ML112440100

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

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

See also: IR 05000293/2011012

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 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