08

-02.04)

a. Inspection Scope

The inspection procedure specified an assessment of in situ screening criteria to assure consistency between assumed nondestructive examination flaw sizing accuracy and data from the EPRI examination technique specification sheets. The inspection procedure also specified assessment of appropriateness of tubes selected for in situ pressure testing, observation of in situ pressure testing, and review of in situ pressure test results. No conditions were identified that warranted in situ pressure testing. The steam generators are original construction steam generators.

The inspectors reviewed both the licensee site

-validated and qualified acquisition and analysis technique sheets used during this refueling outage and the qualifying EPRI examination technique specification sheets to verify that the essential variables regarding flaw sizing accuracy, tubing, equipment, technique, and analysis had been identified and qualified through demonstration.

Wolf Creek completed steam generator eddy current inspections for Refueling Outage 18 on April 12, 2011. In accordance with the EPRI steam generator examination guidelines, bobbin coil inspections were expanded in steam generator B due to inspection results associated with wear at anti

-vibration bar locations that resulted in a C-2 condition. In accordance with the EPRI guidelines, another 20 percent of the tubing in steam generator B was inspected. No other scope expansions were required.

In accordance with Technical Specification 5.5.9.c, nine tubes in steam generator B, three tubes in steam generator C, and three tubes in steam g enerator D were plugged based on inspection results indicating they contained flaws with a depth equal to or exceeding 40 percent of the nominal tube wall thickness. The damage mechanism associated with each of the pluggable indications was wear at anti

-vibration bar locations. No tubes in steam generator A require d plugging. No new corrosion damage mechanisms were identified.

The following secondary side maintenance and inspections were performed:

Sludge lancing of all four steam generators. The amount of sludge removed from each steam generator was:

steam generator A, 26 lbs; steam generator B, 34 lbs; steam generator C, 30 lbs; and steam generator D, 27.5 lbs. Foreign object search and retrieval of all four steam generators to locate, identify, and retrieve foreign objects present on the steam generator tube sheet. Foreign object search and retrieval was also performed to inspect for any possible loose parts identified during the eddy current program. Minor foreign objects were identified and addressed within the corrective action program and plant procedures. Visual examination and eddy current testing verified that no degradation was associated with any tubes surrounding the foreign objects.

In-bundle inspection of stea m generator s B and C to inspect the condition of the top of the tube sheet and to augment the foreign object search and retrieval effort. No anomalies were identified during these inspections and the information will be used for trending and to plan future maintenance operations.

Upper steam drum inspection in steam generator s B and C to evaluate the condition of the upper steam drum components with regard to damage of any kind. Ultrasonic testing was also performed on locations susceptible to erosion on the feeding in steam generator s B and C. No anomalies were identified and the information will be used for trending and to plan future maintenance

. Specific documents reviewed during this inspection are listed in the attachment.

These actions constitute completion of the requirements for Section 71111.08-02.0 4. b. No findings were identified.

Findings

.5 Identification and Resolution of Problems (71111.08

-02.05) a. The inspectors reviewed 99 condition reports which dealt with inservice inspection activities and found the corrective actions for inservice inspection issues were appropriate. The specific condition reports reviewed are listed in the documents reviewed section. From this review the inspectors concluded that the licensee ha d an appropriate threshold for entering inservice inspection issues into the corrective action program and ha d procedures that direct a root cause evaluation when necessary. The licensee also had an effective program for applying industry inservice inspection operating experience. Specific documents reviewed during this inspection are listed in the attachment.

Inspection Scope These actions constitute completion of the requirements for Section 71111.08

-02.0 5. b. No findings were identified.

Findings

1R11 Licensed Operator Requalification Program

.1 On June 14, 2011, the inspectors observed a crew of licensed operators in the plant's simulator to verify that operator performance was adequate, evaluators were identifying

and documenting crew performance problems

and training was being conducted in accordance with licensee procedures. The inspectors evaluated the following areas

Inspection Scope Licensed operator performance Crew's clarity and formality of communications Crew's ability to take timely actions in the conservative direction Crew's prioritization, interpretation, and verification of annunciator alarms

Crew's correct use and implementation of abnormal and emergency procedures Control board manipulations

Oversight and direction from supervisors Crew's ability to identify and implement appropriate technical specification actions and emergency plan actions and notifications Compliance with assumptions for manual action timing in Chapter 15 of the USAR The inspectors compared the crew's performance in these areas to preestablished operator action expectations and successful critical task completion requirements. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one quarterly licensed

-operator requalification program sample as defined in Inspection Procedure 71111.11. b. No findings were identified.

Findings

1R12 Maintenance Effectiveness

a. The inspectors evaluated degraded performance issues involving the following risk significant system

Inspection Scope Vital switchgear air conditioning units The inspectors reviewed events such as where ineffective equipment maintenance has resulted in valid or invalid automatic actuations of engineered safeguards systems and independently verified the licensee's actions to address system performance or condition problems in terms of the following:

Implementing appropriate work practices Identifying and addressing common cause failures

Scoping of systems in accordance with 10 CFR 50.65(b)

Characterizing system reliability issues for performance Charging unavailability for performance

Trending key parameters for condition monitoring Ensuring proper classification in accordance with 10 CFR 50.65(a)(1) or

-(a)(2)

Verifying appropriate performance criteria for structures, systems, and components classified as having an adequate demonstration of performance through preventive maintenance, as described in 10 CFR 50.65(a)(2), or as requiring the establishment of appropriate and adequate goals and corrective actions for systems classified as not having adequate performance, as described in 10 CFR 50.65(a)(1)

The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the system. In addition, the inspectors verified maintenance effectiveness issues were entered into the corrective action program with the appropriate significance characterization. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one quarterly maintenance effectiveness sample s as defined in Inspection Procedure 71111.12-05. b. No findings were identified.

Findings

1R13 Maintenance Risk Assessments and Emergent Work Control

a. The inspectors reviewed licensee personnel's evaluation and management of plant risk for the maintenance and emergent work activities affecting risk

-significant and safet y-related equipment listed below to verify that the appropriate risk assessments were performed prior to removing equipment for work:

Inspection Scope June 3, 2011, Component cooling water train A while train B was inoperable

The inspectors selected these activities based on potential risk significance relative to the Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that licensee personnel performed risk assessments as required by 10 CFR 50.65(a)(4) and that the assessments were accurate and complete. When licensee personnel performed emergent work, the inspectors verified that the licensee personnel promptly assessed and managed plant risk. The inspectors reviewed the scope of maintenance work, discussed the results of the assessment with the licensee's probabilistic risk analyst or shift technical advisor, and verified plant conditions were consistent with the risk assessment. The inspectors also reviewed the technical specification requirements and inspected portions of redundant safety systems, when applicable, to verify risk analysis assumptions were valid and applicable requirements were met. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one maintenance risk assessment and emergent work control inspection sample as defined in Inspection Procedure 71111.13 05.

b. No findings were identified.

Findings

1R15 Operability Evaluations

a. The inspectors reviewed the following issues: Inspection Scope March 12, 2011, Emergency diesel generator A jacket water leak May 18, 2011, Source range NI-31 high counts after loss of cavity cooling June 16, 2011, Essential service water system flaw evaluations The inspectors selected these potential operability issues based on the risk significance of the associated components and systems. The inspectors evaluated the technical adequacy of the evaluations to ensure that technical specification operability was properly justified and the subject component or system remained available s o that no unrecognized increase in risk occurred. The inspectors compared the operability and design criteria in the appropriate sections of the technical specifications and USAR to the licensee personnel's evaluations to determine whether the components or systems were operable. Where compensatory measures were required to maintain operability, the inspectors determined whether the measures in place would function as intended and were properly controlled. The inspectors determined, where appropriate, compliance with bounding limitations associated with the evaluations. Additionally, the inspectors also reviewed a sampling of corrective action documents to verify that the licensee was identifying and correcting any deficiencies associated with operability evaluations. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of three operability evaluations inspection samples as defined in Inspection Procedur e 71111.15-04. b. No findings were identified.

Findings

R18 Plant Modifications

.1 Temporary Modifications

a. To verify that the safety functions of important safety systems were not degraded, the inspectors reviewed the temporary modification identified as TMO 10-017, component cooling water modification to radioactive waste building.

Inspection Scope The inspectors reviewed the temporary modification and the associated safety

-evaluation screening against the system design bases documentation, including the USAR and the technical specifications, and verified that the modification did not adversely affect the system operability/availability. The inspectors also verified that the installation and restoration were consistent with the modification documents and that configuration control was adequate. Additionally, the inspectors verified that the temporary modification was identified on control room drawings, appropriate tags were placed on the affected equipment, and licensee personnel evaluated the combined effects on mitigating systems and the integrity of radiological barriers.

These activities constitute completion of one sample for temporary plant modifications as defined in Inspection Procedure 71111.18-05. b. No findings were identified.

Findings

.2 a. Permanent Modifications

The inspectors reviewed key parameters associated with energy needs, materials, timing, heat removal, control signals, licensing basis, and failure modes for the permanent modification identified as the source range gamma metri cs equivalency to Westinghouse detectors

. Inspection Scope The inspectors verified that modification preparation, staging, and implementation did not impair emergency/abnormal operating procedure actions, key safety functions, or operator response to loss of key safety functions; systems, structures and components' performance characteristics still meet the design basis; the modification design assumptions were appropriate; and licensee personnel identified and implemented appropriate corrective actions associated with permanent plant modifications. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one sample for permanent plant modifications as defined in Inspection Procedure 71111.18-05.

b. No findings were identified.

Findings

1R19 Postmaintenance Testing

a. The inspectors reviewed the following postmaintenance activities to verify that procedures and test activities were adequate to ensure system operability and functional capability:

Inspection Scope November 17, 2010, Startup feedwater pump testing after rebuild March 7, 2011, Feedwater regulating bypass valve controller setting adjustment April 1, 2011, Solid state protection system train B after Westinghouse card testing May 17, 2011, Offsite power to engineered safety features transformer A after replacement of Raychem splices June 12, 2011, Component cooling water to thermal barrier heat exchangers after flow balance Procedure SYS EG-205 June 24, 2011, Main turbine overspeed testing after replacement The inspectors selected these activities based upon the structure, system, or component's ability to affect risk. The inspectors evaluated these activities for the following: The effect of testing on the plant had been adequately addressed; testing was adequate for the maintenance performed Acceptance criteria were clear and demonstrated operational readiness; test instrumentation was appropriate

The inspectors evaluated the activities against the technical specifications, the USAR , 10 CFR Part 50 requirements, licensee procedures, and various NRC generic communications to ensure that the test results adequately ensured that the equipment met the licensing basis and design requirements. In addition, the inspectors reviewed corrective action documents associated with postmaintenance tests to determine whether the licensee was identifying problems and entering them in the corrective action program and that the problems were being corrected commensurate with their importance to safety. Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of six postmaintenance testing inspection sampl e s as defined in Inspection Procedure 71111.19-05. b.

Introduction.

The inspectors identified a finding of very low safety significance (Green) involving the failure to follow the requirements of Procedure AP 1 6E-002, "Post Maintenance Testing Development

," for the startup feedwater pump.

Findings

Description.

On November 4-6, 2010 , Wolf Creek worker s performed maintenance on the startup feedwater pump to replace a leaking pump casing gasket. Workers disassembled the pump per the vendor manual instructions and found a split casing gasket and both mechanical seals darkened and cracked from overheating. The pump was reassembled using new parts including bearings, O

-rings, mechanical seals, and casing gasket. The service water cooling lines were also replaced. Wolf Creek Procedure AP 16E-002, "Post Maintenance Testing Development," states that it provides guidelines to develop test criteria for plant equipment after the completion of maintenance activities. The procedure also ensure s proper testing is implemented to prove components, systems, and sub

-systems perform as designed after the completion of maintenance activities.

Furthermore, Procedure AP 16E-002, Revision 9C, step 6.2 and attachments, requires that when a pump is disassembled or replaced, the postmaintenance testing include s a pump-flow capacity test be conducted to determine the capability of the pump to produce the required flow rates with in the range of differential pressure limits. Also, it requires that vibration, bearing

, and lubrication temperatures, motor current, external leakage, and lubrication level are satisfactory.

The inspectors reviewed root cause corrective action 25817-02-14 which created Procedure STN AE-007 to test the pump with no local actions and ensure a minimum recirculation flow of 60,000 pounds per hour for pump protection. The inspectors did not find a discussion of adequate flow. On November 17, 2010, Wolf Creek conducted surveillance Procedure STN AE-007, "Startup Main Feed water Pump Operational Test,"

following the pump reassembly. The only acceptance criteria listed in this procedure was that the motor

-driven feedwater pump start ed from the control room with no local operator action. The test contained no acceptance criteria to ensure that after the completion of maintenance activities, the pump could produce the required flow rates for either low power or emergency operations.

The purpose of the motor

-driven startup feedwater pump is to provide heated feedwater to the steam generators during plant startup and shutdown operations. The startup feedwater pump is a horizontal, multi

-stage, centrifugal pump with a rated maximum flow rate of 260,000 pounds per hour.

Maximum flow through the startup feedwater pump suction lines is limited to 230,000 pounds per hour to prevent excessive tube vibration in the steam generator blowdown regenerative heat exchanger. According to Wolf Creek training materials, Form APF 30E-004-01, Revision 2, "Main Feedwater System," the required steam generator flow rate during plant startup is 200,000 pounds per hour. This is based on the maximum steam generator blowdown rate, the heat lost to ambient surroundings from all main steam lines, and the maximum heatup rate of all main steam lines and the turbine stop valve heat. The startup feed pump is also used in Emergency Management Guideline FR-H1, "Response to Loss of Secondary Heat Sink," step 17, to feed the steam generator until the steam generator level is restored to greater than the minimum level for ensuring an adequate heat sink. The success criteria in emergency operating procedures for feedwater is based on 270,000 pounds per hour for auxiliary feedwater or greater than 6 percent narrow range steam generator level. The emergency operating procedure setpoint document requires 250,000 pounds per hour from each motor-driven auxiliary feedwater pump.

This issue is captured in Condition Report 39494. Wolf Creek issued a new work package to conduct a single

-point pump capacity test and complete the required postmaintenance testing in accordance with Procedure AP 16E-002. Wolf Creek also found that there was not a technical basis for the blowdown heat exchanger vibrations which previously limited the allowable flow through the pump to approximately 200,000 pounds per hour. Wolf Creek initial calculations

, pending final review, show that the pump would be capable of enough flow to provide a heat sink.

Analysis.

The failure to follow Procedure AP 16E-002 for developing test criteria for plant equipment after the completion of maintenance activities is a performance deficiency. The finding is more than minor because it is associated with the Mitigating Systems Cornerstone attribute of equipment performance and it adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609.04, the inspectors determined that the finding had very low safety significance (Green) because it did not result in a loss of system safety function, an actual loss of safety function of a single train for greater than its technical specification allowed outage time, or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding had a cross-cutting aspect in the area of problem identification and resolution. Specifically, Wolf Creek created a testing procedure in response to a root cause evaluation, but did not consider acceptance criteria to ensure that the pump performs acceptably

P.1(d).

Enforcement.

Enforcement action does not apply because the performance deficiency did not involve a violation of regulatory requirements. This finding is of very low safety significance and the issue was entered into the licensee's corrective action program as Condition Report 39494: FIN 05000482/201100 3-0 6 , "Inadequate Acceptance Criteria for Post maintenance Testing of the Startup Feedwater Pump."

1R20 Refueling and Other Outage Activities

a. The inspectors reviewed the outage safety plan and contingency plans for the refueling outage, conducted from March 19 through June 22, 2011 , and June 26-29, 2011, to confirm that licensee personnel had appropriately considered risk, industry experience, and previous site

-specific problems. The inspectors determined that the plan ensured sufficient defense in depth. During the refueling outage, the inspectors observed portions of the shutdown and cooldown processes and monitored licensee controls over the outage activities listed below.

Inspection Scope Configuration management maintains defense in depth , commensurate with the outage safety plan, and in compliance with technical specifications.

Clearance activities were properly tagged and equipment configured to safely support work.

Installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication, accounting for instrument error.

Status and configuration of electrical systems to ensure that technical specifications and outage safety

-plan requirements were met, and controls over switchyard activities.

Monitoring of decay heat removal processes, systems, and components.

Verification that outage work was not impacting the ability of the operators to operate the spent fuel pool cooling system.

Reactor water inventory controls, including flow paths, configurations, and alternative means for inventory addition, and controls to prevent inventory loss.

Controls over activities that could affect reactivity.

Maintenance of secondary containment as required by the technical specifications.

Refueling activities, including fuel handling and sipping to detect fuel assembly leakage.

Startup and power ascension, tracking of startup prerequisites, walkdown of containment to verify that debris removal, and reactor physics testing.

Licensee identification and resolution of problems related to refueling outage activities.

Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of one refueling outage sample and one forced outage inspection sample as defined in Inspection Procedure 71111.20-05. b. No findings were identified.

Findings

R22 Surveillance Testing

a. The inspectors reviewed the USAR, procedure requirements, and technical specifications to ensure that the surveillance activities listed below demonstrated that the systems, structures, and/or components tested were capable of performing their intended safety functions. The inspectors either witnessed or reviewed test data to verify that the significant surveillance test attributes were adequate to address the following:

Inspection Scope Preconditioning

Evaluation of testing impact on the plant

Acceptance criteria

Test equipment

Procedures Jumper/lifted lead controls Test data Testing frequency and method supported operability or functionality Test equipment removal

Restoration of plant systems Fulfillment of ASME code requirements

Updating of performance indicator data Engineering evaluations, root causes, and bases for restoring systems, structures, and components not meeting acceptance criteria were correct Reference setting data Annunciator and alarm setpoints The inspectors also verified that licensee personnel identified and implemented any needed corrective actions associated with the surveillance testing.

The following surveillance testing was observed:

September 8, 2010, Main steam valve AB-V85 inservice valve test April 24, 2011, Residual heat removal room cooler B test April 26, 2011, Filling, venting, and void surveillance of safety injection May 13, 2011, Tan

-delta testing of offsite power underground cables May 14, 2011 , STS PE-018, Containment integrated leak rate test May 18, 2011, Containment isolation valve EJHV8811B inservice test May 24, 2011, Filling, venting, and void surv e illance of residual heat removal train B June 15, 2011, STS IC

-615A, Safety injection signal slave relay test Specific documents reviewed during this inspection are listed in the attachment.

These activities constitute completion of eight surveillance testing inspection sample s as defined in Inspection Procedure 71111.22-05. b. No findings were identified.

Findings

.2 Surveillance Testing Associated with TI

2515/177, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems a. " When reviewing the April 26, 2011, filling, venting, and void surveillance of safety injection and the May 24, 2011, filling, venting, and void surveillance of residual heat removal train B surveillances listed above in Section

==1R22 .1, the inspectors verified that the procedures were acceptable for

(1) testing with shutdown==

operation, maintenance, and subject system modifications; (2)void determination and elimination methods; and

(3) post-event evaluation

. Inspection Scope The inspectors reviewed the procedures used for conducting surveillance tests and the determination of void volumes to ensure that the acceptance criteria were satisfied and would be reasonably assured to remain satisfied until the next scheduled surveillance test (TI 2515/177, Section 04.03.a). Also, the inspectors reviewed procedures used for filling and venting following conditions which may have introduced voids into the subject systems to verify that the procedures acceptably addressed testing for such voids and provided acceptable processes for their reduction or elimination (TI 2515/177, Section 04.03.b). Specifically, the inspectors verified that:

Gas intrusion prevention, refill, venting, monitoring, trending, evaluation, and void correction activities were acceptably controlled by ap proved operating procedures (TI 2515/177, Section 04.03.c.1)

Procedures ensured the system did not contain voids that may jeopardize operability (TI 2515/177, Section 04.03.c.2)

Procedures established that void criteria were satisfied and will be reasonabl y ensured to be satisfied until the next scheduled void surveillance (TI 2515/177, Section 04.03.c.3)

The licensee entered changes into the corrective action program as needed to ensure acceptable response to issues. In addition, the inspectors confirmed that a clear schedule for completion is included for corrective action program entries

that have not been completed (TI 2515/177, Section 04.03.c.5)

Procedures included independent verification that critical steps were completed (TI 2515/177, Section 04.03.c.6) The inspectors verified the following with respect to surveillance and void detection:

Specified surveillance frequency was consistent with technical specification requirements (TI 2515/177, Section 04.03.d.1)

Surveillance frequencies were stated or, when conducted more often than required by technical specifications, the process for their determination was described (TI 2515/177, Section 04.03.d.2)

Surveillance methods were acceptably established to achieve the needed accuracy (TI 2515/177, Secti on 04.03.d.3)

Surveillance procedures included up

-to-date acceptance criteria (TI 2515/177, Section 04.03.d.4)

Procedures included effective follow

-up actions when acceptance criteria are exceeded or when trending indicates that criteria may be approached before the next scheduled surveillance (TI 2515/177, Section 04.03.d.5)

Measured void volume uncertainty was considered when comparing test data to acceptance criteria (TI 2515/177, Section 04.03.d.6)

Venting procedures and practices utilized criteria such as adequate venting durations and observing a steady stream of water (TI 2515/177, Section 04.03.d.7)

An effective sequencing of void removal steps was followed to ensure that gas does not move into previously filled system volumes (TI 2515/177, Section 04.03.d.8)

Qualitative void assessment methods included expectations that the void will be significantly less than allowed by acceptance criteria (TI 2515/177, Section 04.03.d.9)

Venting results were trended periodically to confirm that the systems are sufficiently full of water and that the venting frequencies are adequate. The inspectors also verified that records on the quantity of gas at each location are maintained and trended as a means of pre

-emptively identifying degrading gas accumulations (TI 2515/177, Section 04.03.d.10)

Surveillances were conducted at any location where a void may form, including high points, dead legs, and locations under closed valves in vertical pipes (TI 2515/177, Section 04.03.d.11)

The licensee ensured that systems were not preconditioned by other procedures that may cause a system to be filled, such as by testing, prior to the void surveillance (TI 2515/177, Section 04.03.d.12)

Procedures included gas sampling for unexpected void increases if the source of the void is unknown and sampling is needed to assist in determining the source (TI 2515/177, Section 04.03.d.13)

The inspectors verified the following with respect to filling and venting:

Revisions to fill and vent procedures to address new vents or different venting sequences were acceptably accomplished (TI 2515/177, Section 04.03.e.1)

Fill and vent procedures provided instructions to modify restoration guidance to address changes in maintenance work scope or to reflect different boundaries from those assumed in the procedure (TI 2515/177, Section 04.03.e.2)

The inspectors verified the following with respect to void control: Void removal methods were acceptably addressed by approved procedures (TI 2515/177, Section 04.03.f.1)

The licensee had reasonably ensured that the residual heat removal pump is free of damage following a gas

-related event in which pump acceptance criteria was exceeded (TI 2515/177, Section 04.03.f.2)

Documents reviewed are listed in the attachment to this report.

This inspection effort counts towards the completion of TI 2515/177. See Section

4OA5 for additional information.

b. No findings were identified.

Findings Cornerstone: Emergency Preparedness

1EP4 Emergency Action Level and Emergency Plan Changes

a. The inspector performed an in

-office review of the Wolf Creek APF 06-002-01, "Emergency Action Levels," Revision 15A. This revision made two administrative changes to EAL

-6, "Loss of Electrical Power/Assessment Capability." The change included replacing the abbreviation "D/Gs" with the capitalized and bolded wording "DIESEL GENERATORS," and capitalizing and bolding the wording "NB TRANSFORMERS

." Inspection Scope This revision was compared to its previous revision, to the criteria of NUREG

-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," Revision 1, and to the standards in 10 CFR 50.47(b) to determine if the revision adequately implemented the requirements of 10 CFR 50.54(q). This review was not documented in a safety evaluation report and did not constitute approval of licensee

-generated changes; therefore, this revision is subject to future inspection.

These activities constitute completion of one sample as defined in Inspection Procedure 71114.04-05.

b. No findings were identified.

Findings

4OA1 Performance Indicator Verification

.1 Data Submission Issue

a. The inspectors performed a review of the performance indicator data submitted by the licensee for the 1st Quarter 20 11 performance indicators for any obvious inconsistencies prior to its public release in accordance with Inspection Manual Chapter 0608, "Performance Indicator Program."

Inspection Scope This review was performed as part of the inspectors' normal plant status activities and, as such, did not constitute a separate inspection sample.

b. No findings were identified.

Finding s

.1 Reactor Coolant System Specific Activity (BI01)

a. The inspectors sampled licensee submittals for the reactor coolant system specific activity performance indicator for the period from the 2nd Quarter 20 10 through the 1 st Quarter 20 11. The inspectors used definitions and guidance contained in NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6 to determine the accuracy of the performance indicator data reported during those periods. The inspectors reviewed the licensee's reactor coolant system chemistry samples, technical specification requirements, issue reports, event reports

, and NRC integrated inspection reports for the period of April 1, 2010 , through March 31, 2011 , to validate the accuracy of the submittals. The inspectors also reviewed the licensee's condition report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. In addition to record reviews, the inspectors observed a chemistry technician obtain and analyze a reactor coolant system sample. Specific documents reviewed are described in the attachment to this report.

Inspection Scope

These activities constitute completion of one reactor coolant system specific activity sample as defined in Inspection Procedure 71151-05. b. No findings were identified.

Findings

.2 Reactor Coolant System Leakage (BI02)

a. The inspectors sampled licensee submittals for the reactor coolant system leakage performance indicator for the period from the 2 nd Quarter 2010 through the 1 st Quarter 2011. The inspectors used definitions and guidance contained in NEI Document 99-02, "Regulatory Assessment Performance Indicator Guideline," Revision 6 to determine the accuracy of the performance indicator data reported during those periods. The inspectors reviewed the licensee's operator logs, reactor coolant system leakage tracking data, issue reports, event reports

, and NRC integrated inspection reports for the period of April 1, 2010 , through March 31, 2011 , to validate the accuracy of the submittals. The inspectors also reviewed the licensee's condition report database to determine if any problems had been identified with the performance indicator data collected or transmitted for this indicator and none were identified. Specific documents reviewed are described in the attachment to this report.

Inspection Scope

These activities constitute completion of one reactor coolant system leakage sample as defined in Inspection Procedure 71151-05.

b. No findings were identified.

Findings

4OA2 Identification and Resolution of Problems

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness, Public Radiation Safety, Occupational Radiation Safety, and Physical Protection

.1 Routine Review of Identification and Resolution of Problems

a. As part of various baseline inspection s discussed in previous sections, the inspectors reviewed issues to verify that they were being entered into the Wolf Creek corrective action program at an appropriate threshold. The inspectors verified the program to be addressing issues in a timely manner as well as identifying and correcting adverse trends. The inspectors reviewed attributes that included

Inspection Scope Complete and accurate identification of the problem Timely correction, commensurate with the safety significance Evaluation and disposition of performance issues, generic implications, common causes, contributing factors, root causes, extent of condition reviews, a nd previous occurrences reviews Classification, prioritization, focus, and timeliness of corrective actions.

Minor issues entered into the licensee's corrective action program because of the inspectors' observations are included in the attached list of documents reviewed.

These reviews for the identification and resolution of problems did not constitute any additional inspection samples. They were considered a part of the inspections performed during the quarter and documented in Section 1 of this report.

b. No findings were identified.

Findings

.2 Daily Corrective Action Program Reviews

a. The inspectors performed a daily screening of items entered into the licensee's corrective action program through review of the Wolf Creek's daily corrective action documents.

Inspection Scope The inspectors performed these daily reviews as part of their plant status monitoring activities and did not constitute any separate inspection samples.

b. No findings were identified.

Findings

.3 Semi-Annual Trend Review

a. The inspectors performed a review of the Wolf Creek corrective action program and associated documents to identify trends that could represent a more significant safety issue. The inspectors focused their review on repetitive equipment issues, but also considered the results of daily corrective action item screening s, licensee trending efforts, and licensee human performance results. The inspectors considered the 6-month period of January through June 2011 although some examples expanded beyond those dates where necessary. Inspection Scope The inspectors also reviewed issues documented outside the normal corrective action program in major equipment problem lists, repetitive and/or rework maintenance lists, departmental problem/challenges lists, system health reports, quality assurance audit/surveillance reports, self

-assessment reports, and maintenance rule assessments. The inspectors compared and contrasted their results with the conclusions reached in the Wolf Creek corrective action program trending reports.

The inspectors reviewed corrective actions for problem identification and resolution and human performance cross-cutting themes. These activities constitute completion of a one semi-annual trend inspection sample as defined in Inspection Procedure 71152-05. b. No findings were identified.

Findings

.4 Selected Issue Follow

-up Inspection a. During a review of Wolf Creek corrective action program items, the inspectors noted a condition report documenting over drilling of stud holes on a feedwater regulating valve body. The inspectors reviewed vendor manuals and station procedures for drilling and installing Heilicoil inserts.

The inspectors reviewed vendor calculations for the strength of the joint.

The inspectors interviewed engineers regarding the procedure and determined that work performed was consistent with vendor instructions.

Inspection Scope These activities constitute completion of one in-depth problem identification and resolution sample as defined in Inspection Procedure 71152-05. b. No findings were identified.

Findings

4OA3 Event Follow

-up (71153)

.1 March 19, 2011, Safety Injection

a. Inspection Scope

On March 19, 201 1 , with the reactor in hot standby , the inspector s responded to the control room when Wolf Creek received a safety injection signal for a rapid steamline pressure decrease. The inspectors reviewed control room logs and plant computer data. The inspectors interviewed control room operators about the conditions leading up to the event as well as the plant response. The inspectors reviewed plant operating practices regarding methods of feedwater heating, main steam procedures, and emergency operating procedures. From interviews with several members of the operating crew and plant data before and after the event, the inspectors independently reviewed the sequence of events:

The crew assumed the watch in Mode 1 and reduced reactor power to Mode 3 for Refueling Outage 18.

On March 18, 2011, SYS AE-200, "Feedwater Preheating During Plant Startup and Shutdown," Revision 29, was entered for the plant shutdown.

At midnight, the turbine was tripped in accordance with procedure.

At 12:07 a.m., feedwater temperature and steam flows begin oscillating.

Over the next hour, feedwater temperature and steam flows oscillated. It was later determined that the oscillations were due to manual control of FB

-PIC 300 combined with solenoid valve air leakage. This action was not peer

-checked and control room supervision was not made aware. PIC

-300 controls valve FB

-17A which admits steam to the high pressure feedwater heaters. This is a larg e steam demand. The heaters had several temperature swings. This was not identified until after the safety injection.

At 12:37 a.m., March 19, the reactor entered Mode 2

. At 12:54 a.m., March 19, the reactor entered Mode 3. At 1:00 a.m., letdown automatically isolated at 17 percent pressurizer level due to a cooldown in progress

.

At 1:01 a.m., reactor coolant temperature could not be maintained

, operators shut the main steam isolation valves to stop the cool down.

Reactor coolant system temperature subse quently recovered to 560°F. Main feedwater pump A turbine was subsequently tripped from the control room.

At 2:21 a.m., feedwater preheating was secured using SYS AE-200. Feedwater temperature decreases. With main steam isolation valves shut, the feedwater heaters continued to draw steam from the main steam header.

Steam line temperature decreases

. At 3:19 a.m., operators re

-open ed the main steam isolation valve bypass valves

. To open the main steam isolation valves, operators entered Procedure SYS AB-120. This procedure is intended for use in Mode 4 with a maximum steam line pressure of 300 psi. Steam line pressure was approximately 1000 psi. Precaution 4.5 and step 6.14.2 require that main steam isolation valve differential pressure be less than 20 psi to open a main steam isolation valve.

Senior reactor operators and management oversight mark these step s as not applicable

. At 4:04 a.m., upon opening main steam isolation valve C, a negative steam line pressure rate on steam line C triggered an automatic safety injection signal.

Operators entered EMG E-0, "Response to Reactor Trip or Safety Injection."

The pressurizer power-operated relief valve began cycling due to the pressure increase from the high head centrifugal charging pumps adding inventory to the reactor coolant system.

At 4:11 a.m., the safety injection signal was reset and Technical Specification 3.0.3 was entered for both trains of emergency core cooling system inoperable because automatic safety injection signal was blocked.

At 4:12 a.m., high pressure injection was terminated when the boron injection tank valves were shut.

At 4:18 a.m., pressurizer power-operated relief valve stops cycling and closes. At 4:23 a.m., Wolf Creek transitioned to Procedure EMG ES-03, "Safety Injectio n Termination."

At 4:44 a.m., normal letdown flow from the reactor coolant system was re

- established to reduce pressurizer level from a high of 88 percent. At 5:20 a.m., Wolf Creek completed Procedure EMG ES

-03 and transitioned to Procedure OFN EM

-024 "Safety Injection Recovery

."

At 5:54 a.m., Wolf Creek notified the headquarters operations officer of the event by making event notification 46685 per 10 CFR 50.72(b)(2)(iv)(A) which is a 4-hour report for emergency core cooling system discharge into the reactor coolant system.

At 6:39 a.m., both reactor trip breakers were closed using SYS SF

-120 and Technical Specification 3.0.3 was exited.

At 11:21a.m., Wolf Creek updated event notification 46685 with additional information regarding safety system actuation and loss of an accident mitigation safety system after the inspectors identified that these 8

-hour reporting requirements may also apply to this event. Condition report 34995 was written for the potentially missed reports.

Wolf Creek subsequently left the main steam isolation valves shut and cooled the plant using the atmospheric relief valves to a temperature where the residual heat removal system could be placed in service.

b. Findings

Introduction.

A Green self

-revealing cited violation of Technical Specification 5.4.1.a, "Administrative Procedures

," was reviewed involving the failure to correct a previous violation for an inadequate main steam system procedure. Specifically, Procedure SYS AB-120 was not corrected to establish appropriate conditions to open a main steam isolation valve. The inadequate procedure resulted in a safety injection

. Description

. The inspectors reviewed a March 5, 2010, event involving excessive steam generator level swell and feedwater isolation following opening of a main feedwater isolation valve described in Condition Report 23938 and noncited violation NCV 05000482/2010004

-01. Wolf Creek determined the cause of the March 5, 2010, P-14 feedwater isolation was an inadequate means of determining the pressure difference across the main steam isolation valves using control room pressure indicators.

Procedure SYS AB-120 , "Main Steam and Steam Dump Startup and Operation,"

Revision 24, used an acceptance criterion of less than 25 psi differential pressure to allow opening of a main steam isolation valve. The procedure directed the operators to determine valve differential pressure using control room indicators before opening the main steam isolation valves. The control room instruments have ranges from 0 to 1300 psi or greater with a 25 psi scale and are accurate to within plus or minus 25 to 38 psi. The inspectors concluded the apparent cause evaluation in Condition Report 23938 appropriately determined that instrument uncertainty equal to or greater than the procedure's acceptance criteria was not reasonable. Subsequently, Wolf Creek revised Procedure SYS AB-120 to direct operators to determine differential pressure using locally installed instruments in lieu of the control room pressure indicators, but this change was only implemented for steam line pressures below 300 psi. Additional changes were made to several procedures which reduced the allowable steam generator level band when opening a main steam isolation valve.

Procedure SYS AB-120 revisions did not address steam pressures above 300 psi nor were its precautions and limitations updated to reflect main control board instrumentation accuracy. On March 19, 2011, Wolf Creek was in Mode 3 shutting down for a refueling outage. The steam header pressure was 1000 psi. At 1:01 a.m., operators shut all main steam isolation valves due to an excessive cooldown. Several hours later, the operators began to open the valves using Procedure SYS AB-120. A tighter acceptance criteri on of 20 psi differential pressure was specified in the procedure before opening a main steam isolation valve.

Wolf Creek operators did not use local instruments as specified by the procedure. Instead they used control room instruments to determine main steam line pressures on both side of the main steam isolation valves without considering that the instrument uncertainty exceeded the range of acceptance criteria.

While the control room pressure and temperature instruments indicated that the differential pressure w as acceptable, actual differential pressure was about 200 psi. When main steam isolation valve C was opened, a safety injection signal occurred. The inspectors reviewed corrective actions for Procedure SYS AB-120 and found several missed opportunities to correct the deficiency. On October 18, 2010, Condition Report 29168 was written stating "Guidance for opening MSIVs not good above 35 psi steam press

," as its problem description. Wolf Creek reviewed Condition Report 30453 which responded to noncited violation NCV 05000482/2010004

-01 and appropriately concluded that the evaluation was flawed for two reasons. First, Condition Report 30453 failed to incorporate the instrument uncertainty issue previously identified in Condition Report 29168 into the precautions and limitations of Procedure SYS AB-120. Second, Condition Report 30453 failed to address the full range of anticipated plant conditions which may require opening a main steam isolation valve, specifically steam pressures above 300 psi. The inspectors concluded the failure to implement comprehensive corrective actions to address the March 5, 2010, event directly contributed to the March 19, 2011, inadvertent safety injection event and constituted a failure to restore compliance for noncited violation NCV 05000482/2010004

-01.

The inspectors reviewed the safety impact of the safety injection transient on Wolf Creek. Actual safety impacts included a waterhammer on the main steam lines. This caused a partial failure of main steam isolation valve actuator to bonnet gaskets. The pressurizer power-operated relief valve 455 cycled seven times.

Main feedwater was lost when the feedwater isolation valves received a close signal from the safety injection. Emergency core cooling system injection check valve BB8948C experienced body

-to-bonnet gasket leakage. The pressurizer started at 17 percent level and filled to 88 percent level until letdown was reestablished. Inadvertent safety injection has the potential to challenge the pressurizer safety valves and escalate to a loss of coolant accident if not terminated.

Analysis.

The failure to correct deficiencies in Procedure SYS AB-120 for steam pressures above 300 psi was a performance deficiency. The inspectors determined that this finding was more than minor because it impacted the equipment performance attribute for the Initiating Events Cornerstone and it affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safet y functions during shutdown as well as power operations. Specifically, this issue relates to the configuration control attribute for shutdown equipment alignment. The inspectors evaluated the significance of this finding using Inspection Manual Chapter 0609.04. Assuming worst case degradation, the finding resulted in exceeding the technical specification limit for reactor coolant system leakage due to the pressurizer power-operated relief valve cycling. Therefore, the inspectors screened the finding to a Phase 2 review by the senior reactor analyst.

The senior reactor analyst used the Wolf Creek SPAR Model and concluded that the incremental core damage probability was 3.7E-7 , Green. The inspectors found that the cause of the finding has a cross

-cutting aspect in the area of problem identification and resolution associated with the corrective action program. Specifically, several evaluations failed to include an adequate extent of condition review that identified that the procedures were inadequate for opening a main steam isolation valve at system pressures above 300 psi P.1(c). Enforcement

. Technical Specification 5.4.1.a requires that procedures be established, implemented, and maintained covering the activities described in Regulatory Guide 1.33, Revision 2, Appendix A. Regulatory Guide 1.33, Revision 2, Appendix A, Section 3.i , requires procedures for the startup, operation, and shutdown for the main steam system.

Wolf Creek Procedure SYS AB-120, "Main Steam and Steam Dump Startup and Operation," Revision 27 , implements these requirements for the main steam system.

Contrary to the above, from March 5, 2010, to March 19, 201 1, Wolf Creek Procedure SYS AB

-120 had not been maintained to cover activities for the startup, operation and shutdown of the main steam system. Specifically, Procedure SYS AB-120, Revision 27, contained inadequate steps necessary to open a main steam isolation valve without causing a safety injection signal.This issue and the corrective actions are being tracked by the licensee in Condition Report 34964. Due to the licensee's failure to restore compliance from previous NCV 05000482/2010004

-01 within a reasonable time after the violation was identified, this violation is being cited as a Notice of Violation consistent with Section 2.3.2 of the Enforcement Policy: VIO 05000482/2011003

-0 7 , "Failure to Correct Procedure for Opening Main Steam Isolation Valves" (EA-11-149).

.2 March 21, 2011, Low Temperature Overpressure System Actuation

.

a. Inspection Scope

On March 21, 2011, Wolf Creek was shutdown for a refueling outage.

While cleaning the r eactor coolant system, operators failed to maintain reactor coolant system pressure below 350 psi. When charging was increased for the clean

-up , the low temperature overpressure setpoint was exceeded causing pressurizer power-operated relief valve 455 to lift three times. The inspectors interviewed reactor operators, reviewed control room logs, procedures, pressure and temperature limits report, Licens e Amendment 130, and plant computer data.

b. Findings

Introduction.

The inspectors reviewed a self

-revealing Green noncited violation of Technical Specification 5.4.1.a , "Procedures," for failure to maintain pressure below the low pressure overpressure protection setpoint.

Description

. On March 21, 2011, Wolf Creek was adjusting the chemical and volume control system to inject hydrogen peroxide into the reactor coolant system to induce a crud burst to reduce system radioactivity for the refueling outage. Letdown flow was at approximately 63 gpm. The unit was in Mode 5 with the pressurizer solid and maintaining reactor coolant system temperature at 160°F and 350 psig pending reactor coolant system cleanup. The pres surizer is considered 'solid' when it is water filled because water is not compressible when compared with a gas bubble. Charging and letdown were in the process of being increased in order to increase the rate of reactor coolant system cleanup. At 2:52 p.m., power-operated relief valve 455A cycled three times over the following 4 minutes when reactor coolant system pressure increased to its lift setpoint of 415 psig. Reactor coolant system pressure control was subsequently reestablished at 350 psig when letdown flow was increased to approximately 120 g pm. During interviews, the operators stated that the charging header controller was adjusted before letdown, and that it was sluggish at the low pressure.

The procedure only state d to maintain pressure and did not provide specific guidance. At the time, operators had a band of 330

-350 psig to maintain, and the operator s stated that the normal charging pump controller was sluggish at its reduced operating pressure. The operator s stated that the charging pump controller was increased three times and on the third time, a large increase in charging was received.

The inspectors reviewed plant computer data and found that when charging header pressure was initially increased without increasing letdown flow from the residual heat removal system, the reactor coolant system pressure rapidly increased. As the 4 minute event progressed, the normal charging pump controller was adjusted several times while letdown was progressively increased. Charging header pressure and flow drove the increases in reactor coolant system pressure. The three lifts of the power operated relief valve were due to the changes in charging header pressure with a solid pressurizer.

The inspectors reviewed Procedures GEN 00-006 and SYS BG-120 and found that they did not contain any precautions or limitations regarding the reactor coolant system pressure response to a sluggish charging controller with a water

-solid plant. There were no instructions that letdown should have been increased first and to adjust charging second, to match. Procedure GEN 00-006 , step 6.44.8.3 only stated to maintain a pressure band of 325

-350 psig when adjusting charging flow. Although the procedures had several steps to maximize letdown and charging for reactor coolant system clean

-up, there were no specific steps on how to perform this, and there were no continuous action steps or precautionary steps to prevent over

-pressurizing the reactor coolant system. The inspectors reviewed the "Just in Time Training

" for the refueling outage and identified that it contained guidance on raising letdown to 120 g pm and subsequently taking the plant solid. It did not contain guidance or lessons on manipulating letdown with the plant solid.

With the reactor coolant system pressure at the upper end of the band specified in Procedure GEN 00-006, letdown would be appropriate to adjust first to prevent the lifting of relief valves.

If the reactor coolant system was solid at the lower end of the pressure band specified in Procedure GEN 00-006, adjusting charging first would be appropriate to avoid a decrease in reactor coolant system pressure that could meet the reactor coolant pump trip criteria.

Analysis.

Failure to maintain pressure below the power operated relief valve setpoint was a performance deficiency. The performance deficiency was more than minor because it impacted the Initiating Events Cornerstone objective of configuration control to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The significance of the finding was determined using Inspection Manual Chapter 0609, "Significance Determination Process," Appendix G, Checklist 2, and determined to be of very low safety significance (Green), because it did not cause the loss of mitigating capability of core heat removal, inventory control, power availability, containment control, or reactivity control. Additionally, the finding also did not cause any low temperature overpressure technical specifications to be exceeded. The inspectors found that the cause of the finding had a cross

-cutting aspect in the area of human performance. Specifically, operators had to rely on skill of the craft when procedures should have supplied more instruction for manipulating charging and letdown with the pressurizer water solid

H.2.c]. Enforcement

. Wolf Creek Technical Specification 5.4.1.a, "Procedures," requires, in part, that written procedures shall be established, implemented and maintained for the activities recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, "General Plant Operating Procedures,"

Section 2.j , requires procedures for hot standby to cold shutdown. Procedure GEN 00-006, "Hot Standby to Cold Shutdown," Revision 76, implements this procedure. Procedure GEN 00

-006, Step 6.44.8.3 required the licensee to maintain a pressure band below 350 psig when manipulating charging flow. Contrary to the above, on March 21, 2011, Wolf Creek did not implement Procedure GEN 00-006 , step 6.44.8.3 , to maintain a pressure band below 350 psig when manipulating charging flow. Because the finding is of very low safety significance and has been entered into the licensee's corrective action program as Condition Report 35244, this violation is being treated as a noncited violation consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000482/201100 3-0 8 , "Failure to Maintain Reactor Coolant System Pressure Below Relief Valve Setpoint

."

.3 April 5, 2011, Vital Switchgear Room Fire

a. Inspection Scope

The inspectors responded to a fire in the switchgear rooms and to the control room. The inspector s interviewed fire brigade leaders and the control room shift manager regarding the fire alarms and the fire brigade response and examined the damage inside of nonvital inverter PN009. The inspectors observed postfire actions to ventilate the area to remove the smoke and Halon.

b. Findings

Introduction

. T he inspectors reviewed a self

-revealing Green noncited violation of License Condition 2.C.5 for failure to implement adequate fire impairments which affected both trains of vital ac and dc switchgear.

Description

. On March 26, 2011, Wolf Creek implemented a breach authorization requiring a continuous fire watch because the doors between vital ac and dc switchgear rooms were propped open. The se doors are 3

-hour fire barriers. This was done to allow the train B air conditioning unit and ventilation to provide cooling t o the train A switchgear in accordance with Procedure SYS GK-200, "Inoperable Class IE A/C Unit." With the train A air conditioning unit out of service, two sets of double doors were propped open between vital ac switchgear trains A and B on the 2000' elevation of the control building. On April 5, 2010, Wolf Creek completed preventive maintenance on nonvital inverter PN009 which is located in the 2000' elevation train A vital switchgear room. Wolf Creek was preparing to test nonvital inverter PN009 and reenergized it for about 20 minutes. Two electricians were at the inverter cabinet in the train A vital switchgear room when smoke began emanating from the top of the cabinet. The electricians shut off the dc input and opened the ac output breakers on the lower door of the cabinet. The Halon actuation alarm sounded indicating that Halon would discharge into the room in 30 seconds. One electrician told the fire watch that it was necessary to evacuate. The two electricians and the fire watch were egressing through the north missile door when the Halon system discharged. The breached doors between ac switchgear rooms were not shut. The fire brigade responded and removed an extension cord and shut the doors between the vital ac switchgear rooms. The fire brigade found only smoke and Halon in the rooms and no fire at PN009. Subsequent examination by Wolf Creek and the vendor found that vendor errors in labeling the terminals caused an excessive current in an adjacent transformer which caused the fire. Both transformers were replaced. The vendor stated that no other damage occurred. Condition Report 36719 was written on the inadequate fire watch response.

The inspector s interviewed the April 5 fire watch and found that he thought Halon was going to discharge into both the train s A and B vital switchgear rooms. Thus, he would have to egress through the north missile door and not to the train B switchgear room. He understood his duty to shut the doors upon alarm, but indicated that he would not be able to remove the extension cord, shut the doors, and exit within 30 seconds. The fire watch stated that removing the extension cord and shutting the doors would likely take 3 to 4 minutes. The inspectors found that the Halon system was designed to discharge into the switchgear room with the alarming smoke detectors. The fire watch also stated that he left the room without shutting the doors because the electricians instructed him to leave the room prior to the Halon actuation.

The inspectors found that the only written instructions for fire watches was the statement on the fire impairment which said "Per AP 10-104, section 6.1.9 (SYS GK-200), in case of fire or Halon discharge, close doors 3301 1 & 33023 and exit area and notify control room." Wolf Creek relies on training and reading of the fire impairment to understand the compensatory action. The inspectors reviewed the design of the 1301 Halon system and found that the system was sized to extinguish a fire in one switchgear room only.

The inspectors found that with doors 33011 and 33023 (each a set of double doors) open between vital switchgear rooms, the Halon system would not have been successful at extinguishing a fir e. The inspectors reviewed written statements from the fire brigade, the fire watch, and the electricians. The inspectors reviewed Procedure AP 10-107, "Fire Protection Impairment Control ," and Procedure APF 10-104-01, "Breach Authorization Permit

," and found that the requirements of the breach permit were not met because the fire watch failed to close doors 33011 and 33023 during an actual fire. Procedure AP 10-104 , step 5.62 , states, in part, that the bound ary watch must be able to clear any cord or tool crossing a breached barrier and to notify the control room if any condition in which a breached barrier cannot be closed within the time requirements. The inspectors reviewed form APF 10-104-01, breach authorization, for the 2000' and 2016' elevation switchgear rooms and found no quantitative timing requirements for closing the doors. The inspectors concluded that a 30

-second acceptance criterion wa s critical because open doors would prevent the Halon system from reaching the necessary concentration to extinguish a fire. The inspectors found that the breach permit was not met because the fire watch did not close the doors and that the breach permit was inadequate because it did not contain timed acceptance criteria necessary to ensure the success of the Halon system. On April 12, 201 1 , the inspectors interviewed a different vital ac switchgear room fire watch and found that the watch was also not clear on their duty to shut the doors regardless of what other workers tell them to do. The fire watch did have correct knowledge of the Halon system, the 30

-second delay between the alarm and discharge, and what room to egress to depending on the fire location. The inspectors shared this with the outage control c enter. On April 14, 201 1, Wolf Creek inhibited the Halon systems for the r od-drive motor generator set room, all vital dc switchgear rooms, and the vital ac switchgear rooms. Wolf Creek judged it more important to ensure that the fire watches shut the fire doors rather than have an ineffective automatic Halon system actuation.

On April 13, 201 1 , Wolf Creek initiated new training for all fire watches to ensure they had copies of the breach permits. The inspectors interviewed fire watches on April 14, 201 1 , and found that the watches had a complete and correct understanding of their duties.

The inspectors found that the inability to close these fire doors was identified in Refueling Outage 16 in Condition Report 2008-1357. Actions included protective equipment and training to remove cables crossing the open doors, but the inspectors concluded that those corrective actions did not ensure proper fire watch actions on April 5, 2011. Corrective actions included "

APF 10-104-06 to include Special Requirements for Boundary Watch

." Although a new section of the breach impairments was created, it was typically not utilized when breaching vital switchgear doors.

As an extent of condition review, the inspectors reviewed previous fire impairments for Procedure SYS GK-200 for open fire doors on the 2000' and 2016' control building elevations. The inspectors found that Procedure SYS GK-200 had been implemented 3 times over the prior year of operation representing approximately 36 days of impairments for both trains of ac and dc switchgear and batteries.

Analysis.

The failure to implement adequate fire watches that ensured the success of the Halon system was considered a performance deficiency. The performance deficiency was considered more than minor because it impacted the Initiating Events Cornerstone and its objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations.

Specifically, the fire area of the protection against external factors attribute was impacted. The inspectors used Inspection Manual Chapter 0609.04 to screen the finding to Inspection Manual Chapter 0609, Appendix F , because the fire protection defense-in-depth strategies involving automatic suppression, fire barriers, administrative controls were degraded. Because the subject finding was not clearly covered by the approach used in Appendix F, the senior reactor analyst performed a Phase 3 analysis. The doors were open for 36 days, so a 3 6-day exposure period (EXP) was used. The FI), Severity Factor (P SF) and the probability of manual suppression before damage (P MS). The Fire Mitigation FM) was calculated as follow s: FM FI

• P SF

• P MS

• EXP = 2.0 x 10

-2/year

• 0.1

• 0.1

• 36 days ÷ 365 days/year

= 1.97 x 10

-5 The analyst assumed that if the fire grew to a point that it could spread to the opposite train, it would actuate the opposite train's Halon system and cause an isolation of all ventilation. However, there was no credible source of flammable materials that would cause the growth of the fire into the opposite train's switchgear. Therefore, the analyst quantified the conditional core damage probability (CCDP) for the failure of Switchgear NB01 using the Standardized Plant Analysis Risk Model for Wolf Creek Station, Revision 8.15. The resulting CCDP was 8.3 x 10-4. The final change in core damage frequency FM

• CCDP = 1.97 x 10

-5

• 8.3 x 10

-4 = 1.6 x 10

-8 Therefore, this finding was determined to be of very low safety significance (Green).

The inspectors found that the cause of the finding had a cross-cutting aspect in the area of problem identification and resolution. Specifically, corrective actions from 2008 ineffective fire watches did not prevent recurrence of the April 5, 2011 , inadequate fire watch P.1.d]

. Enforcement

. License condition 2.C.(5) states, in part, that the licensee shall maintain in effect all provisions of the approved fire protection program as described in the Standardized Nuclear Unit Power Plant System USAR for the facility through Revision 17, the Wolf Creek site addendum through Revision 15 , and as approved in the safety evaluation report through Supplement 5, Amendment s 191 and 193. AP 10-100, fire protection program, states, in part

, that AP 10-104, "Breach Authorization," is part of the fire protection program. Procedure AP 10-104 , step 5.62 , states, in part, that the boundary watch must be able to clear any cord or tool crossing a breached barrier and to notify the control room if any condition in which a breached barrier cannot be closed within the time requirements.

Procedure AP 10-104 , steps 6.1.8 and 6.1.9, require, in part, that a continuous fire watch shall be established for the vital switchgear rooms because open doors will reduce Halon concentration and expose redundant trains to the same fire. Contrary to the above, prior to April 14, 2011, the licensee failed to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the licensee used an ineffective fire barrier breach permit system that did not ensure that the Halon systems would effectively extinguish fires because the fire watches could not clear any cord or tool crossing a breached barrier and did not notify the control room of a condition in which a breached barrier could not be closed within the time requirements. The licensee entered this issue into their corrective action program as Condition Report 36719. Because this violation was of very low safety significance and it was entered into the corrective action program, this violation is being treated as a noncited violation, consistent with the NRC Enforcement Policy , Section 2.3.2: NCV 05000482/2011003

-0 9 , "Inadequate Fire Watch Defeats Halon Fire Suppression in Vital Switchgear Rooms During Fire."

.4 (Closed) Licensee Event Report (LER) 2006-003-00, Indications Discovered on Pressurizer during Preplanned Inservice Inspections

On October 11, 2006, during Refueling Outage 15, engineering personnel performing preplanned inservice examination of the pressurizer nozzle to safe end dissimilar metal welds identified five circumferential flaw indications.

Three indications were located in the surge nozzle dissimilar metal weld, one indication was in the safety nozzle C dissimilar metal weld, and one indication was in the relief nozzle dissimilar metal weld. The locations were all part of the reactor coolant system pressure boundary.

There was no evidence of reactor coolant system pressure boundary leakage.

The most probable mechanism responsible for the indications was primary water stress corrosion cracking.

Wolf Creek Generating Station was in Mode 5, cold shutdown.

Weld overlay repairs of the flaw indications were performed prior to the unit's return to power operations.

The inspectors reviewed LER 05000482/200 6-00 3-00 to verify that the cause was identified and that corrective actions were appropriate.

This LER is closed.

.5 (Closed) Notice of Violation VIO 05000482/2010006-05 , Failure to

Correct NRC Identified NCV Apparent Cause Evaluation Vice Root Cause Evaluation for Essential Service Water The violation involved the failure to perform an adequate cause evaluation and to take corrective actions to preclude repetition for a significant condition adverse to quality. Although determined to be of very low safety significance (Green), this violation was cited in Notice of Violation 05000482/2010006

-05 because not all of the criteria specified in Section 2.3.2 of the NRC Enforcement Policy were satisfied (EA 160). Specifically, the Wolf Creek Generating Station failed to restore compliance within a reasonable time for a previously NRC identified noncited violation as documented in NRC Inspection Report 05000482/2009007

-03. The inspectors reviewed the corrective actions completed by the licensee and verified that the cause was identified and that corrective actions were appropriate.

This violation is closed.

4OA5 Other Activities

.1 a. (Closed) NRC TI 2515/177, Managing Gas

Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems (NRC Generic Letter 2008

-01) As documented in Sections

1R04

.1 and 1R22 of this report, the inspectors confirmed the acceptability of the described actions for the high pressure safety injection system and the containment spray system. This inspection effort counts towards the completion of TI 2515/177 which is closed in this inspection report.

Inspection Scope The inspectors evaluated whether the licensee maintained documents, installed system hardware, and implemented actions with the information provided in their response to NRC Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems

." Specifically, the inspectors verified that the licensee has implemented or was in the process of implementing the commitments, modifications, and programmatically controlled actions described in the response to Generic Letter 2008-01. The inspectors conducted their review in accordance with T I 2515/177 and considered the site

-specific supplemental information provided by the Office of Nuclear Reactor Regulation (NRR) to the inspectors.

b. The inspectors reviewed the licensing basis, design, testing, and corrective actions as specified in the TI. The specific items reviewed and any resulting observations are documented below.

Inspection Documentation Licensing Basis

. The inspectors reviewed selected portions of licensing basis documents to verify that they were consistent with the NRR assessment report and that the licensee properly processed any required changes. The inspectors reviewed selected portions of technical specifications, technical specification bases, and the U SAR. The inspectors also verified that applicable documents that described the plant and plant operation, such as calculations, piping and instrumentation diagrams, procedures, and corrective action program documents, addressed the areas of concern and were changed, if needed, following plant changes. The inspectors confirmed that the licensee performed surveillance tests at the frequency required by the technical specifications. The inspectors verified that the licensee tracked their commitment to evaluate and implement any changes that will be contained in the technical specification task force traveler.

Design The inspectors verified that the licensee ha d identified the applicable gas intrusion mechanisms for their plant.

. The inspectors reviewed selected design documents, performed system walkdowns, and interviewed plant personnel to verify that the licensee addressed design and operating characteristics. Specifically:

The inspectors verified that the licensee had established void acceptance criteria consistent with the void acceptance criteria identified by NRR. If NRR acceptance criteria were not met, then the inspectors verified that the licensee has justified the deviations. The inspectors also confirmed that the range of flow conditions evaluated by the licensee was consistent with the full range of design basis and expected flow rates for various break sizes and locations.

The inspectors noted that the licensee used the methods developed by Westinghouse to estimate the suction voids emergency core cooling system pumps. Westinghouse documented their review and test results in WCAP-16631-NP, "Testing and Evaluation of Gas Transport to the Suction of ECCS Pumps." Wolf Creek used WCAP-16631-NP to show that GOTHIC can acceptably predict quantitative void transport behavior. However, the inspectors noted that test configuration and conditions differed from actual plant configuration and conditions. These methods relied on industry testing documented by Westinghouse and used the GOTHIC computer code to better estimate the impacts resulting from voiding in the emergency core cooling systems.

The licensee had received analyses for their facility based upon the simplified equation developed by Westinghouse, which would more accurately estimate the void sizes allowed on the suction of the emergency core cooling pumps without affecting operability. In addition, the license had received a revised estimate of water hammer effects developed by Fauske on the pump discharges for their emergency core cooling systems. These analyses would replace the use of GOTHIC. These analyses allow for a more realistic estimate of void sizes on both the suction and discharge of the emergency core cooling system pumps. The licensee had not accepted these analyses at the time of this inspection.

The inspectors discussed with NRR that the licensee had used these methods.

The ultimate acceptability of these methods required further evaluation by NRR to:

(1) better understand the acceptability of the application of the revised test results contained in WCAP

-16631-NP to void assessment analysis; (2)better understand and evaluate the use of the simplified equation; and (3)assess potential generic implications. The licensee documented these outstanding issues in Condition Report 39943.

The inspectors selectively review ed applicable documents, including calculations, and engineering evaluations with respect to gas accumulation in the emergency core cooling systems. Specifically, the inspectors verified that these documents addressed venting requirements, aspects where pipes were normally void such as some spray piping inside containment, void control during maintenance activities, and the effect of debris on strainers in containment emergency sumps causing accumulation of gas under the upper elevation of strainers and the impact on the required net positive suction head.

The inspectors conducted a walk down of selected regions of the emergency core cooling systems in sufficient detail to assess the licensee's walk downs. The inspectors completed a full containment spray system alignment as documented in Section

1R04 .

The inspectors also verified that the information obtained during the licensee's walkdown was consistent with the items identified during the inspectors

' independent walk down.

The inspectors verified that piping and instrumentation diagrams and isometric drawings that describe the residual heat removal and safety injection system configurations. The review of the selected portions of isometric drawings considered the following

1. High point vents were identified.

2. High points without vents were recognizable.

3. Other areas where gas could accumulate and potentially impact operability, such as at orifices in horizontal pipes, isolated branch lines, heat exchangers, improperly sloped piping, and under closed valves, were described in the drawings or in referenced documentation. 4. Horizontal pipe centerline elevation deviations and pipe slopes in nominally horizontal lines that exceed specified criteria were identified.

5. All pipes and fittings were clearly shown.

6. The drawings were up to date with respect to recent hardware changes, and that any discrepancies between as

-built configurations and the drawings were documented and entered into the corrective action program for resolution.

The inspectors verified that the licensee had completed their walkdowns and selectively verified that the licensee identified discrepant conditions in their corrective action program and appropriately modified affected procedures and training documents. The inspectors determined that the licensee appropriately considered the differing gas intrusion mechanisms with one exception. The inspectors noted that the licensee failed to analyze whether vortexing would occur in their containment spray additive tank. The details of this issue are described in Section 4OA5.1.e of this report.

Testing. The inspectors reviewed selected surveillance, postmodification test, and postmaintenance test procedures and results implemented during power and shutdown operations to verify that the licensee had approved and used procedures that addressed gas accumulation and/or intrusion into the subject systems. This review included the verification of procedures used for conducting surveillances and determination of void volumes to ensure that the licensee satisfied the established void criteria with reasonable assurance until the next scheduled void surveillance. Also, the inspectors reviewed procedures used for filling and venting following conditions that may have introduced voids into the subject systems to verify that the procedures addressed testing for such voids and provided processes for their reduction or elimination. The inspectors observed the performance of the emergency core cooling system void surveillance as documented in Section

1R22 .

Corrective ActionsBased on this review, the inspectors concluded that reasonable assurance exists the licensee will continue to implement the requirements of Generic Letter 2008-01 and will complete all outstanding items.

T hi s TI is closed.  : The inspectors reviewed selected actions from the 2011 assessment review and sampled other corrective action program documents to assess how effectively the licensee addressed the issues in their corrective action program associated with Generic Letter 2008-01. In addition, the inspectors verified that the licensee implemented appropriate corrective actions for condition reports identified in the 9-month and supplemental responses. The inspectors determined that the licensee had initiated a large number of corrective actions in response to previous events at their facility. The inspectors determined that the licensee had effectively implemented the actions required by Generic Letter 2008-01.

1. Findings Introduction

. The inspectors identified a Green noncited violation of 10 CFR Part 50, Appendix B, Criterion III, "Design Control" for the failure to translate the design basis into instructions, procedures, and drawings. The inspectors found that the licensee failed to assess whether vortexing occurred in the containment spray additive tank during a design-basis accident.

DescriptionThe system used an eductor driven by discharge flow from both of the containment spray pumps to draw sodium hydroxide from the single chemical additive tank during a design-basis accident. Vacuum breakers allow ed air into the tank as the liquid drained. . The inspectors evaluated licensee activities related to evaluation of gas intrusion into their emergency core cooling systems. The inspectors questioned whether air entrainment in the containment spray system, as a result of vortexing in the containment spray additive tank, affected the ability of the containment spray system to remain full of water and meet the accident flow requirements. The licensee did not have a calculation to determine whether vortexing would occur in their containment spray additive tank at the required design flow rates. The licensee initiated Condition Report 38715 to document this deficiency; initiated actions to calculate the effects of vortexing in the containment spray additive tank during design basis flows; and established a Mode 3 restraint related to completing the calculation to ensure that containment spray would be operable as required.

Calculati on EN M-024, "Critical Submergence in Containment Spray Additive Tank (TEN01) to Avoid Vortex," Revision 0, concluded that vortexing would not occur.

Analysis.

Failure to implement design control measures to analyze whether containment spray piping remained full of water was a performance deficiency. This finding was more than minor because it affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of the containment spray system to respond to initiating events and prevent undesirable consequences. Specifically, the inspectors had reasonable doubt on the capability of the containment spray system to properly inject because of vortexing in the containment spray additive tank. The inspectors performed the significance determination using Inspection Manual Chapter 0609.04. The finding was determined to be of very low safety significance (Green) because it was a design or qualification deficiency confirmed not to result in loss of operability or functionality. Although the failure to have this calculation had existed since original construction, the inspectors determined this finding reflected current performance since the licensee was required to evaluate likelihood of tanks allowing gas intrusion into the emergency core cooling systems in response to Generic Letter 2008-01, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems." Consequently, this finding had problem identification and resolution cross-cutting aspects associated with the corrective action program in that the licensee did not evaluate thoroughly the potential for gas intrusion from all possible tanks P.1(c). Enforcement

.2 (Closed) NRC T

I 2515/183, Followup to the Fukushima Daiichi Nuclear Station Fuel Damage Event

. Titl e 10 CFR Part 50, Appendix B, Criterion III , requires, in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program, was identified. Specifically, the design capability of the containment spray system requires that the system be full of water in order to achieve and maintain the design rate of flow. Contrary to the above, as of May 6, 2011, the licensee had not verified the adequacy of the design capability of the containment spray system to remain full of water through design review, calculation, or testing. Specifically, the licensee had not analyzed whether vortexing in the containment spray additive tank would affect system flow. The analysis demonstrated that no air should be entrained as a result of vortexing. The licensee documented this issue in Condition Report 38715. Because this finding was of very low safety significance and has been entered into the corrective action program, it is being treated as a noncited violation consistent with Section 2.3.2 of the NRC Enforcement Policy: NCV 05000482/2011003

-1 0 , "Failure to Analyze for Vortexing in Containment Spray Additive Tank."

a. Inspection Scope

The inspectors assessed the activities and actions taken by the licensee to assess its readiness to respond to an event similar to the Fukushima Daiichi nuclear plant fuel damage event. This included (1)an assessment of the licensee's capability to mitigate conditions that may result from beyond design basis events, with a particular emphasis on strategies related to the spent fuel pool, as required by NRC Security Order Section B.5.b issued February 25, 2002, as committed to in severe accident management guidelines, and as required by 10 CFR 50.54(hh); (2)an assessment of the licensee's capability to mitigate station blackout conditions, as required by 10 CFR 50.63 and station design bases; (3)an assessment of the licensee's capability to mitigate internal and external flooding events, as required by station design bases; and

(4) an assessment of the thoroughness of the walkdowns and inspections of important equipment needed to mitigate fire and flood events, which were performed by the licensee to identify any potential loss of function of this equipment during seismic events possible for the site.

b. Findings and Observations

NRC Inspection Report 05000482/2011008 (ML11133A354) documented detailed results of this inspection activity. Following issuance of the report, the inspectors conducted additional follow-up on the following seven selected issues.

1. Extensive damage mitigation guideline procedures specify that if the control room staff and field operators are compromised, then the shift security commander becomes the incident coordinator until an operator can be found.

The inspectors identified that shift security commanders are not trained on reactor technology and mitigating systems, therefore it is not reasonable to assume they would have a sufficient knowledge base or decision making ability to direct technical response to an extensive damage situation.

The licensee entered the issue into their corrective action program and is in the process of conducting additional procedural and technical training for security commanders.

The inspectors reviewed licensee extensive damage mitigation guidelines in greater detail and compared them to the requirements of 10 CFR 50.54(hh)(2) as well as to the expectations outlined in the NRC Staff Guidance for Use in Achieving Satisfactory Compliance with February 25, 2002, Order Section B.5.b dated February 25, 2005 , and determined that Wolf Creek

's procedures meet agency expectations in that they direct security commanders to seek out persons with the best technical expertise available. This issue of concern is closed with no finding.

2. The licensee identified that extensive damage mitigation guidelines procedures to refill the refueling water storage tank are not viable because the connection point is not readily accessible.

The licensee entered this issue into their corrective action program and is evaluating potential design changes to resolve this concern.

The inspectors reviewed the applicable extensive damage mitigation attachments which direct refilling of the refueling water storage tank in greater detail and compared them to the requirements of 10 CFR 50.54(hh)(2) as well as to the expectations outlined in the NRC Staff Guidance for Use in Achieving Satisfactory Compliance with February 25, 2002, Order Section B.5.b , dated February 25, 2005 , and determined that these procedures do not meet regulatory requirements for compliance with Order 10 CFR 50.54(hh)(2)because the expectation element could not be effectively implemented using existing or readily available resources and because personnel safety concerns associated with the expectation element had not been addressed.

There is no guidance as to how the connection is to be accessed, nor is the required equipment needed access and work safely at heights pre

-staged in advance. This issue of concern is documented as a licensee identified violation in Section 4O A7.1 of this report.

3. The licensee identified that extensive damage mitigation guideline procedures require additional precautionary guidance to prevent excessive reactor coolant system depressurization which could compromise natural circulation core cooling. The licensee entered this issue into their corrective action program and is evaluating procedural enhancements to remedy this concern.

The inspectors reviewed the applicable licensee extensive damage mitigation attachments which direct actions which can cool and depressurize the reactor coolant system and determined that this issue of concern was an enhancement only and not a violation of regulatory requirements. Since operators reviewing these procedures identified the same concerns and because the licensee has entered this issue in their corrective action program this issue of concern is closed with no finding.

4. The licensee identified that alternate power sources specified by extensive damage mitigation guidelines procedures are not properly staged in advance.

Additional technical guidance on the configuration and use of these sources needs to be added to the extensive damage mitigation guidelines procedures. The licensee entered this issue into their corrective action program and is evaluating alternative equipment and procedural enhancements to resolve this concern. The inspectors reviewed the applicable licensee extensive damage mitigation attachments which direct the use of alternate dc sources in greater detail and compared them to the requirements of 10 CFR 50.54(hh)(2) as well as to the expectations outlined in the NRC Staff Guidance for Use in Achieving Satisfactory Compliance with February 25, 2002, Order Section B.5.b dated February 25, 2005 , and determined that these procedures do not meet regulatory requirements for compliance with 10 CFR 50.54(hh)(2)because the expectation element could not be effectively implemented using existing or readily available resources. Specifically

, the components are not pre

-staged in advance. This issue of concern is documented as a licensee identified violation in Section 4OA7.1 of this report.

5. During walkdowns with the inspector, nuclear station operators failed to promptly locate certain station blackout emergency operating procedure components in the plant. The inspectors determined that this was due to inadequate training, lack of specific procedural guidance, and over

-reliance on a computer database of equipment locations.

The computer database would be unavailable during an actual station blackout.

The licensee agreed with this characterization and entered this issue into their corrective action program

. 6. The inspectors determined that this issue of concern was a performance deficiency and a violation of Technical Specificatio n 5.4.1.a which requires, in part, that specified station procedures be established, implemented, and maintained. The inspectors determined that all of the components operators failed to identify for local actions were backed up by components which would fail safe in a loss of ac power event and therefore did not have the potential to further complicate that event.

The inspectors evaluated the issue using Inspection Manual Chapter 0612 , Appendix B , "Issue Screening

," and determined the failure to comply with Technical Specificati on 5.4.1.a constituted a violation of minor significance that is not subject to enforcement action in accordance with the NRC's Enforcement Policy.

The inspectors also found that Wolf Creek had completed appropriate corrective actions in this area. This issue of concern is closed as a n NRC identified minor violation.

7. The licensee identified that some fire protection equipment is not stored in seismic or tornado qualified locations.

The licensee identified that the water supply pumps and piping used for fire protection and extensive damage mitigation guideline actions is not seismic or tornado qualified.

The licensee also identified that equipment used to access underground diesel storage tanks is not seismic or tornado qualified; also the tanker truck used to refill the diesel

-driven fire pump and fire truck is not parked in a seismic or tornado qualified building.

The licensee entered these issues into their corrective action program.

8. The inspectors reviewed the requirements of 10 CFR 50.54(hh)(2) as well as to the expectations outlined in the NRC Staff Guidance for Use in Achieving Satisfactory Compliance with February 25, 2002, Order Section B.5.b , dated February 25, 2005, and determined that those regulatory requirements apply only to fire and explosion events, not to earthquakes and tornadoes. Because Wolf Creek identified this issue and entered into their corrective action program and because this issue of concern has no associated violation of regulatory requirements, it does not meet the criteria of a finding under the Inspection Manual Chapter 0612. This issue of concern is closed with no finding

. 9. The condensate storage tank used in station blackout response and extensive damage mitigation guideline procedures is not seismic or tornado qualified. The licensee entered the issue into their corrective action program. The inspectors found that the safety

-related source, from the essential service water system, would not be impacted.

The inspectors reviewed applicable sections of Wolf Creek's U SAR and determined that this issue is within the boundaries of Wolf Creek's NRC

-approved design bases. This issue of concern is closed with no finding.

.3 (Closed) NRC T

I 2515/184, Availability and Readiness Inspection of Severe Accident Management Guidelines (SAMGs)

The inspectors reviewed the licensee's severe accident management guidelines (SAMGs), implemented as a voluntary industry initiative in the 1990's, to determine

(1) whether the SAMGs were available and updated, (2)whether the licensee had procedures and processes in place to control and update its SAMGs, (3)the nature and extent of the licensee's training of personnel on the use of SAMGs, and (4)licensee personnel's familiarity with SAMG implementation.

The results of this review were provided to the NRC task force chartered by the Executive Director for Operations to conduct a near

-term evaluation of the need for agency actions following the Fukushima Daiichi fuel damage event in Japan. Plant

-specific results for Wolf Creek were provided as Enclosure 14 to a memorandum to the Chief, Reactor Inspection Branch, Division of Inspection and Regional Support, dated May 27, 2011 (ML111470264).

.4 (Closed) TI 2515/172, Reactor Coolant System Dissimilar Metal Butt Welds

a. Inspection Scope

Portions of T I 2515/172 were previously performed at Wolf Creek Nuclear Generating Station, during Refueling Outage s 15, 16, and 17. The results of those inspections are documented in NRC Inspection Reports 05000482/200 6 005, 05000482/2008003, 05000482/20090 05 and 05000482/2011003

, respectively.

Specific documents reviewed during this inspection are listed in the attachment. This unit has the following dissimilar metal butt welds:

COMPONENT ID DESCRIPTION MRP-139 CATEGORY BASELINE EXAM RV-301-121-A Loop 1 Outlet Nozzle to Safe

-End Weld D April 2005 RF14 RV-301-121-B Loop 2 Outlet Nozzle to Safe

-End Weld D April 2005 RF14 RV-301-121-C Loop 3 Outlet Nozzle to Safe

-End Weld D April 2005 RF14 RV-301-121-D Loop 4 Outlet Nozzle to Safe

-End Weld D April 2005 RF14 COMPONENT ID DESCRIPTION MRP-139 CATEGORY BASELINE EXAM RV-302-121-A Loop 1 Inlet Nozzle to Safe

-End Weld E April 2005 RF14 RV-302-121-B Loop 2 Inlet Nozzle to Safe

-End Weld E April 2005 RF14 RV-302-121-C Loop 3 Inlet Nozzle to Safe-End Weld E April 2005 RF14 RV-302-121-D Loop 4 Inlet Nozzle to Safe

-End Weld E April 2005 RF14 TBB03-1-W /MW7090-WOL-DM Pressurizer Surge Nozzle to Safe-End Weld F October 2006 RF15 TBB03-2-W /MW7089-WOL-DM Pressurizer Spray Nozzle to Safe-End Weld B October 2006 RF15 TBB03-3-A-W /MW7086-WOL-DM Pressurizer Safety Nozzle A to Safe-End Weld B October 2006 RF15 TBB03-3-B-W /MW7087-WOL-DM Pressurizer Safety Nozzle B to Safe-End Weld B October 2006 RF15 TBB03-3-C-W /MW7088-WOL-DM Pressurizer Safety Nozzle C to Safe-End Weld F October 2006 RF15 TBB03-4-W /MW7085-WOL-DM Pressurizer Relief Nozzle to F October 2006 RF15 COMPONENT ID DESCRIPTION MRP-139 CATEGORY BASELINE EXAM Safe-End Weld 1. Licensee's Implementation of the Materials Reliability Program (MRP

-139) Baseline Inspections (03.01)

The inspectors reviewed records of structural weld overlays and nondestructive examination activities associated with the licensee's pressurizer and hot leg structural weld overlay mitigation effort. The baseline inspections of the pressurizer dissimilar metal butt welds were completed as noted in the table above.

The pressurizer dissimilar metal butt welds had full structural weld overlay applied in Refueling Outage 15. The first Component ID in the preceding table was the designation prior to the overlay; the latter Component ID is the current weld designation (after overlay).

The licensee requested the deviations from the MRP

-139 baseline inspection requirements. These locations are now examined in accordance with the approved alternative of relief request I3R

-05. The licensee did not take any other deviations from the baseline inspection requirements of MRP

-139, and all other applicable dissimilar metal butt welds were scheduled in accordance with MRP

-139 guidelines.

2. Volumetric Examinations (03.02)

The results of these inspections are documented in NRC Inspection Reports 05000482/2006005, 05000482/2008003, and 05000482/2009005.

3. Weld Overlays (03.03)

Only the pressurizer nozzles have been mitigated. The mitigation type was full structural weld overlay applied in Refueling Outage 15. A pre-service exam in accordance with relief request I3R-05 was performed. An inservice exam on the MRP-139 , category F welds was performed in Refueling Outage 16 in accordance with I3R-05. This examination also falls within the guidelines of MRP

-139 for category F welds. 4. Mechanical Stress Improvement (03.04)

The licensee did not employ a mechanical stress improvement process.

5. Inservice Inspection Program (03.05)

The licensee has prepared an MRP-139 inservice inspection program.

All the welds in the MRP

-139 inservice inspection program are appropriately categorized in accordance with MRP

-139. The inservice inspection frequencies are consistent with the inservice inspection frequencies called for by MRP

-139.

b. No findings were identified.

Findings

4OA6 Meetings Exit Meeting Summary

On April 1, 2011, the inspector s presented the inservice inspection results to Mr. S. Hedges, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors telephonically re

-exited with Mr. Hedges, Site Vice President, and other members of the licensee's staff on June 16, 2011

. The inspectors acknowledged review of proprietary material during the inspection which was returned to the licensee

. On April 5, 2011, the Deputy Director of the Division of Reactor Projects conducted a regulatory performance meeting in conjunction with t he public annual assessment meeting with Mr. M. Sunseri, President and Chief Executive Officer, and other members of the licensee staff to review the corrective actions related to the previously White performance indicators for unplanned scrams per 7000 critical hours, unplanned scrams with complications, and safety system functional failures.

On May 6, 2011, the inspectors presented the inspection results to Mr. M. Sunseri, President and Chief Executive Officer, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary

. On June 8, 2011, the inspector communicated the results of the in

-office inspection of changes to the licensee's emergency plan to Mr. T. East, Superintendent, Emergency Planning, and to Mr. W. Muilenburg, Licensing Engineer, of the licensee's staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

On July 13, 2011, the inspectors presented the inspection results to Mr. S. Hedges, Site Vice President, and other members of the licensee staff. The licensee acknowledged the issues presented. The inspector asked the licensee whether any materials examined during the inspection should be considered proprietary. Although proprietary information was used during the inspection, it was returned to the licensee or destroyed

.

4OA7 Licensee-Identified Violations

The following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of Section 2.3.2 of the NRC Enforcement Policy for being dispositioned as noncited violation s.

.1 Title 10 CFR 50.54(hh)(2)(ii) states: "Each licensee shall develop and implement guidance and strategies intended to maintain or restore core cooling,

containment, and spent fuel pool cooling capabilities under the circumstances associated with loss of large areas of the plant due to explosions or fire, to include strategies in the following area of operations to mitigate fuel damage." On April 13, 2011, while performing procedure reviews as part of industry

-wide self-assessments in response to the core damage events at Fukushima Daiichi, Wolf Creek engineers identified two instances of mitigating strategy procedures which did not contain sufficient information to accomplish those strategies successfully. The first example was the ability to refill the refueling water storage tank, and the second example involved flashing the diesel generator field using alternate dc sources. The s e issues were documented in the licensee's corrective action program as Condition Report 37374. The inspectors evaluated these findings under Inspection Manual Chapter 0609 , Appendix L, and determined these findings to be of very low safety significance because the findings did not involve unrecoverable unavailability of multiple mitigating strategies such that spent fuel pool cooling, injection to the reactor vessel, or injection to steam generators cannot occur, or unrecoverable unavailability of on

-site, self

-powered, portable pumping capability, or substantial inability to perform command and control enhancements.

.2 Title 1 0 CFR Part 50, Appendix

B, Criterion XI, "Test Control," requires, in part, that a test program be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in the applicable design documents.

On May 13, 2011, Wolf Creek identified a noncited violation of 10 CFR Part 50 , Appendix B, Criterion XI, test control for stroking residual heat removal containment sump valve 8811B prior to its as-found diagnostic test. Wolf Creek stroked the valve for a clearance order and as such, preconditioned the valve prior to its test. Plant computer data from this stroke, data from the diagnostic stroke, and valve disassembly showed no deficiencies.

Using Inspection Manual Chapter 0609.04, the inspectors determined the finding to be of very low safety significance because it was confirmed not to result in the loss of operability or functionality.

This issue is captured in Condition Report 37244.

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

G. Beckett, Superintendent, Support Engineering

P. Bedgood, Manager, Radiation Protection

R. Evenson, Requalification Program Supervisor

J. Harris, System Engineer

S. Hedges, Site Vice

President

S. Henry, Operations Manager

R. Hobby, Licensing Engineer

D. Hooper, Supervisor, Regulatory Affairs

T. Just, Senior Technician, Chemistry

J. Keim, Support Engineering Supervisor

S. Koenig, Manager, Corrective Actions

M. McMullen, Technician, Engineering

C. Medency, Supervisor, Radiation Protection

W. Muilenburg, Licensing Engineer

R. Murray, Simulator Supervisor

B. Norton, Manage

r, Integrated Plant Scheduling

J. Pankaskie, Engineering Supervisor

G. Pendergrass, Director of Engineering

L. Rockers, Licensing Engineer

G. Sen, Regulatory Affairs Manager

R. Smith, Plant Manager

M. Sunseri, President and Chief Executive Officer

J. Truelove, Supervisor, Chemistry

J. Weeks, System Engineer

M. Westman, Training Manager

NRC Personnel

D. Loveless, Senior Reactor Analyst

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000482/2011003

-0 7 VIO Failure to Correct Procedure for Opening Main Steam Isolation Valves (EA-11-149) (Section 4OA3.1)

Attachment

Opened and Closed

05000485/2011003

-0 1 NCV No Procedure for Debris in Transformed and Tank Yards Prior to Severe Weather

(Section 1R01)

05000482/2011003

-02 NCV Failure to

Properly Establish Clearance Order Boundary Isolation Resulting in

Loss of Component Cooling Water Inventory (Section 1R04)

05000482/2011003

-03 NCV Failure to Assure Fillet Weld Met Size Requirements on Train B Charging Header Vent Line

(Section 1R08

.1)

05000482/2011003

-04 NCV Failure to Assure Separation of Stainless Steel and Carbon Steel Grinding and Cutting Equipment

(Section 1R08

.1)

05000482/2011003

-05 NCV Failure to Assure Configuration Control of Safety

-Related Systems (Section 1R08

.3)

05000482/201100

3-06 FIN Inadequate Acceptance Criteria for Postmaintenance Testing of the Startup Feedwater

Pump (Section

1R19)

05000482/201100

3-08 NCV Failure to Maintain Reactor Coolant System Pressure Below Relief Valve Setpoint (Section

4OA3.2)

05000482/2011003

-09 NCV Inadequate

Fire Watch Defeats Halon

Fire Suppression in

Vital Switchgear

Rooms During Fire (Section

4OA3.3)

05000482/2011003

-10 NCV Failure to

Analyze for

Vortexing in

Containment

Spray Additive Tank (Section

4OA5.1)

05000482-2515/177 TI Managing Gas Accumulation in Emergency Core Cooling, Decay

Heat Removal, and Containment Spray Systems (NRC Generic Letter 2008

-01) (Section

4OA5.1)

Closed

05000482/2006

-003-00 LER Indications Discovered on Pressurizer during Preplanned In

-service Inspections

(Section 4OA3.4)

05000482/20 10 00 6-0 5 VIO Notice Of Violation EA

-10-160, Failure to correct NRC identified NCV. Apparent Cause Evaluation vice Root Cause Evaluation fo

r Essential Service Water

(Section 4OA3.5)

Attachment

05000482-2515/183 TI Followup to the Fukushima Daiichi Nuclear Station Fuel Damage Event (Section 4OA5.2)

05000482-2515/1 84 TI NRC Temporary Instruction 2515/184, Availability and Readiness Inspection of Severe Accident Management Guidelines (SAMGs)

(Section 4OA5.3)

05000482/2515/172

TI Temporary Instruction 2515/172, Reactor Coolant System

Dissimilar Metal Butt Welds (Section 4OA5.4)