ML092240087

IR 05000482-09-006 on 02/23/09 - 07/01/09; Wolf Creek Generating Station, Focused Baseline Inspection in Response to the Identification of the Potential to Void the Suction Headers of Both Trains of the Residual Heat Removal System on 08/01

ML092240087
Person / Time
Site:	Wolf Creek
Issue date:	08/12/2009
From:	Vincent Gaddy NRC/RGN-IV/DRP/RPB-B
To:	Muench R A Wolf Creek
References
IR-09-006
Download: ML092240087 (34)

See also: IR 05000482/2009006

Text

August 12, 2009

Rick A. Muench, President and Chief Executive Officer

Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839

Subject: WOLF CREEK GENERATING STATION - NRC FOCUSED BASELINE INSPECTION REPORT 05000482/2009006 Dear Mr. Muench: On July 1, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed a focused baseline inspection at your Wolf Creek Generating Station. This inspection examined activities

associated with the station's identification of a potential issue involving the likelihood of steam voiding the suction headers of both trains of the residual heat removal system if system actuation were required for injection or recirculation during Mode 3 operations. The genesis of this issue involved the station's practice of using both trains of the residual heat removal system for shutdown cooling while in Mode 4, with reactor coolant system temperature greater than

240°F, without providing adequate cooling of the suction headers to ensure that steam voiding would not occur if the residual heat removal system was needed for emergency core cooling system injection or recirculation. The NRC's initial evaluation of this issue using the criteria in NRC Management Directive 8.3, "NRC Incident Investigation Program," determined that the estimated Incremental Conditional Core Damage Probability was in the overlap region between a special inspection and an augmented inspection. However, it was determined that the model utilized likely over estimated the risk since this model was based on full power operations. Therefore, based on

management discretion, a decision was made that, although the risk for this event was in the overlap region, a focused baseline inspection would be performed since the risk for this issue was likely overestimated. The enclosed report documents the inspection results, which were discussed at the exit meeting on July 9, 2009, with Mr. Sunseri, Vice President Operations and Plant Manager, and other members of your staff. The inspection 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 inspection team reviewed selected procedures and records, observed activities, and interviewed personnel. This report documents two NRC identified findings of very low safety significance (Green). Both these findings were determined to involve violations of NRC requirements. However, because of their 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 I.A.1 of the NRC Enforcement Policy. If you contest the noncited violations in this report, you should UNITED STATESNUCLEAR REGULATORY COMMISSIONREGION IV612 EAST LAMAR BLVD, SUITE 400ARLINGTON, TEXAS 76011-4125

Wolf Creek Nuclear Operating Corp. - 2 - 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, 76011-4125; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Wolf Creek Generating

Station. In addition, if you disagree with the characterization of 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 Wolf Creek Generating Station. The information you provide will be considered in accordance with Inspection Manual Chapter 0305. In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, and its enclosure, will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room). Sincerely, /RA/

Vincent G. Gaddy, Chief, Project Branch B Division of Reactor Projects

Docket: 50-482 Licenses: NPF-42 Enclosure:

Enclosure: NRC Inspection Report 05000482/2009006 w/Attachment: Supplemental Information

cc w/Enclosure:

Vice President Operations/Plant Manager Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839

Jay Silberg, Esq. Pillsbury Winthrop Shaw Pittman LLP 2300 N Street, NW Washington, DC 20037

Supervisor Licensing Wolf Creek Nuclear Operating Corporation P.O. Box 411 Burlington, KS 66839

Chief Engineer Utilities Division Kansas Corporation Commission 1500 SW Arrowhead Road Topeka, KS 66604-4027

Office of the Governor State of Kansas Topeka, KS 66612-1590

Attorney General 120 S.W. 10th Avenue, 2nd Floor Topeka, KS 66612-1597

Wolf Creek Nuclear Operating Corp. - 3 - County Clerk Coffey County Courthouse 110 South 6th Street Burlington, KS 66839

Chief, Radiation and Asbestos

Control Section Bureau of Air and Radiation Kansas Department of Health and Environment 1000 SW Jackson, Suite 310

Topeka, KS 66612-1366 Chief, Technological Hazards Branch FEMA, Region VII 9221 Ward Parkway Suite 300 Kansas City, MO 64114-3372

Wolf Creek Nuclear Operating Corp. - 4 - Electronic distribution by RIV: Regional Administrator (Elmo.Collins@nrc.gov) Deputy Regional Administrator (Chuck.Casto@nrc.gov) DRP Director (Dwight.Chamberlain@nrc.gov) DRP Deputy Director (Anton.Vegel@nrc.gov) DRS Director (Roy.Caniano@nrc.gov)

DRS Deputy Director (Troy.Pruett@nrc.gov) Senior Resident Inspector (Chris.Long@nrc.gov) Resident Inspector (Charles.Peabody@nrc.gov) Site Secretary (Shirley.Allen@nrc.gov) Branch Chief, DRP/B (Vincent.Gaddy@nrc.gov)

Senior Project Engineer, DRP/B (Rick.Deese@nrc.gov) Public Affairs Officer (Victor.Dricks@nrc.gov) Team Leader, DRP/TSS (Chuck.Paulk@nrc.gov) RITS Coordinator (Marisa.Herrera@nrc.gov) Regional Counsel (Karla.Fuller@nrc.gov) Congressional Affairs Officer (Jenny.Weil@nrc.gov) OEMail Resource

Only inspection reports to the following: DRS STA (Dale.Powers@nrc.gov) OEDO RIV Coordinator (Leigh.Trocine@nrc.gov) ROPreports

File located: R:\_REACTORS\_WC\2009\WC 2008-06 RP-JEJ Adams.doc ML 092240087 SUNSI Rev Compl. Yes No ADAMS Yes No Reviewer Initials Publicly Avail Yes No Sensitive Yes No Sens. Type Initials RI:DRP/E RI:DRS/EB1 SPE:DRP/B SRA:DRS/E JEJosey MRYoung RWDeese MRunyan VGG for /RA/ /RA/ /RA/ 08/12/09 07/20/09 07/24/09 07/20/09 C:DRP/B VGGaddy /RA/ 08/12/09 OFFICIAL RECORD COPY T= Telephone E= E-mail F = Fax

- 1 - Enclosure U.S. NUCLEAR REGULATORY COMMISSION REGION IV

Docket: 50-482 License: NPF-42 Report: 05000482/2009006 Licensee: Wolf Creek Operating Corporation Facility: Wolf Creek Generating Station Location: 1550 Oxen Lane SE Burlington, Kansas Dates: February 23 through July 1, 2009

Inspectors: J. Josey, Resident Inspector, Arkansas Nuclear One, Projects Branch E M. Runyan, Senior Reactor Analyst

M. Young, Reactor Inspector A. Zoulis, Reliability and Risk Analyst, NRR/DRA/APOB

Approved By: V. G. Gaddy, Chief, Project Branch B, Division of Reactor Projects

- 2 - Enclosure SUMMARY OF FINDINGS IR 05000482/2009006; 02/23/09 - 07/01/09; Wolf Creek Generating Station, Focused Baseline Inspection in response to the identification of the potential to void the suction headers of both trains of the residual heat removal system on August 1, 2008. This report covered a 5-day period (February 23-27, 2009) of onsite inspection, with in office review through July 1, 2009. The focused baseline inspection team consisted of one resident inspector, one reactor inspector, and one senior reactor analyst. One Green noncited violation of significance was identified as well as one Green noncited Severity Level IV violation. The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using NRC Inspection Manual Chapter 0609, "Significance Determination Process." Findings for

which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The 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. A. NRC-Identified Findings and Self-Revealing Findings

Cornerstone: Mitigating Systems

• Green. The inspectors identified a noncited violation of Technical Specification 5.4.1, "Procedures," associated with the licensee's failure to ensure that adequate procedures were available for changing modes of operation of the residual heat removal system from shutdown cooling to emergency core cooling system operation. Specifically, station procedures allowed the residual heat removal system to be realigned to the emergency core cooling system mode of

operation following operation in the shutdown cooling mode with suction temperatures as high as 350°F without properly cooling the entire suction header. This resulted in both trains of the residual heat removal system being inoperable during periods of operation in Modes 3 and 4. This issue was entered into the licensee's corrective action program as Condition Reports 2008-3810 and 2008-4997. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and

it directly affected the 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, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because this finding represented a loss of

safety function of the residual heat removal system.

- 3 - Enclosure The inspectors performed a Phase 2 analysis using Appendix A, "Determining the Safety Significance of Reactor Inspection Findings for At-Power Situations," of Inspection Manual Chapter 0609, "Significance Determination Process," and the plant specific Phase 2 presolved tables and worksheets for Wolf Creek. The inspectors determined that the Phase 2 presolved tables and worksheets did not contain appropriate target sets to accurately estimate the risk input of the finding.

Therefore, it was determined that a Phase 3 analysis was required. Senior risk analysts performed a Phase 3 analysis of this issue. The estimated Conditional Core Damage Probability was determined to be 2.84E-7, and the

estimated Conditional Large Early Release Probability was determined to be 2.72E-9. Based on these results, the finding was determined to be of very low safety significance. This finding was determined to have a crosscutting aspect in the area of Problem Identification and Resolution associated with the corrective action program P.1(c), in that the licensee failed to appropriately and thoroughly evaluate

problems such that the resolutions address the causes (Section 2.2). Cornerstone: Miscellaneous

• Severity Level IV. The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73, "Licensee Event Report System," associated with the licensee's failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria as specified. Specifically,

on December 8, 2008, the licensee completed analysis of an issue associated with the residual heat removal system which determined that both trains of the system were inoperable when suction side temperature exceeded 249°F. Based on the results of this analysis as well as plant operating history, it was determined that the licensee failed to report instances where the system was

operated in a condition prohibited by technical specifications, and a loss of safety function of the system existed between March 20, 2008, and December 8, 2008. The licensee entered this issue into their corrective action program as Condition Reports 2009-1261 and 2009-1326 and Action Requests 15244, 17776, and 15306. The inspectors reviewed this issue in accordance with Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue

because the NRC's regulatory ability was affected. Specifically, the NRC relies on licensee to identify and report conditions or events meeting the criteria specified in regulations in order to perform its regulatory function, and when this is not done, the regulatory function is impacted. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated in accordance with the NRC Enforcement Policy. The finding was reviewed by NRC management and, because the violation was determined to be of very low safety significance, was not repetitive or willful, and was entered into the corrective action program, this violation is being treated as a Severity Level IV noncited violation consistent with the NRC

Enforcement Policy. This finding was determined to have a crosscutting aspect in the area of Problem Identification and Resolution associated with the

- 4 - Enclosure corrective action program in that the licensee failed to appropriately and thoroughly evaluate for reportability aspects all factors and time frames associated with the inoperability of residual heat removal system when suction temperatures were above 249°F P.1(c)(Section 2.1). B. Licensee-Identified Violations

None.

- 5 - Enclosure Report Details

1.0 Focused Baseline Inspection Scope

The NRC conducted a focused baseline inspection at the Wolf Creek Generating Station

to better understand the identification of a potential issue involving the likelihood of steam voiding the suction headers of both trains of the residual heat removal system if system actuation were required for injection or recirculation during Mode 3 operations. The genesis of this issue involved the stations practice of using both trains of the residual heat removal system for shutdown cooling while in Mode 4, with reactor coolant

system temperature greater than 240°F, without providing adequate cooling of the suction headers to ensure that steam voiding would not occur if the residual heat removal system was needed for emergency core cooling system injection or recirculation. On August 1, 2008, station personnel generated Condition Report 2008-810 to identify this issue and evaluate potential system impacts due to the potential inoperability of the residual heat removal system. Subsequently, the stations evaluation determined that the historical operating practices for the residual heat removal system had resulted in

past inoperability of both trains of the system. This resulted in the licensee issuing Licensee Event Report 5000482/2008008-00 in October 2008. NRC management's initial evaluation of this issue, using the criteria in NRC Management Directive 8.3, "NRC Incident Investigation Program," determined that the estimated incremental conditional core damage probability was in the overlap region between a special inspection and an augmented inspection. However, it was determined that the model utilized to analyze this issue likely overestimated the risk since this model was based on station full power operations. Therefore, based on

management discretion, a decision was made that, although the risk for this event was in the overlap region due to the likelihood of overestimation, a focused baseline inspection would be performed to determine the full extent of this issue including determination of risk. For this inspection, the Focused Baseline Inspection team used NRC Inspection Procedure 7111115, "Operability Evaluations"; Procedure 71152, "Identification and

Resolution of Problems"; and Procedure 71153, "Followup of Events and Notices of Enforcement Discretion." The team reviewed station procedures, corrective action documents, engineering evaluations and design documentation for the residual heat removal system as well as interviewing various station personnel regarding this issue. The team also reviewed the licensee's root cause analysis, extent of condition

evaluation, immediate and long term corrective actions, and industry operating experience. A list of the specific documents reviewed is provided in Attachment 1.

Background Gas accumulation or voiding of safety-related fluid systems can cause air binding in pumps or water hammer events in piping systems. Instances of gas accumulation or voiding in safety-related fluid systems have occurred on several instances in the nuclear industry, and as a result, the NRC has published 20 information notices, 2 generic letters, and a NUREG related to this issue, as well as interacting with the nuclear

- 6 - Enclosure industry in relation to these publications and in response to gas accumulation/voiding events. It is important that systems relied upon to mitigate accidents and events are able to perform their designed safety function. Specifically, a fluid system whose successful operation is dependant upon the proper operation of a pump to be able to inject water should be sufficiently filled to ensure that it can reliably perform its intended function under all accident and nonaccident conditions as required. Inadequate control of gas introduction or void formation in a fluid system can have the following safety implications:

• The introduction of gases into a pump can cause the pump to become air bound which results in little to no flow being generated by the pump, rendering the pump inoperable. An air bound pump can become damaged quickly, thereby eliminating the possibility of recovering the pump during an event by venting the

pump casing and suction piping.

• Gas introduction into a pump can render a pump inoperable, even if the gas does not air bind the pump. This occurs when there is gas accumulation in the pump casing which reduces the pump's discharge pressure and flow capacity to the point that the pump can no longer perform its design safety function.

• Void formation and gas accumulation can also result in a system pressure transient event known as water hammer. This is most commonly seen in the discharge piping, but can also occur in the suction piping. This phenomenon

occurs when a pressure surge or wave is generated when a fluid in motion is forced to suddenly stop or change direction. Specifically, when there is a rapid venting or void collapse in a system, followed by a rapid refill of the piping with water, there is the potential to have water hammer due to the system configuration.

• Time needed to vent and refill voided discharge piping could delay delivery of water from the system beyond the timeframe assumed in the facilities safety analysis. 1.1 Event Summary

On January 18, 2008, the licensee initiated Condition Report 2008-0164 to address NRC Generic Letter 2008-001, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems." On March 21, 2008, station personnel generated Condition Report 2008-0989 based on questions raised by an individual from the Callaway Plant, the sister unit to Wolf Creek who was performing a benchmarking trip at the Wolf Creek Station. After observing the

reactor coolant system cooldown in preparation for a refueling outage, the individual observed that station procedures allowed for both residual heat removal trains to be aligned in shutdown cooling mode with reactor coolant system temperature above 260°F, which was different than what was allowed by Callaway station procedures. Specifically, the individual noted that in Callaway's last operating cycle station procedures were changed so that only one residual heat removal train could be aligned for shutdown cooling with reactor coolant system temperature above 260°F, and the other train must

- 7 - Enclosure remain aligned to the refueling water storage tank, which was the emergency core cooling system injection lineup. The basis for this change was due to a concern of potential flashing in the residual heat removal systems suction piping if the pressure of the system was reduced following realignment to the refueling water storage tank. The licensee's evaluation of this issue concluded that the current practices associated with the residual heat removal system were acceptable and allowed by technical specifications. This was based on the licensee's review and interpretation of the facilities technical specifications requirements for residual heat removal system alignment, verification that station procedures required cooldown of the residual heat

removal suction prior to alignment to the refueling water storage tank, and information contained in Westinghouse Document WCAP-12476, "Evaluation of LOCA During Mode 3 and 4 Operation for Westinghouse NSSS." On May 10, 2008, following maintenance to correct flange leaks on the residual heat removal Pump B discharge flange and refueling water storage tank check valve, the system was aligned to the reactor coolant system, as part of the retest, to place reactor

coolant system pressure on the affected joints. Subsequently, the pump was secured and the train was realigned to take suction from the refueling water storage tank. At this point the licensee attempted to perform ultrasonic testing of the residual heat removal piping to check for voids, but found that the piping was too hot to attach the required instrumentation. The licensee decided to vent the piping in an effort to reduce

temperature and vented a mixture of steam and water for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> before the suction piping became water solid. The licensee initiated Station Work Order 08-306203-000 to perform troubleshooting to determine if the suction piping temperature was below saturation temperature where the recirculation line taps into the system. (This issue was not entered into the licensee's corrective action program.) During the station's evaluation of Condition Report 2008-0164, an operations representative identified a concern with the potential for steam binding. Specifically, steam voiding concerns that had been identified during restoration of the residual heat

removal system at the end of Refueling Outage 16, on May 10, 2008, which could happen any time the station enters Mode 3, combined with the findings from the generic letter review prompted the initiation of another condition report to review these concerns. On August 1, 2008, station personnel initiated Condition Report 2008-3810 to address concerns that had been identified during the review for Generic Letter 2008-01 regarding potential void formation. This condition report questioned the past operability of the residual heat removal system when aligned in the injection mode with suction piping temperature as high as 350°F, as well as the current design adequacy to ensure cooling

of suction piping using the mini-flow recirculation line. An evaluation was requested to determine the effects of potential steam voiding in the residual heat removal suction piping when realigning the system from reactor coolant system cooling to emergency core cooling system injection while transitioning from Mode 4 to Mode 3. On September 23, 2008, Wolf Creek completed their evaluation of the potential voiding issue identified in Condition Report 2008-3810. The conclusions that were reached were recirculation cannot be relied upon to cool the water in the isolated suction line, the residual heat removal system would not have functioned if a loss of coolant accident had occurred in Mode 3 with elevated suction piping fluid temperature, and the residual heat removal train used for shutdown cooling should be secured, or put in service, only at a

- 8 - Enclosure temperature of 240°F to ensure operability. Based on these conclusions, on October 3, 2008, Wolf Creek submitted Licensee Event Report 05000482/2008008-00 in accordance with 10 CFR 50.73. On October 10, 2008, the licensee initiated Condition Report 2008-4997, "Missed Opportunity to Resolve RHR Suction Piping Issue," for the purpose of determining why two separate conditions initiated for apparently the same issue came to different conclusions. This condition report was also used to perform a root cause analysis of the potential voiding issue associated with the residual heat removal system as well as performing a past operability review.

On December 5, 2008, Wolf Creek completed their evaluations as directed by Condition Report 2008-4997. The evaluations concluded:

• From a past operability perspective, the residual heat removal system must be considered inoperable any time the plant was in Mode 4 with the reactor coolant system suction isolation valves open and reactor coolant system temperature

was above 249.1°F.

• The residual heat removal system must be considered inoperable in Mode 3 during plant heatup for a period of 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> following isolation from the reactor coolant system. This is based on the amount of time it would take the suction piping and fluid to cool down to 225°F.

• From the perspective of past functionality, the residual heat removal system would not have been functional during a small break loss of coolant accident in Modes 3 or 4, nor a large break loss of coolant accident in Mode 3 for injection or recirculation.

Based on these conclusions, on January 30, 2009, Wolf Creek submitted an updated Licensee Event Report 05000482/2008008-01. The team constructed the following time line of events relative to the issue: Date Details

March 1990 Wolf Creek received Westinghouse Report WOG-90-048, "Residual heat removal System Operability During a Mode 4 LOCA." The purpose of this report was to detail the efforts taken

by Westinghouse to evaluate residual heat removal system operability and potential water hammer concerns following a loss of coolant accident during Mode 4 operations. The report stated that the concern of hot residual heat removal pump suction fluid being "trapped" by the refueling water storage tank to residual heat removal system check valve, and the condition of rapid depressurization of saturated water during a pump start producing

a large void fraction in the suction piping. This was an informational report that required no response and no action on behalf of the licensee.

- 9 - Enclosure June 1990 Wolf Creek received Letter OG-90-30, "Shutdown LOCA Concerns that Relate to the Interim Guidance." The purpose of this letter was to inform all Westinghouse Owner Group utilities of the new shutdown loss of coolant accident concerns identified since the interim guidance was issued in 1987. This letter formally identified the two new concerns previously identified in the

Westinghouse Report WOG-0-48, "Residual Heat Removal System Operability During a Mode Loss of Coolant Accident," report: * Operability concerns for the residual heat removal pump

• Residual heat removal pumps discharge cross-tie valve opened vs. closed issue

1991 Westinghouse Report WCAP-12476, "Evaluation of LOCA During Modes 3 and 4 Operation for Westinghouse NSSS," was submitted to the NRC for approval as a method to resolve the Modes 3 and 4 loss of coolant accident issue. The analysis

provided by this evaluation was instrumental in many actions taken by the licensee in regard to Modes 3 and 4 loss of coolant accident conditions.

January 1992 Station Procedure SYS EJ-120, "Placing RHR System in Safety Injection Standby Condition," was released. February 1993 Station Procedure OFN BB-031, "Shutdown LOCA," was released. This procedure contained a requirement to cooldown the residual heat removal system to less than 270°F prior to aligning it in the injection mode. This was accomplished by increasing component cooling water flow to the residual heat removal heat exchanger until residual heat removal temperature was less than 270°F. This procedure was developed from Abnormal Response Guideline 2,

which was distributed by the Westinghouse Owners Group. April 1993 Wolf Creek received Westinghouse's Nuclear Safety Advisory Letter NSAL-93-004, "RHR Operations as Part of the ECCS During Plant Startup," dated April 20, 1993, which reiterated the concern of flashing in the residual heat removal suction line, and provided an assessment of the safety significance.

Letter NSAL-93-004 also contained recommended actions to mitigate the condition:

• Assure that the residual heat removal system suction piping is sufficiently cooled before entering Mode 3

• Force cool the residual heat removal system piping

• Isolate the residual heat removal system from the reactor coolant system at a low enough temperature at which the

- 10 - Enclosure residual heat removal system pump suction pressure is above the saturation pressure corresponding to the fluid temperature

In response Wolf Creek generated Industry Technical Information Program 02324 to evaluate Letter NSAL-93-004. The purpose of

this evaluation was to ensure the information would receive the proper technical review and subsequent organizational actions if required. The stated recommendation of this Industry Technical Information Program review was to "Compare the Westinghouse recommendations to the operating procedures. Ensure that

adequate guidance was available to preclude the potential for forming steam voids in the residual heat removal system upon entry into Mode 3."

June 1993 Following verification that relevant station procedures contained provisions for forced cooling the residual heat removal suction piping, operations requested an evaluation of Industry Technical

Information Program 02324 from system engineering to identify any potential issues that needed to be addressed. (The team subsequently determined that system engineering had not evaluated Industry Technical Information Program 02324, safety analysis instead had performed the evaluation). August 25, 1993 Engineering's review and evaluation of Industry Technical Information Program 02324 was completed, and Industry Technical Information Program 02324 was closed. (Subsequently, the team determined that the review operations requested by engineering consisted of a review of the residual heat removal system piping and instrumentation drawing, a review of relevant

procedures and a teleconference with operations, the industry technical information program coordinator and engineering. During this process, operations was asked if the issue of the Nuclear Safety Advisory Letter had been adequately addressed in the procedures, and an affirmative reply was received. Based on this response from operations, the recommendation was that Industry Technical Information Program 02324 should be closed). 1995 The NRC placed review and approval of Westinghouse Report WCAP-12476, "Evaluation of LOCA During Mode 3 and 4 Operation for Westinghouse NSSS," on hold pending resolution of the shutdown risk review program. 1999 The Westinghouse Owners Group requested that Westinghouse Report WCAP-12476, "Evaluation of LOCA During Mode 3 and 4 Operation for Westinghouse NSSS," be withdrawn from NRC review, which was agreed to in 2000.

- 11 - Enclosure January 18, 2008 Wolf Creek initiated Condition Report 2008-0164, "NRC Generic Letter 2008-001," to address concerns identified in this generic letter. March 21, 2008 Wolf Creek initiated Condition Report 2008-000989, "Evaluate if Callaway limitation on RHR suction temperature applies to WCGS." This condition report was written to evaluate why Wolf Creek and Callaway treat the Mode 4 alignment of residual heat removal in shutdown cooling differently. May 8, 2008 Wolf Creek initiated Condition Report 2008-2187, "Draining of 'B' Residual heat removal Pump." This condition report was written to

document that, while draining the residual heat removal Pump B and associated suction piping to correct flange leaks on the residual heat removal Pump B discharge flange and refueling water storage tank check valve, steam was released into the room. Initially, the residual heat removal Pump B had been lined

up, and in service, in the shutdown cooling mode providing cooling to the reactor coolant system. Upon identification of the flange issue, the pump had been secured, the reactor coolant system suction isolation valves were shut, and the pump was run on mini-flow recirculation until pump discharge temperature was

140°F and then the pump was secured. Condition report initiator postulated: "One explanation to getting steam out of the drain line is that water captured in the line was still at 325°F and flashed to steam as draining occurred. When the residual heat removal pump is run in the recirculation mode, there is approximately 15 feet of suction piping that is being recirculated. Upstream of the recirculation line return is approximately 120 feet of piping that

would not see cooling effect of the recirculation flow. As this hot water was depressurized, it would turn to steam until the drain was uncovered and steam allowed to escape." The licensee did not investigate the reason for steam formation in the residual heat removal suction piping. May 10, 2008 Following maintenance to correct flange leaks on the residual heat removal Pump B discharge flange and refueling water storage tank check valve, the system was aligned to the reactor coolant system to retest the affected joints at reactor coolant system pressure. The pump was secured and the train was subsequently realigned to take suction from the refueling water storage tank.

The licensee attempted to perform ultrasonic testing of the residual heat removal piping to check for voids, but found that the piping was too hot to attach the required instrumentation. The licensee decided to vent the piping in an effort to reduce temperature. No condition report was written for this issue. May 23, 2008 Wolf Creek completed evaluation of Condition Report 2008-0989. The result of this evaluation stated that the current practice of using both residual heat removal trains for cooldown was

- 12 - Enclosure acceptable, as allowed by technical specifications and supported by historical operating experience and Westinghouse Report WCAP-12476. August 1, 2008 Wolf Creek initiated Condition Report 2008-3810, "Evaluate potential steam voiding in RHR suction while transitioning to Mode 3," to address concerns that had been identified during the review for Generic Letter 2008-01 regarding potential void formation. This condition report questioned the past operability of the residual heat removal system when aligned in the injection

mode with suction piping temperature above 260°F as well as the current design adequacy to ensure cooling of suction piping using recirculation flow. An evaluation was requested to determine the effects of potential steam voiding in the residual heat removal suction piping when realigning the system from reactor coolant

system cooling to emergency core cooling system injection while transitioning from Mode 4 to Mode 3. September 23, 2008 Wolf Creek completed their evaluation of Condition Report 2008-3810. The conclusions that were reached were: recirculation cannot be relied upon to cool the water in the isolated suction line, the residual heat removal system would not have

functioned if a loss of coolant accident had occurred in Mode 3, and the residual heat removal train used for shutdown cooling should be secured, or put in service, only at a temperature of 240°F to ensure operability. October 3, 2008 Wolf Creek submitted Licensee Event Report 5000482/2008008-00 in accordance with 10 CFR 50.73. October 10, 2008 Wolf Creek initiated Condition Report 2008-4997, "Missed opportunity to resolve RHR suction piping issue." This condition report was initiated to determine why there were different

responses for the same issue in Condition Reports 2008-0989 and 2008-3810. This condition report was also used to perform a root cause analysis of the potential voiding issue. December 5, 2008 Wolf Creek completed their evaluation of Condition Report 2008-004997. January 30, 2009 Wolf Creek submitted revised Licensee Event Report 5000482/2008008-01 because further evaluation provided additional detail to the safety significance and root cause of the issue. 1.2 Root Cause and Corrective Action Assessment

a. Root Cause Analysis

- 13 - Enclosure The inspectors reviewed and assessed the licensee's root cause analysis for technique, technical accuracy, thoroughness, and corrective actions proposed and taken. The inspectors reviewed the scope and process used by licensee personnel to identify the root cause of the potential to have void formation in the suction piping of the residual heat removal system when transitioning from Mode 4 to Mode 3 with fluid temperatures above 225°F. The inspectors compared information gained through inspection to the

event information and assumptions made in the root cause analysis. The inspectors interviewed licensee personnel, reviewed logs, and system design information. The inspectors also evaluated the licensee's extent of condition review. The licensee entered the potential voiding issue into the corrective action program as Condition Report 2008-3810, "Evaluate Potential Steam Voiding in RHR Suction While Transitioning to M-3," to address concerns identified during their review of Generic Letter 2008-01 regarding potential void formation in safety-related fluid systems. This condition report questioned the past operability of the residual heat removal system

when aligned in the injection mode with suction piping temperature above 260°F, as well as the current design adequacy to ensure cooling of suction piping using recirculation flow. The licensee classified this as a nonsignificant broke-fix condition report, and an evaluation was requested to determine the effects of potential steam voiding in the residual heat removal suction piping when realigning the system from reactor coolant

system cooling to emergency core cooling system injection while transitioning from Mode 4 to Mode 3. Through this evaluation the licensee determined that:

• If the residual heat removal system is aligned to the emergency core cooling system injection mode with suction fluid temperature near 350°F, the water in the suction piping will remain hot for a considerable amount of time. If, while in this condition, a loss of coolant accident was to occur and the safety injection system

initiated, the residual heat removal pumps would start which would cause pressure in the suction piping to decrease; and this correlates to a lowering of the saturation pressure for the corresponding suction piping fluid temperature. When the suction piping pressure is lowered below the saturation pressure for the corresponding temperature, this would cause the hot pressurized water to

flash to steam, and as long as the pressure in the suction piping is higher than the static head of the refueling water storage tank on the supply check valve, the check valve will not open and no injection flow will occur. This would result in the steam void continuing to expand and extending to the pump suction and steam binding the pump.

• Using the mini-flow recirculation line for cooling the suction piping of the residual heat removal train following alignment for emergency core cooling system injection can not be relied upon. Specifically, there is approximately 140 feet of

piping upstream of the mini-flow recirculation line, with a significant portion in the vertical orientation, and this configuration prevents mini-flow recirculation water from mixing with the stagnant hot water in the suction piping.

• The residual heat removal train should be secured, or put in service, only at a reactor coolant system temperature of 240°F (including instrument uncertainties).

The licensee subsequently initiated Condition Report 2008-4997, "Missed Opportunity to Resolve RHR Suction Piping Issue," for the purpose of determining why two separate

- 14 - Enclosure Condition Reports 2008-0989, "Evaluate if Callaway Limitation on RHR Suction Temperature Applies to WCGS"; and 2008-3810, "Evaluate Potential Steam Voiding in RHR Suction While Transitioning to M-3"; initiated for apparently the same issue came to different conclusions. This condition report was also used to perform a root cause analysis of the potential voiding issue associated with the residual heat removal system as well as a past operability review. Through the evaluations that the licensee

performed, they concluded that:

• From a past operability perspective, the residual heat removal system must be considered inoperable any time the plant was in Mode 4 with the reactor coolant system suction isolation valves open and reactor coolant system temperature was above 249.1°F.

• The residual heat removal system must be considered inoperable in Mode 3 during plant heatup for a period of 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> following isolation from the reactor coolant system. This is based on the amount of time it would take the suction piping and fluid to cool down to 225°F.

• From the perspective of past functionality, the residual heat removal system would not have been functional during a small break loss of coolant accident in Modes 3 or 4, nor a large break loss of coolant accident in Mode 3 for injection or recirculation.

Ultimately, through this review the licensee determined that: the direct cause of the

potential voiding issue was that the organization failed to take steps necessary to preclude voiding in the residual heat removal system and the root cause was the residual heat removal system design was not adequate to support all three modes of residual heat removal operation without adversely impacting each other. The licensee also identified as a contributing cause, for the contradictory findings in the condition report evaluations, and the missed opportunities to ensure residual heat

removal train operability was the unrecognized complexity of the residual heat removal systems suction design characteristics which led to a failure by operations and engineering to perform an adequate evaluation of the impact of hot water in the suction piping and the affect this had on operability of the residual heat removal pumps. As a basis for this contributing cause, the licensee identified that the mitigation strategy

chosen to preclude the potential for the flashing of water in the residual heat removal pump suction line was based on analysis and recommendations provided by Westinghouse. The strategy chosen (forced cooling using mini-flow recirculation) best fit the plant system configuration, and Wolf Creek had failed to recognize that the unique design characteristics of the residual heat removal suction piping created an unanalyzed condition for flashing in the piping and subsequent voiding in the pump. The design configuration creates an anomaly in regard to normally accepted standards of water

hammer and flashing criterion. What was not realized was the complexity of the design configuration dynamics of the suction piping for the residual heat removal system in regard to flashing and voiding issues. The current technical evaluation performed to determine residual heat removal operability and the potential impacts of voiding required extensive research, numerous man-hours, numerous personnel, and assistance from

Westinghouse. This issue is a potential voiding concern with the impact and further understanding of voiding only recently being recognized.

- 15 - Enclosure The licensee performed an extent of condition review that examined the challenges in system operation with respect to maintaining the residual heat removal systems suction piping sub-cooled. In this review the licensee determined that the direct extent of condition had been addressed by reviewing the operation of the residual heat removal system under all plant evolutions and operational modes. Specifically, the licensee reviewed all evolutions where maintaining the residual heat removal pump suction piping

water sub-cooled could be challenged by changing system dynamics. During this review, the licensee identified a previously unrecognized concern associated with the automatic swap-over of residual heat removal pump suction from the refueling water storage tank to the containment sump in recirculation mode. The licensee also performed an extent of cause review for the identified root cause. Through this review, the licensee identified that this operability concern is unique to the residual heat removal system because of the three functions that this system is called to perform. Furthermore, the licensee also determined that the other emergency core

cooling system pumps are capable of operating with temperatures of up to 300°F in the recirculation mode and the design does not indicate susceptibility to the identified cause. The inspectors determined that the cause evaluation for the potential voiding issue associated with the residual heat removal system was generally thorough. However, the inspectors determined that in some areas the root cause analysis was narrowly focused and lacked technical rigor when evaluating some aspects of the causes of this issue. Specifically, while the inspectors agreed that system design was a factor in this issue, it was noted that there was a significant amount of industry information available to the licensee that both identified the deficient system design and, if appropriately evaluated,

would have identified the specific issue associated with use of the mini-flow recirculation and its potential system impact. In particular, Nuclear Safety Advisory Letter NSAL-93-004, "RHR Operations as Part of the ECCS During Plant Startup," dated April 20, 1993, was issued by Westinghouse to reiterate the concern of flashing in the residual heat removal suction line while in Modes 3 and 4 operation and provided an assessment of the safety significance. This

advisory described a concern associated with system operation in Mode 4 where the residual heat removal pumps are lined up and taking suction from the reactor coolant system then secured, isolated, and realigned to take suction from the refueling water storage tank prior to transition to Mode 3. In this scenario the pump suction piping and fluid could be at elevated temperatures (as high as 350°F) for some time after Mode 3 is entered, and if a safety injection system actuation occurred, and sufficient time had not elapsed to allow cooling of the system piping by conduction and convection, the pumps

suction pressure could be lowered below the saturation pressure for the corresponding temperature. This would result in fluid in the suction piping flashing to steam and potentially rendering the system inoperable. The inspectors noted that Nuclear Safety Advisory Letter NSAL-93-004 also contained recommended actions to mitigate the condition:

• Assure that the residual heat removal system suction piping is sufficiently cooled before entering Mode 3

• Force cool the residual heat removal system piping

- 16 - Enclosure

• Isolate the residual heat removal system from the reactor coolant system at a low enough temperature at which the residual heat removal system pump suction pressure is above the saturation pressure corresponding to the fluid temperature

However, the advisory specifically identified in the technical evaluation section that use of the mini-flow recirculation method force cools the piping downstream of the mini-flow return and only provides cooling of the water upstream of the mini-flow connection by

means of conduction and convection. As such, the inspectors determined that inadequate engineering evaluations, which had been performed by the licensee, both historically and recently, was another cause of this

issue. Of note, the inspectors identified that the licensee has an ongoing engineering improvement plan from other similar issues associated with engineering rigor, which was still in process at the end of the inspection and tracking the completion of this initiative was credited by the licensee as the corrective action for the contributing cause. Also, the inspectors considered the evaluation to be narrowly focused and lacking in technical rigor with respect to the extent of condition review. Specifically, during their

review, the inspectors noted that Station Procedure AP 28A-100, Section 4.5.1, Revision 7, "Condition Reports," extent of condition is defined as, "The extent to which the actual condition exists or can exist in other plant processes, equipment or human performance. The objective is to reasonably bound the condition in regards to the relative risk it creates for the station." Accordingly, the inspectors determined that the

extent of condition review performed by the licensee was narrowly focused on only the residual heat removal system, and as such, was not a true extent of condition as defined by the station procedure. b. Corrective Actions

The inspectors evaluated the scope, adequacy, and timeliness of the licensee's corrective measures that were both planned and implemented in response to the potential steam voiding issue associated with the residual heat removal system. The inspectors concluded that the actions both planned and implemented by the licensee were appropriate to address the identified issue, to prevent recurrence, and were

consistent with the safety significance of the issue. These corrective actions included:

• Issuing essential reading to the operational crews to keep them aware of the operational changes to the residual heat removal system in Modes 3 and 4

• Issuing operating experience to the industry detailing the concern with the residual heat removal system in Modes 3 and 4

• Revising station calculations to preclude steam voiding

• Revising station operating procedures to only use one train of residual heat removal for reactor coolant system temperature control and to also disable the auto swap-over feature for the train that is used prior to initiation of shutdown cooling.

- 17 - Enclosure 1.3 Related Operating Experience

The team noted that that there was a significant amount of applicable industry information available to the licensee that identified the deficient system design as well as identifying the specific issue associated with use of the mini-flow recirculation and its potential system impact. However, inadequate evaluation of this information resulted in inappropriate implementation of actions by the station. The team determined the licensee's Industry Operating Experience Program previously lacked rigor when evaluating industry operating experience and its applicability to the facility. Specifically, the licensee performed an inadequate evaluation of Nuclear Safety Advisory Letter 93-004, "RHRS Operation as Part of the ECCS During Plant Startup." The team determined that significant improvements have been made to the stations program and procedures pertaining to assessing industry operating experience. Specifically, the licensee has enhanced the following:

• "The initial screening determines if the document could potentially impact the safety or reliability of Wolf Creek Generating Station and insures that the document is entered in the Corrective Action Program."

• "The Supervisor of Improvement Program shall ensure the periodic (at least once per 18 months) performance of effectiveness reviews monitor the success of the Industry Operating Experience Program in attaining its desired objectives and improvements."

• "The first significant operating experience effectiveness review is to be performed one year after completion of all corrective and preventative actions and each subsequent effectiveness review is to be scheduled every 24 months thereafter."

• "A significant operating experience effectiveness review shall be completed on all identified recommendations every six years."

The team noted that the licensee had performed an Industry Operating Experience Re-Evaluation Project to resolve the extent of condition and extent of cause in the quality of evaluations between January 1, 2003, and July 31, 2008. This was performed due to

Corrective Action 4543 from Condition Report 2008-000717. The project consisted of reviewing a sample of 104 from a total of 451 evaluations with an acceptance standard of four or less defective evaluations. The definition of a defective evaluation is when the entire product is considered unacceptable. An evaluation team, which consisted of maintenance, operations, licensing, engineering, and operating experience personnel,

identified eight defective evaluations. Subsequently, an expert panel consisting of a manager of Regulatory Affairs, supervisor of root cause/corrective action, and supervisor of operations reviewed the eight defectives and the definition of defective, and concluded that only four of the eight operating experience evaluations met the defective definition. Thus, four defective evaluations were identified in the project; therefore,the

licensee concluded that no extent of condition or extent of cause was needed.

The team did not review current station evaluations of industry operating experience; however, the team reviewed the four operating experience evaluations that were screened out by the expert panel during the Industry Operating Experience

- 18 - Enclosure Re-Evaluation Project. Subsequently, the inspectors identified a recent industry operating experience evaluation, as documented in Condition Report 2007-2656, which was completed in December 2008. This condition report was written to evaluate Information Notice 2007-01, "Recent OE Concerning Hydrostatic Barriers." The evaluation section of the condition report takes credit for a corrective action associated with Condition Report 2008-3745. When the inspectors reviewed this condition report, it

was determined that this corrective action did not address all aspects which were identified in Information Notice 2007-01. Therefore, the team questioned the evaluation done in Condition Report 2007-2656. The licensee was informed of the teams questions and subsequently the licensee's corrective action, licensing, and operating experience personnel reviewed and determined that the evaluation of the Information

Notice 2007-01 in Condition Report 2007-2656 was inadequate and wrote Condition Report 2009-000939 to address this concern. 1.4 Potential Generic Issues

The team evaluated the circumstances associated with the potential voiding issue and assessed the root cause analysis. Along with this, the team interviewed numerous licensee personnel and reviewed industry operating experience, evaluations the station had performed to analyze this issue as well as NRC generic communications with the goal of identifying any potentially generic issues that should be addressed as a result of this event. The team concluded that, while there is a potential for voiding to occur in any fluid system at any facility, there are no potentially previously unrecognized generic concerns associated with this issue. The team also noted that the licensee has issued an

operating experience report to the industry for future reference. 1.5 Event Precursors

The team performed a review of the licensee's corrective action program documents associated with the residual heat removal system as well as conducting interviews with

station personnel to determine if any previous issues associated with the system could have been viewed as event precursors to the potential voiding condition identified by the licensee. During this review, the team considered previously encountered issues where steam voiding was identified in the suction piping of the residual heat removal system. The inspectors identified two previous events, which had not been recognized by the

licensee in their evaluation, that were indicative of the potential voiding issue. Specifically:

• On May 8, 2008, Wolf Creek initiated Condition Report 2008-2187, "Draining of 'B' RHR Pump," to document that, while attempting to drain the residual heat removal Pump B and associated suction piping to allow performance of corrective maintenance, steam was released into the room that resulted in a personnel contamination event. This condition report established that, prior to the draining evolution of the residual heat removable, Pump B had been in

service providing cooling the reactor coolant system in the shutdown cooling mode. In preparation for the maintenance, the pump had been secured in accordance with Station Procedure SYS EJ-321, Revision 7, "Shut Down of a Residual heat removal Train," of which step 6.2.3 required that the pump be run

- 19 - Enclosure in the recirculation mode until the discharge temperature indicates less than 270°F (the licensee ran the pump in this mode until the discharge temperature indicated 140°F and then secured the pump). Then licensee personnel began draining the pump and piping. Initially, water issued from the pipe but then steam began to issue from the piping which caused the tubing to come out of the floor drain and an individual was contaminated. The condition report initiator

postulated in the condition description of the condition report that: "One explanation to getting steam out of the drain line is that water captured in the line was still at 325 degrees and flashed to steam as draining occurred. When the residual heat removal pump is run in the recirculation mode, there is approximately 15 feet of suction piping that is being recirculated. Upstream of

the recirculation line return is approximately 120 feet of piping that would not see cooling effect of the recirculation flow. As this hot water was depressurized, it would turn to steam until the drain was uncovered and steam allowed to escape." However, the licensee did not investigate the reason for steam formation in the residual heat removal suction piping; instead the focus of this condition report was to determine a better method to secure the drain hoses.

• On May 10, 2008, following maintenance to correct flange leaks on the residual heat removal Pump B discharge flange and refueling water storage tank check valve, the system was aligned to the reactor coolant system, as part of the retest, to place reactor coolant system pressure on the affected joints. Subsequently,

the pump was secured and the train was realigned to take suction from the refueling water storage tank. At this point the licensee attempted to perform ultrasonic testing of the residual heat removal piping to check for voids, but found that the piping was too hot to attach the required instrumentation. The licensee decided to vent the piping in an effort to reduce temperature and vented a mixture

of steam and water for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> before the suction piping became water solid. Though the team noted that there was some discussion and consideration of operability, they determined that the methods and assumptions being used were determined to not be valid. Therefore, it is uncertain if the system would have functioned properly if needed. The licensee did not

investigate this issue any further, nor did they enter this into their corrective action program. Based on this, the team determined that there had been recent event precursors documented by the licensee in various facility databases. As such, the team concluded that due to the lack of a questioning attitude, the licensee had failed to recognize and/or thoroughly evaluate the underlying condition associated with

why steam was vented from the residual heat removal system when it was not expected based on system conditions. As such, this lack of questioning attitude resulted in the licensee's failure to recognize and analyze pertinent information associated with prior issues which were precursors to the issue identified on August 1, 2008. 1.6 Reportability Review

The licensee evaluated the potential voiding condition associated with the residual heat removal system and determined that this was reportable to the NRC in accordance with 10 CFR 50.73(a)(2)(i)(B) as a 60-day report because it represented an operation or

condition prohibited by technical specifications at the station. As such, Licensee Event

- 20 - Enclosure Report 05000482/2008008-00 was submitted on October 3, 2008. This report contained a summary of the initial information known by the licensee at the time of submission. Further evaluation conducted by the licensee provided additional details relative to the safety significance of this issue as well as determining the root cause of the event and past operability of the system. Accordingly, the licensee submitted revised Licensee Event Report 05000/2008008-01 on January 1, 2009. The team reviewed the licensee event reports and determined that the identified aspects of the licensee's reportability determination were correct. However, while reviewing the licensee's past operability determination contained in the root cause analysis, the team

noted the following conclusions:

• From a past operability perspective, the residual heat removal system must be considered inoperable any time the plant was in Mode 4 with the reactor coolant system suction isolation valves open and reactor coolant system temperature was above 249.1°F.

• The residual heat removal system must be considered inoperable in Mode 3 during plant heatup for a period of 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> following isolation from the reactor coolant system. This is based on the amount of time it would take the suction piping and fluid to cool down to 225°F.

• From the perspective of past functionality, the residual heat removal system would not have been functional during a small break loss of coolant accident in Modes 3 or 4, nor a large break loss of coolant accident in Mode 3 for injection or recirculation. Based on this and information contained in the licensee's root cause analysis, the team determined that on March 20, 2008, while in Mode 4 performing a plant cool down for Refueling Outage 16, the licensee had operated the residual heat removal system in a condition prohibited by technical specifications. The team also determined that the

licensee's operation of the residual heat removal system on March 20, 2008 and on May 10, 2008, resulted in a condition that prevented the residual heat removal system from performing its safety function. As such, the team noted that the revised Licensee Event Report 05000/2008008-01 did not identify these reportable conditions, nor had the licensee submitted a separate licensee event report to inform the NRC of the instances that had been identified. Therefore, the team concluded that the licensee had failed to report instances where the residual heat removal system was operated in a condition

prohibited by technical specifications, and a loss of safety function of the system existed between March 20, 2008 and December 8, 2008. The team informed the licensee of their concern. The licensee subsequently entered this into their corrective action program as Condition Reports 2009-1261, and 2009-1326 and Action Requests 15244, 17776, and 15306. The team determined that the licensee's failure to properly report when the station was operated in a condition prohibited by technical specifications and there was a loss of safety function of the residual heat removal system was a violation of 10 CFR 50.73, "Licensee Event Report System." Details associated with this violation are described in

Section 2.1 of this report.

- 21 - Enclosure 2.0 Focused Baseline Inspection Findings

2.1 Failure to Report Conditions Prohibited By Technical Specifications

Introduction. The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73 for failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria. Description. On September 23, 2008, the licensee completed an evaluation of a potential steam voiding issue associated with the residual heat removal systems suction piping that could occur when transitioning from Mode 4 to Mode 3 with elevated fluid temperatures. Based on the results of this evaluation, the licensee determined that both trains of the residual heat removal system had been inoperable during the startup from Refueling Outage 16, on May 10, 2008. Specifically, the residual heat removal system

would not have functioned if a loss of coolant accident had occurred in Mode 3 due to elevated suction piping fluid temperature following transition to Mode 3 from Mode 4. As a result, on October 3, 2008, the licensee submitted Licensee Event Report 05000482/2008008-00, in accordance with 10 CFR 50.73(a)(2(i)(B), to report an operation or condition prohibited by plant technical specifications. Based on further evaluation conducted by the licensee, additional details relative to the safety significance of the potential steam voiding issue and past operability of the residual heat removal system were identified. Accordingly, the licensee submitted

revised Licensee Event Report 05000/2008008-01 on January 1, 2009, to provide the NRC with this additional information that had been learned relative to the residual heat removal systems operation on May 10, 2009. The inspectors reviewed the licensee event reports that had been submitted to the NRC. During this review, the inspectors determined that the licensee had correctly identified and evaluated a reportability aspect during their review. However, while reviewing the licensee's past operability determination contained in the root cause analysis, the inspectors noted the following conclusions:

• From a past operability perspective, the residual heat removal system must be considered inoperable any time the plant was in Mode 4 with the reactor coolant system suction isolation valves open and reactor coolant system temperature

was above 249.1°F.

• The residual heat removal system must be considered inoperable in Mode 3 during plant heatup for a period of 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> following isolation from the reactor coolant system. This is based on the amount of time it would take the suction piping and fluid to cool down to 225°F.

• From the perspective of past functionality, the residual heat removal system would not have been functional during a small break loss of coolant accident in Modes 3 or 4, nor a large break loss of coolant accident in Mode 3 for injection or

recirculation. Based on this and information contained in the licensee's root cause analysis, the inspectors determined that on March 20, 2008, while in Mode 4 performing a plant cool down for Refueling Outage 16, the licensee had operated the residual heat removal

- 22 - Enclosure system in a condition prohibited by technical specifications as well. The inspectors also determined that the licensee's operation of the residual heat removal system on March 20, 2008, and on May 10, 2008, resulted in a condition that prevented the residual heat removal system from performing its safety function. As such, the inspectors noted that both of these issues were reportable as defined by 10 CFR 50.73, and the revised Licensee Event Report 05000/2008008-01 did not identify these

reportable conditions, nor had the licensee submitted a separate licensee event report to inform the NRC of the instances that had been identified. Therefore, the inspectors concluded that the licensee had failed to report instances where the residual heat removal system had been operated in a condition prohibited by technical specifications and a loss of safety function of the system existed between March 20, 2008, and

December 8, 2008. The inspectors informed the licensee of their concerns. The licensee initiated Condition Report 2009-1261 and Action Requests 15244, 17776, and 15306 to address this

concern. Analysis. The inspectors reviewed this issue in accordance with Inspection Manual Chapter 0612 and the NRC Enforcement Manual. Through this review, the inspectors determined that traditional enforcement was applicable to this issue because the NRC's regulatory ability was affected. Specifically, the NRC relies on licensee to identify and report conditions or events meeting the criteria specified in regulations in order to

perform its regulatory function, and when this is not done, the regulatory function is impacted. The inspectors determined that this finding was not suitable for evaluation using the significance determination process, and as such, was evaluated in accordance with the NRC Enforcement Policy. The finding was reviewed by NRC management and because the violation was determined to be of very low safety significance, was not repetitive or willful, and was entered into the corrective action program, this violation is being treated as a Severity Level IV noncited violation consistent with the NRC

Enforcement Policy. This finding was determined to have a crosscutting aspect in the area of Problem Identification and Resolution associated with the corrective action program P.1(c), in that the licensee failed to appropriately and thoroughly evaluate for reportability aspects all factors and time frames associated with the inoperability of residual heat removal system when suction temperatures were above 249°F. Enforcement. Title 10 CFR 50.73(a)(1) requires, in part, that licensees shall submit a licensee event report for any event of the type described in this paragraph within 60 days after the discovery of the event. Title 10 CFR 50.73(a)(2)(i)(B) requires, in part, that the licensee report any operation or condition prohibited by the plant's technical specification. Contrary to the above, it was determined that the residual heat removal system had been operated in a condition prohibited by technical specifications during the

cool down for Refueling Outage 16 on March 20, 2008; and the licensee failed to submit a licensee event report or include this information in revised Licensee Event Report 05000482/2008008-01, submitted on January 1, 2009. This finding was determined to be applicable to traditional enforcement because the failure to report conditions or events meeting the criteria specified in regulations affects the NRC's regulatory ability. The finding was evaluated in accordance with the NRC's Enforcement Policy. The finding was reviewed by NRC management and because the violation was

of very low safety significance, was not repetitive or willful, and was entered into the corrective action program, this violation is being treated as a Severity Level IV noncited violation, consistent with the NRC Enforcement Policy: NCV 05000482/2009006-01,

- 23 - Enclosure "Failure to Report Conditions Prohibited by Technical Specifications, and Safety System Functional Failures."

2.2 Inadequate Procedures for Mode Shifting of the Residual Heat Removal System

Introduction. The inspectors identified a noncited violation of Technical Specification 5.4.1, "Procedures," associated with the licensee's failure to ensure that adequate procedures were available for changing modes of operation of the residual heat removal system from shutdown cooling to emergency core cooling system operation.

Description. On January 18, 2008, the licensee initiated Condition Report 2008-0164 to address NRC Generic Letter 2008-001, "Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems," concerns. The purpose of this condition report was to evaluate the licensing basis, design, testing, and corrective action programs for the emergency core cooling systems, residual heat removal system, and containment spray system to ensure that gas accumulation is maintained less than the amount that challenges operability of these systems, and that appropriate action is taken when conditions adverse to quality are identified. On May 10, 2008, following maintenance to correct flange leaks on the residual heat removal Pump B discharge flange and refueling water storage tank check valve, the system was aligned to the reactor coolant system, as part of the retest, to place reactor coolant system pressure on the affected joints. Subsequently, the pump was secured and the train was realigned to take suction from the refueling water storage tank. At this point, the licensee attempted to perform ultrasonic testing of the residual heat removal

piping to check for voids, but found that the piping was too hot to attach the required instrumentation. The licensee decided to vent the piping in an effort to reduce temperature and vented a mixture of steam and water for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> before the suction piping became water solid. The licensee initiated Station Work Order 08-306203-000 to perform troubleshooting to determine if the suction piping

temperature was below saturation temperature where the recirculation line taps into the system. Subsequently, during the stations evaluation of Condition Report 2008-0164, an operations representative identified a concern with the potential for steam binding. Specifically, steam voiding concerns that had been identified during restoration of the residual heat removal system at the end of Refueling Outage 16, on May 10, 2008, which could happen any time the station enters Mode 3, combined with the findings from the generic letter review, prompted the initiation of another condition report to review

these concerns. On August 1, 2008, station personnel initiated Condition Report 2008-3810 to address concerns that had been identified during the review for Generic Letter 2008-01 regarding potential void formation. This condition report questioned the past operability of the residual heat removal system when aligned in the injection mode with suction piping temperature as high as 350°F, as well as the current design adequacy to ensure cooling of suction piping using the mini-flow recirculation line. An evaluation was requested to determine the effects of potential steam voiding in the residual heat removal suction

piping when realigning the system from reactor coolant system cooling to emergency core cooling system injection while transitioning from Mode 4 to Mode 3.

- 24 - Enclosure On September 23, 2008, the licensee completed their evaluation of the potential voiding issue and concluded that recirculation can not be relied upon to cool the water in the isolated suction line, the residual heat removal system would not have functioned if a loss of coolant accident had occurred in Mode 3 with elevated suction piping fluid temperature, and the residual heat removal train used for shutdown cooling should be secured, or put in service, only at a temperature of 240°F to ensure operability. On October 10, 2008, the licensee initiated Condition Report 2008-4997, "Missed Opportunity to Resolve RHR Suction Piping Issue," for the purpose of determining why two separate conditions initiated for apparently the same issue came to different

conclusions. This condition report was also used to perform a root cause analysis of the potential voiding issue associated with the residual heat removal system as well as performing a past operability review.

On December 5, 2008, the licensee completed their evaluations as directed by Condition Report 2008-4997. Through the evaluations that the licensee performed, they concluded that:

• From a past operability perspective, the residual heat removal system must be considered inoperable any time the plant was in Mode 4 with the reactor coolant system suction isolation valves open and reactor coolant system temperature was above 249.1°F.

• The residual heat removal system must be considered inoperable in Mode 3 during plant heatup for a period of 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> following isolation from the reactor coolant system. This is based on the amount of time it would take the suction piping and fluid to cool down to 225°F.

• From the perspective of past functionality, the residual heat removal system would not have been functional during a small break loss of coolant accident in Modes 3 or 4, nor a large break loss of coolant accident in Mode 3 for injection or

recirculation.

Ultimately, through this review, the licensee determined that the direct cause of the potential voiding issue was that the organization failed to take steps necessary to preclude voiding in the residual heat removal system and the root cause was the

residual heat removal system design was not adequate to support all three modes of residual heat removal operation without adversely impacting each other. The licensee also identified as a contributing cause, for the contradictory findings in the condition report evaluations and the missed opportunities to ensure residual heat removal train operability was the unrecognized complexity of the residual heat removal systems suction design characteristics which led to a failure by operations and engineering to perform an adequate evaluation of the impact of hot water in the suction piping and the

affect this had on operability of the residual heat removal pumps. The inspectors reviewed the licensee's root cause analysis for this issue. While the inspectors agreed that system design was a factor in this issue, they however noted that there was a significant amount of industry information available to the licensee that both identified the deficient system design and, if appropriately evaluated, would have identified the specific issue associated with use of the mini-flow recirculation and its

potential system impact. As such, the inspectors determined that engineering

- 25 - Enclosure evaluations, which had been performed by the licensee, both historically and recently, was another cause of this issue. Of note, the inspectors identified that the licensee has an ongoing engineering improvement plan from other similar issues associated with engineering rigor, which is still in process and tracking the completion of this initiative was credited by the licensee as the corrective action for the contributing cause. Analysis. The licensee's failure to ensure that adequate procedures were available for changing modes of operation of the residual heat removal system from shutdown cooling to emergency core cooling system operation was a performance deficiency. The finding was more than minor because it was associated with the equipment performance

attribute of the Mitigating Systems Cornerstone and it directly affected the 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, "Significance Determination Process," Phase 1 Worksheet, the inspectors concluded that a Phase 2 evaluation was required because this finding represented a

loss of safety function of the residual heat removal system. The inspectors performed a Phase 2 analysis using Appendix A, "Determining the Safety Significance of Reactor Inspection Findings for At-Power Situations," of Inspection Manual Chapter 0609, "Significance Determination Process," and the plant specific Phase 2 presolved tables and worksheets for Wolf Creek. The inspectors determined that the Phase 2 presolved tables and worksheets did not contain appropriate target sets to accurately estimate the risk input of the finding. Therefore, it was determined that a Phase 3 analysis was required. Senior risk analysts performed a Phase 3 analysis of this issue. The estimated Conditional Core Damage Probability was determined to be 2.84E-7, and the estimated Conditional Large Early Release Probability was determined to be 2.72E-9. Based on these results, the finding was determined to be of very low safety significance, Green. The complete Phase 3 analysis is available from the Publicly Available Records component of NRC's document systems (ADAMS) as ML091760764. This finding was determined to have a crosscutting aspect in the area of Problem Identification and Resolution associated with the corrective action program P.1(c), in that the licensee failed to appropriately and thoroughly evaluate problems such that the resolutions address the causes. Enforcement. Technical Specifications, Section 5.4.1, "Procedures," requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Regulatory Guide 1.33, Appendix A, Section 3.c, requires, in part, that instructions for changing modes of operation of the residual heat removal system should be prepared. Contrary to the above from 1992 through December 2008, the

licensee failed to provide adequate instructions for changing modes of operation of the residual heat removal system. Specifically, station procedures allowed the residual heat removal system to be realigned to the emergency core cooling system mode of operation when the system was not able to perform its safety function. Because this violation was of very low safety significance and it was entered into the licensee's corrective action program as Condition Reports 2008-3810 and 2008-4997, this violation

- 26 - Enclosure is being treated as a noncited violation, consistent with Section VI.A of the NRC Enforcement Policy: NCV 05000482/2009006-02, "Inadequate Instructions for Changing Modes of Operation of the Residual Heat Removal System." 4OA6 Meetings Exit Meeting Summary

On February 27, 2009, prior to the teams departure from the facility, an inspection debrief was conducted with Mr. R.A. Muench, President and CEO, and other members of the licensee staff to apprise them of the teams results to date and to explain that the inspection would continue with in office review pending resolution of all questions.

On July 9, 2009, the team conducted a telephonic exit meeting to present the inspection results to Mr. Matt Sunseri, Vice President of Operations and Plant Manager, and other members of the licensee staff. The licensee acknowledged the issues presented. The team acknowledged review of proprietary material, as part of the inspection but no proprietary information was included in the report.

A-1 Attachment SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

R. Muench, President and Chief Executive Officer M. Sunseri, Vice President Operations and Plant Manager S. Hedges, Vice President Oversight G. Pendergrass, Manager, System Engineering T. Garrett, Vice President Engineering

G. Neisis, Manager Design S. Henry, Manager Operations R. Flannigan, Manager Regulatory Affairs D. Hooper, Supervisor Licensing W. Muilenburg, Licensing J. Hsen, Safety Analysis D. Erbe, Manager Security

S. Skidmore, Corrective Actions F. Laflin, Chief Engineer J. Patel, Engineering Supervisor W. Ketchum, Supervisor Fuels/Probabilistic Safety Analysis L. Parmenter, Assistant to Operations Manager

T. Card, Supervisor System NSSS S. Koenig, Manager Corrective Actions J. Harris, System Engineer D. Garrison, Operations Support NRC Personnel

J. Josey, Resident Inspector M. Young, Reactor Inspector

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and Closed

05000482/2009006-01 NCV Failure to Report Conditions Prohibited by Technical Specifications and Safety System Functional Failures05000482/2009006-02 NCV Inadequate Instructions for Changing Modes of Operation of the Residual Heat Removal System

Closed None

A-2 Attachment LIST OF DOCUMENTS REVIEWED

PROCEDURES

NUMBER TITLE REVISION SYS BG-216 Reactor Make-up Control System Alternate Operation

24 EMG C-11 Loss of Emergency Coolant Recirculation 20 EDMG-T01 EDMG Tool Box 4 EMG C-13 Control Room Response to Sump Blockage 2 OFN EJ-015 Loss of RHR Cooling 15A SYS EJ-121 Startup of a RHR Train in Cooldown Mode 21 SYS EJ-120 Startup of a Residual Heat Removal Train 51 GEN 00-006 Hot Standby to Cold Shutdown 68 GEN 00-008 Reduced Inventory Operations 18A AP 20E-001 Industry Operating Experience Program 12 AP 28A-100 Condition Reports 7 OFN EJ-40 CL Recirc During Mode 3, With Accumulators Isolated, Mode 4, 5 or 6

2 EMG ES-11 Post LOCA Cooldown and Depressurization 14 EMG ES-12 Transfer to Cold Leg Recirculation 12 GEN 00-002 Cold Shutdown to Hot Standby 67A SYS EJ-320 Placing RHR System In Safety Injection Standby Condition

32 EP-01-2.1-1 Emergency Action Levels 10 EPP 01-2.1 Emergency Classification 18 OFN BB-031 Shutdown LOCA 9 STS EJ-100A RHR System Inservice Pump A Test 23 STS EJ-100B RHR System Inservice Pump B Test 19 SYS EJ-321 Shutdown of a Residual Heat Removal Train 23 SYS EJ-323 RHR System Depressurization 9 OFN NB-34 Loss of All AC Power - Shutdown Conditions 5 ALAR 00-050C RHR Loop 2 Flow Lo 10 ALAR 00-49C RHR Loop 2 Flow Lo 11

A-3 Attachment CALCULATIONS

NUMBER TITLE REVISION AN-01-025 No Title 0

AN-97-027 Time To Boil In The Core and Core Uncovery In The Event of a Loss of RHR Cooling During Refueling 9

0 CONDITION REPORTS

2007-2162

2007-2656 2008-0164

2008-0717

2008-0989 2008-2187 2008-0717 2008-0989

2008-2187 2008-2262 2008-3745 2008-3810

2008-4997

2008-4997 2008-5912 2008-5913

2008-5915

2008-5917 2009-0939 2009-1261

PERFORMANCE IMPROVEMENT REQUEST

PIR 99-0228 PIR 2004-2440 MISCELLANOUS

NUMBER TITLE REVISION NSAL-93-004 RHRS Operation as Part of the ECCS During Plant Startup 0 LER 2008-008-00 Potential for Residual Heat Removal Trains to be Inoperable during Mode Change

0 LER 2008-008-00 Potential for Residual Heat Removal Trains to be Inoperable during Mode Change

1 ITIP 02324 Westinghouse Letter SAP-93-706 (4-29-93): RHR Operation As Part Of The ECCS During Plant Startup (Residual Heat Removal) (NSAL-93-004)

0 ITIP 05342 Westinghouse InfoGram IG-04-6: Reactor Trip Breaker Auto Shunt Trip Test Panel. February 25, 2009 ITIP 05288 SER 3-04 - Reactor Overpower Events Associated with Ultrasonic Feedwater Flow Measurement Systems February 25, 2009 SEL 2009-135 Self Assessment Plan "Industry Operating Experience Program"

Assessment 92 Assessment/Audit Detail Report "Industry Operating Experience Program" September 21, 2007

A-4 Attachment NUMBER TITLE REVISION Quick Hit Detail Report

1369 IOE Re-Evaluation Project LTR-LIS-09-361 Engineering Report Wolf Creek Generating Station Modes 3 and 4 Loss-of-Coolant Accident Analysis For Residual Heat Removal Operability Study June 5, 2009 SY1300600 Emergency Core Cooling System 18 SY1300500 Residual Heat Removal 15

SY1505900 Feedwater System 16

SY1303200 Containment System 15

SY1300600 Emergency Core Cooling System 18

SY1300400 Chemical Volume and Control System 22