IR 05000247/2006004
| ML063170143 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 11/13/2006 |
| From: | Cobey E Reactor Projects Branch 2 |
| To: | Dacimo F Entergy Nuclear Operations |
| cobey e w | |
| References | |
| FOIA/PA-2007-0080 IR-06-004 | |
| Download: ML063170143 (36) | |
Text
November 13, 2006 Mr. Fred Site Vice President
Entergy Nuclear Operations, Inc.
Indian Point Energy Center
295 Broadway, Suite 1
P.O. Box 249
Buchanan, NY 10511-0249SUBJECT:INDIAN POINT NUCLEAR GENERATING UNIT 2 - NRC INTEGRATED INSPECTION REPORT NO. 05000247/2006004
Dear Mr. Dacimo:
On September 30, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Indian Point Nuclear Generating Unit 2. The enclosed integrated inspection report
documents the inspection results, which were discussed on October 4, 2006, with
Mr. Paul Rubin and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed
personnel.
Based on the results of this inspection, two findings of very low safety significance (Green) were identified. These findings did not involve violations of NRC requirements.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure, and your response (if any) will be available electronically for public inspection in the
NRC Public Document Room or from the Public ly Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is accessible from the NRC Web Site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/Eugene W. Cobey, Chief Reactor Projects Branch 2
Division of Reactor Projects F. Dacimo2 Docket No. 50-247 License No. DPR-26 Enclosure: Inspection Report No. 05000247/2006004 w/Attachment: Supplemental Information cc w/encl:
G. J. Taylor, Chief Executive Officer, Entergy Operations
M. R. Kansler, President, Entergy Nuclear Operations Inc. (ENO)
J. T. Herron, Senior Vice President and Chief Operations Officer (ENO)
C. Schwarz, Vice President, Operations Support (ENO)
P. Rubin, General Manager Operations (ENO)
O. Limpias, Vice President, Engineering (ENO)
J. McCann, Director, Licensing (ENO)
C. D. Faison, Manager, Licensing (ENO)
M. J. Colomb, Director of Oversight (ENO)
J. Comiotes, Director, Nuclear Safety Assurance (ENO)
P. Conroy, Manager, Licensing (ENO)
T. C. McCullough, Assistant General Counsel, Entergy Nuclear Operations, Inc.
P. R. Smith, President, New York State Energy, Research and Development Authority
P. Eddy, Electric Division, New York State Department of Public Service
C. Donaldson, Esquire, Assistant Attorney General, New York Department of Law
D. O'Neill, Mayor, Village of Buchanan
J. G. Testa, Mayor, City of Peekskill
R. Albanese, Four County Coordinator
S. Lousteau, Treasury Department, Entergy Services, Inc.
Chairman, Standing Committee on Energy, NYS Assembly
Chairman, Standing Committee on Environmental Conservation, NYS Assembly
Chairman, Committee on Corporations, Authorities, and Commissions
M. Slobodien, Director, Emergency Planning
B. Brandenburg, Assistant General Counsel
Assemblywoman Sandra Galef, NYS Assembly
County Clerk, Westchester County Legislature
A. Spano, Westchester County Executive
R. Bondi, Putnam County Executive
C. Vanderhoef, Rockland County Executive
E. A. Diana, Orange County Executive
T. Judson, Central NY Citizens Awareness Network
M. Elie, Citizens Awareness Network
D. Lochbaum, Nuclear Safety Engineer, Union of Concerned Scientists
Public Citizen's Critical Mass Energy Project
M. Mariotte, Nuclear Information & Resources Service
F. Zalcman, Pace Law School, Energy Project
L. Puglisi, Supervisor, Town of Cortlandt
Congresswoman Sue W. Kelly
Congresswoman Nita Lowey
Senator Hillary Rodham Clinton
Senator Charles Schumer
J. Riccio, Greenpeace
A. Matthiessen, Executive Director, Riverkeeper, Inc.
M. Kaplowitz, Chairman of County Environment & Health Committee
SUMMARY OF FINDINGS
IR 05000247/2006004; 07/01/2006 - 09/30/2006; Indian Point Nuclear Generating Unit 2;
Event Follow Up.
The report covered a three-month period of inspection by resident inspectors and regional specialist inspectors. Two Green findings were identified. The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process" (SDP). 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 3, dated July 2000. A.
NRC-Identified and Self-Revealing Findings
Cornerstone: Initiating Events
- Green.
A Green self-revealing finding was identified because Entergy failed to develop adequate procedures for governing the response to a loss of both heater drain tank pumps and to an approaching rod insertion limit (RIL) alarm condition.
Specifically, the procedure governing operator actions during a loss of heater drain tank pumps did not specify for the operators to reset the steam dumps following the rapid downpower. The alarm response procedure for the approaching rod insertion limit condition directed the operators to place the rod control system in manual to stop further automatic inward rod motion. This impacted operators ability to add negative reactivity and control the transient.
Entergy entered these procedural deficiencies into their corrective action program and is evaluating the appropriate steps to correct the procedural deficiencies.
The inspectors determined that this finding is greater than minor because it is associated with the Procedure Quality attribute of the Initiating Events cornerstone; and, it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions.
Specifically, the procedural inadequacies complicated operator actions to a rapid downpower, resulted in a manual reactor trip when the operators determined that they did not have sufficient control of the transient, and could impact other accident sequences requiring negative reactivity addition. The inspectors evaluated this finding using Phase I of IMC 0609, Appendix A, "Significance
Determination of Reactor Inspection Findings for At-Power Situations," and determined it to be of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be unavailable. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that plant operating procedures were adequate to ensure operators could appropriately respond to a rapid downpower transient. (Section 4OA3.1)
- Green.
A Green self-revealing finding was identified because Entergy failed to develop an accurate procedure for calibration of the steam dump loss of load controller. This resulted in the steam dumps failing to operate properly during a plant transient, complicating operator response, and leading to a manual reactor v trip. Following identification of the issue, Entergy entered the issue into the corrective action program, corrected the procedural deficiency, and re-calibrated the controller.
The inspectors determined that this finding is greater than minor because it is associated with the Procedural Quality attribute of the Initiating events cornerstone; and, it impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions.
Specifically, the inadequacy in Entergy's calibration procedure caused the steam dumps to operate improperly during a plant transient and contributed to a reactor trip. The inspectors evaluated this finding using Phase I of IMC 0609,
Appendix AProperty "Inspection Manual Chapter" (as page type) with input value "NRC Inspection Manual 0609,</br></br>Appendix A" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process., "Significance Determination of Reactor Inspection Findings for
At-Power Situations," and determined it to be of very low safety significance because it did not contribute to the likelihood of both a reactor trip and the likelihood that mitigation equipment or functions would be available. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that the procedure for calibration of the steam dump loss of load controller was accurate, in that, it specified incorrect settings for the controller. (Section 4OA3.2)
B. Licensee-Identified Violations
None.
REPORT DETAILS
===Summary of Plant Status
Indian Point Nuclear Generating Unit 2 began the inspection period operating at full power. On August 23, 2006, the reactor was manually tripped following the loss of both heater drain tank
pumps and a malfunction in the steam dump valve control system. Full power was restored on August 25, 2006, and the plant continued to operate at or near full power for the remainder of
the inspection period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity1R01Adverse Weather Protection (71111.01 - 2 samples)
a. Inspection Scope
The inspectors completed the following two adverse weather protection samples.
- Because thunderstorms with potential tornados were forecast in the vicinity of the facility for July 12, 2006, the inspectors reviewed Entergy's preparations for
inclement weather conditions. The inspectors walked down portions of the
service water system, the gas turbines, and emergency diesel generators.
These systems were selected because their safety-related functions could be
affected by adverse weather. The inspectors reviewed documents listed in the and observed plant conditions, evaluating those conditions using
criteria documented in OAP-008, "Severe Weather Preparations," Revision 0.
The inspectors also toured the plant grounds for loose debris, which could
become missiles during a tornado, and ascertained if operators could access
controls and indications for those systems required for safe control of the plant.*The inspectors reviewed and verified Entergy's completion of the operations department warm weather preparation checklist contained in procedure OAP-
008, "Severe Weather Preparations." The inspectors reviewed the procedural
limits and actions associated with elevated temperatures, walked down
accessible areas of plant structures to assess the effectiveness of the ventilation
systems, and reviewed calculations supporting equipment operability. The
walkdowns included discussions with operations and engineering personnel to
ensure that they were aware of temperature restrictions and required actions.
The documents reviewed as part of this inspection are listed in the Attachment.
b. Findings
No findings of significance were identified.
2 Enclosure1R04Equipment Alignment (71111.04Q - 3 samples)
a. Inspection Scope
The inspectors performed three partial system walkdowns to verify the operability of redundant or diverse trains and components duri ng periods of system train unavailability or following periods of maintenance. The inspectors referenced the system procedures, the Updated Final Safety Analysis Report (U FSAR), and system drawings in order to verify that the alignment of the available train was proper to support its required safety
functions. The inspectors also reviewed applicable condition reports and work orders to
ensure that Entergy had identified and properly addressed equipment discrepancies that
could potentially impair the capability of the available train. The documents reviewed are
listed in the Attachment. The inspectors performed a partial walkdown of the following
systems which represents three inspection samples: *21 and 23 component cooling water pump trains with 22 component cooling water pump out of service for maintenance;*21 and 23 safety injection pump trains with the 22 safety injection pump out of service for maintenance; and*21 and 23 emergency diesel generators with the 22 emergency diesel generator out of service for maintenance.
b. Findings
No findings of significance were identified.
1R05 Fire Protection (71111.05Q - 7 samples)
a. Inspection Scope
The inspectors conducted tours of the seven areas listed below to assess the material condition and operational status of fire protection features. The inspectors verified that
combustibles and ignition sources were controlled in accordance with Entergy's
administrative procedures; fire detection and suppression equipment was available for
use; that passive fire barriers were maintained; and that compensatory measures for
out-of-service, degraded, or inoperable fire protection equipment were implemented in
accordance with Entergy's fire plan. The inspectors used procedure ENN-DC-161, "Transient Combustible Program," in performing the inspection. The inspectors
evaluated the fire protection program against the requirements of license condition 2.k.
The documents reviewed are listed in the Attachment.
3 Enclosure This inspection satisfied seven inspection samples for fire protection tours. The areas inspected included: *Fire zone 1;*Fire zone 9;
- Fire zone 10;
- Fire zones 23 and 62A;
- Fire zone 650;
- Fire zone 6A; and
- Fire zones 22 and 63A.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures
(71111.06 - 1 sample)
a. Inspection Scope
The inspectors reviewed Indian Point Unit 2's Individual Plant Examination of External Events and the UFSAR concerning external flooding events. The inspection included a
walkdown of accessible areas of the plant to look for potential susceptibilities to external
flooding and to verify the assumptions included in the site's external flooding analysis.
The inspectors also reviewed relevant abnormal operating and emergency plan
procedures. The documents reviewed are listed in the Attachment. This inspection
represented one inspection program sample.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance (71111.07A - 1 sample)
a. Inspection Scope
The inspectors reviewed the testing and evaluation of test results for the containment fan cooler units. Procedure 2-PT-Q16, "Containment Fan Cooler Unit Cooling Water Flow
Test," Revision 0, is performed on a quarterly basis to verify safety-related unit cooler
flow requirements. In addition, Entergy conducts an inspection of each unit every six
years. Visual and eddy current inspections are conducted. The inspectors reviewed
performance data to verify that heat exchanger operation was consistent with design.
The inspectors evaluated heat sink performance against Generic Letter 89-13, "Service
Water System Problems Affecting Safety Related Equipment," and the UFSAR. The documents reviewed are listed in the Attachment. This review represents one inspection sample.
b. Findings
No findings of significance were identified.
4 Enclosure1R11Licensed Operator Requalification Program (71111.11Q - 1 sample, 71111.11S - 1sample).1Quarterly Inspection (1 sample)===
a. Inspection Scope
===On August 29, 2006, the inspectors observed licensed operator simulator training to assess operator performance during several scenarios to verify that operator
performance was adequate and that evaluators were identifying and documenting crew
performance problems. The inspectors evaluated the performance of risk significant
operator actions, including the use of emergency operating procedures. The inspectors
assessed the clarity and effectiveness of communications, the implementation of
appropriate actions in response to alarms, the performance of timely control board
operation and manipulation, and the oversight and direction provided by the shift
manager. The inspector also reviewed simulator fidelity to evaluate the degree of
similarity to the actual control room. Licensed operator training was evaluated against
the requirements of 10 CFR 55, "Operators' Licenses." The documents reviewed are
listed in the Attachment. This observation of operator simulator training constituted one
inspection sample.
b. Findings
No findings of significance were identified..2Biennial Inspection (1 sample)===
a. Inspection Scope
The following inspection activities were performed using NUREG-1021, Revision 9,"Operator Licensing Examination Standards for Power Reactors," Inspection Procedure 71111.11, "Licensed Operator Requalification Program," NRC Inspection
Manual Chapter 0609, Appendix I, "Operator Requalification Human Performance
Significance Determination Process (SDP)," and 10 CFR 55.46 Simulator Rule (sampling basis) as acceptance criteria.
The inspectors reviewed documentation of operating history since the last requalification program inspection. The inspectors also discussed facility operating events with the
resident staff. Documents reviewed included NRC inspection reports, plant performance
insights, licensee event reports (LERs), and licensee condition reports (CRs) that
involved human performance issues for licensed operators to ensure that operational
events were not indicative of possible training deficiencies.
The inspectors reviewed three exam sets (i.e., weeks 2, 5, and 6) for both the
comprehensive reactor operator and senior reactor operator biennial written exams, as
well as scenarios and job performance measures administered during this current exam cycle to ensure the quality of these exams met or exceeded the criteria established in
the Examination Standards and 10 CFR 55.59.
5 Enclosure During the onsite weeks of this inspection, the inspectors observed the administration of operating examinations to operating crew s "B" and "C." The operating examinations consisted of three simulator scenarios for each crew and five job performance measures
administered to each individual. The documents that were reviewed are listed in the
.
Conformance with Simulator Requirements Specified in 10 CFR 55.46 For the site specific simulator, the inspectors observed simulator performance during the conduct of the examinations, and discrepancy reports to verify compliance with the
requirements of 10 CFR 55.46.
The inspectors reviewed simulator maintenance, testing and control procedures and discussed simulator maintenance, testing, configuration control and machine operation
with members of the simulator maintenance staff. The inspectors also reviewed a
sample of simulator tests including transients, core performance, computer real time, steady state, and scenario-based tests and verified that a sample of completed simulator
deficiency item condition reports from the past two-year period effectively addressed the
described issues. The simulator tests reviewed are listed in the Attachment.
Conformance with operator license conditions was verified by reviewing the following records:*A sample of two remediation training packages for the past two-year training cycle was reviewed. *The inspectors conducted a review of proficiency watch-standing and reactivation records including a sample of licensed operator reactivation records and a
random sample of watch-standing documentation for time on shift to verify
currency and conformance with the requirements of 10 CFR 55.
Licensee's Feedback System. The inspectors interviewed instructors, training and operations management personnel, and four operators (i.e., two reactor operators and
two senior reactor operators) for feedback regarding the implementation of the licensed
operator requalification program to ensure the requalification program was meeting their
needs and responsive to their noted deficiencies and recommended changes.
The results of the annual operating exam were assessed to ensure that pass rates were consistent with the guidance of NRC Manual Chapter 0609, Appendix I, "Operator
Requalification Human Performance Significance Determination Process (SDP)." The
inspectors verified that: *Crew failure rate on the dynamic simulator was less than 20%(Failure rate was 0.0%.);*Individual failure rate on the dynamic simulator test was less than or equal to 20% (Failure rate was 0.0%.);
6 Enclosure*Individual failure rate on the walkthrough test (JPMs) was less than or equal to 20% (Failure rate was 0.0%.); and*More than 75% of the individuals passed all portions of the exam (100.0% of the individuals passed all portions of the exam).
b. Findings and Observations
No findings of significance were identified.
1R12 Maintenance Effectiveness (71111.12Q - 3 samples)
a. Inspection Scope
The inspectors reviewed performance-based problems involving selected in-scope structures, systems, or components (SSC s) to assess the effectiveness of the maintenance program. Reviews focused on:*Proper Maintenance Rule Scoping in accordance with 10 CFR 50.65;*Characterization of reliability issues;
- Changing system and component unavailability;
- 10 CFR 50.65 (a)(1) and (a)(2) classifications;
- Identifying and addressing common cause failures;
- Trending of system flow and temperature values;
- Appropriateness of performance criteria for SSCs classified (a)(2); and
- Adequacy of goals and corrective actions for SSCs classified (a)(1).
The inspectors reviewed system health reports, maintenance backlogs, and Maintenance Rule basis documents. The inspectors evaluated the maintenance program
against the requirements of 10 CFR 50.65. The documents reviewed are listed in the
. The following three maintenance rule samples were reviewed:*118 volts alternating current (VAC) instrument supply inverters;*Steam generator atmospheric dump valves; and
- 345 kilo-volts alternating current (kVAC) electrical distribution system.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13 - 4 samples)
a. Inspection Scope
The inspectors reviewed the following four activities to verify that the appropriate risk assessments were performed prior to removing equipment for work. The inspectors
verified that risk assessments were performed as required by 10 CFR 50.65(a)(4), and
were accurate and complete. When emergent work was performed, the inspectors
verified that the plant risk was promptly reassessed and managed. The documents 7 Enclosure reviewed are listed in the Attachment. The following activities represent four inspectionsamples:*Work order (WO)-IP2-06-01431, 138 kVAC cross-tie line 33332L out of service with the 32 emergency diesel generator out of service;*WO-IP2-06-27268, steam dump controller troubleshooting and repair activities;
- Condition Report (CR)-IP2-06-5300, indicated pressurizer pressure lowering due to suspected loss of inventory in pressurizer reference leg; and *WO-IP2-06-28160, 22 auxiliary boiler feedwater pump cooling water supply valve repair.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations (71111.15 - 4 samples)
a. Inspection Scope
The inspectors reviewed operability determinations to assess the acceptability of the evaluations; when needed, the use and control of compensatory measures; and
compliance with Technical Specifications. The inspectors' review included a verification
that the operability determinations were made as specified by ENN-OP-104, "Operability
Determinations." The technical adequacy of the determinations was reviewed and
compared to the Technical Specifications, UFSAR, and associated design basis
documents. The documents reviewed are listed in the Attachment. The following four
evaluations were reviewed and each constituted inspection program samples:*Condition Report (CR)-IP2-2006-04402, environmental qualification of post-accident monitoring instrumentation;*CR-IP2-06-05241, 22 auxiliary boiler feedwater pump following five minute run with degraded bearing cooling water;*CR-IP2-04-06167 and -06776, degradation in service water supply to 21 and 23 emergency diesel generators; and*CR-IP2-06-04581, low air pressure to 21 emergency diesel generator right hand air starting motor.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing (71111.19 - 6 samples)
a. Inspection Scope
The inspectors reviewed post maintenance test procedures and associated testing activities for selected risk significant mitigating systems to assess whether the effect of
maintenance on plant systems was adequately addressed by control room and
engineering personnel. The inspectors verified that test acceptance criteria were clear, 8 Enclosure demonstrated operational readiness and were consistent with design basis documentation; that test instrumentation had current calibrations and the range and
accuracy for the application; and that tests were performed, as written, with applicable
prerequisites satisfied. Upon completion, the inspectors verified that equipment was
returned to the proper alignment necessary to perform its safety function.
Post-maintenance testing was evaluated against the requirements of 10 CFR 50, Appendix B Criterion XI, "Test Control." The documents reviewed are listed in the
. The following post maintenance test activities were reviewed and represent
six inspection program samples:*WO-IP2-06-000954, steam generator atmospheric dump valve PCV-1134 following in-service test failure;*WO-IP2-05-23299, 21 auxiliary boiler feedwater pump following maintenance;
- WO-IP2-05-21977, 22 safety injection pump following preventive maintenance; and*WO-IP2-05-24095, 23 component cooling water pump following preventive maintenance; *WO-IP2-05-23473, gas turbine 3 following maintenance; and
- WO-IP2-04-30457, safety injection valve SI888A following maintenance.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing (71111.22 - 6 samples)
a. Inspection Scope
The inspectors witnessed performance of surveillance tests and/or reviewed test data of selected risk-significant systems, struct ures, and components to assess whether the systems, structures, and components satisfied Technical Specifications, UFSAR, Technical Requirements Manual, and Entergy procedure requirements. The inspectors
verified that test acceptance criteria were clear, demonstrated operational readiness and
were consistent with design basis documentation; that test instrumentation had current
calibrations and the range and accuracy for the application; and that tests were
performed, as written, with applicable prerequisites satisfied. Upon surveillance test
completion, the inspectors verified that equipment was returned to the status specified to
perform its safety function. The inspectors evaluated the surveillance tests against the
requirements in Technical Specifications. The following surveillance tests were
reviewed and represented six inspection program samples:*2-PT-2M2A, "RPS Logic Train 'A' Actuation Logic and Trip Actuation Device Operational Test (TADOT) (> 25% Reactor Power)," Revision 0;*2-PT-Q030B, "21 Component Cooling Water Pump," Revision 13;
- 2-PT-M7, "Analog Rod Position Functional," Revision 27;
- PT-M38C, "Gas Turbine No. 3," Revision 4;
- 2-PT-M021A, "Emergency Diesel Generator 21 Load Test," Revision 15; and
- PT-MT55, "Fire Door Surveillance," Revision 12.
b. Findings
No findings of significance were identified.
1R23 Temporary Plant Modifications (71111.23 - 2 samples)
a. Inspection Scope
The inspectors assessed the adequacy of the 10 CFR 50.59 evaluations for these temporary modifications; that the installation was consistent with the modification
documentation; that the drawings and procedures were updated as applicable; and that
the post-installation testing was adequate. The inspectors assessed the temporary
modification, any planned compensatory actions, and reviewed drawings to evaluate any
potential impact on equipment indications, alarms, or protective functions The
documents reviewed are listed in the Attachment. This inspection satisfied two
inspection samples for temporary modifications. The following modifications were
reviewed:*TM-2-2006-0084, "Isolation of Service Water Leak Upstream of System Drain Valve SWN-76-1;" and*SOP-28.22, "Remove Saturation Monitor Core Exit Thermocouple Input From Scan."
b. Findings
No findings of significance were identified. Cornerstone: Emergency Preparedness1EP5Correction of Emergency Preparedness Weaknesses and Deficiencies (71114.05- 1 sample)
a. Inspection Scope
A region-based specialist inspector conducted an inspection of Entergy's corrective actions related to the current Indian Point alert and notification system, and also of the
progress made in the design and installation of the new siren system. The inspection
was conducted onsite September 13 and 14, 2006, and in the Region I office the week
of September 25, 2006, per the baseline inspection program deviation authorized by the
NRC Executive Director of Operations in a memorandum approved on October 31, 2005.The inspector discussed and reviewed two events during the quarter that involved the loss of the ability to actuate the sirens: one occurred on August 2, 2006, as the result of
a system computer hard drive failure; and one occurred on September 7, 2006, as the result of a human performance issue which occurred during work to prepare for the
installation of the new siren system. The inspector reviewed the condition reports and
the corrective actions, for these two events. To assess the effectiveness of the 10 Enclosure corrective actions and the performance of Entergy's siren system, the inspector observed the performance of the quarterly full siren test conducted on September 13, 2006. The inspector monitored the test from the Indian Point Energy Center Emergency
Operations Facility (EOF) and observed the licensee's communication with the four local
counties, the initiation of the siren system actuation, and the reception and logging of
siren feedback information to determine the performance of the sirens.
Subsequent to the onsite portion of the inspection, the inspector learned of an additional failure of the siren system on September 19, 2006, as the result of the computer
software database failing to reconnect following a preventive maintenance reboot of the
siren system computer. The inspector conducted an initial in-office review of the event, but the licensee's corrective actions were not finalized prior to the end of the inspection
period. The inspector interviewed the project manager for the new siren system to understand Entergy's progress towards meeting the milestone dates required by the NRC's
Confirmatory Order dated January 31, 2006. While on site, the inspector observed the
progress of Entergy's installation of the new siren control system in the EOF. The
inspector also reviewed the proposed final Indian Point Energy Center Prompt Alert and
Notification System Design Report which Entergy had submitted to the New York State
Emergency Management Office on September 28, 2006, for Department of Homeland Security review and approval.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
[OA]4OA1Performance Indicator Verification (71151 - 1 sample)
a. Inspection Scope
The inspectors reviewed performance indicator (PI) data for the below listed cornerstones and used Nuclear Energy Institute (NEI) 99-02, "Regulatory Assessment
Performance Indicator Guidance," to verify individual PI accuracy and completeness.
Initiating Events Cornerstone*Unplanned Transients per 7000 Critical Hours
The inspectors reviewed a selection of LERs, operator log entries, monthly operating reports, and PI data sheets to determine whether Entergy adequately identified the
number of unplanned power changes greater than 20 percent that occurred from
July 2004 to June 2006. This number was compared to the number reported for the PI
during the current quarter. The inspectors also verified the accuracy of the number of
critical hours reported. In addition, the inspectors also interviewed Entergy personnel
associated with the PI data collection, evaluation, and distribution.
11 Enclosure4OA2Identification and Resolution of Problems (71152 - 1 sample).1Annual Sample Review
a. Inspection Scope
The inspector conducted a detailed review of CR-IP2-2006-02133 which was written to address severe degradation of a non-safety-related service water pipe in the steam
generator blowdown tank (SGBT) room. This issue was selected for review because
failure of the pipe could have affected safety-related equipment. The inspector reviewed
Entergy's causal analysis, operability evaluation, and the assigned corrective actions.
The inspector toured the SGBT room and observed the repaired section of the degraded
pipe with the service water system engineer. The inspector reviewed the design change
package for the repair and the non-destruction examination records of the welding for
the piping repair.
b. Findings and Observations
There were no findings or observations identified associated with the reviewed sample.
.2 Routine PI&R Program Review
a. Inspection Scope
As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"
and in order to help identify repetitive equipment failures or specific human performance
issues for follow-up, the inspectors performed a daily screening of all items entered into
Entergy's corrective action program. The review was accomplished by accessing
Entergy's computerized database for CRs and attending CR screening meetings.
b. Findings
No findings of significance were identified.
4OA3Event Followup (71153 - 1 sample).1Unit 2 Manual Reactor Trip - August 23, 2006
a. Inspection Scope
The inspectors responded to a manual reactor trip following a loss of both heater drain tank pumps and the subsequent plant transient that occurred on August 23, 2006 and
observed operator actions. Control room oper ators initiated a manual reactor trip in accordance with directions provided by the control room supervisor following an
unexpected response in the steam dump contro l system. The inspectors discussed the trip with operations, maintenance, and engineering personnel to ensure the event was
understood and to assess followup actions. The inspectors reviewed operator actions
taken in accordance with licensee procedures and reviewed unit and system indications
to verify that actions and system responses were as expected. The inspectors 12 Enclosure discussed the trip with Entergy's root cause analysis team and assessed the team's actions to gather, review, and assess information leading up to and following the reactor
trip. The inspectors also reviewed the initial investigation report and root cause
determination to assess the detail of the review and the adequacy of the root cause and
proposed corrective actions prior to unit restart. The inspectors also reviewed the initial
licensee notification to verify that it met the requirements specified in NUREG-1022, "Event Reporting Guidelines."
b. Findings
.1 Introduction
- A Green self-revealing finding was identified because Entergy failed to develop an adequate procedure for governing the response by operators to a loss of
both heater drain tank pumps and to an approaching rod insertion limit (RIL) alarm
condition.
Description
- On August 23, 2006, both heater drain tank pumps tripped due to a low heater drain tank level condition resulting from a level controller failure. The loss of
these pumps led to a reduction of feed flow to the steam generators and required a
reduction in power to mitigate lowering level in the steam generators. Per procedure
2-AOP-FW-1, "Loss of Main Feedwater," the operators rapidly reduced power to 77
percent (%). During the transient, the high pressure steam dumps and control rods
operated to maintain average coolant temperature at the programmed value. Due to
control rod insertion, an "approaching RIL" alarm was received in the control room. Per
alarm response procedure 2-ARP-SAF, the operators placed the control rods in manual, which stopped automatic control rod insertion. Following the initial power reduction, operators noted that axial flux distribution was outside the value required by Technical
Specification 3.2.3, "Power Distribution Limits - Axial Flux Difference," and determined
that a power reduction to less than 50% was required. During the subsequent power
reduction, operators noted a significant reduction in turbine power with no operator
action. Operators determined that they did not have sufficient control of the transient, and manually tripped the reactor.
The inspectors reviewed plant transient data, operator actions, and the plant operating procedures used during the transient. The inspectors noted that the procedure flow path
for a loss of both heater drain tank pumps in 2-AOP-FW-1, "Loss of Main Feedwater,"
did not direct operators to reset the high pressure steam dump arming signal following
the rapid downpower. The steam dumps are designed to arm on a 10 percent step load
drop or a 5 percent per minute power reduction, and then modulate open to control
average coolant temperature. Failure to reset the arming signal resulted in the steam
dumps actuating during the second downpower, when they should not have been
required. In addition, the inspectors noted that placing the rod control system in manual
when the "approaching RIL" alarm actuated, limited the operators' ability to maintain
proper plant control during the transient. If left in automatic, the control rods would have
driven in and added negative reactivity on an in creasing average coolant temperature.
With the control rods in manual, negative reac tivity is provided by boron addition, which does not provide as timely a response as the control rods. The inspectors determined
that these procedural deficiencies complicated the plant transient and impacted the
operators' ability to effectively respond to plant conditions.
13 Enclosure
Analysis:
The inspectors determined that Entergy's failure to develop adequate procedures to respond to a loss of both heater drain tank pumps and an "approaching
RIL" alarm is a performance deficiency. IP-SMM-AD-102, "Indian Point Energy Center
Implementing Procedure Preparation, Review, and Approval" requires that a technical
review be conducted for site-specific implementing procedures. This review is required
to verify the adequacy and technical accuracy of the subject procedure. It is reasonable
that Entergy should have identified this procedural inadequacy. Traditional enforcement
does not apply since there were no actual safety consequences or potential for
impacting the NRC's regulatory function, and the finding was not the result of any willful
violation of NRC requirements or Entergy's procedures.
The inspectors determined that this finding was greater than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone; and, it impacted the cornerstone objective of limiting the likelihood of those events that upset
plant stability and challenge critical safety functions. Specifically, the procedural
inadequacies complicated operator actions to a rapid downpower, resulted in a manual reactor trip when the operators determined that they did not have sufficient control of the
transient, and could impact other accident sequences requiring negative reactivity
addition. The inspectors evaluated this finding using Phase 1 of Inspection Manual
Chapter (IMC) 0609, Appendix A, "Significance Determination of Reactor Inspection
Findings for At-Power Situations," and determined it to be of very low safety significance
because it did not contribute to the likelihood of both a reactor trip and the likelihood that
mitigation equipment or functions would be unavailable.
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that plant operating procedures 2-
AOP-FW-1 and 2-ARP-SAF were adequate to ensure operators could appropriately
respond to a rapid downpower transient.
Enforcement
- Because this finding is related to a procedural deficiencies associated with the non-safety related heater drain and feedwater systems, and the non-safety
related portion of the rod control system, no violation of regulatory requirements occurred:
FIN 05000247/2006004-01, Inadequate Operating Procedure for Loss of Both Heater Drain Tank Pumps.
Entergy entered this issue into the corrective action program as corrective action number IP2-2006-05065. Entergy is evaluating the appropriate steps to correct the procedural
deficiencies.
.2 Introduction
- A Green self-revealing finding was identified because Entergy failed to develop an adequate procedure for calibrating the steam dump loss of load controller.
An error in the calibration procedure caused the steam dumps to operate improperly
during a plant transient and contributed to a manual reactor trip.
Description
- On August 23, 2006, the plant experienced a loss of both heater drain tank pumps, which necessitated a rapid reduction in plant power. During the load reduction, the steam dump system failed to operate as expected. Specifically, the steam dump valves opened earlier and more rapidly than expected. This resulted in a large deviation
between reactor power and generator load.
14 Enclosure The steam dump loss of load controller is used, in conjunction with the rod control system, to control average coolant temperature following large changes in secondary
load. The controller receives an arming signal when turbine load drops rapidly, and then
sends a signal to the steam dump valves, causing them to modulate open if average coolant temperature exceeds the reference temperature by a set value. The controller is
designed to have a lead time of 10 seconds, which allows the controller to respond to
anticipated changes in average coolant temperature and start to open before the set
error threshold is reached.
Following the reactor trip, the inspectors reviewed calibration procedures for the steam dump loss of load controller. It was identified that the calibration procedures
inappropriately specified setting up the controller with a lead gain of 10, rather than a
lead time of 10 seconds. The lead gain is a measure of the magnitude of controller
output change for a given controller input change. As a result, the steam dump system
response to changes in average coolant temperature was amplified by a factor of four, resulting in significant complications during the power reduction, and the subsequent
Analysis:
The inspectors determined that Entergy's failure to develop a technically accurate procedure for calibration of the steam dump loss of load controller is a
performance deficiency. IP-SMM-AD-102, "I ndian Point Energy Center Implementing Procedure Preparation, Review, and Approval" requires that a technical review be
conducted for site-specific implementing procedures. This review is required to verify
the adequacy and technical accuracy of the subject procedure. It is reasonable that
Entergy identify this deficiency since the correct calibration settings of this controller
were specified in Westinghouse analyses. Traditional enforcement does not apply since
there were no actual safety consequences or potential for impacting the NRC's
regulatory function, and the finding was not the result of any willful violation of NRC
requirements or Entergy's procedures.
The inspectors determined that this finding was greater than minor because it was associated with the Procedure Quality attribute of the Initiating Events cornerstone; and
it impacted the cornerstone objective of limiting the likelihood of those events that upset
plant stability and challenge critical safety functions. Specifically, the inadequacy in
Entergy's calibration procedure caused the steam dumps to operate improperly during a
plant transient and contributed to a reactor trip. This inspectors evaluated this finding
using Phase 1 of IMC 0609, Appendix A, "Significance Determination of Reactor
Inspection Findings for At-Power Situations," and determined it to be of very low safety
significance because it did not contribute to the likelihood of both a reactor trip and the
likelihood that mitigation equipment or functions would be unavailable.
The inspectors determined that this finding had a cross-cutting aspect in the area of human performance because Entergy did not ensure that the procedure for calibration of
the steam dump loss of load controller was accurate, in that, it specified incorrect
settings for the controller.
Enforcement
- Because this finding is related to a procedural deficiency associated with the non-safety related steam dump system, no violation of regulatory requirements occurred:
FIN 05000247/2006004-02, Inadequate Procedure for Calibrating the Steam Dump Loss of Load Controller.
15 Enclosure Entergy entered this issue into the corrective action program as corrective action number IP2-2006-05065. Entergy corrected the procedure, and correctly calibrated and set up
the steam dump loss of load controller prior to plant start up..3(Closed) LER 05000247/2006001-00 , Manual Reactor Trip Due to Multiple Dropped Control Rods Caused by Loss of Control Rod Power Due to Personnel Error On March 1, 2006, plant operators initiated a manual reactor trip following indications of 12 dropped control rods. The control rods dropped when a contractor employee
assembling scaffolding in the cable spreading room inadvertently bumped a disconnect
switch which interrupted power to rod control power supply cabinet 1AC. Entergy
operators appropriately responded to the reactor trip, actions were taken to strengthen
control of scaffold construction activities, and the issue was entered into the corrective
action program (CR-IP2-06-01012). The inspectors previously reviewed this issue in
Inspection Report 05000247/2006-002, and issued NCV 05000247/2006002-04, "Scaffolding Control Issue Results in Reactor Trip," because Entergy failed to adequately
assess and manage the risk associated with scaffolding construction activities. The LER
was reviewed by the inspectors and no additional findings of significance were identified.
This LER is closed.4OA5Other Activities.1Groundwater Contamination Investigation
a. Inspection Scope
Inspection of the groundwater contamination investigation at Indian Point Energy Center was authorized by the NRC Executive Director of Operations in a Reactor Oversight Process (ROP) deviation memorandum approved on October 31, 2005 (ADAMS
Accession number ML053010404). Accordingly, oversight of licensee progress has
been conducted throughout this inspection period consisting of weekly discussions with
the licensee on groundwater investigation status and bi-weekly communications with
Federal, State, and local government stakeholders. In addition, NRC continued to split
samples of offsite, site boundary and other selected monitoring wells with the licensee in
order to verify the licensee's sample results.
b.Results The NRC's assessment of the licensee's sample data indicated that the licensee continued to report sample results that were consistent with NRC results. The Oak
Ridge Institute for Science and Education, Environmental Site Survey and Assessment
Program (ORISE/ESSAP) sample results are available in ADAMS under Accession
Number ML062720227. To date, sample results from site boundary wells and off-site
environmental groundwater sampling locations have not indicated any detectable plant-
related activity..2(Closed) URI 05000247/2003003-04 , Electrical Calculation Reconstitution to Support Off-Site Power Design Basis (SAT Load Tap Changer)
In August 2002, Entergy identified a concern associated with operability of the off-site electrical distribution system following a safety injection system actuation. Specifically, preliminary calculations showed that under certain expected grid voltage conditions, the
voltage on the safety buses could drop below the degraded voltage relay setpoint for
greater than 10 seconds. This would result in loss of the off-site power supply to the
safety buses. Entergy identified a number of non-conservative assumptions in their
original degraded grid calculation, including a failure to account for the fast bus transfer
30 seconds after the safety injection signal, instrument tolerances for the degraded
voltage relays, and the neutral position of the load tap changer. The inspectors reviewed calculation FEX-00143-01, "Indian Point Unit 2 (IP2) Load Flow Analysis of the Electrical Distribution Sy stem," Revision 1, which analyzed bus voltages for limiting transients. The inspectors verified that for the transients analyzed, voltage
would either remain above the degraded voltage relay setting, or that operation of the
load tap changer would restore voltage to above the degraded voltage relay reset point
prior to off-site power separating from the buses. The inspectors confirmed that the non-
conservative assumptions in the original calculation had been appropriately addressed
in the revised calculation. The inspectors did not identify any issues associated with
historical plant risk, or any issues meeting reportability requirements. This issue is
closed. 4OA6Meetings, Including Exit
Exit Meeting Summary
On October 4, 2006, the inspectors presented the inspection results to Mr. Paul Rubin and other Entergy staff members, who acknowledged the inspection results presented.
Entergy acknowledged that no proprietary information was involved.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Entergy Personnel
- F. Dacimo, Site Vice President
- P. Rubin, General Manger of Plant Operations
- E. O'Donnell, U2 Operations Manager
- B. Christman, Manager of Training and Development
- S. Davis, Superintendent of Operator Training
- A. Singer, Operations Training Supervisor
- D. Eccleston, Senior Operations Instructor
- E. Goetchius, Senior Operations Instructor
- D. Huntington, Senior Operations Instructor
- R. Robenstein, Simulator Support Supervisor
- J. Gullick, Senior Simulator Specialist
- J. Rowland, Senior Simulator Specialist
- T. Beasley, System Engineer
- J. Kayani, System Engineer
- B. Meek, Maintenance Supervisor
- J. Bubniak, Senior Engineer
- R. Scalone, Performance Engineering Supervisor
- N. Azevedo, Code Program Supervisor
- D. Gaynor, Senior Engineer
- R. Lee, Senior Design Engineer
- G. Dahl, Licensing Engineer
- R. Mann, Programs and Components Engineer
- T. Pepe, Programs and Components Engineer
- J. Joy, Programs and Components Engineer
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
05000247/2006004-01FINInadequate Operating Procedures for Loss
of Both Heater Drain Tank Pumps (Section
1R13)05000247/2006004-02FINInadequate Procedure for Calibrating the
Steam Dump Loss of Load Controller
(Section 1R13)
Closed
05000247/2006001-00LERManual Reactor Trip Due to Multiple
Dropped Control Rods Caused by Loss of
Control Rod Power Due to Personnel Error
(Section 4OA3)
Attachment05000247/2003003-04URIElectrical Calculation Reconstitution to
Support Off-Site Power Design Basis (SAT
Load Tap Changer) (Section 4OA5)