IR 05000293/2011012

IR 05000293-11-012, on 05/16/2011 - 07/20/2011, Pilgrim Nuclear Power Station, Inspection Procedure 93812, Special Inspection

ML112440100
Person / Time
Site:	Pilgrim
Issue date:	09/01/2011
From:	Christopher Miller Division of Reactor Safety I
To:	Smith R G Entergy Nuclear Operations
References
EA-11-174 IR-11-012
Download: ML112440100 (37)

Text

,2+**ti September 1, 2011EA-11-174Mr. Robert G. SmithSite Vice PresidentEntergy Nuclear Operations, Inc.Pilgrim Nuclear Power Station600 Rocky Hill RoadPlymouth, MA 02360-5508PILGRIM NUCLEAR POWER STATION . NRC SPECIAL INSPECTIONREPORT 05000293/2011012: PRELIMINARY WHITE FINDING

Dear Mr. Smith:

On July 2Q,2011, the U.S. Nuclear Regulatory Commission (NRC) completed a SpecialInspection at your Pilgrim Nuclear Power Station (PNPS). The inspection was conducted inresponse to the May 10,2011, reactor scram event that occurred due to an unrecognizedsubcriticality and subsequent unrecognized return to criticality. The NRC's initial evaluation ofthis event satisfied the criteria in NRC Inspection Manual Chapter (lMC) 0309, "Reactivelnspection Decision Basis for Reactors," for conducting a Special Inspection. The SpecialInspection Team (SlT) Charter (Attachment 2 of the enclosed report) provides the basis andadditional details concerning the scope of the inspection. The enclosed inspection reportdocuments the inspection results, which were discussed at the exit meeting on July 20,2011,with you and other members of your staff.The inspection team examined activities conducted under your license as they relate to safetyand compliance with Commission rules and regulations and with the conditions of your license.The inspection team reviewed selected procedures and records, observed activities, andinterviewed personnel. In particular, the inspection team reviewed event evaluations, causalinvestigations, relevant performance history, and extent of condition to assess the significanceand potential consequences of issues related to the May 10 event.The inspection team concluded that the plant operated within acceptable power limits, and noequipment malfunctioned during the power transient and subsequent reactor scram.Nonetheless, the inspection team identified several issues related to human performance andcompliance with conduct of operations and reactivity control standards and procedures thatcontributed to the event. The enclosed chronology (Attachment 3 of the enclosed report)provides additional details regarding the sequence of events.

R, Smith 2This report documents one finding that, using the reactor safety Significance DeterminationProcess (SDP), has preliminarily been determined to be White, or of low to moderate safetysignificance. The finding involves the failure of Pilgrim personnel to implement conduct ofoperations and reactivity control standards and procedures during a reactor startup, whichcontributed to an unrecognized subcriticality followed by an unrecognized return to criticality andsubseq uent reactor scram.This finding was assessed using NRC IMC 0609, Appendix M, "Significance DeterminationProcess Using Qualitative Criteria," because probabilistic risk assessment tools were not wellsuited to evaluate the multiple human performance errors associated with this issue.Preliminarily, the NRC has determined this finding to be of low to moderate safety significancebased on a qualitative assessment. There was no significant impact on the plant following thetransient because the event itself did not result in power exceeding license limits or fueldamage. Additionally, interim corrective actions were taken, which included removing thePilgrim control room personnel involved in the event from operational duties pendingremediation, providing additional training for operators not involved with the event, and providingincreased management oversight presence in the Pilgrim control room while long termcorrective actions were developed.The finding involved one apparent violation (AV) of NRC requirements regarding TechnicalSpecification 5.4, "Procedures," that is being considered for escalated enforcement action inaccordance with the NRC's Enforcement Policy, which can be found on NRC's website athttp://www. nrc.qov/read inq-rom/doc-col lections/enforcemenU.ln accordance with NRC IMC 0609, we will complete our evaluation using the best availableinformation and issue our final determination of safety significance within 90 days of the date ofthis letter. The SDP encourages an open dialogue between the NRC staff and the licensee;however, the dialogue should not impact the timeliness of the staff's final determination. Beforewe make a final decision on this matter, we are providing you with an opportunity to (1) attend aRegulatory Conference where you can present to the NRC your perspective on the facts andassumptions the NRC used to arrive at the finding and assess its significance, or (2) submityour position on the finding to the NRC in writing. lf you request a Regulatory Conference, itshould be held within 30 days of your response to this letter, and we encourage you to submitsupporting documentation at least one week prior to the conference in an effort to make theconference more efficient and effective. lf a Regulatory Conference is held, it will be open forpublic observation. lf you decide to submit only a written response, such submittal should besent to the NRC within 30 days of your receipt of this letter. lf you decline to request aRegulatory Conference or submit a written response, you relinquish your right to appeal the finalSDP determination, in that by not doing either, you fail to meet the appeal requirements statedin the Prerequisite and Limitation Sections of Attachment 2 of IMC 0609.Please contact Mr. Donald E. Jackson by telephone at (610) 337-5306 within 10 days from theissue date of this letter to notify the NRC of your intentions. lf we have not heard from youwithin 10 days, we will continue with our significance determination and enforcement decision.The final resolution of this matter will be conveyed in separate correspondence. Because the NRC has not made a final determination in this matter, no Notice of Violation isbeing issued for this inspection finding at this time. Please be advised that the number andcharacterization of the apparent violation described in the enclosed inspection report maychange as a result of further NRC review.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be available electronically for public inspection in theNRC Public Document Room and from the Publicly Available Records (PARS) component ofNRC's document system, Agencywide Documents Access and Management System (ADAMS),ADAMS is accessible from the NRC Website at http://www.nrc.qov/readinq-rm/adams.html (thePublic Electronic Reading Room).Division of Reactor SafetyDocket No. 50-293License No. DPR-35

Enclosure:

Inspection Report 05000293/201 1012

w/Attachments:

Supplemental Information (Attachment 1 )Special Inspection Team Charter (Attachment 2)Detailed Sequence of Events (Attachment 3)Appendix M Table 4.1 (Attachment 4)cc w/encl: Distribution via ListServ

Sincerely,&

R, SmithBecause the NRC has not made a final determination in this matter, no Notice of Violation isbeing issued for this inspection finding at this time. Please be advised that the number andcharacterization of the apparent violation described in the enclosed inspection report maychange as a result of further NRC review.In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, itsenclosure, and your response (if any) will be available electronically for public inspection in theNRC Public Document Room and from the Publicly Available Records (PARS) component ofNRC's document system, Agencywide Documents Access and Management System (ADAMS).ADAMS is accessible from the NRC Website at http://www.nrc.qov/readinq-rm/adams.html (thePublic Electronic Reading Room).

Sincerely,/RA/Christopher G. Miller, DirectorDivision of Reactor SafetyDocket No. 50-293License No. DPR-35

Enclosure:

lnspection Report 05000293/201 1012

w/Attachments:

Supplemental Information (Attachment 1 )Special Inspection Team Charter (Attachment 2)Detailed Sequence of Events (Attachment 3)Appendix M Table 4.1 (Attachment 4)cc w/encl: Distribution via ListServDistribution: See next pageSUNSI Review Complete: rrm* (Reviewer's Initials)DOCUMENT NAME: G:\DRS\Operations Branch\M0KINLE\lPilgrim SIT June 20'11\lnspection Report Drafts\Pilgrim SIT Concurrence\Pilgrim2011 SIT Report Final.docxAfter declaring this document "An Official Agency Record" it will be released to the Public.MLI12440't00To teceivs a coov of this documGnt. indicate in the box: "C" = CoDy without attachmenvenclosure "E" = Copy with attachmenvenclosure 'N" = NoOFFICE RIiDRSRI/DRSRI/ORARI/DRPRI/DRPNAME RMcKinley/rrm*Prior concurrenceDJackson/dej*Prior concurrenceDHolody/aed for*Prior concurrenceRBellamy/tcs for*Prior concurrenceDRoberts/djr-Prior concurrenceDATE 08t19t1108t19t1108t19t1108/ t1108/30/1 1OFFICE RI/DRSNAME CMiller/cgmDATE 08131111OFFICIAL RECORD COPY Docket No.:License No.:Report No,:Licensee:Facility:Location:Dates:Team Leader:Team:Approved By:U. S. NUCLEAR REGULATORY COMMISSIONREGION I50-293DPR-3505000293/2011012Entergy Nuclear Operations, IncPilgrim Nuclear Power Station (PNPS)600 Rocky Hill RoadPlymouth, MA 02360May 16 through July 20,2011R. McKinley, Senior Emergency Response CoordinatorDivision of Reactor SafetyB. Haagensen, Resident Inspector, Division of Reactor ProjectsD. Molteni, Operations Engineer, Division of Reactor SafetyDonald E. Jackson, ChiefOperations BranchDivision of Reactor SafetyEnclosure

SUMMARY OF FINDINGS

lR 0500029312011012; 0511612011 - 071201201 1; Pilgrim Nuclear Power Station (PNPS);lnspection Procedure 93812, Special Inspection.A three-person NRC team, comprised of two regional inspectors and one resident inspector,conducted this Special lnspection. One finding with potentialfor greater than Green safetysignificance was identified. The significance of most findings is indicated by their color (Green,White, Yellow, or Red) using lnspection Manual Chapter (lMC) 0609, "SignificanceDetermination Process" (SDP). The cross-cutting aspect was determined using IMC 0310,"Components Within the Cross-Cutting Areas." Findings for which the SDP does not apply maybe Green or be assigned a severity level after NRC management review. The NRC's programfor overseeing the safe operation of commercial nuclear power reactors is described inNUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.NRC ldentified and Self Revealing Findings

Cornerstone: Initiating Events. Preliminary

White: A self-revealing finding was identified involving the failure of Pilgrimpersonnel to implement conduct of operations and reactivity control standards andprocedures during a reactor startup, which contributed to an unrecognized subcriticalityfollowed by an unrecognized return to criticality and subsequent reactor scram.The significance of the finding has preliminarily been determined to be White, or of low tomoderate safety significance. The finding is also associated with one apparent violationof NRC requirements specified by Technical Specification 5.4, "Procedures." There wasno significant impact on the plant following the transient because the event itself did notresult in power exceeding license limits or fuel damage. Additionally, interim correctiveactions were taken, which included removing the Pilgrim control room personnel involvedin the event from operational duties pending remediation, providing additional training foroperators not involved with the event, and providing increased management oversightpresence in the Pilgrim control room while long term corrective actions were developed.Entergy staff entered this issue, including the evaluation of extent of condition, into itscorrective action program (CR-PNP-2011-2475) and performed a Root Cause Evaluation(RcE).The finding is more than minor because it was associated with the Human Performanceattribute of the Initiating Events cornerstone and affected the cornerstone objective oflimiting the likelihood of those events that upset plant stability and challenge criticalsafety functions during power operations. Specifically, the failure of Pilgrim personnel toeffectively implement conduct of operations and reactivity control standards andprocedures during a reactor startup caused an unrecognized subcriticality followed by anunrecognized return to criticality and subsequent reactor scram. Because the findingprimarily involved multiple human performance errors, probabilistic risk assessment toolswere not well suited for evaluating its significance. The inspection team determined thatthe criteria for using IMC 0609, Appendix.M, "Significance Determination Process Usinglll

Qualitative Criteria," were met, and the finding was evaluated using this guidance, asdescribed in Attachment 4 to this report. Based on the qualitative review of this finding,the NRC has preliminarily concluded that the finding was of low to moderate safetysignificance (preliminary White).The inspection team determined that multiple factors contributed to this performancedeficiency, including: inadequate enforcement of operating standards, failure to followprocedures, and ineffective operator training. The Entergy RCE determined that theprimary cause was a failure to adhere to established Entergy standards andexpectations due to a lack of consistent supervisory and management enforcement. Theinspection team concluded that the finding had a cross-cutting aspect in the HumanPerformance cross-cutting area, Work Practices component, because Entergy did notadequately enforce human error prevention techniques, such as procedural adherence,holding pre-job briefs, self and peer checking, and proper documentation of activitiesduring a reactor startup, which is a risk significant evolution. Additionally, licensedpersonnel did not effectively implement the human performance prevention techniquesmentioned above, and they proceeded when they encountered unceftainty andunexpected circumstances during the reactor startup H.4(a). (Section 2)ivEnclosure 1.

REPORT DETAILS

Backoround and Description of EventIn accordance with the Special Inspection Team (SlT) Charter (Attachment 2), theinspection team conducted a detailed review of the May 10, 2011, reactor scram event atPilgrim Nuclear Power Station, including a review of the Pilgrim operators' response tothe event. The inspection team gathered information from the plant process computer(PPC) alarm printouts and parameter trends, interviewed station personnel, observed on-going control room activities, and reviewed procedures, logs, and various technicaldocuments to develop a detailed timeline of the event (Attachment 3).On May 10,2011, following a refueling outage, the reactor mode switch was taken tostartup at 0626, and control rod withdrawal commenced at 0641. The control room crewconsisted of the following personnel (additional licensed operators were present in thecontrol room conducting various startup related activities):o Assistant Operations Manager (AOM-Shift) - Senior Line Management oversightr Shift Manager (SM)- management oversight. Reactivity Senior Reactor Operator (SRO/Control Room Supervisor (CRS) -command and controlo Assistant Control Room Supervisor (ACRS). Reactor Operator At-The-Controls (RO-ATC)o Reactor Operator (Verifier)

• ATC verifierr Reactor Engineer (RE). RE in TrainingAt 1212, the reactor was made critical when control rod 38-19 was moved to position 12.Power continued to rise to the point of adding heat (POAH), and the POAH wasachieved at 1227. Once the POAH was achieved, the RO-ATC operator inserted rod38-19 to position 10 to obtain lntermediate Range Monitor (lRM) overlap correlationdata. Following the data collection, the RO-ATC operator withdrew rod 38-19 back toposition 12.At approximately 1231, the Reactivity SRO/CRS and the RO-ATC operator wererelieved by other licensed operators who continued with plant startup. The crewwithdrew control rods to establish a moderator heat-up rate. The RO-ATC operatorwithdrew control rods 14-35, 38-35 , 14-19 and 22-43 from position 08 to 12 withoutincident.The RO-ATC operator continued with the rod withdrawal sequence and tried to withdrawcontrol rod 30-11 from position 08 to 12, but the control rod would not move usingnormal notch withdraw commands. The RO-ATC then attempted to withdraw control rod30-11 using a "double-clutch" maneuver in accordance with procedures; however, thecontrol rod inadvertently inserted and settled at position 06. As stated during interviewswith the NRC inspectors, the RO-ATC operator, the ATC verifier, and the ReactivitySRO/CRS all saw the control rod in the incorrect position. However, the operators didnot enter and follow Pilgrim Nuclear Power Station (PNPS) Procedure 2.4.11, "ControlRod Positioning Malfunctions" as required. This procedure required the operators toassess the amount of the mispositioning to determine the appropriate course of remedial 2action before proceeding, and it also required the issue to be documented in a conditionreport. The operators did not perform an assessment, and they moved the control rodback to position 08 and ultimately to position 12, which was the correct final position inaccordance with reactor engineering maneuvering instructions. During interviews withthe NRC inspectors, the three operators each indicated that there was confusion in theirmind regarding whether or not the control rod met the definition of a mispositionedcontrol rod because the control rod was only out of position by one notch from the initialposition, but none of the operators referred to the procedure, and there was nodiscussion or challenge regarding the proper course of action among the operators. Thecondition was not logged, and a condition report was not generated until the issue wasidentified by NRC inspectors. In addition, the problem of the mispositioned control rodwas not discovered by the licensee during the post trip review.Following withdrawal of the five control rods (ten control rod notches), the RO-ATCobserved the process computer displaying a high short{erm (five minute average)moderator heat-up rate reading of 18'F per 5 minutes that he mistakenly believedcorresponded to an hourly heat-up rate of 216"Flhr (the actual hourly heatup rate was50"F/hr). The heat-up rate concern was discussed among the SM, Reactivity SRO/CRS,RO-ATC operator, Verifier and AOM-Shift. After the discussion, the SM directed thecrew at the controls to insert control rods to reduce the heat-up rate. This direction didnot include specific guidance or limitations regarding the number of control rod notchesto insert, At this point, the AOM-Shift and SM left the front panels area of the controlroom.The RE and RE-in-training were working at their computer terminals in the control roomperforming procedurally required calculations related to the startup. The REs had beenoccupied with these tasks from the time criticality had been achieved and had not beenconsulted on the plan to insert control rods to reduce the heatup rate. The RE-in-training overheard the operator conversation about inserting control rods. He informedthe RE, who in turn, questioned the SM about the decision to insert rods. The SMresponded that the actions were necessary to control heat-up rate. No furtherdiscussion occurred between the SM and the RE regarding the number of controlrods/notches to be used to control the heat-up rate or if there was a need to modify thereactor maneuvering plan. During interviews with the NRC inspectors, the SM and theAOM-Shift stated that they both discussed that there was a need to be careful to avoidtaking the reactor subcritical and that the action of inserting control rods had thepotential to cause the reactor to become subcritical. However, this important informationwas never communicated to any of the operators at the controls, including at the timewhen the SM directed the at-the-controls crew to insert control rods to reduce theheat-up rate.As a result of the previous control rod withdrawal, moderator temperature was 40"Fhigher than it was at initial criticality resulting in slightly increased control rod worth.The crew did not factor this increased control rod worth into their decision regarding thenumber of control rod notches to insert.Over the next three minutes, the RO-ATC operator proceeded to re-insert the followingcontrol rods from positions 12 to 8 (10 notches total) that had been previously withdrawnEnclosure 2.3to establish the heat-up: 30-1 1 , 22-43, 14-19,38-35 and 14-35. At the end of the rodinsertion evolution, the SM directed the Reactivity SRO/CRS and the RO-ATC operatorto keep reactor power on IRM range 7. This communication was not acknowledged bythe RO-ATC operator. During interviews with the NRC inspectors, none of the operatorsrecalled receiving such instructions. The SM then left the control room to take a break.The AOM-Shift left the controls area to get lunch in the control room kitchen.As a result of the control rod insertions, reactor power lowered, thus requiring theRO-ATC operator to range the lRMs down to range 7 and then to range 6. The reactorhad become subcritical, but the crew did not recognize the change in reactor status.Approximately four minutes after the control rods were inserted to reduce the heat-uprate, the RO-ATC operator observed the process computer displaying a 0'F/hr heat-uprate. At this time, the SRO who had previously been relieved, returned and re-assumedhis role as Reactivity SRO/CRS. The Reactivity SRO/CRS and the RO-ATC operatordecided to once again withdraw control rods to re-establish the desired heat-up rate.Three of the same control rods (14-35, 38-35, and 14-19) were withdrawn from positions8-12 resulting in a rising IRM count rate that was observed by the operators. However,the crew did not recognize that the reactor status had changed from subcritical to critical.At this point, the AOM-Shift returned to the reactor panel area. The RO-ATC operatorcontinued rod withdrawal with control rod 22-43 from position 08 to 10. The RO-ATCoperator and the Verifier ranged the lRMs up as reactor power increased. The RO-ATCoperator then withdrew control rod 22-43 from position 10 to 12. The operators did notrecognize the increasing rate of change in IRM power.Finally, the RO-ATC operator selected and withdrew control rod 30-11 from position 8 to10. At 1318, IRM readings rose sharply and an IRM Hi-Hi flux condition wasexperienced on both Reactor Protection System (RPS) channels resulting in anautomatic reactor scram at approximately 1

.7 o/o reactor power.Operator Human PerformanceInspection ScopeThe inspection team interviewed the Pilgrim control room personnel that responded tothe May 10,2011, event including the SM, AOM-Shift, CRS, ACRS, RO-ATC, ROverifier, and the REs to determine whether these personnel performed their duties inaccordance with plant procedures and training. The inspection team also reviewednarrative logs, sequence of events and alarm printouts, condition reports, PPC trenddata, procedures implemented by the crew, and procedures regarding the conduct ofoperations.a.Enclosure

4b. Findinqs/ObservationsFailure to lmplement Procedures durinq Reactor StartupIntroduction: A self-revealing finding was identified involving the failure of Pilgrimpersonnel to implement conduct of operations and reactivity control standards andprocedures during a reactor startup, which contributed to an unrecognized subcriticalityfollowed by an unrecognized return to criticality and subsequent reactor scram. Thesignificance of the finding has preliminarily been determined to be White, or of low tomoderate safety significance. The finding is also associated with one apparent violationof NRC requirements specified by Technical Specification 5.4, "Procedures."Description: On May 10,2011, following a refueling outage, operators were in theprocess of conducting a reactor startup. During the course of the startup, multiplelicensed operators failed to implement written procedures as described below:. Entergy procedure EN-OP-1 15, "Conduct of Operations," Revision 10, Section 4.0,states that the SM is to "provide oversight of activities supporting complex andinfrequently performed plant evolutions such as plant heat-up [and] startup."Additionally, the SM is responsible for ensuring "conservative actions are takenduring unusual conditions ... when dealing with reactivity control," However, the SMdid not oversee the activities in progress during reactor heatup and left the controlroom when the heat-up rate was being adjusted with control rod insertion, The SMdid not ensure the actions taken to reestablish or adjust the reactor heatup ratewere conservative nor did he reinforce those actions with the operating crew.r Entergy procedure EN-OP-1 15, "Conduct of Operations," Revision 10, Section 4.0,states that the CRS is required to "Ensure Pre-Evolution Briefings are held [and]plant operations are conducted in compliance with administrative and regulatoryrequirements." PNPS procedure 1.3.34, "Operations Administrative Policies andProcedures," Revision 1 17, Section 6.10.1

.1 states, "All complex or infrequentlyperformed activities warrant a pre-evolution briefing." Section 6,10.1.1[8] lists anInfrequently Performed Tests or Evolutions Briefing as one type of pre-evolutionbriefing, and Section 6.10.1 .1 [4] states, "lnfrequently Performed Tests or EvolutionsBriefings for the performance of Procedures classified as "lnfrequently PerformedTests or Evolutions" (IPTE) should be performed with Senior Line Manager oversightas specified in EN-OP-116, "lnfrequently Performed Tests or Evolutions." EntergyProcedure EN-OP-116, Revision 7, Attachment 9.1 identifies "Reactor Startup" as anIPTE. However, in this case, the licensee conducted a reactor startup withoutperforming an IPTE briefing or any other type of pre-evolution briefing as defined inPNPS procedure 1.3.34. lt is noteworthy to point out that an IPTE briefing packagewas previously prepared, approved, and scheduled; however, the IPTE briefing wasnever performed as required by the procedures described above. In addition, anIPTE briefing was also not performed for the startup following this event. Finally, theCRSs did not ensure the administrative requirements of the conduct of operationsprocedures or the regulatory requirement to implement the control rod mispositioningprocedure were met. This issue was identified by the NRC inspectors.Enclosure

5Entergy procedure EN-OP-115, "Conduct of Operations," Revision 10, Section 5.2,states control room operators are required to "develop and implement a plan thatincludes contingencies and compensatory measures" and when implementing thoseplans the "crew ... continuously evaluates the plan for changing conditions" and"Human Performance (HU) tools (..., peer/cross-checking, oversight, questioningattitude, etc.) are utilized ..." In addition, "When the control room team is faced witha time critical decision: Use all available resources...do not proceed in the face ofuncertainty..." However, the control room operators failed to develop contingencyplans or compensatory measures for adjusting reactor heat-up rate or addressinghigher than expected reactor heat-up rates. The crew also failed to develop orimplement contingencies for control rods which were difficult to maneuver when theywere at low reactor power. Additionally, the use of human performance tools wasineffective in addressing the actions or conditions that led to the unexpected reactorheatup rate and the mispositioning of control rod 30-11. Specifically, failures in theuse of peer checking and questioning the conditions that led to the unexpectedreactor heat-up rate directly contributed to the mispositioned control rod and thesubsequent reactor scram. Lastly, the control room team did not use all availableresources by involving Reactor Engineering staff in its decision-making, andproceeded in the face of uncertainty by failing to consider the consequences of thereactivity changes.Entergy procedure EN-OP-115, "Conduct of Operations," Revision 10, Section 5.4states that reactor operators are expected to perform reactivity manipulations "in adeliberate, carefully controlled manner while the reactor is monitored to ensure thedesired result is obtained." However, the reactor operators did not adequatelymonitor the conditions of the reactor while attempting to establish and adjust thereactor heat-up rate. Although the reactor operators were watching the response ofboth the lRMs and the computer point displaying a five minute average reactorheatup, they were moving control rods faster than the plant temperature couldrespond and therefore taking actions to continue control rod movement before thedesired result of their manipulations could be assessed. Additionally, after insertingcontrol rods to adjust the reactor heat-up rate, the operators had sufficient indicationsthat the reactor was significantly subcritical as evidenced by the required rangingdown of lRMs, the drop in Source Range Monitor (SRM) count rates, andestablishing a negative reactor period. The operator's failure to adequately monitorthe status of the reactor led to an unrecognized subcritical condition and subsequentreturn to criticality resulting in an eventual reactor scram.PNPS procedure 1.3.34, "Operations Administrative Policies and Procedures,"Revision 1 17, Section 6.7.5 states, "Any relief occurring during the shift (eithershort-term or for the remainder of the shift) will be recorded in the CRS log." ltfurther states, "...a verbal discussion of plant status and off-normal conditions mustbe conducted." However, several people in watch standing positions changed fromthe start of the shift, but none of those changes were entered into the control roomlog. In addition, when the ACRS was turning over to the CRS, there was nodiscussion of the mispositioning of control rod 30-11.Enclosure

6. PNPS Procedure 2.4.11, "Control Rod Positioning Malfunctions," Revision 35,Section 5.4 defines a mispositioned control rod as "a control rod found to be left in aposition other than the intended position $ a control rod that moves more than onenotch beyond its intended position." Attachment 4 Step [3] and Step [a] of the sameprocedure requires the operators to assess the degree of mispositioning and take theappropriate remedial action depending on the degree of mispositioning. Attachment4 Step [5] also states, "lf the control rod is determined to be mispositioned, thenrecord the event as a condition report." In this case, the RO-ATC attempted towithdraw control rod 30-11 from position 08 to position 10 (intended position), but therod inadvertently insertbd to position 06. Upon recognizing the error, the operatorsdid not enter the procedure when control rod 30-11 was found to be left in a positionother than the intended position and which was more than one notch from theintended position. The operators did not assess the amount of the control rodmispositioning in accordance with the procedure, nor was there any discussion aboutthe mispositioning on the crew. Furthermore, the event was not logged, nor was acondition report generated. Instead, the operators did not enter and follow theprocedure, and they continued on with the startup in the face of uncertainty. Thisissue was not detected during the licensee posttrip review. lt was identified by theNRC inspectors.o PNPS Procedure 2.1.1, "Startup from Shutdown," Revision 173, Page 53, Caution 2states, "ln the event the reactor goes subcritical after achieving initial criticality, thenreturn to step [53] and re-perform the steps to restore the Reactor to a criticalcondition." In addition, PNPS Procedure 2.1.4, "Approach to Critical," Revision 26,Section 5.0 states, "ln the event the reactor goes subcritical after achieving initialcriticality, then with Reactor Engineering guidance, re-perform Section 7.0 Steps [6]and [7] to restore the Reactor to a critical condition." However, the operators did notrecognize that the reactor had become subcritical and did not re-perform theprocedural steps mentioned above to restore the reactor to a critical condition in acontrolled manner under the guidance of Reactor Engineering. There was sufficientinformation available to the operators to identify that the reactor had becomesubcritical. In addition, REs were available in the control room, but they were notconsulted by the operators.Analvsis: The inspection team determined that the failure of Pilgrim personnel toimplement conduct of operations and reactivity control standards and procedures duringa reactor startup was a performance deficiency that was reasonably within Entergy'sability to foresee and prevent. The finding is more than minor because it was associatedwith the Human Performance attribute of the Initiating Events cornerstone and affectedthe cornerstone objective of limiting the likelihood of those events that upset plantstability and challenge critical safety functions during power operations. Specifically, thefailure of Pilgrim personnel to effectively implement conduct of operations and reactivitycontrol standards and procedures during a reactor startup caused an unrecognizedsubcriticality followed by an unrecognized return to criticality and subsequent reactorscram.Enclosure

7The inspection team determined that multiple factors contributed to this performancedeficiency including: inadequate enforcement of operating standards, failure to followprocedures, and ineffective operator training. The Entergy RCE documented that theprimary cause was a failure to adhere to established Entergy standards and expectationsdue to a lack of consistent supervisory and management enforcement. In addition, theEntergy RCE specified a number of condition reports and self assessment reports writtenin the months preceding this event that demonstrated that the performance deficiencyexisted over an extended period of time and affected all operating crews. While theperformance deficiency manifested itself during this particular low power event, therewas the potential for the performance deficiency to result in a more consequential eventunder different circumstances.Because the finding primarily involved multiple human performance errors, probabilisticrisk assessment tools were not well suited for evaluating its significance. The inspectionteam determined that the criteria for using IMC 0609, Appendix M, "SignificanceDetermination Process Using Qualitative Criteria," were met, and the finding wasevaluated using this guidance as described in Attachment 4 to this report. Based on thequalitative review of this finding, the NRC concluded that the finding was preliminarily oflow to moderate safety significance (preliminary White). The completed Appendix Mtable is attached to this report (Attachment 4). There was no significant impact on theplant following the transient because the event itself did not result in power exceedinglicense limits or fuel damage. Additionally, interim corrective actions were taken, whichincluded removing the Pilgrim control room personnel involved in the event fromoperational duties pending remediation, providing additional training for operators notinvolved with the event, and providing increased management oversight presence in thePilgrim control room while long term corrective actions were developed.This finding had a cross-cutting aspect in the Human Performance cross-cutting area,Work Practices component, because Entergy management and supervision did notadequately enforce human error prevention techniques, such as procedural adherence,holding pre-job briefs, self and peer checking, and proper documentation of activitiesduring a reactor startup, which is a risk significant evolution. Additionally, licensedpersonnel did not effectively implement the human performance prevention techniquesmentioned above, and they proceeded when they encountered uncertainty andunexpected circumstances during the reactor startup [H.a(a)].Enforcement: Technical Specification 5.4, "Procedures," states, in part, that writtenprocedures shall be established, implemented, and maintained covering the applicableprocedures recommended in Appendix "A" of Regulatory Guide (RG) 1.33, February,1978. RG 1.33, Appendix "A," requires that typical safety-related activities listed thereinbe covered by written procedures. Contrary to the above, on May 10,2011, as reflectedin the examples listed in the description section of this finding, the licensee failed toimplement safety-related procedures related to RG 1.33, Appendix "A," Paragraph 1,"Administrative Procedures;" Paragraph 2, "General Plant Operating Procedures;" and,Paragraph 4, "Procedures for Startup, Operation, and Shutdown of Safety-Related BWRSystems."Enclosure 3.IFollowing a review of the event, the licensee documented the condition in the correctiveaction program (CR-PNP-2011-2475). There was no significant impact on the plantfollowing the transient because the event itself did not result in power exceeding licenselimits or fuel damage. Additionally, interim corrective actions were taken, which includedremoving the Pilgrim control room personnel involved in the event from operationalduties pending remediation, providing additional training for operators not involved withthe event, and providing increased management oversight presence in the Pilgrimcontrol room while long term corrective actions were developed.Pending determination of final safety significance, this finding with the associatedapparent violation will be tracked as AV 05000293/2011012-01, Failure to lmplementGonduct of Operations and Reactivity Gontrol Procedures during Reactor Startup.Fitness for DutvInspection ScopeThe inspection team interviewed the control room personnel that were directly involvedwith the May 10,2011, reactor scram event as well as management personnel involvedwith the immediate post event investigation. The inspection team also reviewed EntergyFitness for Duty (FFD) program requirements contained in the corporate and siteprocedures.Fi nd i nos/ObservationsNo findings were identified.TraininqInspection ScopeThe inspection team interviewed personnel, reviewed simulator modeling andperformance, and reviewed training material related to Just in Time Training (JITT)material for the initial and subsequent startups, remedial training for the operatorsinvolved with the event, and training plans for startups and reactivity maneuvers.Fi nd i nqs/ObservationsNo findings were identified.The inspection team observed that the JITT training that was provided prior to the initialstartup was very limited in scope in that it only covered the approach to criticality up tothe POAH. lt did not cover the full range of reactor heat-up, and it covered very littleOperating Experience. In addition, several operators that were directly involved with thisevent did not attend the JITT training including the SM, the ACRS who temporarilyrelieved the CRS prior to the scram, and the RO who was at the controls when thescram occurred.a.h4.a.b.Enclosure 5.IOrqanizational Responselmmediate ResponseInspection ScopeThe inspection team interviewed personnel, reviewed various procedures and records,and observed control room operations to assess immediate response of stationpersonnel to the reactor scram event.Fi nd i nqs/ObservationsNo findings were identified.The inspection team observed that Entergy's initial response to the event was notappropriately thorough and was narrowly focused. lmmediately foilowing the event,operators were debriefed in an attempt to ascertain the cause of the event. Initially,Entergy personnel focused on a potential IRM malfunction as the potential cause of theevent despite the fact that multiple IRM channels accurately tracked reactor power alongwith operator reactivity inputs. lmmediate post event interviews with the crew did notprobe human error as a potential cause even though the SM, the AOM-Shift, and theREs had expressed concerns just prior to the scram regarding the insertion of controlrods so near the point of criticality. Operators involved with the event were dismissed forthe day as the investigation continued to incorrectly focus on equipment malfunction asthe most likely cause of the event. Several hours passed before it became clear to sitemanagement that human error was the cause of the event. As a result, the operatorsinvolved with the event were not thoroughly interviewed to ensure that all of the humanperformance aspects were fully understood prior to proceeding with the next startup. Inaddition, the inspection team identified that the posttrip review failed to identify that acontrol rod had been mispositioned just prior to the scram and that an IPTE briefing hadnot been conducted for the startup. Consequently, additional human performanceissues were not evaluated, and the licensee again failed to perform an IPTE briefingprior to the subsequent startup as required by Entergy procedures.Post-Event Root Cause Evaluation and ActionsInspection ScopeThe inspection team reviewed Entergy's Root Cause Evaluation (RCE) report for theevent to determine whether the causes and associated human performance issues wereproperly identified. Additionally, the inspection team assessed whether interim andplanned long term corrective actions were appropriate to address the cause(s).61a.b.5.2a.Enclosure b.10Find inqs/ObservationsNo findings were identified.The RCE was thorough and appeared to identify the underlying causal factors. Theassociated proposed corrective actions appeared to adequately address the underlyingcausal factors. Entergy identified the root cause as a lack of consistent supervisory andmanagement enforcement of administrative procedure requirements and managementexpectations for command and control, roles and responsibilities, reactivitymanipulations, clear communications, proper briefings, and proper turnovers.The RCE also identified contributing causes including weaknesses in monitoring plantstatus and parameters as well as weaknesses in operator proficiency with regards to lowpower operations.Meetinqs. Includinq ExitExit Meetino SummarvOn July 20,2011, the inspection team discussed the inspection results with Mr. R. Smith,Site Vice President, and members of his staff. The inspection team confirmed thatproprietary information reviewed during the inspection period was returned to Entergy.40A6Enclosure Enterov PersonnelR. SmithJ. DreyfussD. NoyesJ. MacdonaldR. ProbascoJ. CoutoS. AndersonT. TomonJ. ByronJ. HayhurstS. BethayJ. LynchT. WhiteF. McGinnisR. ByrneV. FallacaraS. ReininghausJ. HouseV. MagnattaR. ParanjapeA,1-1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACTS

ite Vice PresidentGeneral Manager Plant OperationsManager, OperationsAssistant Manager, OperationsShift Manager, OperationsShift Supervisor, OperationsShift Supervisor, OperationsReactor Operator, OperationsReactor Operator, OperationsReactor Operator, OperationsDirector, Nuclear Safety AssuranceManager, LicensingManager, Quality AssuranceEngineer, LicensingSenior Engineer, LicensingDirector, EngineeringManager, TrainingSupervisor, Operations TrainingLead lnstructor, Operations TrainingReactor Engineer

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Ooened05000293/2011012-01 AV Failure to lmplement Conduct of Operations andReactivity Control Procedures during ReactorStartup (Section 2)

LIST OF DOCUMENTS REVIEWED

Procedures:1.3.34, "Operations Administrative policies and Procedures," Revision 1 191 .3.37 , "Post-Trip Reviews," Revision 271.3,63, "Conduct of Event Review Meetings," Revision 251.3.109, "lssue Management," Revision 82.1.1, "Startup from Shutdown," Revision 1732.1.4, "Approach to Critical," Revision 262.1.7, "Vessel Heat-up and Cool Down," Revision 542.4.11, "Control Rod Positioning Malfunctions," Revision 352.4.11.1, "CRD System Malfunctions," Revision 21Attachment 1

A-1-2SUPPLEMENTAL INFORMATIONNOP96A3, "Reactivity Management Peer Panel," Revision 10EN-FAP-AD-OO1, "Fleet Administrative Procedure (FAP) Process," Revision 0EN-FAP-OM-006, "Working Hour Limits for Non-Covered Workers," Revision 2EN-FAP-OP-008, "Reactivity Management Performance Indicator Program," Revision 0EN-FAP-OP-01 1, "Operator Human Performance Indicator Program," Revision 0EN-HU-102, "Human Performance Tools," Revision 5EN-HU-103, "Human Performance Error Reviews," Revision 4EN-NS-102, "Fitness for Duty Program," Revision 9EN-OM-119, "On-Site Safety Review Committee," Revision 7EN-OM-123, "Fatigue Management Program," Revision 3EN-OP-103, "Reactivity Management Program," Revision 5EN-OP-1 15, "Conduct of Operations," Revision 10EN-OP-1 16, "lnfrequently Performed Tests of Evolutions," Revision 7EN-RE-214, "Conduct of Reactor Engineering," Revision 0EN-RE-215, "Reactivity Maneuver Plan," Revision 1EN-RE-219, "Startup sequence Criticality Controls (BWR)," Revision 0Condition Reports:CR-PNP-2011-02475 and associated Root Cause Evaluation Report, Revision 1CR-PNP-201

1-02488cR-PNP-2011-02493cR-PNP-2011-02504CR-PNP-201

1-02506CR-PNP-2011-02546CR-PNP-201

1-02568CR-PNP-2011-02572cR-PNP-2011-02577CR-PNP-201

1-03598Self Assessments:LO-PNPLO-2009-00071, "Focused Assessment on Reactivity Management"LO-PNPLO-2010-00106, "Snapshot Assessment on Reactivity Management ProcedureRevision lmplementation"LO-PNPLO-2010-00106, "Snapshot Assessment on SOER 07-01 Recommendation 4 ReactivityManagement Operations Training"Technical Specifications:3.5.C, "HPCI System"3.5.D,'RCIC System"5.4.1, "PROCEDURES"Traininq Material:lnstructional Module, Reactor Startup and Criticality (& Main Turbine Overspeed) Just in TimeTraining used for

0511012011 and

0511112011 Startup JITTInstructional Module, Reactor Startup and Criticality May 2011 Just in Time Training used for051

1812011 Startup JITTAttachment 1

A-1-3SUPPLEMENTAL INFORMATIONJust in Time Training PowerPoint used for 05/1812011 Startup JITTlnstructor Lesson Plan JITT

RFO 18 Hydro 2.1 .8.5Simulator JITT Reactor Shutdown 2.1.5 and Vessel Cooldown 2.1.7 , Revised 0410112011Simulator JITT Reactor Shutdown 2.1.5 and Vessel Cooldown 2.1 .7 , Revised 0211912011Training Schedules for Outage Training Cycle

0311412011 -0410712011Training Schedules for Training Cycle 020211312011 -0211712011Training Schedules for Training Cycle 01

1112212010 - 0112212011Training Records and Remediation Training for Current Licensed Operatorslnitial License Class 2009-2011 Class ScheduleO-RO-03-02, "Reactor Plant Startup Certification Unit Guide," Revision 10O-RO-03-01-19, "Reactivity Management and Control Instructor/Student Guide," Revision 2O-RO-03-01 -20, "Simulator Scenario, Operations Standards," Revision 0O-RO-03-02-01, "lnstructional Module - Day One Cold Reactor Startup," Revision 7O-RO-03-02-02, "lnstructional Module - Day Two Hot Reactor Startup," Revision 7O-RO-03-02-03, "lnstructional Module * Day Three Cold Reactor Startup," Revision 3O-RO-03-02-04, "lnstructional Module - Day Four Hot Reactor Startup," Revision 3O*RO-03-02-05, "lnstructional Module - Day Five Cold Reactor Startup," Revision 3O-RO-03-02-06, "lnstructional Module - Day Six Cold Reactor Startup," Revision 3O-RO-03-02-07, "lnstructional Module - Day Seven 905 Certification Practice," Revision 3O-RO-03-02-08, "lnstructional Module * Day Eight 905 Certification Practice," Revision 2O-RO-03-02-09, "lnstructional Module - Day Nine Reactor Power Operations," Revision 1O-RO-03-02-51, "lnstructional Module - SOER 90-3 Nuclear Instrument Miscalibration,"Revision 3Miscellaneous:Crew Briefing Sheet from May 10,2011 SCRAMOperations Section Standing Order 11-03OSRC Meeting 2011-008 Meeting MinutesPost-Trip Review Package from May 10,2011 SCRAM with Attachments and Supporting Data"EN-OP-116 Attachment 9,3 ITPE Supplemental Controls," developed for Post-RefuelingOutage StartupReactor Engineer's calculations pertaining to criticality prior to the reactor SCRAMeSOMS Control Room Logs from

0510912011 through 0511112011SRM and Moderator Temperature Traces with Calculated SRM Period 0511012011Control Room Personnel Chart Dayshift 0511012011Control Rod Notch History from Reactor Critical to Reactor SCRAM 0511012011Control Rod Notch Worth Calculations for 05/1012011 Reactor StartupPower Maneuver Plan Cycle 19-01Attachment 1

A-1-4SUPPLEMENTAL INFORMATION

LIST OF ACRONYMS

ACRS Assistant Control Room SupervisorADAMS Agency-wide Documents Access and Management SystemAOM Assistant Operations ManagerATC At the ControlsAV Apparent ViolationBOP Balance of PlantCCDP Conditional Core Damage ProbabilityCFR Code of Federal RegulationsCR Condition ReportCRD Control Rod DriveCRS Control Room SupervisorDRP Division of Reactor ProjectsDRS Division of Reactor SafetyFFD Fitness for DutyHEP Human Error ProbabilityHPCI High Pressure Coolant InjectionHUR Heatup RateIMC lnspection Manual ChapterIPTE Infrequently Performed Tests or EvolutionsIRM Intermediate Range MonitorJITT Just in Time TrainingNRC Nuclear Regulatory CommissionOPS

MGR Operations Manager

PARS Publicly Available RecordsPD Performance DeficiencyPNPS Pilgrim Nuclear Power StationPOAH Point of Adding HeatPPC Plant Process ComputerPRA Probabilistic Risk AssessmentRCE Root Cause EvaluationRCIC Reactor Core lsolation CoolingRE Reactor EngineerRG Regulatory GuideRO Reactor OperatorRO-ATC Reactor Operator at the ControlsRPS Reactor Protection SystemSDP Significance Determination ProcessSM Shift ManagerSRI Senior Resident InspectorSRM Source Range MonitorSRO Senior Reactor OperatorSIT Special Inspection TeamSTA Shift Technical AdvisorTS Technical SpecificationAttachment 1

A-2-1SPECIAL

INSPEC TION

TEAM [[]]

CHARTE RUNITED STATESN

UCLEAR [[]]

REGULA TORY COMMISSIONREGION I475 ALLENDALE ROADKING

OF [[]]

PRUSSI A.

PA 19406-1415

MEMORA NDUM TO:SPECIAL INSPECTION

TEAM [[]]

CHARTE RMay 13, 2011Samuel

L. Hansell Jr., ManagerSpecial Inspection TeamRaymond R. McKinley, LeaderSpecial Inspection TeamChristopher G. Miller, Director /

RA /Division of Reactor SafetyDarrell

J. Roberts, Director /

RA by Paul Krohn Acting For/Division of Reactor ProjectsSPECIAL INSPECTION

TEAM [[]]

CHARTE R -PILGRIM NUCLEAR

POWER [[]]

STATIO N OPERATORPERFORMANCE

DURING [[]]

REACTO R STARTUPON

MAY 1Q.2011

FROM [[:SUBJECT:In accordance with lnspection Manual Chapter (lMC) 0309, "Reactive Inspection DecisionBasis for Reactors," a Special Inspection Team (SlT) is being chartered to evaluate operatorperformance and organizational decision-making associated with a reactor scram that occurredduring a startup on May 10,2011, The decision to conduct this special inspection was based onmeeting the deterministic criteria (the event involved questions or concerns pertaining tolicensee operational performance) and risk criteria specified in Enclosure 1 of]]

IMC 0309. Thecalculable increase in conditional core damage probability (

CCDP ), which was in the low E-6range, was based on application of an Initiating Event Analysis in Sapphire 8 due to the reactorscram, which was then modified for the conditions of the reactor when the transient occurred,The

SIT will expand on the event follow-up inspection activities started by the residentinspectors and augmented by a Division of Reactor Projects (

DRP) inspector who wasdispatched to the site soon after the event. The Team will review the causes of the event, andEntergy's organizational and operator response during and after the event, The Team willAttachment 2t rt *.r. i

A-2-2SPECIAL

INSPEC TION

TEAM [[]]

CHARTE Rperform interviews, as necessary, to understand the scope of operator actions performed duringthe event. The Team will also assess whether the

SIT should be upgraded to an AugmentedInspection Team in accordance with

IMC 0309.The inspection will be conducted in accordance with the guidance contained in

NRC [[InspectionProcedure 93812, "Special Inspection," and an inspection report will be issued within 45 daysfollowing the final exit meeting for the inspection.The Special Inspection willcommence on May 16, 2411. The following personnel have beenassigned to this effort:Manager: Samuel L. Hansell, Jr., Branch ChiefOperations Branch,]]

DRS , Region ITeam Leader:Team Members:Enclosure: Special Inspection Team CharterRaymond

R. McKinley, Senior Emergency Response CoordinatorPlant Support Branch,

DRS , Region IBrian

C. Haagensen, Millstone Power Station Resident InspectorDivision of Reactor Projects,

DRP , Region IDavid

L. Molteni, Operations EngineerOperations Branch,

DRS, Region IAttachment 2

A-2-3SPECIAL

INSPEC TION

TEAM [[]]

CHARTE [[RSpecial Inspection Team CharterPilgrim Nuclear Power StationOperator Performance During ReactorStartup May 10,2011Backqround:During startup from a refueling outage, Entergy operators withdrew rods to criticality theafternoon of May 10,2011 and continued to withdraw control rods to the point of adding heat(approximately 1o/o power). While continuing to increase power, operators identified a higherthan expected heat-up rate (HUR) with a five minute average]]

HUR that, if allowed to continue,would have resulted in exceeding the technical specification limit. Operators made the ControlRoom Supervisor (

CRS ) and Shift Manager (SM) aware of the condition and proceeded to insertfive control rods (two notches each) to lower the

HUR [[to approximately 65"F/hr. At the time, itwas not identified by the operators, reactor engineers or management oversight in the controlroom that the control rod insertions brought the reactor to a subcritical state (approximately0.35% subcritical by later calculations). After reducing the]]

HUR , the operators (withoutrecognition of the subcritical reactor condition), proceeded to withdraw the five control rods backto their previous position. While withdrawing the fifth control rod back to its original position, thereactor experienced a full

SCRAM on Intermediate Range Monitor (l

RM ) Hl-Hl flux signals. Allrods inserted and equipment responded as expected.Pilgrim initially investigated potential equipment related causes for the automatic scram ascommunicated to the

NRC on the afternoon of May 10,2011. Subsequent analysis revealedthat human performance errors made by the operators were the cause of the scram.

NRC [[wasinformed of this in the early morning hours of May 11,2011. Entergy is continuing itsinvestigation of the operator actions taken during this event. Entergy suspended thequalifications of the operators and the Shift Manager directly involved with the event while theinvestigation continues. Additional actions have been taken by Entergy that include morerestrictive controls on reactivity additions following a negative reactivity insertion of any kind,briefing to other operating crews regarding the event, and initiation of a root cause evaluation.The Pilgrim resident inspectors and a resident inspector from a different site provided follow-upto this event under the Reactor Oversight Process (ROP) baseline inspection program,Basis for the Formation of the SIT:The]]

IMC [[0309 review concluded that one of the deterministic criteria was met due to questionsor concerns pertaining to licensee operational performance. This criterion was met based onhuman performance errors that occurred and led to the unanticipated automatic reactor scram.The human performance errors included:. Reactor operators were focused on monitoring heatup rate (]]

HUR)without appropriatefocus on power level throughout the startup event;. Reactor operators and control room supervision did not have proper sensitivity for theimpacts from negative reactivity insertions with the reactor at low power conditions;Attachment 2

A-2-4SPECIAL

INSPEC TION

TEAM [[]]

CHARTE R. The operators did not identify or utilize available plant indications that indicated thereactor was subcritical;. Reactor operators did not follow shift manager instructions to maintain reactor powerwithin the current

IRM power band while addressing the elevated

HUR [[;. Operators and control room supervision did not engage reactor engineering staff withregard to planned rod movement after the reactor was made subcritical; ando Prior to the identification of the unexpected HUR, reactor operators did notimplemenVenter the required abnormal operating procedure for a mispositioned controlrod (Rod 30-1 1).In accordance with]]

IMC 0309, the event was evaluated for risk significance because onedeterministic criterion was met, A Region I

SRA [[evaluated the transient (reactor scram)fromlow reactor power using the Initiating Event Assessment feature of Saphire 8. The lE-Transbasic event probability was set to 1.0 and all other initiating events were set to zero. Theresulting dominant core damage sequences were subsequently evaluated by the]]

SRA toaccount for the low reactor power conditions and alternating current (

AC ) power being suppliedby off-site sources at the time of the event. The resulting conditional core damage probability(CCDP)was conservatively estimated in the low E-6 range, which is the overlap region betweenan

SIT and No Additional inspection required. The dominant core damage sequences involvefailure of direct current (

DC) power sources and failure of residual heat removal. However, withthe low decay heat load following the refuel outage, these core damage sequences represent aconservative estimate of risk.Additionally, this event involved multiple licensed operators not recognizing the reactivity statusof an operating reactor during startup and demonstrating a poor understanding of reactorphysics in a low power condition. In light of the aforementioned human performance errors, andconsistent with the risk evaluation and Section 4.04, Region I has decided to initiate an SlT.Obiectives of the Special Inspection:The Team will review the causes of the event, and Entergy's organizational and operatorresponse during and following the event. The Team will perform interviews, as necessary, tounderstand the scope of operator actions performed during the event.To accomplish these objectives, the Team will:1. Develop a complete sequence of events including follow-up actions taken byEntergy, and the sequence of communications within Entergy and to the NRCsubsequent to the event;2. Review and assess crew operator performance and crew decision making, includingadherence to expected roles and responsibilities, the use of the command andcontrol elements associated with reactivity manipulations, the use of procedures, theuse of diverse instrumentation to assess plant conditions, response to alarms andoverall implementation of operations department and station standards;Attachment 2

A-2-5SPECIAL

INSPEC TION

TEAM [[]]

CHARTE [[REvaluate the extent of condition with respect to the other crews;Review the adequacy of operator requalification training as it relates to this event,including the integration of newly licensed operators into the operator requalificationtraining program;Review the adequacy of the preparation by the operations staff for the reactor startupincluding training prior to the evolution and briefings by the operations staff.Review the adequacy of the simulator to model the behavior of the current reactorcore during startup activities and the current adequacy of the simulator for use inreactor startup training ;Assess the decision making and actions taken by the operators and stationmanagement during the initial and subsequent reactor startup to determine if thereare any implications related to safety culture;Review and assess the effectiveness of Entergy's response to this event andcorrective actions taken to date. This includes overall organizational response, andadequacy of immediate, interim and proposed longterm corrective actions. This willalso include evaluation of the root cause analysis when developed by the licensee;9. Review the adequacy of the Entergy and Site fitness for duty processes andprocedures when a human performance error has occurred;10. Evaluate Entergy's application of pertinent industry operating experience, includingINPO]]

SOER 10-2, "Engaged, Thinking Organizations,"

INPO SOER 07-1, "ReactivityManagement," and other recent events involving reactivity management errors toassess the effectiveness of any actions taken in response to the operatingexperience; and11. Document the inspection findings and conclusions in a Special Inspection Team finalreport within 45 days of inspection completion.Guidance:Inspection Procedure 93812, "Special Inspection", provides additional guidance to be used bythe SlT. Team duties will be as described in Inspection Procedure 93812. The inspectionshould emphasize fact-finding in its review of the circumstances surrounding the event. Safetyconcerns identified that are not directly related to the event should be reported to the Region Ioffice for appropriate action.The Team will conduct an entrance meeting and begin the inspection on May 16,2011. Whileon-site, the Team Leader will provide daily briefings to Region I management, who willcoordinate with the Office of Nuclear Reactor Regulation to ensure that all other pertinentparties are kept informed. The Team will also coordinate with the Region I State Liaison OfficerAttachment 23.4.5.6.7.8.

A-2-6SPECIAL

INSPEC TION

TEAM [[]]

CHARTE Rto implement the Memorandum of Understanding between the

NRC [[and the State ofMassachusetts to offer observation of the inspection by representatives of the state. A reportdocumenting the results of the inspection will be issued within 45 days following the final exitmeeting for the inspection.Before the end of the first day onsite, the Team Manager shall provide a recommendation to theRegional Administrator as to whether the]]

SIT should continue or be upgraded to an AugmentedInspection Team response.This Charter may be modified should the Team develop significant new information thatwarrants review.Attachment 2

A,3-1DETAILED

SEQUEN CE

OF [[]]

EVENTS [[May 10,2011, Reactor Scram EventThe team constructed the sequence of events from a review of control room narrative logs, plantprocess computer (PPC) data (alarm printout, sequence of event printout, plant parametergraphs) and plant personnel interviews.TimeEvent05/09/11TwoSessionsJust In Time Training (JITT) was conducted for the reactor startup. Certain keymembers of the operating crew that were directly involved with this event were notpresent for the training including the Shift Manager (SM), the Assistant ControlRoom Supervisor (ACRS) who temporarily relieved the Control Room Supervisor(CRS) prior to the scram, and the Reactor Operator who was at the controls (RO-ATC)when the scram occurred.05/10/1 10626The reactor mode switch was moved to the startup position.-0630The oncoming day shift operators received a reactor maneuvering plan briefing.The Reactor Engineers (REs) led the brief.0641Operators commenced control rod withdrawal.0700The day shift operating crew assumed the shift, and control rod withdraw continues.1212The reactor became critical.1227The point of adding heat was reached.-1231The]]

CRS was relieved for lunch by the

ACRS. The oncoming

CRS providing therelief did not receive Just In Time Training (

JITT ), nor did he participate in thereactor maneuvering plan briefing.-1231The RO-ATC was relieved for lunch by the Licensed Operator previously assignedas the

ATC verifier. The oncoming

RO -ATC providing the relief did not receive JustIn Time Training (JITT), but he did participate in the reactor maneuvering planbriefing.-1231A Licensed Operator previously assigned to other startup activities was reassignedto fill the role of

ATC verifier. This individual received

JITT training, and he alsoreceived a separate reactor maneuvering plan briefing from a

RE upon arriving towork at approximately 1 100.1246The

RO-ATC withdrew 5 rods 2 notches to establish a heat-up rate.Attachment 3

A-3-2DETAILED

SEQUEN CE

OF [[]]

EVENTS incorrectly inserted one notchto position 06. The RO-ATC does not discuss the control rod mispositioning errorwith the crew.1257The]]

ATC verifier and

CRS [[also saw control rod 30-11 move incorrectly to position06, but the control rod mispositioning error is not discussed.1302The RO-ATC then withdraws control rod 30-11 from position 06 to position 12.-1 305The crew observes that the 5 minute average reactor coolant heat-up rate is 18"Fover the 5 minute period, and the crew determines that this corresponded to a216'Flhour heat-up rate. In actuality, the 5 minute average heat-up rate reflectedthe instantaneous heat-up rate. The actual hourly heat-up rate was 50'F/hour.The crew informs the]]

SM of the perceived heat-up rate.-1 306The

SM directed the RO-ATC to insert control rods to reduce the heat-up rate, butthe

SM did not specify the number of control rods or notches to insert.1307The

RO -ATC begins to drive 5 rods 2 notches into the core to the reduce heatuprate.-1 308The

RE question the

SM regarding the decision to insert control rods, and the SMtold the

RE that the insertion was needed to control the heat-up rate. There wasno further discussion.-1 309The Assistant Operations Manager (

AOM -Shift) cautioned the

SM that there wasthe potential to drive the reactor sub-critical by inserting control rods and that theyneeded to be careful. The

SM [[also recalled being concerned about the potential todrive the reactor sub-critical. The operating crew at the controls was not madeaware of these concerns.1310Control rod insertion is stopped. The control rods are now at the same position aswhen the reactor initially became critical; however, moderator temperature is now40"F higher than it was at initial criticality. The higher moderator temperature inconjunction with the control rod insertion rendered the reactor sub-critical, but theoperators were not aware of this.-1310The]]

SM left the control room to take a break, and the

AOM-Shift left the controlsarea to get his lunch in the control room kitchen.Attachment 3

A-3-3DETAILED

SEQUEN CE

OF [[]]

EVENTS TimeEvent-1311The operators range down the Intermediate Range Monitors (lRMs)two decadesfrom Range 8 to Range 6 in response to the lowering neutron flux.-1312The original

CRS returns from break and resumes duties as

CRS as well asresponsibility for the reactivity maneuver as the Reactivity SRO.1313After observing a O"F/hour heat-up rate, the

CRS directs the

RO [[-ATC to resumecontrol rod withdrawalto establish a positive heat-up rate. The RO-ATC begins towithdraw 5 rods 2 notches each to restore the heat-up rate.1315While notch withdrawing control rod 14-19 from position 08 to position 12, IRMreadings begin to rise again requiring the operators to range up on the lRMs inresponse to the rising neutron flux. The reactor has returned to a critical condition,but the operators are not aware of the change in reactor status with regards tocriticality.1316The RO-ATC notch withdraws control rod 22-43 from position 08 to position 12resulting in a more rapid rise in]]

IRM readings, The reactor period was calculated tobe 40 seconds during the post trip review.-1318The

RO -ATC attempts to notch withdraw control rod 30-11 from position 08 toposition 10 resulting in a sharp rise in

IRM readings.1318The reactor automatically scrammed on

IRM [[high-high flux level prior to completingthe withdrawal of rod 30-1 1 to position 10. Post event analysis determined that thereactor period was approximately 20 seconds, and that the scram occurred atapproximately 1.7o/o equivalent Average Power Range Monitor (APRM) power.-1320The]]

RE stated that he recognized that the operators had caused the reactor scramby withdrawing rods to criticality.1 345The crew debriefed the events leading up to the reactor scram.-1400The

RE participated in a conference call with the fuels group in Jackson (corporatereactor engineering staff) to discuss the event. The

RE informed the conferencecall participants that the reactor scram had been caused by human error.-1 600The

RE participated in a conference callwith General Electric (GE) to discuss theevent.-1 630The

RE informed the Director of Engineering that the reactor scram was caused byhuman error.-1 700The

RE informed the General Manager Plant Operations (GMPO) that the reactorscram was caused by human error. The

GMPO asked the

RE to draft a memodescribing what happened and send it to him.Attachment 3

A-3-4TimeEvent1730The

GMPO met with the Operations Manager (

OPS MGR) and the operatorsinvolved in the re-criticality to discuss the events.-1 900After shift turnover, the Assistant Operations Manager (AOM) recognized thathuman error was the cause of the scram. Equipment issues had been ruled out.-1 930To*2200The

GMPO recalls meeting with the

OPS MGR,

RE and corporate core designgroup to discuss issues associated with the scram. The

GMPO [[indicated that histeam was certain that the scram was caused by a human performance / knowledgedeficiency problem.-2330The Operations Manager (OPS MGR) prepared a written briefing for the crew onthe event.5t111110030An On-site Safety Review Committee (OSRC) conference callwas convened toreview the event and evaluate a recommendation to restart the reactor.01 30The]]

OSRC recommended restarting the reactor. The

GMPO was briefed regardingthe

OSRC recommendations.0200The

GMPO approved restarting the reactor. He directed the

OPS [[]]

MGR to call theNRC Senior Resident lnspector (SRl).0200The

OPS [[]]

MGR called the

SRI to inform him of the decision to restart the plant. The

OPS [[]]

MGR informed the

SRI that the cause of the scram was due to human error.0215The

SRI called the

NRC Region I Division of Reactor Projects (DRP) Branch Chiefto inform him of the decision to restart the plant. The

SRI then responded to thesite to observe the startup.-0300The reactor mode switch was placed in the startup position.-0300The

SRI arrives onsite.*0300The

DRP Branch Chief called the

GMPO to discuss the decision to restart thereactor.DETAILED SEQUENCE

OF [[]]

EVENTSAttachment 3

A-4-1IMC 0609,

APPEND [[IX M,Qualitative Decision-Making Attributes forTABLE 4.1NRC Management ReviewDecision AttributeApplicabletoDecision?Basis for Input to Decision - Provide qualitativeand/or quantitative information for managementreview and decision making.Finding can be boundedusing qualitative and/orquantitative information?NoIMC 0609 Appendix G is not appropriate since theconditions for reactor shutdown operations were notmet. The at-power safety Significance DeterminationProcess,]]

IMC 0609 Appendix A, quantitative analysismethodology is not adequate to provide reasonableestimates of the finding's significance. Furthermore, the

SDP [[does not model errors of commission and doesnot provide a method of accurately estimating changesto the human error probabilities caused for errors ofomission. As a result, no quantitative risk evaluationcan be performed for this finding.lmproper use and execution of procedures coupled withweak work control practices has the potential toincrease the human error probability (HEP) for creditedoperator actions. The probabilistic risk assessmentmodels are highly sensitive to small variations in HEPchanges. The existing]]

PRA research does not currentlysupport a method for varying the performance shapingfactors in response to defined error forcing contexts. ltis not possible to calculate a valid single point riskestimate. Human performance is a very largecontributor to

PRA [[uncertainty.Defense-in-Depthaffected?YesThe term "defense in depth" is commonly associatedwith the maintenance of the integrity and independenceof the three fission product barriers as well asemergency response actions. In addition, redundantand diverse safety systems, including trained licensedoperators conducting operations in accordance withapproved station procedures that were developedunder an approved quality control program are integralto maintaining a "defense in depth." While an automaticreactor scram was initiated as designed to protect thecore during this event, the fuel barrier was not actuallycompromised by the crew's actions since the automaticprotective action was successful.However, this performance deficiency revealedorganizational and human performance weaknesseswhich eroded defense in depth. The operating crewAttachment 4]]

IMC 0609,

APPEND IX M,

TABLE [[4.1plays a vital role in the maintenance of "defense indepth" from the perspective that they directly operatestation controls. Human errors can lead toconsequences that have the potential to compromisethe three fission product barriers. The commission ofmultiple unforeseen human errors in a short period oftime during the reactor startup degraded the operator'sperformance as an important "defense in depth" barrier.These operator human performance errors resulted in achallenge to the automatic Reactor Protection Systemwhich successfully terminated the event in thisparticular case.Performance Deficiencyeffect on the SafetyMargin maintained?YesThis performance deficiency had the potential toadversely affect the margin of safety. In this particularevent, the failure to implement conduct of operationsand reactivity control standards and procedures led to areactor protection set-point being exceeded, causing areactor scram. In fact, non-conservative operatoractions led to an unrecognized subcriticality followed byan unrecognized return to criticality. These operatoractions caused a rapid rise in neutron flux and reactorpower such that the]]

IRM Hl-Hl neutron flux reactor tripset point was exceeded resulting in an automaticreactor scram,In this case, the

IRM Hl-Hl neutron flux

RPS protectivefunction successfully terminated the event andprevented exceeding fuel barrier design safety marginand the potential for subsequent fuel barrier damage. ltshould also be noted that the Average Power RangeMonitor (APRM) Low Power

RPS set point wasavailable as a backup to the

IRM trip function. TheAPRM Low Power set point will initiate a reactor scramat less than or equal to 15% power whenever the modeswitch is

NOT in "

RUN".While there was no reduction in the quantitative designmargin, there was a qualitative reduction in the safetymargin as there is an expectation that the operators willmaintain an understanding of the status of the reactorand approach criticality in a deliberate and carefullycontrolled manner. ln this case, the operators lostsituational awareness regarding the status of thereactor and subsequently initiated incorrect actions thatled to an unrecognized subcriticality followed by anAttachment 4

A-4-3unrecognized return to criticality resulting in anautomatic reactor scram.The extent theperformance deficiencyaffects other eq uipment.YesThe inspectors reviewed the Entergy root causeevaluation team report and determined that theunderlying causes of this performance deficiency existacross the Operations organization, This includesweaknesses in oversight, human performancebehaviors, as well as operator knowledge, skills, andabilities deficiencies associated with low power reactorphysics and operations in the

IRM [[range. lt should benoted that the performance deficiency did not degradephysical plant equipment; however, the requirementthat licensed operators conduct licensed activities inaccordance with station approved procedures is integralto maintaining plant safety. Faulty operatorperformance has the potential to adversely affect plantequipment.Degree of degradation offailed or unavailablecomponent(s).N/]]

AN [[/APeriod of time (exposuretime) effect on theperformance deficiency.YesWith respect to the issues underlying this performancedeficiency, the exposure time is indeterminate, butclearly developed over an extended period of time.The Entergy root cause evaluation team determinedthat the causal factors for the event had existed for aconsiderable period of time, but they did not quantifythe exposure time, A number of condition reports werewritten over the last year, including a Fleet Assessmentperformed in February 2011, which identified shortfallsin oversight and adherence to conduct of operationshuman performance standards.This assessment is complicated by the fact that therewere not any apparent significant licensed operatorperformance issues at Pilgrim before this event. ln theHuman Performance cross-cutting area, none of theaspects currently has a theme, nor has there been atheme in the recent past. The behaviors outlined by theperformance deficiency have not been observed by theresident inspector staff prior to this event.IMC 0609, APPENDIX M,]]

TABLE 4.1Attachment 4

IMC 0609, APPENDIX M, TABLE 4.1The likelihood that thelicensee's recoveryactions wouldsuccessfully mitigate theperformance deficiency.YesAlthough "recovery actions" do not equate to "correctiveactions," this section lends itself to a discussion oflicensee corrective action in that completion of theseactions would mitigate the performance deficiency.The licensee's root cause analysis was thorough andappeared to identify all underlying causal factors. Theassociated proposed corrective actions appear toadequately address the undedying causal factors.Short term corrective actions have been completed tocorrect the specific issues associated with this event.Longer term corrective actions are in progress toaddress programmatic weakness in training and humanperformance behaviors.Additional qualitativecircumstancesassociated with thefinding that regionalmanagement shouldconsider in theevaluation process.YesIn this event, there were a significant number of lapsesin operator human performance fundamentals asdescribed in the conduct of operations and reactivitycontrol standards and procedures. These lapses inhuman performance fundamentals degraded individualoperator performance, crew performance, as well asmanagement oversight performance. The lack ofenforcement of, and adherence to, the conduct ofoperations and reactivity control standards andprocedures were identified as the root cause of thereactor scram event.The inspectors, as well as the Entergy root causeevaluation team, determined that the extent of conditionexisted across multiple crews of the Operationsdepartment and has the potential to exist across allPilgrim Nuclear Power Station departments.It should be noted that overall licensee operationalperformance has been acceptable. The plant runs well,and there are few bhallenges to the licensed operatorssince the plant tends to run reliably through theoperating cycle.The inspectors noted that licensee corrective actions tocorrect this performance deficiency prior to this eventwere ineffective, and that this pattern continued tomanifest itself immediately before the reactor scramand in the days immediately following the reactorscram. For example, the Entergy root cause teamidentified a number of condition reports that wereAttachment 4

A-4-5IMC 0609,

APPEND IX M,

TABLE [[4.1written over the past year that identified shortfalls inoversight and adherence to conduct of operationshuman performance standards, Corrective actionswere narrowly focused and failed to arrest thedegrading trend. Inspectors also noted that, during thestartup leading to the reactor scram, there werenumerous lapses in human performance fundamentalsand missed opportunities to correct those behavioraldeficiencies. lmmediately following the reactor scram,the licensee's post trip reviews and]]

OSRC [[reviewsfailed to fully evaluate the extent and scope of thehuman performance and knowledge deficiencies priorto authorizing the restart of the reactor. For instance,NRC inspectors identified that a control rod had beenmispositioned during the startup and that anlnfrequently Performed Test or Evolution (IPTE) briefinghad not been conducted during the initial andsubsequent startups. The control rod mispositioningand failure to perform the]]

IPTE [[briefing were notidentified by the licensee. In addition, in the daysimmediately following the event, inspectors continued toobserve a lack of formality in operator communications,a lack of apparent peer checking, and a number ofcontrol room distractions,While it will clearly take time to fully change thebehaviors associated with this performance deficiency,the inspectors did observe progress being made duringthe inspection. The licensee's Significant Event ReviewTeam (]]

SERT [[) and root cause analysis team performedthorough reviews of the event, and the licensee hasidentified a number of appropriate corrective actionsthat should correct the performance deficiency. Inaddition, licensee line personnel up through senior plantmanagement were interviewed extensively by theinspectors in the days and weeks following the event,and it appears as though the licensee has fullyinternalized the significance of this event.However, while progress is being made to correct theperformance deficiency, add itiona I follow-u pinspection(s) may be warranted to confirm the futureeffectiveness of the licensee's corrective actions.Attachment 4]]