IR 05000483/2006011

IR 05000483-06-011, on 04/11-06/16/2006 for Callaway Plant: Special Inspection to Evaluate Amerenue'S Discovery That Component Cooling Water Flow to the Residual Heat Removal Heat Exchangers Would Not Have Been Established Until After the P

ML061950308
Person / Time
Site:	Callaway
Issue date:	07/14/2006
From:	Jones W B NRC/RGN-IV/DRP/RPB-B
To:	Naslund C D Union Electric Co
References
IR-06-011
Preceding documents:	ML062570547 (14 September 2006)
Download: ML061950308 (35)

Text

July 14, 2006

Charles D. Naslund, Senior Vice President and Chief Nuclear Officer Union Electric Company

P.O. Box 620 Fulton, MO 65251

SUBJECT: CALLAWAY PLANT - NRC SPECIAL INSPECTION REPORT 05000483/2006011

Dear Mr. Naslund:

On April 11-14, 2006, the U.S. Nuclear Regulatory Commission (NRC) conducted a specialinspection at your Callaway Plant. The inspection effort continued with in-office and additional on-site reviews through June 16, 2006. The purpose of the inspection was to evaluate the impact of the discovery that component cooling water would not be established to the residualheat removal heat exchangers until after the postloss-of-coolant accident recirculation phasewas initiated. The enclosed report documents the inspection findings, which were discussed on June 26, 2006, with Mr. Tim Herrmann and members of your staff. The inspection was conducted as a result of your staff's identification, during a plant simulatorexercise, that component cooling water to the residual heat removal heat exchangers would not have been established until the containment recirculation phase of emergency core coolingsystem injection had been initiated. The failure to establish procedures that were consistentwith the safety analysis could have challenged the ability of the emergency core cooling systemin performing its safety functions during the containment recirculation phase. As discussed in detail in the enclosed report, because the underlying safety concern was corrected on March 30, 2006, and does not represent a current safety concern, the inspection focused onthe circumstances that lead up to your staff identifying this condition, AmerenUE's response, including the root cause and extent of condition reviews, and the identification of any genericissues related to the design and operating practices that resulted in this condition. This inspection report documents several opportunities prior to March 27, 2006, includingoperating experience and review of other emergency operating procedure deficiencies, to identify that the established emergency operating procedures did not ensure that the facilitywould be operated in accordance with the safety analysis. In addition, the inspection team identified that, after the condition was identified, the immediate actions that were taken to placethe plant in a configuration to meet the safety analysis did not adequately consider thecomponent cooling water system response to a loss of offsite power. The plant wassubsequently placed in a configuration that supports the design basis component cooling watersystem requirements.

Union Electric Company-2-Based on the results of this inspection, the NRC identified two findings, each evaluated underthe risk significance determination process as having very low safety significance (Green). The NRC also determined that there was a violation associated with each of the findings. Theseviolations are being treated as noncited violations, consistent with Section VI.A of the Enforcement Policy. These noncited violations are described in the subject inspection report.

In addition, a licensee-identified violation, which was determined to be of very low safety significance, is listed in the report. If you contest these violations or the significance of theviolations, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with copies to the Regional Administrator, U.S.

Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive, Suite 400, Arlington, Texas 76011-4005; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at the Callaway Plant facility. Inaccordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and itsenclosure will be available electronically for public inspection in the NRC Public DocumentRoom 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 (thePublic Electronic Reading Room).

Sincerely,/RA/William B. Jones, ChiefProject Branch B Division of Reactor ProjectsDocket: 50-483License: NPF-30

Enclosure:

Inspection Report 05000483/2006011

w/attachments:

Supplemental InformationTimeline Describing CCW to RHR Heat Exchangers ProblemCharter Memorandum dated April 10, 2006cc w/enclosure:Professional Nuclear Consulting, Inc.

19041 Raines Drive Derwood, MD 20855John O'Neill, Esq.Pillsbury Winthrop Shaw Pittman LLP 2300 N. Street, N.W.

Washington, DC 20037 Union Electric Company-3-Keith A. Mills, Supervising Engineer, Regional Regulatory Affairs/

Safety Analysis AmerenUE P.O. Box 620 Fulton, MO 65251Missouri Public Service CommissionGovernor's Office Building 200 Madison Street

P.O. Box 360 Jefferson City, MO 65102H. Floyd GilzowDeputy Director for Policy Missouri Department of Natural Resources

P. O. Box 176 Jefferson City, MO 65102-0176Rick A. Muench, President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation

P.O. Box 411 Burlington, KS 66839Dan I. Bolef, PresidentKay Drey, Representative Board of Directors Coalition for the Environment 6267 Delmar Boulevard University City, MO 63130Les H. Kanuckel, ManagerQuality Assurance AmerenUE P.O. Box 620 Fulton, MO 65251Director, Missouri State Emergency Management Agency

P.O. Box 116 Jefferson City, MO 65102-0116Keith D. Young, ManagerRegulatory Affairs AmerenUE P.O. Box 620 Fulton, MO 65251 Union Electric Company-4-David E. Shafer Superintendent, Licensing Regulatory Affairs AmerenUE P.O. Box 66149, MC 470 St. Louis, MO 63166-6149Certrec Corporation4200 South Hulen, Suite 630 Fort Worth, TX 76109Keith G. Henke, PlannerDivision of Community and Public Health Office of Emergency Coordination 930 Wildwood, P.O. Box 570 Jefferson City, MO 65102Chief, Radiological Emergency Preparedness Section Kansas City Field Office Chemical and Nuclear Preparedness and Protection Division Dept. of Homeland Security 9221 Ward Parkway Suite 300 Kansas City, MO 64114-3372 Union Electric Company-5-Electronic distribution by RIV:Regional Administrator (BSM1)DRP Director (ATH)DRS Director (DDC)DRS Deputy Director (RJC1)Senior Resident Inspector (MSP)Branch Chief, DRP/B (WBJ)Senior Project Engineer, DRP/B (RAK1)Team Leader, DRP/TSS (RLN1)RITS Coordinator (KEG)DRS STA (DAP)J. Lamb, OEDO RIV Coordinator (JGL1)ROPreports CWY Site Secretary (DVY)W. A. Maier, RSLO (WAM)SUNSI Review Completed: ___WBJ_ ADAMS: Yes G No Initials: __WBJ_ Publicly Available G Non-Publicly Available G Sensitive Non-SensitiveR:\_REACTORS\_CW\2006\CW2006011RP-DED.wpdRIV:RI:DRP/BRI:DRS/EB2SRA:DRSC:DRS/EB2C:DRP/BDEDumbacherGAPickDPLovelessLJSmithWBJonesE-WBJonesE- WBJones/RA//RA//RA/7/10/067/10/067/14/067/13/067/13/06OFFICIAL RECORD COPY T=Telephone E=E-mail F=Fax

-1-EnclosureU.S. NUCLEAR REGULATORY COMMISSION REGION IV Docket:50-483License:NPF-30 Report:05000483/2006011 Licensee:AmerenUE Facility:Callaway Plant Location:Junction Highway CC and Highway O Fulton, Missouri Dates:April 11-14, 2006, with additional on-site in-office inspection throughJune 16, 2006Team Leader:D. Dumbacher, Senior Resident Inspector, Project Branch B Inspectors:G. Pick, Senior Reactor Inspector, Engineering BranchD. Loveless, Senior Reactor AnalystApproved By:W. B. Jones, Chief, Project Branch B, Division Reactor Projects

-2-Enclosure

SUMMARY OF FINDINGS

IR 05000483/2006011; 04/11-06/16/06; Callaway Plant: Special Inspection to evaluateAmerenUE's discovery that component cooling water flow to the residual heat removal heat exchangers would not have been established until after the postloss-of-coolant accident recirculation phase was initiated.This report covered the initial on-site inspection conducted April 11-14, 2006, with in-officereview and additional on-site inspection conducted through June 16, 2006, by a special inspection team consisting of one resident inspector, one region-based reactor inspector, and one region-based senior reactor analyst. Two noncited violations were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609, "Significance Determination Process." Findings for which the significance determination process does not apply may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation ofcommercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process,"Revision 3, dated July 2000.A.

Cornerstone: Mitigating Systems

Green.

The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,Criterion XVI, for the failure to take adequate corrective action to prevent recurrence of a significant condition adverse to quality. Specifically, AmerenUE failed to correct the Emergency Operating Procedure deficiencies associated with Final Safety Analysis Report requirements following an April 15, 1998 notification of the same deficiencies at another standardized nuclear unit power plant system plant. At that time AmerenUE didnot identify and correct similar deficiencies involving the component cooling water system support function for residual heat removal heat exchangers. The EmergencyOperating Procedure deficiencies were discovered by plant personnel on March 27, 2006, during a simulator exercise involving the transition to the emergency core cooling system recirculation phase. Problem identification and resolution cr osscutti ng aspectswere identified for the failure to adequately identify and correct Emergency Operating Procedures deficiencies to ensure operation within the design basis. This issue was more than minor because it affected the Mitigating Systems cornerstoneobjective of equipment reliability. The failure to provide for component cooling watersystem flow through the residual heat removal heat exchangers for initial containmentrecirculation could result in a loss of the component cooling water system and thusbecome a much more significant safety concern. AmerenUE's evaluation of the condition was considered for the time allowable to establish component cooling water flow before a loss of the component cooling water system would occur. AmerenUEprovided an evaluation that demonstrated a loss of component cooling water would not occur based on the timing of operator actions. Because the timing did affect the probabilistic risk assessment for human reliability, a Phase 3 risk assessment wasperformed by an NRC senior reactor analyst. The analyst determined that the finding-3-Enclosurewas of very low safety significance,

Green.

AmerenUE entered this issue into their corrective action program as Callaway Action Request 200602565 (Section 03).*Green. The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B,Criterion XVI, for AmerenUE's failure to implement appropriate corrective actions for maintaining component cooling water flow consistent with design basis requirements.

On April 11 and 12, 2006, AmerenUE placed the Train A component cooling water system in a configuration which could result in component cooling water pump runout inthe event of a loss-of-coolant accident coincident with a loss of offsite power.

Crosscutting aspects associated with problem identification and resolution wereidentified for the failure to implement appropriate corrective actions to ensure the component cooling water system remained operable for other design basis events. This issue was more than minor because it affected the Mitigating Systems cornerstoneobjective of equipment reliability in that a loss of one train of the component coolingwater system could cause other mitigating equipment (i.e., pumps and heat exchangers)to fail and thus become a much more significant safety concern. Using the NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Screening Worksheet, the finding was determined to be of very low safety significance because it did not result in a loss of safety function for a single train for greater than its Technical Specification allowed outage time. AmerenUE entered this issue into its corrective action program as Callaway Action Request 200602995 (Section 04.02).B.Licensee-Identified FindingA violation of very low significance, which was identified by AmerenUE, has beenreviewed by the inspectors. Corrective actions taken or planned by AmerenUE have been entered into AmerenUE's corrective action program. This violation and the corrective action tracking number are listed in Section 4OA7 of this report.

-4-Enclosure

REPORT DETAILS

01 Backgr ound01.1Summary of Discovery and Immediate Response to Component Cooling Water (CCW)System Operability for Emergency Core Cooling System (ECCS) ContainmentRecirculation On March 27, 2006, operations personnel were conducting emergency operatingprocedure (EOP) validations on the plant simulator to verify "time critical" manual operator actions. During this activity a senior reactor operator identified a concern with the timing of CCW initiation during ECCS containment recirculation. Although the validation actions were not specifically being conducted to validate the time at which CCW would be initiated, the operator noted that CCW may not be established to theresidual heat removal (RHR) heat exchangers until after the postloss-of-coolant accident(post-LOCA) recirculation phase was automatically initiated. Subsequently, theCallaway Training Department requested that Wolf Creek Generating Station provideinformation on CCW initiation for ECCS recirculation and a calculation for the allowed maximum design basis CCW temperatures from a previous NRC violation (50-482/9812-01).On March 29, 2006, Callaway received the requested information and the review wascompleted on March 30, 2006. Corrective Action Request 200602565 was initiated the same morning. The concern with the timing of CCW initiation during ECCS containment recirculation was then relayed to the Operations shift crew who aligned CCW to the

RHR heat exchangers to provide continuous flow during power operation. In accordance with Management Directive 8.3, " NRC Incident Investigation Program,"the NRC determined that a special inspection was warranted, in part, on the basis of thepotential safety significance of a loss of CCW. AmerenUE established a root cause team and a past operability determination team on April 6, 2006. The NRC chartered aspecial inspection which began on April 11, 2006. The inspection team completed all aspects identified in the charter on June 16, 2006. The team used NRC InspectionProcedure 93812, "Special Inspection," to perform the scope identified in the inspection charter, dated April 10, 2006. The charter may also be found in the NRC PublicDocument Room or from the Publicly Available Records component of NRC's documentsystem (ADAMS) under Accession Number ML061010217. ADAMS is accessible fromthe NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public ReadingRoom). 01.2Impact of CCW Initiation to RHR Heat Exchangers Following Post-LOCA RecirculationPhaseThe Callaway Final Safety Analysis Report (FSAR), Section 9.1.3.2.3 and Table 6.3-8,specified that the operators initiate CCW flow to the RHR heat exchangers as therefueling water storage tank (RWST) level neared the automatic transfer setpoint and prior to the recirculation phase. This was significant because the automatic transfer ofRHR pump suction from the RWST to the containment recirculation sump would introduce hot water, approximately 265F, to the RHR heat exchangers. ContainmentECCS recirculation, without CCW cooling flow to the RHR heat exchangers, would heatup the shell side of the RHR heat exchangers to temperatures in excess of the designbases CCW system temperature and possibly cause boiling of the CCW water. 02Prior Opportunities to Address Emergency Operating Procedure Deficiencies 02.01Generic Communications Related to Containment RecirculationThe following provides a summary of selected generic communications applicable toCallaway ECCS containment recirculation and CCW initiation.

10/14/76Westinghouse issued Letter SLBE 6-803 recommending automatic CCWinitiation to the RHR heat exchangers prior to the swapover point. Callaway plant, owned by Union Electric Company, was part of the Standardized Nuclear Unit Power Plant System (SNUPPS) group.

SNUPPS documented that the manual action was acceptable as operators were expected, with training, to safely perform the requirement and because automatic action would result in additional unnecessary surveillances. The letter stated that automatic function could be backfitted by the NRC at the FSAR st age.5/29/80Westinghouse issued SNUPPS Letter SNP-3346. The letter stated thatCCW must be aligned to the RHR heat exchanger prior to swapover inthe recirculation mode.12/82Generic Letter 82-33, SUPPLEMENT 1 TO NUREG-0737-REQUIREMENTS FOR EMERGENCY RESPONSE CAPABILITY, Section 7.1, established requirements for licensees to reanalyze transients and accidents and prepare technical guidelines. These analyses were to identify critical operator tasks and were to be the basesfor upgraded EOPs. AmerenUE's commitments were to provide a procedures generation package, including a program for validating EOPs.

Callaway had several opportunities to validate that CCW is established to

RHR heat exchangers prior to transfer to the cold leg recirculation phase.02.02Licensee Documents Addressing Callaway Containment ECCS RecirculationThe following provides a summary of selected corrective action and licensing documentsinvolving the Callaway ECCS containment recirculation and CCW initiation.

1980 to 1982Callaway FSAR was issued. FSAR Section 9.1.3.2.3 and Table 6.3-8stated that the CCW initiation must be initiated prior to ECCS recirculation mode swapover.1984Callaway EOPs were initiated and required, in Procedure ES 1.3,"Transfer to Cold Leg Recirculation," that the CCW to the RHR heatexchangers be initiated. The Westinghouse emergency response guideline (ERG), for Procedure E-1, "Loss of Reactor or SecondaryCoolant," did not have a step to open the CCW inlet valves to the RHR heat exchangers. The ERG basis to Procedure ES 1.3, step 2, specifies that the step to align CCW was a "verify" step that assumed previous CCW flow initiation to each RHR heat exchanger. 4/15/1998Callaway initiated a corrective action document, SOS 98-1577, notingthat the NRC had issued Wolf Creek Generating Station a 10 CFR 50.59violation highlighting that late initiation of the CCW to the RHR heatexchangers could result in 270F recirculation sump water beingintroduced to the RHR heat exchangers. Without cooling this could resultin exceeding the design temperature of the CCW system and causeboiling to occur (Wolf Creek Generating Station PIR 973483).5/5/98In response to SOS 98-1577, Callaway recognized that Procedure E-1did not have a step prior to entry to Procedure ES 1.3 and added Step 14 to open the CCW inlet valve to each RHR heat exchanger. The changewas made as a temporary change notice (TCN 98-0427). The added step was not validated to ensure it would address the concern. 9/5/2002Callaway corrective action document Callaway Action Request(CAR) 200205499 stated that the Callaway EOP validation process hadvalidated Westinghouse recommendations in regard to EOP steps to enter cold leg recirculation. The CAR stated that Callaway Plant had nointerim configuration issues and that FSAR Table 6.3.2 commitments for timing actions during the swapover were met. 1/2/2004Callaway corrective action document CAR 200400017 noted that WolfCreek Generating Station required that the CCW inlets to each RHR heatexchanger be opened in 90 seconds or less following the automatic sump swapover. The CAR initiator asked if Callaway had any similar concerns.

The Callaway accident analysis group identified no concerns. 1/27/2005Callaway corrective action document CAR 200500564 stated that FSARTable 6.3-8 assumed that CCW flow is aligned to the RHR heatexchangers before the RWST low-low-1 swapover point is reached. The initiator questioned why the RWST outflow analysis did not explicitly include times to align CCW flow to the RHR heat exchangers. Theresponse to the CAR was that steps not directly associated with the swapover were not appropriate. 03Corrective Actions to Address CCW Initiation on Containment Recirculation

a. Inspection Scope

The inspectors reviewed AmerenUE's actions to evaluate EOP deficiencies prior toidentifying the concern with the timing of CCW initiation during ECCS containment recirculation on March 27, 2006. The team considered whether AmerenUE's corrective action program had opportunities to identify and prevent the EOP deficienciesassociated with ECCS recirculation cooling. Specifically, the inspectors reviewed whether AmerenUE's past reviews adequately considered:1)System safety function - classification and prioritization of the problemcommensurate with its safety significance

2)EOP validation - identification of corrective actions which are appropriatelyfocused to correct the problem3)Licensing bases requirements including 10 CFR 50.59 reviews

4)Operability/reportability issues5)Review of operational experience

b. Findings

Failure to Identify and Correct Inadequate EOPsIntroduction: The inspectors identified a Green noncited violation (NCV) of 10 CFRPart 50, Appendix B, Criterion XVI, for the failure to identify and implement appropriate corrective actions for EOP deficiencies associated with CCW cooling to RHR heatexchangers as required to respond to a large-break LOCA.Description: On March 27, 2006, during performance of EOP validations on the plantsimulator, AmerenUE recognized that CCW would not be established to the RHR heatexchangers until after the post-LOCA recirculation phase was automatically initiated. The automatic transfer of each RHR pump suction path from the RWST to the containment recirculation sump would introduce hot water, approximately 265F, toeach RHR heat exchanger prior to CCW flow being established. This could result in theCCW system exceeding it's design basis maximum temperature.Callaway FSAR, Section 9.1.3.2.3 and Table 6.3-8, required that operators initiate CCWto the RHR heat exchangers as the RWST level neared the automatic transfer setpointand prior to the recirculation phase of a LOCA. The hot water, without CCW cooling flow, would heat up the shell side of the RHR heat exchangers to temperatures inexcess of the design bases CCW system temperature and possibly create boiling of theCCW water in the RHR heat exchangers. Procedure E-1 "Loss of Reactor or SecondaryCoolant," as written, had manual operator actions to align cooled CCW water to the

RHR heat exchangers which could not be performed prior to reaching the RWST lo-lo-1level setpoint. This would cause a delay in cooling hot containment recirculation sump water. AmerenUE reviewed the simulator data and initiated a CAR on March 30, 2006.AmerenUE established a plant lineup that provided continuous CCW flow through each

RHR heat exchanger until a permanent resolution could be established. This addressedthe immediate safety concern. The team verified that the failure to meet assumptions in the accident analyses had no impact on peak containment temperatures and pressuresfor the LOCA accident sequences as peak conditions are mostly a function of the containment spray system function and not the time of initiation of CCW into the RHRheat exchanger.

AmerenUE performed heat transfer calculations and EOP validations to ensure that noboiling of the stagnant CCW water would have occurred prior to initiating CCW water in step 2 of ES 1.3. The heat transfer calculations determined that, over the range ofperformance of different operating crews, 8 to 37 seconds of margin existed between the initiation of opening the CCW valves and the onset of boiling. Based on plantinservice testing, the valves were fully opened in 50 to 51 seconds. As a result of these very low margins of time to boil, AmerenUE performed impact studies associated with the collapse of steam bubble formation and steam slug flow analyses for the tube region of the RHR heat exchangers. These studies resulted in approximately 60 seconds toboil off the volume (approximately 12 percent of the total volume) of CCW water above the heat exchanger tubes. The conclusion was there would be no significant water hammer or steam slug flow forces created by the collapse of steam that would have been formed. The team independently reviewed the calculations and supporting documentation for these conclusions. This review included EPRI-NP-6766, "Water Hammer Prevention, Mitigation and Accommodation," and NRC NUREG-CR-6519,"Screening Reactor Steam/Water Piping Systems for Water Hammer."AmerenUE documented the following opportunities to have identified and implementedappropriate corrective action to address the inadequate EOP and safety system designaspect:*Westinghouse Letters SLBE 6-803 and SNP-3346

• Callaway initial FSAR reviews

• Callaway corrective action documents directly associated with the issue (SOS 98-1577, CAR 200106536, CAR 200400017, CAR 200202808, CAR 200503084, and CAR 200507150)*Operational experience associated with SNUPPs plant (Wolf Creek), NRC SafetySystem Engineering inspection finding (05000482/1998-012) *Two EOP change requests and associated 10 CFR 50.59 screening reviewsassociated with Procedures E-1 and ES 1.3*Wolf Creek corrective action document problem identification Report PIR 973483 In addition, the team considered the following documents in their assessment of theoverall corrective action effectiveness to address the EOP deficiencies associated with containment ECCS recirculation and impact on the supporting safety system designaspect.

• NRC Generic Letter 82-33 response and inclusion into TechnicalSpecification (TS) 5.4.1.b *Reviews associated with the initiation of Callaway EOPs versus WestinghouseERGs*NRC Inspection Report 05000483/2004006 and finding 05000483/2003006-02documented critical operator EOP response times being exceeded. The deficiency resulted in critical operator response times taking longer than assumed in the accident analysis. AmerenUE review identified three similar extent of condition reviews but missed the noncompliance with FSAR assumptions described in this finding.*Callaway corrective action documents directly associated with the issue,CAR 200205499 (Callaway EOP validation process had responded to Westinghouse OE regarding EOP steps to enter cold leg recirculation) and CAR 200500564 (FSAR Table 6.3-8 assumed that CCW flow is aligned to the

RHR heat exchangers before the RWST low-low-1 swapover point is reached)Analysis: In accordance with NRC Inspection Manual Chapter 0612, Section 05.01,"Screen for Performance Deficiencies," the team determined that this issue constituted a performance deficiency because AmerenUE repeatedly failed to identify and correct the issues related to a previous NRC finding (05000483/200306-02), CAR 200500564 andother identified significant conditions adverse to quality. Consequently, AmerenUE had operated the plant for years with the potential for boiling in the shell side of each RHRheat exchanger following a postulated a large-break LOCA. Each missed opportunity to correct inadequate emergency operating procedures was a result of ineffective corrective action reviews, a lack of understanding of the accident analysis and licensing bases, and poor interface between AmerenUE's accident analysis and emergency operating procedures writers groups. Phase 1 Screening Logic, Results, and Assumptions In accordance with NRC Inspection Manual Chapter 0612, Section 05.03, "Screen forMinor Issues," the inspectors determined that the finding was more than minor. This finding was associated with the equipment performance, reliability, attribute of themitigating systems cornerstone and was determined to affect the objective of thatcornerstone. Specifically, the finding could have resulted in the loss of CCW following a postulated large-break LOCA.The inspectors evaluated the issue using the Significance Determination Process (SDP)Phase 1 Screening Worksheet for the Initiating Events, Mitigating Systems, and Barriers Cornerstones provided in NRC Inspection Manual Chapter 0609, Appendix A,"Significance Determination of Reactor Inspection Findings for At-Power Situations."

Following a postulated large-break LOCA, the component cooling water system wouldnot have functioned without quick operator action because of boiling in the RHR systemheat exchanger. This represents a loss of the system safety function. Therefore, thescreening indicated that a Phase 2 estimation was required.

Phase 2 Estimation for Internal EventsIn accordance with NRC Inspection Manual Chapter 0609, Appendix A, Attachment 1,"User Guidance for Determining the Significance of Reactor Inspection Findings for At-Power Situations," the inspectors estimated the risk of the subject finding using theRisk-Informed Inspection Notebook for Callaway, Revision 2. The inspectors made the following assumptions:1)The performance deficiency that resulted in inadequate EOPs (failure toestablish CCW cooling to the RHR heat exchangers until after the post-LOCArecirculation phase was automatically initiated) existed from April 15, 1998, untilMarch 30, 2006, when licensee personnel revised the procedure to correct the deficiency. Therefore, this deficiency affected plant risk for an extended period of time and the Phase 2 exposure window of greater than 30 days was used to estimate the risk impact of the deficiency over a 1-year assessment period.2)The failure to establish CCW cooling to the RHR heat exchangers prior to therecirculation phase, following a postulated large-break LOCA, would have resulted in a complete loss of the CCW system without operator intervention.3)Table 2 of the Risk-Informed Inspection Notebook identified that worksheets forall initiating events except the total loss of service water were applicable when a finding affected the CCW system. However, the senior reactor analystdetermined that this performance deficiency only impacted the plant during alarge-break LOCA. Therefore, none of the sequences on any other worksheet were applicable or quantified.4)Table 1 of the Risk-Informed Inspection Notebook identified that the initiatingevent likelihood for a large-break LOCA having an exposure time window of greater than 30 days was 5. The inspectors noted that the performancedeficiency did not increase the likelihood of a large-break LOCA.5)Given Assumption 2, the inspectors adjusted the low pressure recirculationmitigation in the large-break LOCA worksheet from a credit of 3 to a credit of 0 because cooling would have been lost to the sump without CCW.6)Despite not being required before the recirculation phase began, the actions toestablish CCW cooling to the RHR heat exchangers were proceduralized in theemergency operating procedures. Additionally, operating crews being tested in the plant simulator were able to establish CCW prior to the postulated failure of the CCW system as defined by licensee calculations. Therefore, the inspectorsgave operator recovery credit in the worksheet indicating that sufficient time was available to implement the actions, operators had been trained in the procedures that could be implemented entirely from the main control room, and that nospecial equipment was necessary to complete the actions. Therefore, as defined in NRC Inspection Manual Chapter 0609, Appendix A, Attachment 1,Table 4, "Remaining Mitigation Capability Credit," the inspectors gave aRecovery of Failed Train Credit (PCREDIT) of 1.Based on the above assumptions, only Sequence 1 of the large-break LOCA worksheetwas applicable. The resulting sequences are provided in Table 1 below:Table 1Phase 2 Worksheet ResultsInitiatorSequenceInitiating Event LikelihoodMitigating FunctionsResultLarge-Break LOCA15Operator Recovery of CCW6By application of the counting rule, the internal event risk contribution of this finding tothe change in delta core damage frequency (CDF) was of low to moderate risksignificance (White). The approximate value of this frequency (CDFPHASE 2) wascalculated by the senior reactor analyst to be 3.3 x 10

-6.Phase 3 Analysis Assumption 6 made during the Phase 2 estimation process was overly conservativeand did not completely represent the actual probability that operators would fail toestablish CCW cooling to the RHR heat exchangers prior to the time that the CCWsystem would no longer be capable of performing its intended safety function following apostulated large-break LOCA. Therefore, the senior reactor analyst performed a modified Phase 2 estimation to better indicate the risk of the subject performance deficiency.Internal Initiating Events

The analyst utilized the simplified plant analysis Risk H (SPAR-H) method used by IdahoNational Engineering and Environmental Laboratories (INEEL) during the development of the SPAR models and published in NUREG/CR-6883, INEEL/EXT-02-10307, "TheSPAR-H Human Reliability Analysis Method," as an appropriate tool for evaluating theprobability that operators would establish CCW cooling to the RHR heat exchangers in atimely manner following a postulated large-break LOCA.The probability (PRECOVERY) that operators failed to properly perform the EOPs and/orfailed to perform them prior to the failure of the CCW system upon demand wascalculated to be 2.0 x 10

-2. In calculating this failure probability, the analyst assumedthat the nominal action failure rate of 0.001 should be adjusted by multiplying this nominal rate with the following performance shaping factors:Available Time: 10The available time was barely adequate to complete the action. Licenseeoperating crews in the plant simulator took up to 2-1/2 minutes after the switchover to recirculation to establish CCW flow to both RHR heat exchangers. By licensee calculations, this action would have occurred approximately 1 minute prior to boil off of the CCW water in the isolated RHR heat exchangers.Stress: 2Stress under the conditions postulated would be high. Multiple alarms would beinitiated, causing loud, continuous noise in the main control room. Additionally, the operators would readily identify that a large break had occurred in the reactorcoolant system and would understand that the consequences of their actionswould represent a threat to plant safety.All remaining performance shaping factors were considered to be nominal underthe subject conditions.Using this more realistic operator recovery credit, the analyst recalculated the change incore damage frequency as follows:

CDF = CDFPHASE 2 ÷ PCREDIT

• PRECOVERY = 3.3 x 10

-6 ÷ 0.1

• 2.0 x 10

-2= 6.6 x 10

-7By modification of the Phase 2 estimation and in accordance with NRC InspectionManual Chapter 0609, Appendix A, Attachment 1, Phase 3, "Risk Evaluation Using AnyRisk Basis that Departs from the Phase 1 or Phase 2 Process," the analyst determined that the internal event risk contribution of the subject finding to the CDF was of verylow risk significance (Green). The best estimate value of this frequency was calculated by the senior reactor analyst to be 6.6 x 10

-7.External Events The plant-specific SDP worksheets do not currently include initiating events related tofire, flooding, severe weather, seismic, or other external initiating events. In accordance with Manual Chapter 0609, Appendix A, Attachment 1, step 2.5, "Screen for thePotential Risk Contribution Due to External Initiating Events," experience with using the site-specific Risk-Informed Inspection Notebook has indicated that accounting for external initiators could result in increasing the risk significance attributed to an inspection finding by as much as one order of magnitude. Therefore, the analyst assessed the impact of external initiators because the Phase 2 SDP result provided a risk significance estimation of 7 or greater. However, the analyst determined that the likelihood that an external event could result in a large-break LOCA was so small as to be negligible to the quantification of the risk of the subject performance deficiency.

Potential Risk Contribution from Large Early Release Frequency (LERF)In accordance with Manual Chapter 0609, Appendix A, Attachment 1, step 2.6, "Screenfor the Potential Risk Contribution Due to Large Early Release Frequency (LERF)," the analyst determined that the finding needed to be screened for its potential riskcontribution to LERF using Manual Chapter 0609, Appendix H, "Containment IntegritySignificance Determination Process," because the estimated CDF result provided arisk significance estimation of greater than 1 x 10

-7.According to Appendix H, Section 4.1, the subject performance deficiency represented aType A finding because the finding influenced the likelihood of accidents leading to core damage. As documented in Appendix H, Table 5.1, the only accident sequences thatwould lead to LERF for a pressurized water reactor with a large-dry containment like Callaway's would be steam generator tube ruptures and intersystem LOCAs. Theanalyst noted that the only affected core damage sequence involved a large-break LOCA initiator. These sequences do not typically result in containment bypass accidents.Based on the above, and in accordance with Appendix H, the analyst screened out allaccident sequences related to the finding as not significant to LERF.Conclusion The performance deficiency resulted in a finding that was of very low risk significance(Green). The best estimate change in core damage frequency was 6.6 x 10

-7 ,representing the risk related to internal initiators. The change in risk related to external events, as well as the change in LERF, was determined to provide only negligible increase in risk.The inspection team found that this finding has crosscutting implications in the problemidentification and resolution performance area. AmerenUE's inadequate evaluations resulted in not correcting a licensing basis safety issue.Enforcement: Title 10 of the Code of Federal Regulations, Part 50, Appendix B,Criterion XVI, "Corrective Action," required that conditions adverse to quality are promptly identified and corrected. Further, the requirement states that, in the case ofsignificant conditions adverse to quality, measures shall be taken to ensure that the cause of the condition is determined and corrective action taken to preclude repetition. Contrary to the above, the corrective actions taken for a previous NRC finding(05000483/200306-02) and CAR 200500564, as well as other identified opportunities to correct the deficient EOP, were a significant condition adverse to quality where the measures taken to ensure that the cause of the condition is determined and corrective action taken to preclude repetition were not effective. This finding is a noncited violation (NCV 05000483/2006011-01) consistent with Section VI.A of the NRC EnforcementPolicy. AmerenUE entered this issue into its corrective action program as CAR 200602565.

4Adequacy of Planned or Completed Corrective Actions

a. Inspection Scope

The team reviewed AmerenUE's immediate corrective actions needed to ensure thefunction of the RHR heat exchangers in a large-break LOCA event and those actions toprevent recurrence of a failure of the EOPs to meet licensing bases accident analyses. The corrective actions implemented by AmerenUE involved three sets of actions. Thefirst set of actions was to establish continuous CCW flow through the RHR heatexchangers. This would preserve the FSAR described licensing basis and ensure that the hot containment recirculation sump water does not boil the CCW in the RHR heatexchangers.The second set of actions was to have a root cause team formed to determine theextent of condition of missed licensing bases related requirements as they apply to the EOPs. This team was to provide immediate evaluation and communication of requirements not clearly met and provide input to the root cause determination forCAR 200602565.The third set of actions was to ensure that plant procedures and planned maintenanceassociated with the CCW system were reviewed to ensure compliance with TSs andother aspects of the current licensing bases. The inspectors independently reviewed the adequacy of AmerenUE's initial and plannedcorrective actions.

.1 Upon discovery of the EOP deficiency, AmerenUE placed the plant in a configurationthat ensured adequate ECCS flow to the

RHR heat exchangers during a large-breakLOCA and for containment ECCS recirculation. AmerenUE's root cause team hadsufficient resources allocated and performed a thorough extent of condition review of licensing bases requirements pertaining to the EOPs. Three minor licensing bases conflicts with the EOPs were identified and actions were assigned to immediately resolve the conflicts. These are discussed in Section 04.3 of this report. Initial reviews were performed on plant procedures and planned maintenance but they did not effectively prevent a potential CCW pump runout scenario associated with a LOCA with a loss of offsite power event as discussed in Section 04.2.

.2 Corrective Action to Establish Continuous CCW Flow to RHR Results in Possible CCWRunout Conditions

Introduction:

On April 12, 2006, the inspectors identified a Green, noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, due to AmerenUE's failure to implement and adequately communicate the corrective actions identified to permanently establish CCW to RHR heat exchangers on a large-break LOCA. This failure resulted in potential CCWpump runout conditions for a LOCA with loss of offsite power.

Description:

On March 30, 2006, CAR 200602565 (CCW alignment for large-breakLOCAs), operability determination, operator night orders, and the acceptance criteria forEngineering Procedure ETP-EG-0002, "Component Cooling Water System Flow Verification," Revision 4, had identified that CCW train flow should not be run above 7250 klbm/hr due to potential pump runout concerns. On April 11, 2006, CCW Pump A was started to augment Train A CCW Pump C flow to address a Train A charging pump oil cooler low flow alarm. The operating shift then aligned CCW flow to the spent fuel pool to balance Train A CCW system flows. The combined flow for the two CCWpumps was in excess of 8400 klbm/hr. On April 12, 2006, the inspectors informed AmerenUE operating personnel that this alignment was in conflict with the operabilitydetermination and may not ensure CCW pump operability in a postulated LOCA with aloss of offsite power. The design of the sequencer for the essential Train A 4160 volt bus would only start CCW Pump A. Pump A would experience a runout condition, causing damage to the pump and possible pump failure. On failure of Pump A, Pump C would automatically start and subject Pump C to the same conditions. Pump C's reliable operation would then be challenged. The 10 CFR 50.59 safety evaluation screening for changes to ProcedureOTN-EG-00001, "Component Cooling Water System," Revision 25, on April 7, 2006, discussed potential runout system alignment but did not result in a change to theannunciator response procedure describing what maximum flow for a single CCW pump would be. CAR 200602995 was written to address the inspectors' concerns. This CAR identified that on two occasions, each less than 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> duration on April 11 and 12, 2006, flow in Train A was in excess of 7250 klbm/hr flow. CAR 200602995 also stated that TS 3.7.7, Limiting Condition for Operation, Action A, for an inoperable CCW pumpwas missed but the 72-hour allowed outage time was not exceeded.

Analysis:

The team determined that this finding constituted a performance deficiency. AmerenUE's corrective action for the EOP deficiency resulted in subsequent plant configurations that did not ensure that a single CCW pump would not be subjected topump runout conditions. This issue was more than minor because it affected the Mitigating Systems cornerstone objective of equipment reliability and capability ofsystems that respond to initiating events in that established CCW pump operating flowconditions would not have ensured an operable Train A CCW pump. Using the NRC Inspection Manual Chapter 0609, Phase 1 Screening Worksheet, the finding was determined to be of very low safety significance since it did not result in a loss of safety function for a single train for greater than its TS allowed outage time. The inspection team found that this finding has crosscutting implications in the problemidentification and resolution performance area. The inadequate CCW system flow wasa result of inadequate corrective action described in CAR 200602565.

Enforcement:

Corrective actions for operability of the RHR system prescribed inCAR 200602565 failed to ensure that each train of the CCW system was available toprovide RHR heat exchanger cooling. Title 10 of the Code of Federal Regulations,Part 50, Appendix B, Criterion XVI, "Corrective Action," required that conditions adverse to quality are promptly identified and corrected. Contrary to 10 CFR Part 50, Appendix B, Criterion XVI, on both April 11 and 12, 2006, AmerenUE corrective actionslead to improper CCW system conditions that challenged the CCW train to RHR heatexchanger safety function. This finding is an NCV (NCV 05000483/2006011-02, Inadequate Corrective ActionsResult in Possible CCW Runout Conditions) consistent with Section VI.A of the NRC Enforcement Policy. AmerenUE entered this issue into its corrective action program as CAR 200602995.

.3 Comprehensiveness of the Licensee's Determination of the Extent of Condition

a. Inspection Scope

Through interviews and documentation reviews, the team evaluated thecomprehensiveness of AmerenUE's extent of condition review for the failure to implement FSAR design bases requirements. Specifically, the team assessed whether licensee personnel had adequately reviewed procedures and engineering issues associated with the newly established CCW to RHR heat exchangers alignment. Alsothe inspectors independently reviewed the FSAR and EOPs to assess whether other associated licensing bases requirements were met.The inspectors reviewed the licensing documents identified by AmerenUE that were notdirectly met by the current revision of the Callaway EOPs. These were:*FSAR, Chapter 6, Engineering Safety Features, Section 6.2.2, stated thatcontainment spray cannot be terminated until completion of the injection phase. Procedure E-1, step 7, allowed both containment spray pumps to be turned off, prior to completing the injection phase, provided containment pressure had been reduced below 5.5 psig. Having containment pressure less than 5.5 psig ensured that no adverse impact on dose consequences would occur. *FSAR, Chapter 9, Auxiliary Systems, Section 9.2.1.2.2.3, stated that auxiliaryfeedwater low suction pressure signal opened the essential service water systemisolation valves to ensure essential service water supply to the auxiliaryfeedwater system. The FSAR provided only a percent margin to the ultimateheat sink total volume for essential service water system use and not a specifictime requirement for the operators to secure the auxiliary feedwater use. Thelicense calculated the time based on the percent margin and determined the operators have 65 minutes to complete the task. This was validated as having sufficient time to perform the task to realign the auxiliary feedwater systemsuction away from the ultimate heat sink.

• FSAR, Chapter 15, Safety Analysis, Section 15.6.3.2.2, and Table 15.6.1 statedthat, following a steam generator rupture accident, a cooldown to RHR conditions using the intact steam generator atmospheric steam dumps must be initiated at approximately 60 minutes from the start of the event. Since this occurs after the leak from the ruptured steam generator tubes is stopped, no additional radioactivity is released. This requirement had no logic basesdocumented and is being reviewed by AmerenUE.

b. Observations and Findings

The inspectors found that the corrective actions, in response to normal and off-normalprocedures associated with the CCW system were generally being correctly applied tothe Callaway Plant. CCW annunciator response procedures associated with CCW loads were not appropriately addressed as discussed in the finding in Section 04.02.

The inspectors also found AmerenUE had not fully evaluated smaller and medium break LOCA scenarios or possible CCW pump loss of net positive suction head. AmerenUE provided data and calculations to show that these cases with elevated initial CCWtemperatures still resulted in no significant impact.

.4 Evaluation of Licensee's Initial Root Cause Determination

a. Inspection Scope

The team reviewed AmerenUE's preliminary root cause determination of the failure toimplement FSAR design bases requirements for independence, completeness, and accuracy.

b. Observations and Findings

The inspectors found AmerenUE's direct cause determination to be accurate. Thisinitial licensee report, however, did not emphasize why so many opportunities to identify the issue were missed. Organizational interface ineffectiveness and an inadequate corrective action program allowed AmerenUE's organization to repeatedly miss opportunities to understand an unanalyzed safety issue. Specifically the bases and sequence of FSAR requirements were not researched when questions arose. This lead to inaccurate and incomplete initial reviews by CAR lead responders and prevented licensee management from becoming appropriately engaged. The team noted that AmerenUE had identified three preliminary causes of not havingestablished procedures that would ensure RHR heat exchanger cooling in a LOCA priorto automatic introduction of hot containment recirculation sump water. These were discussed in significant condition adverse to quality CAR 200602565. *The 1984 EOP revision did not match requirements provided in the FSARwording. The FSAR wording had remained unchanged since initial issue.*AmerenUE had at least three opportunities to identify and correct the problem. Reviews of corrective action documents and the emergency procedures were narrowly focused.*Ineffective communication between Callaway and Wolf Creek contributed to thenarrow focus of the reviews and corrective action evaluations.

.5 Discussion of the Potential Impact Associated with Boiling in the RHR Heat Exc

hangerCCW SideThe team reviewed the engineering calculations to evaluate whether the safety functionof heat removal from the containment sump following a LOCA could be achieved.

Specifically, the concern related to aligning CCW to the shell side of the RHR heatexchangers after the RHR suction path was realigned to the containment sump fromthe RWST. Because the containment recirculation sump water would be at a saturation temperature of approximately 265F, boiling of CCW on the shell side of the RHR heatexchangers would occur with no shell side fluid flow. AmerenUE performed four separate calculations that determined:

(1) the temperaturerise in the shell side of the heat exchangers,

(2) the heat exchanger voiding rate,

(3) the magnitude and impact of any resulting water hammer, and

(4) the CCW inlet impingement plate response to a water hammer. The actions to ensure CCW flow is delivered to the shell side of the heat exchangerswere specified in the EOPs; consequently, the delivery of CCW to the shell side of the heat exchangers was a function of operator time. The team determined that AmerenUE used appropriate design inputs, calculationmethodologies, and conservative assumptions. The calculations determined that the time required to boil and subsequently void the shell side of the heat exchangers would have been prevented by prior operator action. Further, although it was anticipated thatCCW would be delivered to the heat exchangers prior to voiding of the heat exchangers shell, AmerenUE demonstrated the collapse of a steam bubble that would void the space above the heat exchanger tubes would create a water hammer of small magnitude. The maximum force predicted for this case was equivalent to approximately 115 psig.

.6 Generic Implications

a. Inspection Scope

The team reviewed AmerenUE's design bases to determine whether generic issuesrelated to the design and operating practices existed with other Callaway systems orother nuclear plants.

b. Observations and Findings

The team, with assistance from AmerenUE, and the NRC's Office of Nuclear ReactorRegulation, determined that other Westinghouse primary water reactor plants withoutautomatic CCW initiation may not be in full compliance with their licensing bases. This issue is being reviewed by the NRC.

OA6Meetings, Including ExitOn April 14, 2006, the team presented the status of the inspection, to date, toMr. Tod Moser, Manager, Plant Engineering. On May 2, 2006, the team leader conducted an exit meeting with Mr. Tim Herrmann,Vice President, Engineering, and other members of his staff.On June 26, 2006, the team leader conducted a supplemental exit meeting withMr. Tim Herrmann, Vice President, Engineering, and other members of his staff.While proprietary information was reviewed, no proprietary information is being retainedor is included in this report. 4OA7Licensee-Identified ViolationsThe following violation of very low safety significance (Green) was identified byAmerenUE and is a violation of NRC requirements which meet the criteria of Section VIof the NRC Enforcement Policy, NUREG-1600, for being dispositioned as NCVs. TS 5.4.1.b, "EOP Program," required that requirements of NUREG 0737 as described inGeneric Letter 82-33 be adhered to, ensuring that applicable accidental analysis licensing bases are correctly translated into emergency procedures. Contrary to this, AmerenUE had not translated the FSAR described licensing basis into EOPs.

Callaway's requirements to initiate CCW flow to the RHR heat exchangers prior to theopening of the containment recirculation sump valves on a postulated large-break LOCAwere not met. Specifically FSAR, Section 9.1.3.2.3, and Table 6.3-8 required thatoperators initiate CCW to the RHR heat exchangers as the RWST level neared theautomatic transfer setpoint prior to the recirculation phase of a LOCA.The particular function to prevent boiling in the RHR heat exchangers on a large-breakLOCA required subsequent analysis to ensure the RHR and CCW functions were notunrecoverable. Through calculations, AmerenUE was able to demonstrate that using the actual EOP step location, operator action occurred in time to prevent boiling.This issue is more than minor because it was similar to Example 3.I of Appendix E ofManual Chapter 0612. It was necessary for AmerenUE to perform a calculation to determine whether the existing EOPs were acceptable. Because there was availablemargin in the time to boil and time to RHR heat exchanger tube uncovery calculations,this issue was confirmed not to involve a loss of function of t he system in accordancewith Part 9900, Technical Guidance, Operability Determination Process for Operabilityand Functional Assessment. Therefore, this issue screens as Green during Phase 1 of the SDP as described in Manual Chapter 0609, Appendix A, Attachment 1.This issue was identified in AmerenUE's corrective action program as CAR 200602565.

ATTACHMENTS:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee Personnel

D. Fuller, System Engineer

B. Huhmann, Supervising Engineer, Nuclear Engineering Systems, Mechanical

M. Jennings, Operating Supervisor

S. Maglio, Superintendent, Systems Engineering

J. Milligan, Shift Manager, Operations

K. Mills, Supervising Engineer, Regional Regulatory Affairs/Safety Analysis

T. Moser, Manager, Plant Engineering

S. Petzel, Engineer, Regional Regulatory Affairs

T. Herrmann, Vice President, Engineering

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened and

Closed

05000483/FIN-2006011-01NCVFailure to Recognize and Correct Inadequate EmergencyProcedures (Section 03)

05000483/FIN-2006011-02NCVInadequate Corrective Actions Result in Possible CCWRunout Conditions (Section 04.2)

LIST OF DOCUMENTS REVIEWED

CalculationsNumberTitleRevisionEJ-M 18(Wolf Creek)RHR Pump Recirculation Operation versus Time ofInitiation of CCW Flow to RHR Heat Exchangers 1C-4176-00-01Callaway RHR Heat Exchanger Shell Side Temperature Rise 0C-4176-00-02Callaway RHR Heat Exchanger Transient Voiding Rate0C-4176-00-03Likelihood and Magnitude of Water Hammers in theCallaway RHR Heat Exc hanger 0M-EG-05,ADD 2Calculate the NPSH Available to the CCW Pumps With theSurge Tank Empty

0BN-16Maximum RWST Transfer Volumes and Swapover Times0

A1-2Attachment 1Callaway Action Requests19910074619860054199801577200106536200202808200205499200400017200500564200503084200507150

200602565200602908200602992200602995DrawingsNumberTitleRevisionFSARFigure 9.2-3, Sheet 1Piping and Instrumentation Diagram - CCW SystemNAFSARFigure 9.2-3, Sheet 2Piping and Instrumentation Diagram - CCW SystemNAM-23EG01(Q)Piping Isometric CCW System, Auxiliary Building, Train A6M-23EG03(Q)Piping Isometric CCW System, Auxiliary Building, Train B7

M-23EG04(Q)Piping Isometric CCW System, Auxiliary Building, Train B3

M-23EG05(Q)Piping Isometric CCW System, to Fuel Building, Train B2

5736Vertical Residual Heat Exchanger Outline Drawing5

5739Vertical Residual Heat Exchanger Details3

5740Vertical Residual Heat Exchanger Details4Miscellaneous DocumentsNumberTitleRevision/DateMemorandum-4176-00-01Summary of Screening Criteria for the Evaluation ofSteam Water Hammer at Power Plants April 12, 200650.59 Screen

RFR 23374 Evaluate the Use of Gothic Software38204SafetyEvaluationKewaunee Nuclear Power Plant - Review for KewauneeReload Safety Evaluation Methods Topical Report WPSRSEM-NPRevision 3,September 10, 2001

Miscellaneous

DocumentsNumberTitleRevision/DateA1-3Attachment 1InspectionReport 05000

2/97-201Wolf Creek Generating Station Design Inspection, February 23, 1998EPRI

NP-6766Water Hammer Prevention, Mitigation andAccommodation July 1992PRAEvaluation Request

06-269Risk Assessment for CCW Flow to RHR HeatExchanger Issue

38819NUREG/CR-6519 Screening Reactor Steam/Water Piping Systems forWater Hammer September

1997PIR 973483Wolf Creek Performance Improvement RequestDescribing USAR and EMG

ES-12 Conflicts

35731DraftRevision 2Event & Causal Factors Chart,

CAR 200602565 CCWFlow Requirements to RHR Heat Exchangers NotMeeting FSAR

38818Formal SafetyEvaluation for

RFR 19025Callaway FSE for Evaluating FSAR Chapters 6 and 15as a Result of Calculation

BN-16, Revision 0 Which Determined the Maximum Times for Swapover of emergency core cooling system and CS Pumps fromInjection Phase to Recirculation Phase for 5 Cases

1998FSARSection 9.1Fuel Pool Cooling System5/97FSARSection 9.2Component Cooling System5/97FSARTable 6.3-8Sequence of Changeover Operation from Injection toRecirculation

5/97Night Order Callaway Night Order, CCW Alignment Requirementsbased on CAR 200602565

38813

Miscellaneous

DocumentsNumberTitleRevision/DateA1-4Attachment 1Night Order Callaway Night Order, CCW Alignment Requirementsbased on CAR 200602565

38806Computer History PrintoutsTypePeriodRHR Pump BCurrent19 minutes on 2/11/2004RWSTTemperature3 years beginning 1/1/2001CCW Flow &Valve Operation50 seconds of operation to show flow versus valve positionProceduresNumberRevisionSubjectE-15Loss of Reactor or Secondary Coolant

E-16Loss of Reactor or Secondary Coolant

OTA-RK-000201Annunciator Response Window 52B for CCW Pump Aor C Pressure LowOTA-RK-000201Annunciator Response Window 54B for CCW Pump Bor D Pressure LowOTN-EG-0000125CCW System

OTN-EG-0000126CCW System

OTO-BB-0000223Rcp Off-Normal

OTO-EG-000019CCW System Malfunction

EOP E-11B2Loss of Reactor or Secondary Coolant ProceduresNumberRevisionSubjectA1-5Attachment 1ES-1.3 ERG(background) NAWestinghouse Owner Group ERG Background for

ES-1.3ES-1.3 ERGNAWestinghouse Owner Group ERG for

ES-1.3ES-1.30Transfer to Cold Leg Recirculation

ES-1.35Transfer to Cold Leg Recirculation

ES-1.36Transfer to Cold Leg RecirculationACRONYMSCARCallaway Action RequestCCWcomponent cooling water

CDFdelta core damage frequency

CFRCode of Federal RegulationsECCSemergency core cooling system

EOPemergency operating procedure

ERGemergency response guideline

ESWessential service water

FSARFinal Safety Analysis Report

INEELIdaho National Engineering and Environmental Laboratories

LERFLarge Early Release Frequency

LOCAloss-of-coolant accident

NCVnoncited violation

RHRresidual heat removal

RWSTrefueling water storage tank

SDPsignificance determination process

SNUPPSStandardized Nuclear Unit Power Plant System

SPARsimplified plant analysis risk

TSTechnical Specification

A2-1Attachment 2TIMELINE DESCRIBING CCW TO RHR HEAT EXCHANGERS PROBLEMOctober 14, 1976Westinghouse issued letter SLBE 6-803 recommending automatic CCWinitiation to the RHR heat exchangers prior to the swapover point.

CallawayPlant, owned by Union Electric Company, was part of the SNUPPS group.

SNUPPS felt that manual action was acceptable as operators are expected to be trained and felt that automatic action would result in additional unnecessary surveillances.

The letter stated that automatic function couldbe backfitted by the NRC at the FSAR st age.May 29, 1980Westinghouse issued SNUPPS Letter

SNP-3346.

It stated that CCW mustbe aligned to the

RHR heat exchangers prior to swapover in the recirculationmode.1980 to 1982Callaway FSAR issued.

In two locations it was stated that the CCW initiationmust be prior to recirculation mode swapover. (Section 9.1.3.2.3 and Table 6.3-8.December 1982Generic Letter 82-33, Section 7.1, established requirements for licensees toreanalyze transients and accidents and prepare technical guidelines.

These analyses were to identify critical operator tasks and were to be the bases forupgraded EOPs.

AmerenUE's EOPs were to provide a procedures generation package, including a program for validating EOPs.

Callaway had several opportunities to validate that CCW is established to

RHR heatexchangers prior to the transfer to the cold leg recirculation phase.June 7, 1905Callaway EOPs were initiated and required, only in Procedure ES 1.3, thatthe CCW to the RHR heat exchangers be initiated.

This was contrary to theFSAR sections requiring prior initiation.

The Westinghouse ERG, for the

Procedure

E-1 "Loss of Reactor or Secondary Coolant," response, also did not have a step to open the CCW inlet valves to the

RHR heat exchangers. The ERG clearly identified, in the basis to Procedure ES 1.3, step 2, that thestep to align CCW was a "verify" step that assumed previous attempts to initiate CCW flow to the RHR heat exchangers.

April 15, 1998Callaway initiated a corrective action document, SOS (previous CARname) 98-1577, noting that the

NRC had issued Wolf Creek a 50.59 violationhighlighting that late initiation of the CCW to the RHR heat exchangers couldresult in 270F recirculation sump water being introduced to the

RHR heatexchangers.

Without cooling, this could result in exceeding the design temperature of the CCW system and cause boiling to occur. (Wolf CreekPIR 973483).

A2-2Attachment 2May 5, 1998 Callaway recognized that Procedure E-1 did not have a step prior to entry toProcedure ES 1.3 and added a step to open the CCW inlet valve to each

RHR heat exchanger.

However the change was made as a temporarychange notice (TCN 98-0427) and the 50.59 screening question addressing whether the change was to a procedure as described in the FSAR was answered "NO."

CAR 200205499Callaway

CAR 200205499 stated that the Callaway EOP procedurevalidation process had validated OE14159 in regard to EOP steps to enter cold leg recirculation.

The CAR stated that Callaway had no interim configuration issues and that FSAR 6.3.2 commitments for timing actions during the swapover were met.January 2, 2004Callaway

CAR 200400017 noted that Wolf Creek nuclear power plantrequired that the CCW inlets to each

RHR heat exchanger be opened in90 seconds or less following the automatic sump swapover.

The CAR

initiator asked if Callaway had any similar concerns and the accident analysis group replied "No." January 27, 2005CAR

200500564 stated that FSAR Table 6.3-8 assumed that CCW flow isaligned to the

RHR heat exchangers before RWST low-low-1 swapover pointis reached.

The initiator questioned why the RWST outflow analysis did not explicitly include times to align CCW flow to the

RHR heat exchangers.

Theresponse to the CAR was that steps not directly associated with the swapover were not appropriate. March 20, 2006Licensed operator retraining to perform EOP validations questioned whetherCCW initiation to RHR heat exchangers was time critical on a large-breakLOCA.March 30, 2006CAR

200602565 was initiated describing the discovery of the simulator EOPvalidation.

The Operations department placed the

RHR heat exchangerCCW alignment in a safe condition.April 7, 2006Operability determination and 50.59 screening for changes toProcedure

OTN-EG-00001 (CCW system) descri be the extent of conditionfor the current CCW to RHR heat exchangers alignment.

Each describes amaximum 7250 klbm/hr flow rate for a single CCW train due to pump runout concerns during a large-break LOCA scenario with loss of offsite power to an engineered safety features bus.

Concern is that only a single CCW pump will be s equenced onto the bus with a CCW system alignment for two pumpoperation.

April 10, 2006Callaway forms root cause and engineering teams to address the EOP/CCWissue.

A2-3Attachment 2April 11, 2006NRC charters a Special Inspection Team to respond to the discovery thatCCW would not be initiated to the RHR heat exchangers prior to autoswapover to the recirculation phase on a large-break

LOCA.April 11, 2006Low flow on the Train A charging pump oil cooler occurs.

Shift operatorsreview annunciator response Procedure

OTA-RK-00020 guidance, start a second Train A CCW pump, and increase Train A flow to 8400 klbm/hr.

April 12, 2006NRC inspector questions the conflict with the operability determination andthe actions by the operating crew in response to the 4/11/06 low CCW flow on the charging pump.

April 13, 2006CAR

200602995 describes two times when the 7250 klb m/hr CCW pumplimit was exceeded.

One was approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> on 4/11/06 and again for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> on 4/12/06.

April 14, 2006Initial onsite inspection completed by NRC team

A3-1Attachment 3April 10, 2006MEMORANDUM TO:David Dumbacher, Resident Inspector, Callaway StationProject Branch B, Division of Reactor ProjectsGreg Pick, Senior Reactor InspectorEngineering Branch 2, Division of Reactor SafetyFROM:Arthur T. Howell III, Director, Division of Reactor Projects /RA/SUBJECT: SPECIAL INSPECTION CHARTER TO EVALUATE CALLAWAY PLANTCOMPONENT COOLING WATER INITIATION TO THE RESIDUAL

HEAT REMOVAL HEAT EXCHANGERS DURING THE INITIAL POST-

LOCA RECIRCULATION PH

ASE A Special Inspection Team is being chartered in response to the discovery that componentcooling water (CCW) would not be established to the residual heat removal (RHR) heatexchangers until after the postloss of coolant accident (LOCA) recirculation phase was initiated. This could lead to a failure of the CCW system and a loss of safety injection and other essentialloads (such as spent fuel pool cooling).

The licensee implemented prompt actions to establish flow to the RHR heat exchangers to restore the safety systems and essential loads to anoperable status.

You are hereby designated as the Special Inspection Team members.

Mr.

Dumbacher is designated as the team leader.A.BasisOn March 30, 2006, the Callaway Plant reported (CAR 200602565) that, during asimulator exercise on March 20, 2006, an operator raised a concern regarding the timeliness of initiation of the CCW flow to the RHR heat exchangers during post-LOCA(large break) recirculation from the containment safety injection sumps.

The licensee identified that the sequence of establishing CCW flow, and the delays in its initiation because of the sequence in the emergency operating procedures, could result in the potential to exceed the CCW design temperature during a large LOCA when containment recirculation is first initiated.

The licensee found during a simulator exercise that CCW flow to the

RHR heat exchangers was not initiated until 4-6 minutesafter containment recirculation flow was first established through the

RHR heatexchangers.

The Final Safety Analysis Report describes that CCW is placed in service prior to refueling water storage tank lo-lo 1 level being reached and the swapover occurring.

The licensee had previously established, through the emergency operating Multiple Addressees- 2 -A3-2Attachment 3procedures, that CCW would be initiated through the

RHR heat exchanger following theswapover to containment recirculation.

The licensee's identification that the CCWsystem may not actually be aligned in sufficient time to ensure adequate cooling of the

RHR heat exchanger resulted in the licensee questioning their ability to meet designbasis requirements.

The license's immediate corrective action included aligning and running the CCW system continuously to ensure that adequate cooling water wasavailable to the RHR heat exchanger in the event of a design basis LOCA event.

This Special Inspection Team is chartered to compare the as-found conditions to thelicensing basis for containment recirculation; determine if there are generic safety implications associated with the timing of CCW initiation post-LOCA through the RHR

heat exchangers; review the identification, evaluation, and determination whether theCCW system and associated safety injection systems were inoperable for thepostrecirculation phase; review the licensee's compensatory measures following discovery of the condition; and review the licensee's calculations regarding the impact of the timing of CCW initiation to the RHR heat exchangers as provided in their emergencyoperating procedures. B.ScopeThe team is expected to address the following:

1.Develop a complete sequence of events related to the discovery of the CCWtiming concern for post-LOCA safety injection and the followup actions taken by the licensee.

2.Compare operating experience involving post-LOCA emergency core coolingsystem (ECCS) cooling requirements to actions implemented at the CallawayPlant.

Review prior opportunities to have addressed EOP and/or design considerations associated with ECCS recirculation cooling requirements, including the effectiveness of those actions.

Determine if there are any generic issues related to the design and operating practices associated with post-LOCA

recirculation and ECCS cooling.

Promptly communicate any potential generic issues to regional management.3.Review the extent of condition determination for this condition and whether thelicensee's actions are comprehensive.

This should include potential for other

EOP validation issues as well as potential ECCS recirculation timing issues.

4.Review the licensee's determination of the cause of any procedural designdeficiencies and/or operating practices that allowed the potential for CCWsystem design temperature to be exceeded.

Independently verify keyassumptions and facts.

If available, determine if the licensee's root cause analysis and corrective actions have addressed the extent of condition for problems with CCW cooling to the safety systems.

Multiple Addressees- 3 -A3-3Attachment 35.Determine if the Technical Specifications were met for the ECCS and CCWsystems following the implementation of compensatory measures.6.Determine if the supporting analyses for the licensee's compensatory measures were made in accordance with 10

CFR 50.59.7.Review the calculations the licensee is developing to evaluate the CCW initiationsequence for post-LOCA ECCS and CCW operability. 8.Collect data necessary to support a risk analysis.

Specifically obtain informationassociated with the degree to which the ECCS and CCW systems would beaffected during post-LOCA recirculation, the break sizes that are affected, thecontainment response, the ability to recover failed pumps and other components,and the dominant accident sequences.C.GuidanceInspection Procedure 93812, "Special Inspection," provides additional guidance to beused by the Special Inspection Team.

Your duties will be as described in InspectionProcedure 93812.

The inspection should emphasize fact-finding in its review of the circumstances surrounding the event.

It is not the responsibility of the team to examinethe regulatory process.

Safety concerns identified that are not directly related to theevent should be reported to the Region IV office for appropriate action.The Team will report to the site, conduct an entrance, and begin inspection no later than April 11, 2006.

While on site, you will provide daily status briefings to Region IVmanagement, who will coordinate with the Office of Nuclear Reactor Regulation, toensure that all other parties are kept informed.

A report documenting the results of theinspection should be issued within 30 days of the completion of the inspection.This Charter may be modified should the team develop significant new information thatwarrants review.

Should you have any questions concerning this Charter, contact me at

(817) 860-8248. cc via E-mail:B. MallettM. Peck

T. GwynnR. Kopriva

A. VegelD. Overland

D. ChamberlainW. Jones

R. CanianoS. O'Connor

L. SmithD. Terao

J. ClarkJ. Donohew

V. DricksM. King

W. Maier Multiple Addressees- 4 -A3-4Attachment 3SUNSI Review Completed: _WBJ_____ADAMS:

Yes G

No

Initials: _WBJ

Publicly Available

G

Non-Publicly Available

G

Sensitive

Non-SensitiveS:\DRP\DRPDIR\CHARTER\Callaway April 2006.wpdML061010217RIV:C:DRP/BDD:DRPD:DRSD:DRPWBJones;df:laoAVegelDDChamberlainATHowell /RA//RA//RA/ /RA/

4/10/064/10/064/10/064/10/06OFFICIAL RECORD COPY T=Telephone

E=E-mail

F=Fax