IR 05000293/2011012
| 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:
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
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
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:
- System" 5.4.1, "PROCEDURES" Traininq Material: lnstructional Module, Reactor Startup and Criticality
- 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
- 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
CCDP Conditional
Core Damage Probability
CR Condition
Report
HUR Heatup Rate IMC lnspection
Manual Chapter IPTE Infrequently
Performed
Tests or Evolutions
IRM Intermediate
Range Monitor
NRC Nuclear Regulatory
Commission
- OPS [[]]
MGR Operations
Manager PARS Publicly Available
Records PD Performance
Deficiency
PRA Probabilistic
Risk Assessment
RCIC Reactor Core lsolation
Cooling
RG Regulatory
Guide
RO-ATC Reactor Operator at the Controls RPS Reactor Protection
System SDP Significance
Determination
Process
SRO Senior Reactor Operator SIT Special Inspection
Team STA Shift Technical
Advisor TS Technical
Specification
A-2-1
- TEAM [[]]
- UNITED [[]]
- UCLEAR [[]]
REGULATORY
- ROAD [[]]
- KING [[]]
- OF [[]]
MEMORANDUM
- TEAM [[]]
CHARTER May 13, 2011 Samuel L. Hansell Jr., Manager Special Inspection
Team Raymond R. McKinley, Leader Special Inspection
Team Christopher
- TEAM [[]]
- POWER [[]]
- DURING [[]]
- 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
- 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
- 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
- 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
- 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 [[]]
- 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
- 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
- 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
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
- 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
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 regarding
the decision to insert control rods, and 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
AOM-Shift
left the controls area to get his lunch in the control room kitchen.Attachment
A-3-3
- 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 as well as responsibility
for the reactivity
maneuver as the Reactivity
SRO.1313 After observing
a O"F/hour heat-up rate, 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 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
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
OSRC recommendations.
200 The GMPO approved restarting
the reactor. He directed the
- OPS [[]]
- OPS [[]]
- OPS [[]]
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
- OF [[]]
EVENTS Attachment
A-4-1
APPENDIX M, Qualitative
Decision-Making
Attributes
for
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
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
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
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,
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
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