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{{Adams
#REDIRECT [[IR 05000293/2011012]]
| number = ML112440100
| issue date = 09/01/2011
| title = IR 05000293-11-012, on 05/16/2011 - 07/20/2011, Pilgrim Nuclear Power Station, Inspection Procedure 93812, Special Inspection
| author name = Miller C
| author affiliation = NRC/RGN-I/DRS
| addressee name = Smith R
| addressee affiliation = Entergy Nuclear Operations, Inc
| docket = 05000293
| license number = DPR-035
| contact person =
| case reference number = EA-11-174
| document report number = IR-11-012
| document type = Inspection Report, Letter
| page count = 37
}}
See also: [[see also::IR 05000293/2011012]]
 
=Text=
{{#Wiki_filter:,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 G. Smith 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.
R. Smith 3 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
TABLE OF CONTENTS SUMMARY OF FlNDlNGS ...........
............
iii REPORT DETAILS..
.................1
1. Background
and Description
of Event .........1
2. Operator Human Performance...,.....,.....
......................3
3. Fitness for Duty.....
....................,8
4. Training....
,................8
5. Organizational
Response............, ...............9
4OAO Meetings, Including
Exit,..........
...,.............10
ATTACHMENT
1 . SUPPLEMENTAL
INFORMATION........
.......,,,..,...,A-1-1
KEY POINTS OF CONTACT LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED LIST OF DOCUMENTS
REVIEWED LIST OF ACRONYMS ATTACHMENT
2. SPECIAL INSPECTION
TEAM CHARTER .....,.,.,...4-2-1
ATTACHMENT
3. DETAILED SEQUENCE OF EVENTS ...A-3-1 ATTACHMENT
4 - IMC 0609 APPENDIX M. TABLE 4.1............
..........A-4-1
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 Enclosure
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 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 Enclosure
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 .7o/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.
Enclosure
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.
Attachment
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
1
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
4 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
1
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
Attachment
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
1
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
Attachment
1
A-2-1 SPECIAL INSPECTION
TEAM CHARTER UNITED STATES N UCLEAR REGULATORY
COMMISSION
REGION I 475 ALLENDALE
ROAD KING OF PRUSSIA. 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 CHARTER -PILGRIM NUCLEAR POWER STATION OPERATOR PERFORMANCE
DURING REACTOR 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
1 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
2 t rt *.r. i
A-2-2 SPECIAL 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
2
A-2-3 SPECIAL 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;
Attachment
2
A-2-4 SPECIAL 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;
Attachment
2
A-2-5 SPECIAL 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
2 3.4.5.6.7.8.
A-2-6 SPECIAL 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
2
A,3-1 DETAILED 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.
1212 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
1 100.1246 The RO-ATC withdrew 5 rods 2 notches to establish
a heat-up rate.Attachment
3
A-3-2 DETAILED 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 REs 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
3
A-3-3 DETAILED 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
20 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.
1 345 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
3
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.
0200 The GMPO approved restarting
the reactor. He directed the OPS MGR to call the NRC Senior Resident lnspector (SRl).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.DETAILED SEQUENCE OF EVENTS Attachment
3
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
4
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
4
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, APPENDIX M, TABLE 4.1 Attachment
4
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
4
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
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