Text

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; Wolf Creek
References
	IR-09-006
	Download: ML092240087 (34)
	v • d • e

UNITED STATES

NUC LE AR RE G UL AT O RY C O M M I S S I O N

R E GI ON I V

612 EAST LAMAR BLVD , SU I TE 400

AR LI N GTON , TEXAS 76011-4125

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 stations 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.

The NRCs 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 Commissions 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

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 NRCs document system (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the

Public Electronic Reading Room).

Sincerely,

/RA/

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 Chief Engineer

Wolf Creek Nuclear Operating Corporation Utilities Division

P.O. Box 411 Kansas Corporation Commission

Burlington, KS 66839 1500 SW Arrowhead Road

Topeka, KS 66604-4027

Jay Silberg, Esq.

Pillsbury Winthrop Shaw Pittman LLP Office of the Governor

2300 N Street, NW State of Kansas

Washington, DC 20037 Topeka, KS 66612-1590

Supervisor Licensing Attorney General

Wolf Creek Nuclear Operating Corporation 120 S.W. 10th Avenue, 2nd Floor

P.O. Box 411 Topeka, KS 66612-1597

Burlington, KS 66839

Wolf Creek Nuclear Operating Corp. -3-

County Clerk Chief, Technological Hazards

Coffey County Courthouse Branch

110 South 6th Street FEMA, Region VII

Burlington, KS 66839 9221 Ward Parkway

Suite 300

Chief, Radiation and Asbestos Kansas City, MO 64114-3372

Control Section

Bureau of Air and Radiation

Kansas Department of Health and

Environment

1000 SW Jackson, Suite 310

Topeka, KS 66612-1366

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

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

-1- 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 licensees 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

licensees 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.

-2- 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

licensees 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

-3- 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.

-4- 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 managements 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 licensees 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

-5- 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 Callaways 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

-6- 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 licensees 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 licensees 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 licensees corrective action program.)

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.

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

-7- 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 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months 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.

-8- 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 Westinghouses 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

-9- 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 Engineerings 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.

- 10 - 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

- 11 - 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

- 12 - Enclosure

The inspectors reviewed and assessed the licensees 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 licensees 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

- 13 - 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 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months 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.

- 14 - 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

- 15 - 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 licensees

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.

- 16 - 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 licensees 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

- 17 - 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 licensees 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 licensees 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

- 18 - 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 licensees 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

- 19 - 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 licensees reportability determination were correct. However, while reviewing the

licensees 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 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months 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 licensees 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

licensees 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 licensees 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.

- 20 - 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

licensees 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 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months 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 licensees 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

- 21 - Enclosure

system in a condition prohibited by technical specifications as well. The inspectors also

determined that the licensees 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 NRCs

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,

- 22 - 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 licensees 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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.