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 UNITED STATES N UCLEAR REGU LATORY COMM ISSION REGION I 475 ALLENDALE ROAD KING OF PRUSSIA. PENNSYLVANIA 19406-1415 September 1, 2011 EA-11-174 Mr. Robert Site Vice President Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360-5508 PILGRIM NUCLEAR POWER STATION . NRC SPECIAL INSPECTION REPORT 05000293/2011012:

PRELIMINARY WHITE FINDING

Dear Mr. Smith:

On July 2Q,2011, the U.S. Nuclear Regulatory Commission (NRC) completed a Special Inspection at your Pilgrim Nuclear Power Station (PNPS). The inspection was conducted in response to the May 10,2011, reactor scram event that occurred due to an unrecognized subcriticality and subsequent unrecognized return to criticality.

The NRC's initial evaluation of this event satisfied the criteria in NRC Inspection Manual Chapter (lMC) 0309, "Reactive lnspection Decision Basis for Reactors," for conducting a Special Inspection.

The Special Inspection Team (SlT) Charter (Attachment 2 of the enclosed report) provides the basis and additional details concerning the scope of the inspection.

The enclosed inspection report documents 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 safety and compliance with Commission rules and regulations and with the conditions of your license.The inspection team reviewed selected procedures and records, observed activities, and interviewed personnel.

In particular, the inspection team reviewed event evaluations, causal investigations, relevant performance history, and extent of condition to assess the significance and potential consequences of issues related to the May 10 event.The inspection team concluded that the plant operated within acceptable power limits, and no equipment malfunctioned during the power transient and subsequent reactor scram.Nonetheless, the inspection team identified several issues related to human performance and compliance with conduct of operations and reactivity control standards and procedures that contributed to the event. The enclosed chronology (Attachment 3 of the enclosed report)provides additional details regarding the sequence of events.

R, Smith 2 This report documents one finding that, using the reactor safety Significance Determination Process (SDP), has preliminarily been determined to be White, or of low to moderate safety significance.

The finding involves the failure of Pilgrim personnel to implement conduct of operations and reactivity control standards and procedures during a reactor startup, which contributed to an unrecognized subcriticality followed by an unrecognized return to criticality and subseq uent reactor scram.This finding was assessed using NRC IMC 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria," because probabilistic risk assessment tools were not well suited 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 significance based on a qualitative assessment.

There was no significant impact on the plant following the transient because the event itself did not result in power exceeding license limits or fuel damage. Additionally, interim corrective actions were taken, which included removing the Pilgrim control room personnel involved in the event from operational duties pending remediation, providing additional training for operators not involved with the event, and providing increased management oversight presence in the Pilgrim control room while long term corrective actions were developed.

The finding involved one apparent violation (AV) of NRC requirements regarding Technical Specification 5.4, "Procedures," that is being considered for escalated enforcement action in accordance with the NRC's Enforcement Policy, which can be found on NRC's website at http://www.

nrc.qov/read inq-rom/doc-col lections/enforcemenU.

ln accordance with NRC IMC 0609, we will complete our evaluation using the best available information and issue our final determination of safety significance within 90 days of the date of this 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.

Before we make a final decision on this matter, we are providing you with an opportunity to (1) attend a Regulatory Conference where you can present to the NRC your perspective on the facts and assumptions the NRC used to arrive at the finding and assess its significance, or (2) submit your position on the finding to the NRC in writing. lf you request a Regulatory Conference, it should be held within 30 days of your response to this letter, and we encourage you to submit supporting documentation at least one week prior to the conference in an effort to make the conference more efficient and effective.

lf a Regulatory Conference is held, it will be open for public observation.

lf you decide to submit only a written response, such submittal should be sent to the NRC within 30 days of your receipt of this letter. lf you decline to request a Regulatory Conference or submit a written response, you relinquish your right to appeal the final SDP determination, in that by not doing either, you fail to meet the appeal requirements stated in 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 the issue date of this letter to notify the NRC of your intentions.

lf we have not heard from you within 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 is being issued for this inspection finding at this time. Please be advised that the number and characterization of the apparent violation described in the enclosed inspection report may change 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, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room and from the Publicly Available Records (PARS) component of NRC'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 (the Public Electronic Reading Room).Division of Reactor Safety Docket No. 50-293 License 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, Smith Because the NRC has not made a final determination in this matter, no Notice of Violation is being issued for this inspection finding at this time. Please be advised that the number and characterization of the apparent violation described in the enclosed inspection report may change 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, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room and from the Publicly Available Records (PARS) component of NRC'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 (the Public Electronic Reading Room).

Sincerely,/RA/Christopher G. Miller, Director Division of Reactor Safety Docket No. 50-293 License 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 ListServ Distribution:

See next page SUNSI Review Complete:

rrm* (Reviewer's Initials)DOCUMENT NAME: G:\DRS\Operations Branch\M0KINLE\lPilgrim SIT June 20'11\lnspection Report Drafts\Pilgrim SIT Concurrence\Pilgrim 2011 SIT Report Final.docx After declaring this document "An Official Agency Record" it will be released to the Public.MLI12440't00 To teceivs a coov of this documGnt.

indicate in the box: "C" = CoDy without attachmenvenclosure "E" = Copy with attachmenvenclosure

'N" = No OFFICE RIiDRS RI/DRS RI/ORA RI/DRP RI/DRP NAME RMcKinley/rrm*

Prior concurrence DJackson/dej*

Prior concurrence DHolody/aed for*Prior concurrence RBellamy/tcs for*Prior concurrence DRoberts/djr-

Prior concurrence DATE 08t19t11 08t19t11 08t19t11 08/ t11 08/30/1 1 OFFICE RI/DRS NAME CMiller/cgm DATE 08131111 OFFICIAL RECORD COPY Docket No.: License No.: Report No,: Licensee: Facility: Location: Dates: Team Leader: Team: Approved By: U. S. NUCLEAR REGULATORY COMMISSION REGION I 50-293 DPR-35 05000293/2011012 Entergy Nuclear Operations, Inc Pilgrim Nuclear Power Station (PNPS)600 Rocky Hill Road Plymouth, MA 02360 May 16 through July 20,2011 R. McKinley, Senior Emergency Response Coordinator Division of Reactor Safety B. Haagensen, Resident Inspector, Division of Reactor Projects D. Molteni, Operations Engineer, Division of Reactor Safety Donald E. Jackson, Chief Operations Branch Division of Reactor Safety Enclosure

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 safety significance was identified.

The significance of most findings is indicated by their color (Green, White, Yellow, or Red) using lnspection Manual Chapter (lMC) 0609, "Significance Determination 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 may be Green or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 4, dated December 2006.NRC ldentified and Self Revealing Findings

Cornerstone: Initiating

Events. Preliminary White: A self-revealing finding was identified involving the failure of Pilgrim personnel to implement conduct of operations and reactivity control standards and procedures during a reactor startup, which contributed to an unrecognized subcriticality followed 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 to moderate safety significance.

The finding is also associated with one apparent violation of NRC requirements specified by Technical Specification 5.4, "Procedures." There was no significant impact on the plant following the transient because the event itself did not result in power exceeding license limits or fuel damage. Additionally, interim corrective actions were taken, which included removing the Pilgrim control room personnel involved in the event from operational duties pending remediation, providing additional training for operators not involved with the event, and providing increased management oversight presence 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 its corrective 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 Performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations.

Specifically, the failure of Pilgrim personnel to effectively implement conduct of operations and reactivity control standards and procedures during a reactor startup caused an unrecognized subcriticality followed by an unrecognized return to criticality and subsequent reactor scram. Because the finding primarily involved multiple human performance errors, probabilistic risk assessment tools were not well suited for evaluating its significance.

The inspection team determined that the criteria for using IMC 0609, Appendix.M, "Significance Determination Process Using lll

Qualitative Criteria," were met, and the finding was evaluated using this guidance, as described in Attachment to this report. Based on the qualitative review of this finding, the NRC has preliminarily concluded that the finding was of low to moderate safety significance (preliminary White).The inspection team determined that multiple factors contributed to this performance deficiency, including:

inadequate enforcement of operating standards, failure to follow procedures, and ineffective operator training.

The Entergy RCE determined that the primary cause was a failure to adhere to established Entergy standards and expectations due to a lack of consistent supervisory and management enforcement.

The inspection team concluded that the finding had a cross-cutting aspect in the Human Performance cross-cutting area, Work Practices component, because Entergy did not adequately enforce human error prevention techniques, such as procedural adherence, holding pre-job briefs, self and peer checking, and proper documentation of activities during a reactor startup, which is a risk significant evolution.

Additionally, licensed personnel did not effectively implement the human performance prevention techniques mentioned above, and they proceeded when they encountered unceftainty and unexpected circumstances during the reactor startup H.4(a). (Section 2)iv

1.

REPORT DETAILS

Backoround and Description of Event In accordance with the Special Inspection Team (SlT) Charter (Attachment 2), the inspection team conducted a detailed review of the May 10, 2011, reactor scram event at Pilgrim Nuclear Power Station, including a review of the Pilgrim operators' response to the 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 technical documents to develop a detailed timeline of the event (Attachment 3).On May 10,2011, following a refueling outage, the reactor mode switch was taken to startup at 0626, and control rod withdrawal commenced at 0641. The control room crew consisted of the following personnel (additional licensed operators were present in the control room conducting various startup related activities):

o Assistant Operations Manager (AOM-Shift) - Senior Line Management oversight r Shift Manager (SM)- management oversight. Reactivity Senior Reactor Operator (SRO/Control Room Supervisor (CRS) -command and control o Assistant Control Room Supervisor (ACRS). Reactor Operator At-The-Controls (RO-ATC)o Reactor Operator (Verifier)

• ATC verifier r Reactor Engineer (RE). RE in Training At 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 was achieved at 1227. Once the POAH was achieved, the RO-ATC operator inserted rod 38-19 to position 10 to obtain lntermediate Range Monitor (lRM) overlap correlation data. Following the data collection, the RO-ATC operator withdrew rod 38-19 back to position 12.At approximately 1231, the Reactivity SRO/CRS and the RO-ATC operator were relieved by other licensed operators who continued with plant startup. The crew withdrew control rods to establish a moderator heat-up rate. The RO-ATC operator withdrew control rods 14-35, 38-35 , 14-19 and 22-43 from position 08 to 12 without incident.The RO-ATC operator continued with the rod withdrawal sequence and tried to withdraw control rod 30-11 from position 08 to 12, but the control rod would not move using normal notch withdraw commands.

The RO-ATC then attempted to withdraw control rod 30-11 using a "double-clutch" maneuver in accordance with procedures; however, the control rod inadvertently inserted and settled at position 06. As stated during interviews with the NRC inspectors, the RO-ATC operator, the ATC verifier, and the Reactivity SRO/CRS all saw the control rod in the incorrect position.

However, the operators did not enter and follow Pilgrim Nuclear Power Station (PNPS) Procedure 2.4.11, "Control Rod Positioning Malfunctions" as required.

This procedure required the operators to assess the amount of the mispositioning to determine the appropriate course of remedial 2 action before proceeding, and it also required the issue to be documented in a condition report. The operators did not perform an assessment, and they moved the control rod back to position 08 and ultimately to position 12, which was the correct final position in accordance with reactor engineering maneuvering instructions.

During interviews with the NRC inspectors, the three operators each indicated that there was confusion in their mind regarding whether or not the control rod met the definition of a mispositioned control rod because the control rod was only out of position by one notch from the initial position, but none of the operators referred to the procedure, and there was no discussion or challenge regarding the proper course of action among the operators.

The condition was not logged, and a condition report was not generated until the issue was identified by NRC inspectors.

In addition, the problem of the mispositioned control rod was not discovered by the licensee during the post trip review.Following withdrawal of the five control rods (ten control rod notches), the RO-ATC observed 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 believed corresponded to an hourly heat-up rate of 216"Flhr (the actual hourly heatup rate was 50"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 the crew at the controls to insert control rods to reduce the heat-up rate. This direction did not include specific guidance or limitations regarding the number of control rod notches to insert, At this point, the AOM-Shift and SM left the front panels area of the control room.The RE and RE-in-training were working at their computer terminals in the control room performing procedurally required calculations related to the startup. The REs had been occupied with these tasks from the time criticality had been achieved and had not been consulted 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 informed the RE, who in turn, questioned the SM about the decision to insert rods. The SM responded that the actions were necessary to control heat-up rate. No further discussion occurred between the SM and the RE regarding the number of control rods/notches to be used to control the heat-up rate or if there was a need to modify the reactor maneuvering plan. During interviews with the NRC inspectors, the SM and the AOM-Shift stated that they both discussed that there was a need to be careful to avoid taking the reactor subcritical and that the action of inserting control rods had the potential to cause the reactor to become subcritical.

However, this important information was never communicated to any of the operators at the controls, including at the time when the SM directed the at-the-controls crew to insert control rods to reduce the heat-up rate.As a result of the previous control rod withdrawal, moderator temperature was 40"F higher 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 the number of control rod notches to insert.Over the next three minutes, the RO-ATC operator proceeded to re-insert the following control rods from positions 12 to 8 (10 notches total) that had been previously withdrawn Enclosure 2.3 to establish the heat-up: 30-1 1 , 22-43, 14-19,38-35 and 14-35. At the end of the rod insertion evolution, the SM directed the Reactivity SRO/CRS and the RO-ATC operator to keep reactor power on IRM range 7. This communication was not acknowledged by the RO-ATC operator.

During interviews with the NRC inspectors, none of the operators recalled 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 the RO-ATC operator to range the lRMs down to range 7 and then to range 6. The reactor had 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-up rate, the RO-ATC operator observed the process computer displaying a 0'F/hr heat-up rate. At this time, the SRO who had previously been relieved, returned and re-assumed his role as Reactivity SRO/CRS. The Reactivity SRO/CRS and the RO-ATC operator decided 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 positions 8-12 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 operator continued rod withdrawal with control rod 22-43 from position 08 to 10. The RO-ATC operator and the Verifier ranged the lRMs up as reactor power increased.

The RO-ATC operator then withdrew control rod 22-43 from position 10 to 12. The operators did not recognize the increasing rate of change in IRM power.Finally, the RO-ATC operator selected and withdrew control rod 30-11 from position 8 to 10. At 1318, IRM readings rose sharply and an IRM Hi-Hi flux condition was experienced on both Reactor Protection System (RPS) channels resulting in an automatic reactor scram at approximately 1

.7 o/o reactor power.Operator Human Performance

Inspection Scope The inspection team interviewed the Pilgrim control room personnel that responded to the May 10,2011, event including the SM, AOM-Shift, CRS, ACRS, RO-ATC, RO verifier, and the REs to determine whether these personnel performed their duties in accordance with plant procedures and training.

The inspection team also reviewed narrative logs, sequence of events and alarm printouts, condition reports, PPC trend data, procedures implemented by the crew, and procedures regarding the conduct of operations.

a.Enclosure 4 b. Findinqs/Observations Failure to lmplement Procedures durinq Reactor Startup

Introduction:

A self-revealing finding was identified involving the failure of Pilgrim personnel to implement conduct of operations and reactivity control standards and procedures during a reactor startup, which contributed to an unrecognized subcriticality followed 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 to moderate safety significance.

The finding is also associated with one apparent violation of NRC requirements specified by Technical Specification 5.4, "Procedures."

Description:

On May 10,2011, following a refueling outage, operators were in the process of conducting a reactor startup. During the course of the startup, multiple licensed 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 and infrequently performed plant evolutions such as plant heat-up [and] startup." Additionally, the SM is responsible for ensuring "conservative actions are taken during unusual conditions

... when dealing with reactivity control," However, the SM did not oversee the activities in progress during reactor heatup and left the control room when the heat-up rate was being adjusted with control rod insertion, The SM did not ensure the actions taken to reestablish or adjust the reactor heatup rate were 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 regulatory requirements." PNPS procedure 1.3.34, "Operations Administrative Policies and Procedures," Revision 1 17, Section 6.10.1

.1 states, "All complex or infrequently

performed activities warrant a pre-evolution briefing." Section 6,10.1.1[8]

lists an Infrequently Performed Tests or Evolutions Briefing as one type of pre-evolution briefing, and Section 6.10.1

.1 [4] states, "lnfrequently

Performed Tests or Evolutions Briefings for the performance of Procedures classified as "lnfrequently Performed Tests or Evolutions" (IPTE) should be performed with Senior Line Manager oversight as specified in EN-OP-116, "lnfrequently Performed Tests or Evolutions." Entergy Procedure EN-OP-116, Revision 7, Attachment 9.1 identifies "Reactor Startup" as an IPTE. However, in this case, the licensee conducted a reactor startup without performing an IPTE briefing or any other type of pre-evolution briefing as defined in PNPS procedure 1.3.34. lt is noteworthy to point out that an IPTE briefing package was previously prepared, approved, and scheduled; however, the IPTE briefing was never performed as required by the procedures described above. In addition, an IPTE briefing was also not performed for the startup following this event. Finally, the CRSs did not ensure the administrative requirements of the conduct of operations procedures or the regulatory requirement to implement the control rod mispositioning procedure were met. This issue was identified by the NRC inspectors.

5 Entergy procedure EN-OP-115, "Conduct of Operations," Revision 10, Section 5.2, states control room operators are required to "develop and implement a plan that includes contingencies and compensatory measures" and when implementing those plans the "crew ... continuously evaluates the plan for changing conditions" and"Human Performance (HU) tools (..., peer/cross-checking, oversight, questioning attitude, etc.) are utilized ..." In addition, "When the control room team is faced with a time critical decision:

Use all available resources...do not proceed in the face of uncertainty..." However, the control room operators failed to develop contingency plans or compensatory measures for adjusting reactor heat-up rate or addressing higher than expected reactor heat-up rates. The crew also failed to develop or implement contingencies for control rods which were difficult to maneuver when they were at low reactor power. Additionally, the use of human performance tools was ineffective in addressing the actions or conditions that led to the unexpected reactor heatup rate and the mispositioning of control rod 30-11. Specifically, failures in the use of peer checking and questioning the conditions that led to the unexpected reactor heat-up rate directly contributed to the mispositioned control rod and the subsequent reactor scram. Lastly, the control room team did not use all available resources by involving Reactor Engineering staff in its decision-making, and proceeded in the face of uncertainty by failing to consider the consequences of the reactivity changes.Entergy procedure EN-OP-115, "Conduct of Operations," Revision 10, Section 5.4 states that reactor operators are expected to perform reactivity manipulations "in a deliberate, carefully controlled manner while the reactor is monitored to ensure the desired result is obtained." However, the reactor operators did not adequately monitor the conditions of the reactor while attempting to establish and adjust the reactor heat-up rate. Although the reactor operators were watching the response of both the lRMs and the computer point displaying a five minute average reactor heatup, they were moving control rods faster than the plant temperature could respond and therefore taking actions to continue control rod movement before the desired result of their manipulations could be assessed.

Additionally, after inserting control rods to adjust the reactor heat-up rate, the operators had sufficient indications that the reactor was significantly subcritical as evidenced by the required ranging down of lRMs, the drop in Source Range Monitor (SRM) count rates, and establishing a negative reactor period. The operator's failure to adequately monitor the status of the reactor led to an unrecognized subcritical condition and subsequent return 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 (either short-term or for the remainder of the shift) will be recorded in the CRS log." lt further states, "...a verbal discussion of plant status and off-normal conditions must be conducted." However, several people in watch standing positions changed from the start of the shift, but none of those changes were entered into the control room log. In addition, when the ACRS was turning over to the CRS, there was no discussion 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 a position other than the intended position $ a control rod that moves more than one notch beyond its intended position." Attachment 4 Step [3] and Step [a] of the same procedure requires the operators to assess the degree of mispositioning and take the appropriate remedial action depending on the degree of mispositioning.

4 Step [5] also states, "lf the control rod is determined to be mispositioned, then record the event as a condition report." In this case, the RO-ATC attempted to withdraw control rod 30-11 from position 08 to position 10 (intended position), but the rod inadvertently insertbd to position 06. Upon recognizing the error, the operators did not enter the procedure when control rod 30-11 was found to be left in a position other than the intended position and which was more than one notch from the intended position.

The operators did not assess the amount of the control rod mispositioning in accordance with the procedure, nor was there any discussion about the mispositioning on the crew. Furthermore, the event was not logged, nor was a condition report generated.

Instead, the operators did not enter and follow the procedure, and they continued on with the startup in the face of uncertainty.

This issue was not detected during the licensee posttrip review. lt was identified by the NRC inspectors.

o PNPS Procedure 2.1.1, "Startup from Shutdown," Revision 173, Page 53, Caution 2 states, "ln the event the reactor goes subcritical after achieving initial criticality, then return to step [53] and re-perform the steps to restore the Reactor to a critical condition." 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 initial criticality, 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 not recognize that the reactor had become subcritical and did not re-perform the procedural steps mentioned above to restore the reactor to a critical condition in a controlled manner under the guidance of Reactor Engineering.

There was sufficient information available to the operators to identify that the reactor had become subcritical.

In addition, REs were available in the control room, but they were not consulted by the operators.

Analvsis:

The inspection team determined that the failure of Pilgrim personnel to implement conduct of operations and reactivity control standards and procedures during a reactor startup was a performance deficiency that was reasonably within Entergy's ability to foresee and prevent. The finding is more than minor because it was associated with the Human Performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations.

Specifically, the failure of Pilgrim personnel to effectively implement conduct of operations and reactivity control standards and procedures during a reactor startup caused an unrecognized subcriticality followed by an unrecognized return to criticality and subsequent reactor scram.Enclosure 7 The inspection team determined that multiple factors contributed to this performance deficiency including:

inadequate enforcement of operating standards, failure to follow procedures, and ineffective operator training.

The Entergy RCE documented that the primary cause was a failure to adhere to established Entergy standards and expectations due to a lack of consistent supervisory and management enforcement.

In addition, the Entergy RCE specified a number of condition reports and self assessment reports written in the months preceding this event that demonstrated that the performance deficiency existed over an extended period of time and affected all operating crews. While the performance deficiency manifested itself during this particular low power event, there was the potential for the performance deficiency to result in a more consequential event under different circumstances.

Because the finding primarily involved multiple human performance errors, probabilistic risk assessment tools were not well suited for evaluating its significance.

The inspection team determined that the criteria for using IMC 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria," were met, and the finding was evaluated using this guidance as described in Attachment 4 to this report. Based on the qualitative review of this finding, the NRC concluded that the finding was preliminarily of low to moderate safety significance (preliminary White). The completed Appendix M table is attached to this report (Attachment 4). There was no significant impact on the plant following the transient because the event itself did not result in power exceeding license limits or fuel damage. Additionally, interim corrective actions were taken, which included removing the Pilgrim control room personnel involved in the event from operational duties pending remediation, providing additional training for operators not involved with the event, and providing increased management oversight presence in the Pilgrim 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 not adequately enforce human error prevention techniques, such as procedural adherence, holding pre-job briefs, self and peer checking, and proper documentation of activities during a reactor startup, which is a risk significant evolution.

Additionally, licensed personnel did not effectively implement the human performance prevention techniques mentioned above, and they proceeded when they encountered uncertainty and unexpected circumstances during the reactor startup [H.a(a)].Enforcement:

Technical Specification 5.4, "Procedures," states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix "A" of Regulatory Guide (RG) 1.33, February, 1978. RG 1.33, Appendix "A," requires that typical safety-related activities listed therein be covered by written procedures.

Contrary to the above, on May 10,2011, as reflected in the examples listed in the description section of this finding, the licensee failed to implement 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 BWR Systems." Enclosure 3.I Following a review of the event, the licensee documented the condition in the corrective action program (CR-PNP-2011-2475).

There was no significant impact on the plant following the transient because the event itself did not result in power exceeding license limits or fuel damage. Additionally, interim corrective actions were taken, which included removing the Pilgrim control room personnel involved in the event from operational duties pending remediation, providing additional training for operators not involved with the event, and providing increased management oversight presence in the Pilgrim control room while long term corrective actions were developed.

Pending determination of final safety significance, this finding with the associated apparent violation will be tracked as AV 05000293/2011012-01, Failure to lmplement Gonduct of Operations and Reactivity Gontrol Procedures during Reactor Startup.Fitness for Dutv Inspection Scope The inspection team interviewed the control room personnel that were directly involved with the May 10,2011, reactor scram event as well as management personnel involved with the immediate post event investigation.

The inspection team also reviewed Entergy Fitness for Duty (FFD) program requirements contained in the corporate and site procedures.

Fi nd i nos/Observations No findings were identified.

Traininq Inspection Scope The inspection team interviewed personnel, reviewed simulator modeling and performance, and reviewed training material related to Just in Time Training (JITT)material for the initial and subsequent startups, remedial training for the operators involved with the event, and training plans for startups and reactivity maneuvers.

Fi nd i nqs/Observations No findings were identified.

The inspection team observed that the JITT training that was provided prior to the initial startup was very limited in scope in that it only covered the approach to criticality up to the POAH. lt did not cover the full range of reactor heat-up, and it covered very little Operating Experience.

In addition, several operators that were directly involved with this event did not attend the JITT training including the SM, the ACRS who temporarily relieved the CRS prior to the scram, and the RO who was at the controls when the scram occurred.a.h 4.a.b.Enclosure 5.I Orqanizational Response lmmediate Response Inspection Scope The inspection team interviewed personnel, reviewed various procedures and records, and observed control room operations to assess immediate response of station personnel to the reactor scram event.Fi nd i nqs/Observations No findings were identified.

The inspection team observed that Entergy's initial response to the event was not appropriately 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 the event despite the fact that multiple IRM channels accurately tracked reactor power along with operator reactivity inputs. lmmediate post event interviews with the crew did not probe human error as a potential cause even though the SM, the AOM-Shift, and the REs had expressed concerns just prior to the scram regarding the insertion of control rods so near the point of criticality.

Operators involved with the event were dismissed for the day as the investigation continued to incorrectly focus on equipment malfunction as the most likely cause of the event. Several hours passed before it became clear to site management that human error was the cause of the event. As a result, the operators involved with the event were not thoroughly interviewed to ensure that all of the human performance aspects were fully understood prior to proceeding with the next startup. In addition, the inspection team identified that the posttrip review failed to identify that a control rod had been mispositioned just prior to the scram and that an IPTE briefing had not been conducted for the startup. Consequently, additional human performance issues were not evaluated, and the licensee again failed to perform an IPTE briefing prior to the subsequent startup as required by Entergy procedures.

Post-Event Root Cause Evaluation and Actions Inspection Scope The inspection team reviewed Entergy's Root Cause Evaluation (RCE) report for the event to determine whether the causes and associated human performance issues were properly identified.

Additionally, the inspection team assessed whether interim and planned long term corrective actions were appropriate to address the cause(s).61 a.b.5.2 a.Enclosure b.10 Find inqs/Observations No findings were identified.

The RCE was thorough and appeared to identify the underlying causal factors. The associated proposed corrective actions appeared to adequately address the underlying causal factors. Entergy identified the root cause as a lack of consistent supervisory and management enforcement of administrative procedure requirements and management expectations for command and control, roles and responsibilities, reactivity manipulations, clear communications, proper briefings, and proper turnovers.

The RCE also identified contributing causes including weaknesses in monitoring plant status and parameters as well as weaknesses in operator proficiency with regards to low power operations.

Meetinqs.

Includinq Exit Exit Meetino Summarv On 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 that proprietary information reviewed during the inspection period was returned to Entergy.40A6 Enclosure Enterov Personnel R. Smith J. Dreyfuss D. Noyes J. Macdonald R. Probasco J. Couto S. Anderson T. Tomon J. Byron J. Hayhurst S. Bethay J. Lynch T. White F. McGinnis R. Byrne V. Fallacara S. Reininghaus J. House V. Magnatta R. Paranjape A,1-1

=SUPPLEMENTAL

INFORMATION=

KEY POINTS OF CONTACT

Site Vice President General Manager Plant Operations

Manager, Operations

Assistant

Manager, Operations

Shift Manager, Operations

Shift Supervisor, Operations

Shift Supervisor, Operations

Reactor Operator, Operations

Reactor Operator, Operations

Reactor Operator, Operations

Director, Nuclear Safety Assurance Manager, Licensing Manager, Quality Assurance Engineer, Licensing Senior Engineer, Licensing Director, Engineering

Manager, Training Supervisor, Operations

Training Lead lnstructor, Operations

Training Reactor Engineer

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Ooened

05000293/2011012-01

AV Failure to lmplement

Conduct of Operations

and Reactivity

Control Procedures

during Reactor Startup (Section 2)

LIST OF DOCUMENTS

REVIEWED Procedures:

1.3.34, "Operations Administrative policies and Procedures," Revision 1 19 1 .3.37 , "Post-Trip Reviews," Revision 27 1.3,63, "Conduct of Event Review Meetings," Revision 25 1.3.109, "lssue Management," Revision 8 2.1.1, "Startup from Shutdown," Revision 173 2.1.4, "Approach to Critical," Revision 26 2.1.7, "Vessel Heat-up and Cool Down," Revision 54 2.4.11, "Control Rod Positioning Malfunctions," Revision 35 2.4.11.1, "CRD System Malfunctions," Revision 21 Attachment

A-1-2 SUPPLEMENTAL

INFORMATION

NOP96A3, "Reactivity Management Peer Panel," Revision 10

EN-FAP-AD-OO1, "Fleet Administrative

Procedure

(FAP) Process," Revision 0

EN-FAP-OM-006, "Working Hour Limits for Non-Covered Workers," Revision 2

EN-FAP-OP-008, "Reactivity Management Performance Indicator Program," Revision 0

EN-FAP-OP-01

1, "Operator Human Performance Indicator Program," Revision 0

EN-HU-102, "Human Performance Tools," Revision 5

EN-HU-103, "Human Performance Error Reviews," Revision 4

EN-NS-102, "Fitness for Duty Program," Revision 9

EN-OM-119, "On-Site Safety Review Committee," Revision 7

EN-OM-123, "Fatigue Management Program," Revision 3

EN-OP-103, "Reactivity Management Program," Revision 5

EN-OP-1 15, "Conduct of Operations," Revision 10

EN-OP-1 16, "lnfrequently Performed Tests of Evolutions," Revision 7

EN-RE-214, "Conduct of Reactor Engineering," Revision 0

EN-RE-215, "Reactivity Maneuver Plan," Revision 1

EN-RE-219, "Startup sequence Criticality Controls (BWR)," Revision 0 Condition Reports:

CR-PNP-2011-02475

and associated Root Cause Evaluation Report, Revision 1

CR-PNP-201

1-02488 cR-PNP-2011-02493

cR-PNP-2011-02504

CR-PNP-201

1-02506 CR-PNP-2011-02546

CR-PNP-201

1-02568 CR-PNP-2011-02572

cR-PNP-2011-02577

CR-PNP-201

1-03598 Self Assessments:

LO-PNPLO-2009-00071, "Focused Assessment on Reactivity Management"

LO-PNPLO-2010-00106, "Snapshot Assessment on Reactivity Management

Procedure

Revision lmplementation"

LO-PNPLO-2010-00106, "Snapshot Assessment on SOER 07-01 Recommendation Reactivity Management 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 Time Training used for

0511012011

and

0511112011

Startup JITT Instructional Module, Reactor Startup and Criticality May 2011 Just in Time Training used for 051

1812011 Startup JITT Attachment

A-1-3 SUPPLEMENTAL

INFORMATION

Just in Time Training PowerPoint used for 05/1812011

Startup JITT lnstructor Lesson Plan JITT

RFO 18 Hydro 2.1 .8.5 Simulator

JITT Reactor Shutdown 2.1.5 and Vessel Cooldown 2.1.7 , Revised 0410112011

Simulator

JITT Reactor Shutdown 2.1.5 and Vessel Cooldown 2.1 .7 , Revised 0211912011

Training Schedules for Outage Training Cycle

0311412011

-0410712011

Training Schedules for Training Cycle 020211312011

-0211712011

Training Schedules for Training Cycle 01

1112212010 - 0112212011

Training Records and Remediation Training for Current Licensed Operators lnitial License Class 2009-2011

Class Schedule O-RO-03-02, "Reactor Plant Startup Certification Unit Guide," Revision 10 O-RO-03-01-19, "Reactivity Management and Control Instructor/Student Guide," Revision 2 O-RO-03-01

-20, "Simulator Scenario, Operations Standards," Revision 0 O-RO-03-02-01, "lnstructional Module - Day One Cold Reactor Startup," Revision 7 O-RO-03-02-02, "lnstructional Module - Day Two Hot Reactor Startup," Revision 7 O-RO-03-02-03, "lnstructional Module * Day Three Cold Reactor Startup," Revision 3 O-RO-03-02-04, "lnstructional Module - Day Four Hot Reactor Startup," Revision 3 O*RO-03-02-05, "lnstructional Module - Day Five Cold Reactor Startup," Revision 3 O-RO-03-02-06, "lnstructional Module - Day Six Cold Reactor Startup," Revision 3 O-RO-03-02-07, "lnstructional Module - Day Seven 905 Certification Practice," Revision 3 O-RO-03-02-08, "lnstructional Module * Day Eight 905 Certification Practice," Revision 2 O-RO-03-02-09, "lnstructional Module - Day Nine Reactor Power Operations," Revision 1 O-RO-03-02-51, "lnstructional Module - SOER 90-3 Nuclear Instrument Miscalibration," Revision 3 Miscellaneous:

Crew Briefing Sheet from May 10,2011 SCRAM Operations Section Standing Order 11-03 OSRC Meeting 2011-008 Meeting Minutes Post-Trip Review Package from May 10,2011 SCRAM with Attachments and Supporting Data"EN-OP-116

9,3 ITPE Supplemental Controls," developed for Post-Refueling Outage Startup Reactor Engineer's calculations pertaining to criticality prior to the reactor SCRAM eSOMS Control Room Logs from

0510912011

through 0511112011

SRM and Moderator Temperature Traces with Calculated

SRM Period 0511012011

Control Room Personnel Chart Dayshift 0511012011

Control Rod Notch History from Reactor Critical to Reactor SCRAM 0511012011

Control Rod Notch Worth Calculations for 05/1012011

Reactor Startup Power Maneuver Plan Cycle 19-01 Attachment

A-1-4 SUPPLEMENTAL

INFORMATION

LIST OF ACRONYMS

ACRS Assistant

Control Room Supervisor

ADAMS Agency-wide

Documents

Access and Management

System AOM Assistant

Operations

Manager

ATC At the Controls

AV Apparent Violation

BOP Balance of Plant

CCDP Conditional

Core Damage Probability

CFR Code of Federal Regulations

CR Condition

Report

CRD Control Rod Drive

CRS Control Room Supervisor

DRP Division of Reactor Projects

DRS Division of Reactor Safety

FFD Fitness for Duty

HEP Human Error Probability

HPCI High Pressure Coolant Injection

HUR Heatup Rate IMC lnspection

Manual Chapter IPTE Infrequently

Performed

Tests or Evolutions

IRM Intermediate

Range Monitor

JITT Just in Time Training

NRC Nuclear Regulatory

Commission

OPS [[]]

MGR Operations

Manager PARS Publicly Available

Records PD Performance

Deficiency

PNPS Pilgrim Nuclear Power Station

POAH Point of Adding Heat

PPC Plant Process Computer

PRA Probabilistic

Risk Assessment

RCE Root Cause Evaluation

RCIC Reactor Core lsolation

Cooling

RE Reactor Engineer

RG Regulatory

Guide

RO Reactor Operator

RO-ATC Reactor Operator at the Controls RPS Reactor Protection

System SDP Significance

Determination

Process

SM Shift Manager

SRI Senior Resident Inspector

SRM Source Range Monitor

SRO Senior Reactor Operator SIT Special Inspection

Team STA Shift Technical

Advisor TS Technical

Specification

A-2-1

SPECIA L INSPECTION

TEAM [[]]

CHARTE R

UNITED [[]]

STATES N

UCLEAR [[]]

REGULATORY

COMMIS SION

REGION I 475

ALLEND ALE

ROAD [[]]

KING [[]]

OF [[]]

PRUSSI A.

PA 19406-1415

MEMORANDUM

TO SPECIAL INSPECTION

TEAM [[]]

CHARTER May 13, 2011 Samuel L. Hansell Jr., Manager Special Inspection

Team Raymond R. McKinley, Leader Special Inspection

Team Christopher

G. Miller, Director /

RA /Division of Reactor Safety Darrell

J. Roberts, Director /

RA by Paul Krohn Acting For/Division of Reactor Projects SPECIAL INSPECTION

TEAM [[]]

CHARTE R -PILGRIM NUCLEAR

POWER [[]]

STATIO N OPERATOR PERFORMANCE

DURING [[]]

REACTO R STARTUP

ON [[]]

MAY 1Q.2011 FROM: SUBJECT: In accordance

with lnspection

Manual Chapter (lMC) 0309, "Reactive

Inspection

Decision Basis for Reactors," a Special Inspection

Team (SlT) is being chartered

to evaluate operator performance

and organizational

decision-making

associated

with a reactor scram that occurred during a startup on May 10,2011, The decision to conduct this special inspection

was based on meeting the deterministic

criteria (the event involved questions

or concerns pertaining

to licensee operational

performance)

and risk criteria specified

in Enclosure

of IMC 0309. The calculable

increase in conditional

core damage probability (CCDP), which was in the low E-6 range, was based on application

of an Initiating

Event Analysis in Sapphire 8 due to the reactor scram, 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 resident inspectors

and augmented

by a Division of Reactor Projects (DRP) inspector

who was dispatched

to the site soon after the event. The Team will review the causes of the event, and Entergy's

organizational

and operator response during and after the event, The Team will Attachment

t rt *.r. i

A-2-2

SPECIA L INSPECTION

TEAM [[]]

CHARTER perform interviews, as necessary, to understand

the scope of operator actions performed

during the event. The Team will also assess whether the SIT should be upgraded to an Augmented Inspection

Team in accordance

with IMC 0309.The inspection

will be conducted

in accordance

with the guidance contained

in NRC Inspection

Procedure

93812, "Special Inspection," and an inspection

report will be issued within 45 days following

the final exit meeting for the inspection.

The Special Inspection

willcommence

on May 16, 2411. The following

personnel

have been assigned to this effort: Manager: Samuel L. Hansell, Jr., Branch Chief Operations

Branch, DRS, Region I Team Leader: Team Members: Enclosure:

Special Inspection

Team Charter Raymond R. McKinley, Senior Emergency

Response Coordinator

Plant Support Branch, DRS, Region I Brian C. Haagensen, Millstone

Power Station Resident Inspector Division of Reactor Projects, DRP, Region I David L. Molteni, Operations

Engineer Operations

Branch, DRS, Region I Attachment

A-2-3

SPECIA L INSPECTION

TEAM [[]]

CHARTER Special Inspection

Team Charter Pilgrim Nuclear Power Station Operator Performance

During Reactor Startup May 10,2011 Backqround:

During startup from a refueling

outage, Entergy operators

withdrew rods to criticality

the afternoon

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 higher than 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 Control Room Supervisor (CRS) and Shift Manager (SM) aware of the condition

and proceeded

to insert five control rods (two notches each) to lower the HUR to approximately

65"F/hr. At the time, it was not identified

by the operators, reactor engineers

or management

oversight

in the control room that the control rod insertions

brought the reactor to a subcritical

state (approximately

0.35% subcritical

by later calculations).

After reducing the HUR, the operators (without recognition

of the subcritical

reactor condition), proceeded

to withdraw the five control rods back to their previous position.

While withdrawing

the fifth control rod back to its original position, the reactor experienced

a full SCRAM on Intermediate

Range Monitor (lRM) Hl-Hl flux signals. All rods inserted and equipment

responded

as expected.Pilgrim initially

investigated

potential

equipment

related causes for the automatic scram as communicated

to the NRC on the afternoon

of May 10,2011. Subsequent

analysis revealed that human performance

errors made by the operators

were the cause of the scram. NRC was informed of this in the early morning hours of May 11,2011. Entergy is continuing

its investigation

of the operator actions taken during this event. Entergy suspended

the qualifications

of the operators

and the Shift Manager directly involved with the event while the investigation

continues.

Additional

actions have been taken by Entergy that include more restrictive

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-up to 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 questions or concerns pertaining

to licensee operational

performance.

This criterion

was met based on human 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

appropriate

focus on power level throughout

the startup event;. Reactor operators

and control room supervision

did not have proper sensitivity

for the impacts from negative reactivity

insertions

with the reactor at low power conditions;

A-2-4

SPECIA L INSPECTION

TEAM [[]]

CHARTER. The operators

did not identify or utilize available

plant indications

that indicated

the reactor was subcritical;. Reactor operators

did not follow shift manager instructions

to maintain reactor power within the current IRM power band while addressing

the elevated HUR;. Operators

and control room supervision

did not engage reactor engineering

staff with regard to planned rod movement after the reactor was made subcritical;

and o Prior to the identification

of the unexpected

HUR, reactor operators

did not implemenVenter

the required abnormal operating

procedure

for a mispositioned

control rod (Rod 30-1 1).In accordance

with IMC 0309, the event was evaluated

for risk significance

because one deterministic

criterion

was met, A Region I SRA evaluated

the transient (reactor scram)from

low reactor power using the Initiating

Event Assessment

feature of Saphire 8. The lE-Trans basic event probability

was set to 1.0 and all other initiating

events were set to zero. The resulting

dominant core damage sequences

were subsequently

evaluated

by the SRA to account for the low reactor power conditions

and alternating

current (AC) power being supplied by 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 between an SIT and No Additional

inspection

required.

The dominant core damage sequences

involve failure of direct current (DC) power sources and failure of residual heat removal. However, with the low decay heat load following

the refuel outage, these core damage sequences

represent

a conservative

estimate of risk.Additionally, this event involved multiple licensed operators

not recognizing

the reactivity

status of an operating

reactor during startup and demonstrating

a poor understanding

of reactor physics in a low power condition.

In light of the aforementioned

human performance

errors, and consistent

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 operator response during and following

the event. The Team will perform interviews, as necessary, to understand

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 by Entergy, and the sequence of communications

within Entergy and to the NRC subsequent

to the event;2. Review and assess crew operator performance

and crew decision making, including adherence

to expected roles and responsibilities, the use of the command and control elements associated

with reactivity

manipulations, the use of procedures, the use of diverse instrumentation

to assess plant conditions, response to alarms and overall implementation

of operations

department

and station standards;

A-2-5

SPECIA L INSPECTION

TEAM [[]]

CHARTER Evaluate 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 requalification

training program;Review the adequacy of the preparation

by the operations

staff for the reactor startup including

training prior to the evolution

and briefings

by the operations

staff.Review the adequacy of the simulator

to model the behavior of the current reactor core during startup activities

and the current adequacy of the simulator

for use in reactor startup training ;Assess the decision making and actions taken by the operators

and station management

during the initial and subsequent

reactor startup to determine

if there are any implications

related to safety culture;Review and assess the effectiveness

of Entergy's

response to this event and corrective

actions taken to date. This includes overall organizational

response, and adequacy of immediate, interim and proposed longterm corrective

actions. This will also 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

and procedures

when a human performance

error has occurred;10. Evaluate Entergy's

application

of pertinent

industry operating

experience, including

INPO [[]]

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

INPO [[]]

SOER 07-1, "Reactivity

Management," and other recent events involving

reactivity

management

errors to assess the effectiveness

of any actions taken in response to the operating experience;

and 11. Document the inspection

findings and conclusions

in a Special Inspection

Team final report within 45 days of inspection

completion.

Guidance: Inspection

Procedure

93812, "Special Inspection", provides additional

guidance to be used by the SlT. Team duties will be as described

in Inspection

Procedure

93812. The inspection

should emphasize

fact-finding

in its review of the circumstances

surrounding

the event. Safety concerns identified

that are not directly related to the event should be reported to the Region I office for appropriate

action.The Team will conduct an entrance meeting and begin the inspection

on May 16,2011. While on-site, the Team Leader will provide daily briefings

to Region I management, who will coordinate

with the Office of Nuclear Reactor Regulation

to ensure that all other pertinent parties are kept informed.

The Team will also coordinate

with the Region I State Liaison Officer Attachment

3.4.5.6.7.8.

A-2-6

SPECIA L INSPECTION

TEAM [[]]

CHARTER to implement

the Memorandum

of Understanding

between the NRC and the State of Massachusetts

to offer observation

of the inspection

by representatives

of the state. A report documenting

the results of the inspection

will be issued within 45 days following

the final exit meeting for the inspection.

Before the end of the first day onsite, the Team Manager shall provide a recommendation

to the Regional Administrator

as to whether the SIT should continue or be upgraded to an Augmented Inspection

Team response.This Charter may be modified should the Team develop significant

new information

that warrants review.Attachment

A,3-1

DETAIL ED SEQUENCE

OF [[]]

EVENTS May 10,2011, Reactor Scram Event The team constructed

the sequence of events from a review of control room narrative

logs, plant process computer (PPC) data (alarm printout, sequence of event printout, plant parameter graphs) and plant personnel

interviews.

Time Event 05/09/11 Two Sessions Just In Time Training (JITT) was conducted

for the reactor startup. Certain key members of the operating

crew that were directly involved with this event were not present for the training including

the Shift Manager (SM), the Assistant

Control Room 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 1 0626 The reactor mode switch was moved to the startup position.-0630 The oncoming day shift operators

received a reactor maneuvering

plan briefing.The Reactor Engineers (REs) led the brief.0641 Operators

commenced

control rod withdrawal.

0700 The day shift operating

crew assumed the shift, and control rod withdraw continues.

212 The reactor became critical.1227 The point of adding heat was reached.-1231 The

CRS was relieved for lunch by the

ACRS. The oncoming CRS providing

the relief did not receive Just In Time Training (JITT), nor did he participate

in the reactor maneuvering

plan briefing.-1231 The RO-ATC was relieved for lunch by the Licensed Operator previously

assigned as the ATC verifier.

The oncoming RO-ATC providing

the relief did not receive Just In Time Training (JITT), but he did participate

in the reactor maneuvering

plan briefing.-1231 A Licensed Operator previously

assigned to other startup activities

was reassigned

to fill the role of ATC verifier.

This individual

received JITT training, and he also received a separate reactor maneuvering

plan briefing from a RE upon arriving to work at approximately

100.1246 The RO-ATC withdrew 5 rods 2 notches to establish

a heat-up rate.Attachment

A-3-2

DETAIL ED SEQUENCE

OF [[]]

EVENTS Time Event 1255 The RO-ATC attempts to withdraw control rod 30-11 from position 08 to position 10 using single notch control, but the control rod does not move, The RO-ATC raised drive water pressure and attempted

several notch withdraw commands, but the control rod failed to move.1257 The RO-ATC attempts to withdraw control rod 30-1 1 from position 08 to position 10 using a "double clutch" maneuver, but the control rod incorrectly

inserted one notch to position 06. The RO-ATC does not discuss the control rod mispositioning

error with the crew.1257 The

ATC verifier and

CRS also saw control rod 30-11 move incorrectly

to position 06, but the control rod mispositioning

error is not discussed.

1302 The RO-ATC then withdraws

control rod 30-11 from position 06 to position 12.-1 305 The crew observes that the 5 minute average reactor coolant heat-up rate is 18"F over the 5 minute period, and the crew determines

that this corresponded

to a 216'Flhour

heat-up rate. In actuality, the 5 minute average heat-up rate reflected the 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.-1 306 The

SM directed the

RO -ATC to insert control rods to reduce the heat-up rate, but the

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

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

RE question the

SM regarding

the decision to insert control rods, and the

SM told the

REs that the insertion

was needed to control the heat-up rate. There was no further discussion.

-1 309 The Assistant

Operations

Manager (AOM-Shift)

cautioned

the SM that there was the potential

to drive the reactor sub-critical

by inserting

control rods and that they needed to be careful. The SM also recalled being concerned

about the potential

to drive the reactor sub-critical.

The operating

crew at the controls was not made aware of these concerns.1310 Control rod insertion

is stopped. The control rods are now at the same position as when the reactor initially

became critical;

however, moderator

temperature

is now 40"F higher than it was at initial criticality.

The higher moderator

temperature

in conjunction

with the control rod insertion

rendered the reactor sub-critical, but the operators

were not aware of this.-1310 The

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

AOM-Shift

left the controls area to get his lunch in the control room kitchen.Attachment

A-3-3

DETAIL ED SEQUENCE

OF [[]]

EVENTS Time Event-1311 The operators

range down the Intermediate

Range Monitors (lRMs)two

decades from Range 8 to Range 6 in response to the lowering neutron flux.-1312 The original

CRS returns from break and resumes duties as

CRS as well as responsibility

for the reactivity

maneuver as the Reactivity

SRO.1313 After observing

a O"F/hour heat-up rate, the

CRS directs the

RO-ATC to resume control rod withdrawalto

establish

a positive heat-up rate. The RO-ATC begins to withdraw 5 rods 2 notches each to restore the heat-up rate.1315 While notch withdrawing

control rod 14-19 from position 08 to position 12, IRM readings begin to rise again requiring

the operators

to range up on the lRMs in response 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 to criticality.

1316 The RO-ATC notch withdraws

control rod 22-43 from position 08 to position 12 resulting

in a more rapid rise in IRM readings, The reactor period was calculated

to be 40 seconds during the post trip review.-1318 The RO-ATC attempts to notch withdraw control rod 30-11 from position 08 to position 10 resulting

in a sharp rise in IRM readings.1318 The reactor automatically

scrammed on IRM high-high

flux level prior to completing

the withdrawal

of rod 30-1 1 to position 10. Post event analysis determined

that the reactor period was approximately

seconds, and that the scram occurred at approximately

1.7o/o equivalent

Average Power Range Monitor (APRM) power.-1320 The RE stated that he recognized

that the operators

had caused the reactor scram by withdrawing

rods to criticality.

345 The crew debriefed

the events leading up to the reactor scram.-1400 The RE participated

in a conference

call with the fuels group in Jackson (corporate

reactor engineering

staff) to discuss the event. The RE informed the conference

call participants

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

in a conference

callwith General Electric (GE) to discuss the event.-1 630 The RE informed the Director of Engineering

that the reactor scram was caused by human error.-1 700 The

RE informed the General Manager Plant Operations (

GMPO ) that the reactor scram was caused by human error. The

GMPO asked the

RE to draft a memo describing

what happened and send it to him.Attachment

A-3-4 Time Event 1730 The GMPO met with the Operations

Manager (OPS MGR) and the operators involved in the re-criticality

to discuss the events.-1 900 After shift turnover, the Assistant

Operations

Manager (AOM) recognized

that human error was the cause of the scram. Equipment

issues had been ruled out.-1 930 To*2200 The

GMPO recalls meeting with the

OPS MGR, RE and corporate

core design group to discuss issues associated

with the scram. The GMPO indicated

that his team was certain that the scram was caused by a human performance

/ knowledge deficiency

problem.-2330 The Operations

Manager (OPS MGR) prepared a written briefing for the crew on the event.5t11111 0030 An On-site Safety Review Committee (OSRC) conference

callwas convened to review the event and evaluate a recommendation

to restart the reactor.01 30 The OSRC recommended

restarting

the reactor. The

GMPO was briefed regarding the

OSRC recommendations.

200 The GMPO approved restarting

the reactor. He directed the

OPS [[]]

MGR to call the

NRC Senior Resident lnspector (

SR l).0200 The

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.0215 The

SRI called the

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

to the site to observe the startup.-0300 The reactor mode switch was placed in the startup position.-0300 The

SRI arrives onsite.*0300 The

DRP Branch Chief called the

GMPO to discuss the decision to restart the reactor.

DETAIL ED SEQUENCE

OF [[]]

EVENTS Attachment

A-4-1

IMC 0609,

APPENDIX M, Qualitative

Decision-Making

Attributes

for

TABLE 4.1

NRC Management

Review Decision Attribute Applicable

to Decision?Basis for Input to Decision - Provide qualitative

and/or quantitative

information

for management

review and decision making.Finding can be bounded using qualitative

and/or quantitative

information?

No IMC 0609 Appendix G is not appropriate

since the conditions

for reactor shutdown operations

were not met. The at-power safety Significance

Determination

Process, IMC 0609 Appendix A, quantitative

analysis methodology

is not adequate to provide reasonable

estimates

of the finding's

significance.

Furthermore, the SDP does not model errors of commission

and does not provide a method of accurately

estimating

changes to the human error probabilities

caused for errors of omission.

As a result, no quantitative

risk evaluation

can be performed

for this finding.lmproper use and execution

of procedures

coupled with weak work control practices

has the potential

to increase the human error probability (HEP) for credited operator actions. The probabilistic

risk assessment

models are highly sensitive

to small variations

in

HEP changes. The existing

PRA research does not currently support a method for varying the performance

shaping factors in response to defined error forcing contexts.

lt is not possible to calculate

a valid single point risk estimate.

Human performance

is a very large contributor

to PRA uncertainty.

Defense-in-Depth

affected?Yes The term "defense in depth" is commonly associated

with the maintenance

of the integrity

and independence

of the three fission product barriers as well as emergency

response actions. In addition, redundant and diverse safety systems, including

trained licensed operators

conducting

operations

in accordance

with approved station procedures

that were developed under an approved quality control program are integral to maintaining

a "defense in depth." While an automatic reactor scram was initiated

as designed to protect the core during this event, the fuel barrier was not actually compromised

by the crew's actions since the automatic protective

action was successful.

However, this performance

deficiency

revealed organizational

and human performance

weaknesses

which eroded defense in depth. The operating

crew Attachment

IMC 0609,

APPENDIX M, TABLE 4.1 plays a vital role in the maintenance

of "defense in depth" from the perspective

that they directly operate station controls.

Human errors can lead to consequences

that have the potential

to compromise

the three fission product barriers.

The commission

of multiple unforeseen

human errors in a short period of time during the reactor startup degraded the operator's

performance

as an important "defense in depth" barrier.These operator human performance

errors resulted in a challenge

to the automatic

Reactor Protection

System which successfully

terminated

the event in this particular

case.Performance

Deficiency

effect on the Safety Margin maintained?

Yes This performance

deficiency

had the potential

to adversely

affect the margin of safety. In this particular

event, the failure to implement

conduct of operations

and reactivity

control standards

and procedures

led to a reactor protection

set-point

being exceeded, causing a reactor scram. In fact, non-conservative

operator actions led to an unrecognized

subcriticality

followed by an unrecognized

return to criticality.

These operator actions caused a rapid rise in neutron flux and reactor power such that the IRM Hl-Hl neutron flux reactor trip set point was exceeded resulting

in an automatic reactor scram, In this case, the

IRM Hl-Hl neutron flux

RPS protective

function successfully

terminated

the event and prevented

exceeding

fuel barrier design safety margin and the potential

for subsequent

fuel barrier damage. lt should also be noted that the Average Power Range Monitor (APRM) Low Power RPS set point was available

as a backup to the IRM trip function.

The

APRM Low Power set point will initiate a reactor scram at less than or equal to 15% power whenever the mode switch is

NOT in "RUN".While there was no reduction

in the quantitative

design margin, there was a qualitative

reduction

in the safety margin as there is an expectation

that the operators

will maintain an understanding

of the status of the reactor and approach criticality

in a deliberate

and carefully controlled

manner. ln this case, the operators

lost situational

awareness

regarding

the status of the reactor and subsequently

initiated

incorrect

actions that led to an unrecognized

subcriticality

followed by an Attachment

A-4-3 unrecognized

return to criticality

resulting

in an automatic reactor scram.The extent the performance

deficiency

affects other eq uipment.Yes The inspectors

reviewed the Entergy root cause evaluation

team report and determined

that the underlying

causes of this performance

deficiency

exist across the Operations

organization, This includes weaknesses

in oversight, human performance

behaviors, as well as operator knowledge, skills, and abilities

deficiencies

associated

with low power reactor physics and operations

in the IRM range. lt should be noted that the performance

deficiency

did not degrade physical plant equipment;

however, the requirement

that licensed operators

conduct licensed activities

in accordance

with station approved procedures

is integral to maintaining

plant safety. Faulty operator performance

has the potential

to adversely

affect plant equipment.

Degree of degradation

of failed or unavailable

component(s).

N/A N/A Period of time (exposure time) effect on the performance

deficiency.

Yes With respect to the issues underlying

this performance

deficiency, the exposure time is indeterminate, but clearly developed

over an extended period of time.The Entergy root cause evaluation

team determined

that the causal factors for the event had existed for a considerable

period of time, but they did not quantify the exposure time, A number of condition

reports were written over the last year, including

a Fleet Assessment

performed

in February 2011, which identified

shortfalls

in oversight

and adherence

to conduct of operations

human performance

standards.

This assessment

is complicated

by the fact that there were not any apparent significant

licensed operator performance

issues at Pilgrim before this event. ln the Human Performance

cross-cutting

area, none of the aspects currently

has a theme, nor has there been a theme in the recent past. The behaviors

outlined by the performance

deficiency

have not been observed by the resident inspector

staff prior to this event.IMC 0609,

APPEND IX M,

TABLE 4.1 Attachment

IMC 0609, APPENDIX M, TABLE 4.1 The likelihood

that the licensee's

recovery actions would successfully

mitigate the performance

deficiency.

Yes Although "recovery

actions" do not equate to "corrective

actions," this section lends itself to a discussion

of licensee corrective

action in that completion

of these actions would mitigate the performance

deficiency.

The licensee's

root cause analysis was thorough and appeared to identify all underlying

causal factors. The associated

proposed corrective

actions appear to adequately

address the undedying

causal factors.Short term corrective

actions have been completed

to correct the specific issues associated

with this event.Longer term corrective

actions are in progress to address programmatic

weakness in training and human performance

behaviors.

Additional

qualitative

circumstances

associated

with the finding that regional management

should consider in the evaluation

process.Yes In this event, there were a significant

number of lapses in operator human performance

fundamentals

as described

in the conduct of operations

and reactivity

control standards

and procedures.

These lapses in human performance

fundamentals

degraded individual

operator performance, crew performance, as well as management

oversight

performance.

The lack of enforcement

of, and adherence

to, the conduct of operations

and reactivity

control standards

and procedures

were identified

as the root cause of the reactor scram event.The inspectors, as well as the Entergy root cause evaluation

team, determined

that the extent of condition existed across multiple crews of the Operations

department

and has the potential

to exist across all Pilgrim Nuclear Power Station departments.

It should be noted that overall licensee operational

performance

has been acceptable.

The plant runs well, and there are few bhallenges

to the licensed operators since the plant tends to run reliably through the operating

cycle.The inspectors

noted that licensee corrective

actions to correct this performance

deficiency

prior to this event were ineffective, and that this pattern continued

to manifest itself immediately

before the reactor scram and in the days immediately

following

the reactor scram. For example, the Entergy root cause team identified

a number of condition

reports that were Attachment

A-4-5

IMC 0609,

APPENDIX M, TABLE 4.1 written over the past year that identified

shortfalls

in oversight

and adherence

to conduct of operations

human performance

standards, Corrective

actions were narrowly focused and failed to arrest the degrading

trend. Inspectors

also noted that, during the startup leading to the reactor scram, there were numerous lapses in human performance

fundamentals

and missed opportunities

to correct those behavioral

deficiencies.

lmmediately

following

the reactor scram, the licensee's

post trip reviews and OSRC reviews failed to fully evaluate the extent and scope of the human performance

and knowledge

deficiencies

prior to authorizing

the restart of the reactor. For instance, NRC inspectors

identified

that a control rod had been mispositioned

during the startup and that an lnfrequently

Performed

Test or Evolution (IPTE) briefing had not been conducted

during the initial and subsequent

startups.

The control rod mispositioning

and failure to perform the IPTE briefing were not identified

by the licensee.

In addition, in the days immediately

following

the event, inspectors

continued

to observe a lack of formality

in operator communications, a lack of apparent peer checking, and a number of control room distractions, While it will clearly take time to fully change the behaviors

associated

with this performance

deficiency, the inspectors

did observe progress being made during the inspection.

The licensee's

Significant

Event Review Team (SERT) and root cause analysis team performed thorough reviews of the event, and the licensee has identified

a number of appropriate

corrective

actions that should correct the performance

deficiency.

In addition, licensee line personnel

up through senior plant management

were interviewed

extensively

by the inspectors

in the days and weeks following

the event, and it appears as though the licensee has fully internalized

the significance

of this event.However, while progress is being made to correct the performance

deficiency, add itiona I follow-u p inspection(s)

may be warranted

to confirm the future effectiveness

of the licensee's

corrective

actions.Attachment

4