ML032110555

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IR 05000247-03-007, on March 30, 2003 - June 28, 2003, Entergy Nuclear Operations, Inc.; Indian Point 2 Nuclear Power Plant; Operability Evaluations; and Problem Identification and Resolution Samples
ML032110555
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 07/30/2003
From: Eselgroth P, Brian Holian
Division Reactor Projects I
To: Dacimo F
Entergy Nuclear Northeast
References
FOIA/PA-2003-0379, FOIA/PA-2003-0388 IR-03-007
Download: ML032110555 (35)
v • d • e


See also: IR 05000247/2003007

Text

July 30, 2003

Mr. Fred Dacimo

Site Vice President

Entergy Nuclear Northeast

Indian Point Nuclear Generating Station

295 Broadway, Suite 1

Post Office Box 249

Buchanan, NY 10511-0249

SUBJECT: INDIAN POINT 2 - NRC INTEGRATED INSPECTION REPORT

050000247/2003007

Dear Mr. Dacimo:

On June 28, 2003, the US Nuclear Regulatory Commission (NRC) completed an inspection at

the Indian Point 2 Nuclear Power Plant. The enclosed integrated inspection report documents

the inspection findings, which were discussed on July 9, 2003, with yourself and other members

of your staff.

The inspection examined activities conducted under your license as they relate to safety and

compliance with the Commissions 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, the inspectors identified two findings of very low safety

significance (Green) which did not present an immediate safety concern. One of the findings

was determined to be a violation of NRC requirements. However, because it was of very low

safety significance and because the issue has been addressed and entered into your corrective

action program, the NRC is treating this issue as a non-cited violation, in accordance with

Section VI.A.1 of the NRCs Enforcement Policy. If you deny this non-cited violation, you should

provide a response with the basis for your denial, within 30 days of the receipt of this letter, to

the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C. 20555-

001; with copies to the Regional Administrator, Region 1; the Director, Office of Enforcement;

and the NRC Resident Inspector at the Indian Point 2 facility.

Since the terrorist attacks on September 11, 2001, NRC has issued five Orders and several

threat advisories to licensees of commercial power reactors to strengthen licensee capabilities,

improve security force readiness, and enhance controls over access authorization. In addition to

applicable baseline inspections, the NRC issued Temporary Instruction 2515/148, "Inspection of

Nuclear Reactor Safeguards Interim Compensatory Measures," and its subsequent revision, to

audit and inspect licensee implementation of the interim compensatory measures required by

order. Phase 1 of TI 2515/148 was completed at all commercial power nuclear power plants

during calender year '02 and the remaining inspection activities for Indian Point 2 were

completed in January 2003. The NRC will continue to monitor overall safeguards and security

controls at Indian Point 2.

Mr. Fred Dacimo 2

In accordance with 10 CFR 2.790 of the NRCs "Rules of Practice," a copy of this letter and its

enclosure will be available electronically for public inspection in the NRC Public Document Room

or from the Publicly Available Records (PARS) component of the NRCs 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). Should you have any questions

regarding this report, please contact Mr. Peter Eselgroth at 610-337-5234.

Sincerely,

/RA/ by

Peter W. Eselgroth

Acting For/

Brian E. Holian, Deputy Director

Division of Reactor Projects

Docket No.50-247

License No. DPR-26

Enclosure: Inspection Report 05000247/2003007

w/Attachment: Supplemental Information

cc w/encl: G. J. Taylor, Chief Executive Officer, Entergy Operations

M. R. Kansler, President - Entergy Nuclear Operations, Inc.

J. Herron, Senior Vice President and Chief Operating Officer

C. Schwarz, General Manager - Plant Operations

D. Pace, Vice President, Engineering

R. Edington, Vice President, Operations Support

J. McCann, Manager, Nuclear Safety and Licensing

J. Kelly, Director, Nuclear Safety Assurance

J. Comiotes, Director, Nuclear Safety Assurance

C. Faison, Manager, Licensing

H. Salmon, Jr., Director of Oversight

J. Fulton, Assistant General Counsel, Entergy Nuclear Operations, Inc.

P. Smith, Acting President, New York State Energy, Research

and Development Authority

J. Spath, Program Director, 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

T. Walsh, Secretary, NFSC, Entergy Nuclear Operations, Inc.

D. ONeill, Mayor, Village of Buchanan

J. G. Testa, Mayor, City of Peekskill

R. Albanese, Executive Chair, Four County Nuclear Safety Committee

S. Lousteau, Treasury Department, Entergy Services, Inc.

Chairman, Standing Committee on Energy, NYS Assembly

Chairman, Standing Committee on Environmental Conservation, NYS Assembly

Mr. Fred Dacimo 3

Chairman, Committee on Corporations, Authorities, and Commissions

M. Slobodien, Director, Emergency Planning

B. Brandenburg, Assistant General Counsel

P. Rubin, Operations Manager

Assemblywoman Sandra Galef, NYS Assembly

C. Terry, Niagara Mohawk Power Corporation

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 Citizens 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, Riverkeepers, Inc.

M. Kapolwitz, Chairman of County Environment & Health Committee

A. Reynolds, Environmental Advocates

M. Jacobs, Director, Longview School

D. Katz, Executive Director, Citizens Awareness Network

P. Gunter, Nuclear Information & Resource Service

P. Leventhal, The Nuclear Control Institute

K. Copeland, Pace Environmental Litigation Clinic

R. Witherspoon, The Journal News

W. DiProfio, PWR SRC Consultant

W. Poole, PWR SRC Consultant

W. Russell, PWR SRC Consultant

W. Little, Associate Attorney, NYSDEC

Mr. Fred Dacimo 4

Distribution w/encl: H. Miller, RA/J. Wiggins, DRA (1)

H. Nieh, RI EDO Coordinator

P. Habighorst, SRI - Indian Point 2

P. Eselgroth, DRP

R. Laufer, NRR

P. Milano, PM, NRR

G. Vissing, PM, NRR (Backup)

W. Cook, DRP

R. Martin, DRP

Region I Docket Room (w/concurrences)

DOCUMENT NAME: C:\ORPCheckout\FileNET\ML032110555.wpd

After declaring this document An Official Agency Record it will be released to the Public. To

receive a copy of this document, indicate in the box: "C" = Copy without

attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy

OFFICE RI/DRP RI/DRP E RI/DRP E

NAME Phabighorst/wac for PEselgroth/wac BHolian/pwe for

for

DATE 07/09/03 07/30 /03 07/30/03

OFFICIAL RECORD COPY

U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No. 50-247

License No. DPR-26

Report No. 05000247/2003007

Licensee: Entergy Nuclear Operations, Inc.

Facility: Indian Point 2 Nuclear Power Plant

Location: Buchanan, New York 10511

Dates: March 30, 2003 - June 28, 2003

Inspectors: Peter Habighorst, Senior Resident Inspector

Lois James, Resident Inspector

Mark Cox, Resident Inspector, IP3

Monica Salter-Williams, Reactor Engineer

Peter Wen, NRR/DRIP/RGEB Inspector

William Cook, Senior Project Engineer

John McFadden, Health Physicist

Greg Smith, Senior Physical Security Inspector

David Silk, Senior Emergency Preparedness Inspector

Approved by: Peter W. Eselgroth, Chief

Projects Branch 2

Division of Reactor Projects

i Enclosure

TABLE OF CONTENTS

TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii

SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iii

Report Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

Summary of Plant Status . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1. REACTOR SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1R01 Adverse Weather Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

1R02 Evaluation of Changes, Tests, or Experiments . . . . . . . . . . . . . . . . . . . . . . . . . 2

1R04 Equipment Alignment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

1R05 Fire Protection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3

1R07 Heat Sink Performance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

1R11 Operator Requalification Inspection . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

1R12 Maintenance Effectiveness . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

1R13 Maintenance Risk Assessment and Emergent Work Activities . . . . . . . . . . . . . . 7

1R14 Personnel Performance During Non-Routine Plant Evolutions and Events . . . . 8

1R15 Operability Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9

1R16 Operator Workarounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

1R19 Post Maintenance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

1R22 Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

1EP4 Emergency Action Level and Emergency Plan Changes . . . . . . . . . . . . . . . . . 13

2. RADIATION SAFETY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13

2OS1 Access Control to Radiologically Significant Areas . . . . . . . . . . . . . . . . . . . . . 13

2OS2 ALARA Planning and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

2OS3 Radiation Monitoring Instrumentation and Protective Equipment . . . . . . . . . . . 15

3. SAFEGUARDS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

3PP4 Security Plan Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

4. OTHER ACTIVITIES (OA) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

4OA1 Performance Indicator Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

4OA2 Identification and Resolution of Problems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18

4OA6 Meetings, Including Exit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

ATTACHMENT: SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . A-2

LIST OF CONDITION REPORTS GENERATED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-3

LIST OF DOCUMENTS REVIEWED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-4

LIST OF BASELINE INSPECTIONS PERFORMED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-7

LIST OF ACRONYMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-7

ii Enclosure

SUMMARY OF FINDINGS

IR 05000247-03-07, on March 30, 2003 - June 28, 2003, Entergy Nuclear Operations, Inc.;

Indian Point 2 Nuclear Power Plant; Operability Evaluations; and Problem Identification and

Resolution Samples.

The report covered a twelve-week period of inspection by resident and announced region-based

and headquarters-based inspectors. Two Green findings, of which one was a non-cited violation

(NCV), were identified. The significance of the findings are indicated by their color (Green,

White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination

Process (SDP). The NRCs 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: Mitigating Systems

Green. The inspector identified that the licensees operability evaluation during a 13.8 KV

system reduced voltage test was not complete. The operability evaluation did not

evaluate accident load carrying capability and it did not address communication protocols

between the distribution company (Con Edison) and Entergy to restore from the test in a

timely manner.

The finding is more than minor because it impacts the attribute of the mitigating system

cornerstone objective. Specifically, the cornerstone objective is to ensure that the 13.8

KV system is capable of performing its safety function during a postulated loss of normal

power event without undesirable consequences. This finding was determined to be of

low safety significance because there was no actual loss of either the 138 KV or 13.8 KV

offsite power supplies during the short duration of the degraded voltage condition on the

13.8 KV feed. (Section 1R15)

Green. The inspector identified a violation of 10 CFR 50, Appendix B, Criterion XVI.

Entergy did not evaluate and take effective corrective actions associated with a material

substitution for the No. 22 component cooling water (CCW) pump inboard bearing oil

level indication system. The substitute bearing oil level indication system contributed to

the failure of the No. 22 CCW pump on December 5, 2002.

This finding is greater than minor since it is associated with the design control attribute of

the mitigating systems cornerstone and affected the cornerstone objective. The

inspectors conducted a Phase 1 SDP screening and determined that the failure to take

effective corrective action on No. 22 CCW pump was of a very low safety significance

since the redundant train components were operable and unaffected by this inadequate

modification. Accordingly, this issue was treated as a Non-cited Violation. (4OA2)

B. License-Identified Violations

None

iii Enclosure

Report Details

Summary of Plant Status

Indian Point Unit 2 began the period at full Rated Thermal Power (RTP) and operated at full

power until April 28, 2003. On April 28, the unit experienced a main turbine trip on over-

frequency and a reactor trip due to a generator load reject as a result of electrical faults on the

off site 345 KV and 138 KV distribution systems (reference report detail 1R14). On May 1, 2003,

the unit was restored back to RTP. On May 24, 2003, following issuance of the Technical

Specification amendment, RTP was increased from 3,071.4 to 3114.4 thermal megawatts. The

unit operated at RTP throughout the remainder of the inspection period.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity and Emergency

Planning

1R01 Adverse Weather Protection

a. Inspection Scope (71111.01)

The inspector reviewed hot weather preparations by the licensee. The inspector selected

licensee controls in the following plant areas: central control room, auxiliary boiler

feedwater pump, intake structure, and the 480 volt switchgear room. Documents

reviewed by the inspector are listed below:

  • SOP 11.1, Ventilation System Operation
  • ARP SCF Window 4-4, 22 ABFP Inlet Valves 1310A/1310B Not Fully Open
  • ARP SJF, Window 4-1, 480 Volt Switchgear Temperature Hi
  • OASL 15.90, Inclement Weather Conditions
  • AOI 11.1, Failure of CCR Air Conditioners/Fan Systems
  • OAD 22, Seasonal Weather Preparation
  • OAD 44, Summer Reliability

Documents

  • System Health reports for the EDG and CCR HVAC Systems
  • List of Unit 2 control room deficiencies
  • Individual Plant Examination of External Events (IPEEE) Table 6.5-1

The inspector walked down selected plant areas to verify availability of ventilation and air

conditioning units, availability of back-up mitigation equipment as defined in AOI 11.1,

and material condition of various temperature sensors and recorders.

b. Findings

No findings of significance were identified.

1R02 Evaluation of Changes, Tests, or Experiments

Enclosure

2

a. Inspection Scope (71111.02)

The inspector reviewed the 10 CFR 50.59 evaluation associated with the increase in TAVG

from 559 0F to 562 0F to verify that this change to the facility and associated procedures,

as described in the Updated Final Safety Analysis Report (UFSAR), was reviewed and

documented in accordance with 10 CFR 50.59 and that the safety issues pertinent to the

change were properly resolved or adequately addressed. This evaluation was selected

based on the safety significance of the changes and the risk to structures, systems, and

components.

The inspector reviewed the licensees evaluation package, 02-0344-PR-02-RE, and

interviewed engineering personnel cognizant of the associated 10 CFR 50.59 evaluation.

The evaluation concluded that the increase to the operating TAVG from 559 0F to 562 0F at

Indian Point Unit 2 does not adversely affect the safe operation of the plant and did not

require a change to the plant Technical Specifications. The inspector observed the

control room activities associated with raising TAVG on April 25, 2002. The inspector

noted that the licensee conducted this evolution in a thorough and deliberate manner.

During the review of associated operating procedures impacted by the TAVG increase, the

inspector questioned two conversion factors (one for the pressurizer level change and

the other for the reactor coolant system mass change) used in the licensees reactor

coolant system leakage surveillance procedure (SOP 1.7, Revision 35). As a result, the

licensee generated a condition report IP2-2003-02567) to provide values that more

accurately reflect the current plant conditions for the reactor water inventory calculation.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment

a. Inspection Scope (71111.04)

On April 18, 2003, the inspector performed a partial system walkdown of the 21 boric

acid transfer pump (BATP) while the 22 BATP was out of service for preventive

maintenance. The purpose of this walkdown was to verify equipment alignment and

identify any discrepancies that could impact the function of emergency boration, thereby

potentially increasing risk. The inspector observed the physical condition of the system

pump and valves and reviewed the operations logs. The inspector used check-off list

(COL) 3.1, Chemical and Volume Control System, for this walkdown and reviewed the

design basis document for the boric acid transfer system to verify the valve positions, as

defined in the COL, were appropriate.

On May 12, 2003, the inspector performed a partial system walkdown of the 21 and 22

safety injection trains while the 23 safety injection pump was being surveillance tested

pursuant to PT-Q29C, 23 Safety Injection Pump. The purpose of this walkdown was to

verify equipment alignment and identify any discrepancies that could adversely impact

the reactor coolant system injection and heat removal functions and thereby potentially

Enclosure

3

increase risk. The inspector observed the physical condition of the subsystems and

reviewed the operations logs. The inspector used COL 10.1.1, Safety Injection System,

and COL 10.1.1.1, Backseated Safety Injection Recirculation Valves, for this walkdown

and reviewed the plant drawings 9321-F-2738, 9321-F-2735-131, and A235296-59 to

verify valve positions, as defined in the COLs, were appropriate. The inspector identified

minor housekeeping and valve and pump seal leakages that were documented in

Condition Report (CR) Nos. IP2-2003-2876, -2877, and -2878 by the licensee.

The inspector performed a partial system walkdown of the 21 emergency diesel

generator (EDG) to evaluate the operability of the starting air system while the 23 EDG

was removed from service for preplanned maintenance. The inspector checked for

correct valve and power alignments by comparing positions of valves, switches, and

electrical power breakers to COL 27.3.1, Diesel Generators, as well as applicable

chapters of the Final Safety Analysis Report (FSAR) to verify proper system alignment.

The inspector also verified starting air system pressure, component labeling, and the

condition of hangers and support installations.

Air compressor operation during the walkdown was observed to ensure that system

vibration and pump leakage was not excessive, and that system operating pressure met

operational and design specifications.

b. Findings

No findings of significance were identified. During the safety injection sub-system

walkdown, the inspector identified an operator work-around involving the 21 safety

injection motor-operated valve (856A) not previously entered into the licensees work-

around system. This minor observation is also discussed in report section 1R16.

1R05 Fire Protection

.1 Fire Protection Tours

a. Inspection Scope (71111.05)

The inspector toured the areas important to plant safety and risk based upon a review of

Section 4.0, Internal Fires Analysis, and Table 4.6-2, Summary of Core Damage

Frequency Contributions from Fire Zones, in the Indian Point 2 Individual Plant

Examination for External Events (IPEEE). The objective of this inspection was to

determine if the licensee had adequately controlled combustibles and ignition sources

within the plant, effectively maintained fire detection and suppression capability, and had

adequately established compensatory measures for degraded fire protection equipment.

The inspector evaluated conditions related to: 1) licensee control of transient

combustibles and ignition sources; 2) the material condition, operational status, and

operational lineup of fire protection systems, equipment and features; and 3) the fire

barriers used to prevent fire damage or fire propagation. The areas reviewed were:

  • Fire Zone 1, Component Cooling Water Pump Room
  • Fire Zone 90A/91A, Spent Fuel Pool Building

Enclosure

4

  • Fire Zone 25, 23 Station Battery
  • Fire Zone 11, Cable Spreading Room
  • Fire Zone 650, Gas Turbine 1 Room
  • Fire Zone 17, Turbine Oil Reservoir Area
  • Fire Zone 47A, 12' and 15' Turbine Building
  • Fire Zone 48A, 3.3' and 15' Turbine Building

Reference material consulted by the inspector included the Fire Protection

Implementation Plan, Pre-Fire Plan, and Station Administrative Orders (SAOs)-700, Fire

Protection and Prevention Policy, SAO-701, Control of Combustibles and Transient Fire

Load, SAO-703, Fire Protection Impairment Criteria and Surveillance, and Calculation

PGI-00433, Combustible Loading Calculation. The inspector identified a number of

minor items related to drawing errors in the pre-fire plan sketch and penetration drawings

and a few minor housekeeping items. The associated condition reports for these minor

issues are identified in the Attachment to this inspection report.

b. Findings

No findings of significance were identified.

.2 Fire Brigade Observation

a. Inspection Scope (71111.05)

On June 13, 2003, the inspectors observed an announced fire brigade drill. The drill was

in accordance with the pre-planned drill scenario for a 21 auxiliary boiler feedwater pump

motor fire. This was a routine training drill for current fire brigade members. The purpose

of this observation was to evaluate the readiness of the licensee's personnel to prevent

and fight fires. The inspector evaluated the following aspects:

  • Protective clothing/turnout gear is properly donned.
  • Self-contained breathing apparatus (SCBA) equipment is properly worn and used.
  • Fire hose lines are capable of reaching all necessary fire hazard locations, are

laid out without flow constrictions, and are simulated being charged with water.

  • Fire area is entered in a controlled manner.
  • Sufficient fire fighting equipment is brought to the scene by the fire brigade.
  • Effective smoke removal operations are simulated.
  • The fire fighting pre-plan strategies are utilized.
  • The licensees pre-planned drill scenario is followed.
  • The drill objectives and acceptance criteria are met.

The inspector reviewed Station Administrative Order (SAO) -706, Fire Brigade

Organization, Operation, and Training, and procedure OASL 15.22, Fire Brigade

Requirements, to confirm the minimum fire brigade manning during the drill was

achieved.

Some minor deficiencies, not impacting the ability of the fire brigade to fight a fire, were

addressed during the drill critique and were entered into the Condition Reporting System

Enclosure

5

(CR-IP2-2003-03778, CR-IP2-2003-03780, CR-IP2-2003-03786, and CR-IP2-2003-

03791).

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance

a. Inspection Scope (71111.07)

The inspector verified that the licensees program was adequate to ensure proper heat

exchanger performance for the Nos. 21 and 22 component cooling water (CCW) heat

exchangers. The inspection consisted of a review of the most recently completed

performance tests, (PT-2Y10A, 21 CCW HX Test, and PT-2Y10B, CCW HX Test,

conducted on October 7, 2002 and October 11, 2002, respectively), examination of the

preliminary engineering calculation No. PGI-00462-01, Component Cooling Water Heat

Exchanger Performance Evaluation, and discussions with the responsible performance

engineer. The inspector noted that these performance tests were conducted just prior to

the Cycle 15 refueling outage in the Fall of 2002, during which both heat exchangers

were opened for a planned clearing and inspection. The inspector also reviewed

completed performance test results for both CCW heat exchangers dating back to July

1991.

b. Findings

No findings of significance were identified.

1R11 Operator Requalification Inspection

a. Inspection Scope (71111.11)

On May 4, 2003, the inspector observed the performance of an operating crew (2C)

during licensed operator re-qualification training. Specifically, the inspector observed a

simulator exam associated with lesson plan ESR-024-007. The inspection was

conducted to assess the adequacy of the training, licensed operator performance,

emergency plan implementation, and the adequacy of the licensees critique.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness

.1 22 EDG Load Test

a. Inspection Scope (71111.12Q)

Enclosure

6

The inspector evaluated Entergys corrective actions for the emergency diesel generator

(EDG) system equipment issues identified during the EDG 22 load test (PT-M21B)

performed on April 23, 2003, to assess the effectiveness of the licensees maintenance

rule implementation. The inspector reviewed the EDG system performance history and

assessed the licensees maintenance rule determination for the equipment issue

involving the two loose fuel line capscrews found during the test (CR IP2-2003-02392).

The inspector reviewed Entergys problem identification and resolution actions for this

issue and evaluated Entergys monitoring, analysis, and disposition of the issues in

accordance with station procedures and 10 CFR 50.65, "Requirements for Monitoring the

Effectiveness of Maintenance." The inspector noted that after correcting the EDG 22

problem, the licensee also checked the same parts in EDG 21 and EDG 23; the problem

observed in EDG 22 was not found in the other two EDGs.

b. Findings

No findings of significance were identified.

.2 Auxiliary Feedwater System

a. Inspection Scope (71111.12Q)

The inspectors evaluated Entergys work practices and preventive maintenance activities

for the auxiliary feedwater system to assess the effectiveness of maintenance activities.

The inspectors reviewed the performance history of the auxiliary feedwater pumps to

assess the adequacy of the licensee's corrective actions and to evaluate Entergys

monitoring, evaluations, and dispositions of issues completed in accordance with station

procedures and the requirements of 10 CFR 50.65, "Requirements for Monitoring the

Effectiveness of Maintenance." The inspectors reviewed the following documents

associated with the system design and licensing basis:

Enclosure

7

Procedures and Documents

  • UFSAR Chapter 10, Steam and Power Conversion System

Quarter 2003

10943; and CR-IP2-2003-00165

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessment and Emergent Work Activities

a. Inspection Scope (71111.13)

The inspector observed selected portions of emergent maintenance work activities to

assess the licensees risk management in accordance with 10 CFR 50.65 (a)(4). The

inspector verified that the licensee took the necessary steps to plan and control emergent

work activities, to minimize the probability of initiating events, and to maintain the

functional capability of mitigating systems. The inspector observed and/or discussed risk

management with maintenance and operations personnel for the following activities:

  • WO IP2-03-05130, troubleshooting power range upper channel high flux deviation

circuits in anticipation of NRC approval of a reactor thermal power uprate

amendment request

  • WO IP2-03-06817, repairs to 22 component cooling water pump discharge check

valve (761B)

reactor protection system bistable setpoints

b. Findings

No findings of significance were identified.

Enclosure

8

1R14 Personnel Performance During Non-Routine Plant Evolutions and Events

.1 Partial Loss of Power Reactor Trip/Turbine Trip

Introduction. On April 28, 2003, offsite electrical disturbances caused a reactor trip and

partial loss of normal power. At 4:41 p.m. a phase A ground fault on 345 KV

transmission line No. Y94 occurred. During the automatic isolation of the Y94 fault, a

138 KV breaker (No. F7) in the Buchanan substation faulted to ground. These two faults

and the response of protective relaying at the Millwood substation, resulted in a loss of

the credited 138 KV power supplies to both Units 2 and 3 for approximately five minutes.

During this period of time, the IP2 main generator was supplying portions of the 345 KV

system through one of its two output breakers. After repeated unsuccessful attempts to

restore the No. Y94 line by the transmission network operator, the remaining 345 KV

output transmission lines for the IP2 generator were lost. This resulted in a full load

reject and trip of the main turbine on over-frequency.

The resultant turbine/reactor trip placed the plant in natural circulation with all three

emergency diesel generators started and two of the four 480 volt safeguards buses

energized by the No. 22 EDG. The event was documented in the licensees corrective

action program via condition report IP2-CR-2003-2511.

This event was similar, with respect to the in-plant consequences, to events on

December 26, 2001 (inspection report 50-247/2001-011) and July 28, 1997 (inspection

report 50-247/97-010).

a. Inspection Scope (71111.14)

The inspector observed operator response to the event, including their use of emergency

operating procedures. The inspector compared the plant response to Updated Final

Safety Analysis Report (UFSAR) section 14.1.13, Turbine Overspeed, UFSAR Section

14.1.8, Loss of Load, and UFSAR Section 14.1.12, Loss of Station Auxiliaries. The

inspector also reviewed licensee corrective actions to prevent recurrence of this event.

The inspector reviewed the transient and compared it to NRC safety evaluation report

dated 1982. The safety evaluation report concluded that no event or condition could

result in the simultaneous or consequential loss of both required circuits from the offsite

power network to the onsite distribution system. The inspector observed the post-trip

review presented to the on-site review committee on April 29, 2003. The inspector

reviewed the post-trip review report as defined in operations administrative directive

(OAD) 23, Post Trip Review and Evaluation Procedure. The inspector also evaluated

equipment issues not associated with the offsite power perturbation.

b. Findings

No findings of significance were identified.

Enclosure

9

.2 Unavailability of Emergency Planning Zone Sirens

a. Inspection Scope

The inspectors evaluated the licensees problem identification and evaluations associated

with a number of emergency planning siren failures, as reported per 10 CFR 50.72

between February 2003, and June 17, 2003. The inspector observed the on-line

monitoring of siren performance at the emergency operations facility, reviewed the

licensees assessment of overall system availability, and discussed proposed corrective

actions with cognizant licensee personnel.

The inspector consulted NRC Manual Chapter 0609, Appendix B for examples of a loss

and/or degraded risk significant planning standard 10 CFR 50.47(b)(5), associated with

the public alert and notification system (ANS). The NRC staff had previously

documented a review of ANS siren failures in report 50-247/2003-003, detail 4OA2.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations

a. Inspection Scope (71111.15)

The inspectors reviewed the below listed condition reports and associated operability

evaluations to ensure that operability was properly justified and that the component or

system remained available, without a significant degradation in performance or

unrecognized operability issue. The inspectors used Technical Specifications, Updated

Final Safety Analysis Report, and design basis documents, as appropriate.

Stop.

Operator

Oil Storage Tanks

Oil

Enclosure

10

b. Findings

Introduction. The inspector identified an incomplete operability evaluation. The

operability evaluation involved the 13.8 KV electrical system during a planned voltage

reduction test.

Description. At 12:55 a.m. on May 28, 2003, the Unit 2 control room was notified by the

Con Edison system operator that a voltage reduction test of 8% would be performed on

the 13.8 KV feeders in the Buchanan Substation. The Con Edison system operator

subsequently notified the control room that the test was completed at 2:24 a.m. The

inspector determined that Entergy received no prior notification of the test and that there

was no evaluation performed by Entergy to determine the impact of the test on 13.8 KV

offsite power operability. Further, Operations Department personnel did not question the

impact of the voltage test on operability of the 13.8 KV system. Based upon the

inspectors expressed concern that an operability evaluation had not been performed in

response to the planned reduced voltage test of the 13.8 KV system, Entergy initiated CR

IP2-2003-3470 to address operability of the system and to evaluate the needed protocols

between Con Edison and Entergy for the coordination of offsite electrical power system

testing.

The licensees operability evaluation determined that if a safety injection and a loss of

offsite power (138 KV system) occurred, the emergency diesel generators would

automatically start and operators would manually place the generators on the safeguards

buses. The operability evaluation also documented that operators would manually tie the

13.8 KV source to safeguard buses. The inspector identified that the operability

evaluation did not evaluate if the reduced voltage condition could support in-plant

accident loads, as defined in the Technical Specification bases. Further, the operability

evaluation did not address communication protocols between the distribution company

and Entergy in the event there is a need to restore the 13.8 KV system during a

postulated loss of normal power (138 KV). Abnormal operating instruction AOI-27.1.1,

Loss of Normal Power, does not tie the 13.8 KV system to the safeguards buses unless

it is predicted that the preferred 138 KV system will not be available for greater than 30

minutes. Notwithstanding, the licensee concluded that there was reasonable assurance

of system operability during the short duration degraded voltage test.

Analysis. The inspectors concluded that the licensees operability evaluation was

incomplete based on the absence of an evaluation of in-plant accident electrical loads

that would be supplied by the 13.8 kV power feed and the absence of established

communication protocols between Entergy and Con Edison for the control of degraded

system voltage testing. The inspectors referenced NRC Generic Letter (GL) 91-18,

Information to Licensees Regarding Two NRC Manual Sections on Resolution of

Degraded and Non-Conforming Conditions and on Operability, in support of their

conclusion. GL 91-18 states, in part, that when a systems capability is degraded to a

point where it cannot perform with reasonable assurance of reliability, the system should

be judged inoperable. Entergy did not document sufficient basis for their operability

evaluation or provide appropriate guidance to plant operators and the distribution

operator in the event of a condition which warranted use of the 13.8 KV electrical power

feed.

Enclosure

11

The inspectors used NRC Manual Chapter 0612, Appendix B, to disposition this issue.

The finding was more than minor because it impacted the attribute of the mitigating

system cornerstone objective. Specifically, the cornerstone objective is to ensure that

the 13.8 KV system is capable of performing its safety function during a postulated loss

of normal power event without undesirable consequences. This finding was determined

to be of low safety significance because the degraded conditions was of a short duration

and there was no actual loss of the normal offsite power supplies (138 KV or 13.8 KV)

during the test. (FIN 50-247/2003-007-01)

Enforcement. The incomplete operability evaluation does not represent a violation of

regulatory requirements.

1R16 Operator Workarounds

a. Inspection Scope (71111.16)

The inspector reviewed the licensees list of active operator burdens to assess the

cumulative effects on system reliability, availability, and potential for mis-operation of a

system. The inspector also toured various areas of the plant to evaluate deficient

conditions and potential impact to operators during EOP or AOP usage. At the time of

the inspection, no operator work-arounds were identified by Entergy. The inspector used

OASL 15.43, Operator Burden Program as a reference for this review.

b. Findings

No findings of significance were identified.

The inspector noted that one deficiency in the primary auxiliary building that impacted

operators during emergency operating procedure usage (ES 1.4, Transfer to Hot Leg

Recirculation) was a failed breaker handle for valve 856A, 21 safety injection cold leg

isolation valve. Entergy added this deficiency as an operator workaround and repaired

the breaker handle on the motor control center during the inspection period.

1R19 Post Maintenance Testing

a. Inspection Scope (71111.19)

The inspector reviewed post-work test (PWT) procedures and associated testing

activities to assess whether: 1) the effect of testing in the plant had been adequately

addressed by control room personnel; 2) testing was adequate for the maintenance work

order (WO) performed; 3) acceptance criteria were clear and adequately demonstrated

operational readiness consistent with design and licensing documents; 4) test

instrumentation had current calibrations, range, and accuracy for the application; and 5)

test equipment was removed following testing.

The selected testing activities involved components that were risk significant as identified

in the IP2 Individual Plant Examination. The regulatory references for the inspection

Enclosure

12

included Technical Specification 6.8.1.a. and 10 CFR 50, Appendix B, Criteria XIV,

Inspection, Test, and Operating Status. The following testing activities were evaluated:

Valve 866D

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing

a. Inspection Scope (71111.22)

The inspector reviewed surveillance test procedures and observed testing activities to

assess whether: 1) the test preconditioned the component tested; 2) the effect of the

testing was adequately addressed in the control room; 3) the acceptance criteria

demonstrated operational readiness consistent with design calculations and licensing

documents; 4) the test equipment range and accuracy was adequate and the equipment

was properly calibrated; 5) the test was performed per the procedure; 6) the test

equipment was removed following testing; and 7) test discrepancies were appropriately

evaluated. The surveillance tests observed were based upon risk significant components

as identified in the IP2 Individual Plant Examination. The regulatory requirements that

provided the acceptance criteria for this review were 10 CFR 50, Appendix B, Criterion V,

Instructions, Procedures, and Drawings, Criterion XIV, Inspection, Test, and Operating

Status, Criterion XI, Test Control, and Technical Specifications 6.8.1.a. The following

test activities were reviewed:

  • PT-Q33A, 21 Charging Pump
  • PI-M2, Containment Building Inspection
  • TOI 213 Tavg Increase to 562 degrees F
  • PT-Q31A, 21 Auxiliary Component Cooling Pump

b. Findings

No findings of significance were identified.

During the monthly containment inspection, the inspectors noted that the non-licensed

operator was not familiar with the location of the seven conoseals on top of the reactor

vessel head. The inspection was re-performed on April 18 and confirmed there was no

leakage from the flanged instrument tubes on top of the vessel head. The visual

inspection of the conoseal flanges was a commitment to NRC Generic Letter 88-05,

Enclosure

13

Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR

Plants.

1EP4 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope (71114-04)

During an in-office inspection conducted May 19 - 20, 2003, the inspectors reviewed

recent changes to Emergency Plan documents as stated in the attachment to this report.

A thorough review was conducted of aspects of the plan relating to the risk significant

planning standards (RSPS), such as classifications, notifications, and protective action

recommendations. A general review was conducted for non-RSPS portions. These

changes were reviewed against 10 CFR 50.54(q) to ensure that the changes do not

decrease the effectiveness of the plan, and that the changes to meet the standards of 10

CFR 50.47(b) and the requirements of Appendix E. All of the changes made to the

Emergency Plan or implementing procedures are subject to future inspections to ensure

that the results of the changes continue to meet NRC regulations.

b. Findings

No findings of significance were identified.

2. RADIATION SAFETY

Cornerstone: Occupational Radiation Safety (OS)

2OS1 Access Control to Radiologically Significant Areas

a. Inspection Scope (71121.01)

The inspector reviewed radiological work activities and practices and procedural

implementation during tours and observations of the facilities. Additionally, the inspector

reviewed procedures, records, and other program documents to evaluate the

effectiveness of access controls to radiologically significant areas.

On April 9, 2003, the inspector toured and observed work activities in selected portions of

the fuel handling building and the chemical systems building in Unit 1, including the area

in the sphere annulus area where the pipe from the north curtain drain was located. On

April 10, the inspector, accompanied by the Technical Support Manager, toured and

observed work activities on various elevations in the primary auxiliary and fuel handling

buildings in Unit 2. At the routine radiologically controlled area (RCA) access control

point, the inspector observed radiation workers logging into the RCA on radiological work

permits (RWPs) using electronic dosimeters and observed radiation workers exiting the

RCA and then logging out of their RWPs. The inspector examined the use of personnel

dosimetry and the radiological briefings for radiation workers.

On May 21, 2003, the inspector toured and observed work activities in selected portions

of the fuel handling building and of the chemical systems building in Unit 1, including the

Enclosure

14

area in the sphere annulus area where the pipe from the north curtain drain was located.

On May 21 and 23, the inspector toured and observed work activities on various

elevations in the primary auxiliary, fuel handling, and maintenance and outage buildings

in Unit 2. On May 22, inspectors examined the outside and inside of the old steam

generator storage building. Also, during these walkdowns, the inspector observed and

verified the appropriateness of the posting, labeling, and barricading of radioactive

material, radiation, contamination, high radiation, and locked high radiation areas. The

inspector reviewed work activities by both radiation workers and radiation protection

technicians for compliance with the RWP requirements and radiological protection

procedures. Specifically, the radiological controls for replacing packing on the No. 22

charging pump, covered by radiological work permit No. 032023, were reviewed and

observed.

The inspector reviewed radiological work activities and practices and procedural

implementation during tours and observations of the facilities and inspected procedures,

records, and other program documents to evaluate the effectiveness of Entergys access

controls to radiologically significant areas.

The inspector performed a selective examination of program documents (reference the

List of Documents Reviewed) to evaluate the adequacy of radiological controls. The

review was against criteria contained in 10 CFR 19.12, 10 CFR 20 (Subparts D, F, G, H,

I, and J), Technical Specifications, and site procedures.

b. Findings

No findings of significance were identified.

2OS2 ALARA Planning and Controls

a. Inspection Scope (71121.02)

The inspector reviewed the effectiveness of Entergys program to maintain occupational

radiation exposure as low as is reasonably achievable (ALARA).

During the inspection week, the inspector discussed the Unit 2 cumulative dose result for

2002 (248 person-rem) and the three-year-average (2000 through 2002) cumulative dose

result for Unit 2 (279.2 person-rem) with the Indian Point Energy Center (IPEC) Technical

Support Manager and the Radiation Protection Manager. The inspector also discussed

the actual versus projected cumulative year-to-date dose results for 2003 for Units 1 and

2 with the Radiation Protection Manager.

During the inspectors tour of Unit 2 on April 10, 2003, accompanied by the Technical

Support Manager, the inspector examined the decontamination efforts accomplished

during the first quarter of the year and reviewed the planned source term and exposure

reduction efforts anticipated over the next five years.

The inspector performed a selective examination of documents (reference the List of

Documents Reviewed) for regulatory compliance and for adequacy of control of radiation

Enclosure

15

exposure. The review was against criteria contained in 10 CFR 20.1101 (Radiation

protection programs), 10 CFR 20.1701 (Use of process or other engineering controls),

and site procedures.

The inspector performed a selective examination of procedures and program documents

(List of Documents Reviewed Attachment) for regulatory compliance and for adequacy of

control of radiation exposure. The review was against criteria contained in 10 CFR

20.1101 (Radiation protection programs), 10 CFR 20.1701 (Use of process or other

engineering controls), and site procedures.

b. Findings

No findings of significance were identified.

2OS3 Radiation Monitoring Instrumentation and Protective Equipment

a. Inspection Scope (71121.03)

The inspector reviewed the program for health physics instrumentation to determine the

accuracy and operability of the instrumentation.

During the plant tours described in Section 2OS1 of this report, the inspector reviewed

field instrumentation utilized by health physics technicians and plant workers to measure

radioactivity and radiation levels, including portable field survey instruments, hand-held

contamination frisking instruments, and continuous air monitors. The inspector also

reviewed installed radiation monitors including whole body friskers, portal monitors, area

monitors, and process monitors. The inspector verified current calibration, source

checking, and proper instrument function. The inspector also identified and noted the

condition, operability, and calibration status of selected installed area and process

radiation monitors and any accessible local indication information for those monitors.

The inspector performed a selective examination of documents (reference the List of

Documents Reviewed) for regulatory compliance and adequacy. The review was against

criteria contained in 10 CFR 20.1501, 10 CFR 20 Subpart H, site Technical

Specifications, and site procedures.

On May 21, the inspector met with the cognizant radiological engineer to discuss the

corrective actions for the issue identified in CR-IP2-2002-04583. The issue involved the

need to periodically evaluate and document the impact that difficult-to-detect

radionuclides have on the detection capabilities and limits of the contamination

monitoring instrumentation in use. The inspector also reviewed selected use/calibration

procedures (reference the List of Documents Reviewed) for this instrumentation.

b. Findings

No findings of significance were identified.

3. SAFEGUARDS

Enclosure

16

3PP4 Security Plan Changes

a. Inspection Scope (71130.04)

An in-office review was conducted of changes to the Security Plan, identified as

Revision 21A, submitted to the NRC on August 16, 2002, in accordance with the

provisions of 10 CFR 50.54(p). The review was conducted to confirm that the changes

were made in accordance with 10 CFR 50.54(p), and did not decrease the effectiveness

of the plan.

The NRC recognizes that some requirements contained in this program plan may have

been superceded by the February 25, 2001 interim compensatory measures order.

An in-office review was conducted of changes to the licensees Training and Qualification

Plan identified as Revision 0. This document was submitted to the NRC on December 9,

2002, in accordance with the provisions of 10 CFR 50.54(p). The review was conducted

to confirm that the changes were made in accordance with 10 CFR 50.54(p), and did not

decrease the effectiveness of the Training and Qualification Plan. The NRC recognizes

that some requirements contained in the Training and Qualification Plan may have been

superceded by the February 2002 Interim Compensatory Measures Order.

b. Findings

No findings of significance were identified.

Enclosure

17

4. OTHER ACTIVITIES (OA)

4OA1 Performance Indicator Verification

The inspector reviewed the licensees performance indicator (PI) data collecting and

reporting process as described in procedure SAO-114, Preparation of NRC and WANO

Performance Indicators. The purpose of the review was to determine whether the

methods for reporting PI data are consistent with the guidance contained in Nuclear

Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator

Guidelines, Revisions 1 and 2. The inspection included a review of the indicator

definitions, data reporting elements, calculation methods, definition of terms, and

clarifying notes for the performance indicators. Plant records and data were sampled

and compared to the reported data. The inspector reviewed the licensees actions to

address and satisfactorily resolve discrepancies in the performance indicator data.

.1 Unplanned Power Changes Greater than 20% over 7000 Critical Hours

a. Inspection Scope (711151)

The inspector performed a periodic review of the 3rd and 4th quarters of 2002 and the 1st

quarter of 2003 performance indicator (PI) data submitted by the licensee for the

unplanned power changes greater than 20% over 7000 critical hours to determine

accuracy and completeness. The inspectors researched the control room operating logs

and the condition reporting system to identify power reductions greater than 20% during

these quarters. The inspectors compared the PI data against the guidance contained in

NEI 99-02.

b. Findings

No findings of significance were identified.

.2 Safety System Unavailability - Auxiliary Feedwater

a. Inspection Scope (71151)

The inspector reviewed Entergys PI data for Auxiliary Boiler Feedwater (ABFW) Safety

System Unavailability to verify that the PI data was accurate and complete. The

inspectors compared the PI data reported by the licensee to information gathered from

the control room logs, condition reports, and work orders for the 2nd, 3rd, and 4th quarters

of 2002. In addition, the inspectors interviewed the system engineers. The inspectors

compared the PI data against the guidance contained in NEI 99-02

b. Findings

No findings of significance were identified.

.3 Reactor Coolant System Specific Activity

Enclosure

18

a. Inspection Scope

The inspector reviewed the PI for reactor coolant system (RCS) specific activity for the

period from January 2002 - March 2003. The RCS specific activity PI is reported as a

percentage of the maximum Technical Specification limit for dose equivalent iodine-131

in micro-Curies per cubic centimeter. For the period reviewed, this PI remained in the

Green band. The inspector reviewed monthly average RCS sample results based upon

daily samples obtained per IPC-S-009-S, Primary Sampling System Sentry. The

inspector also observed a daily sample on May 14, 2003. The inspector compared the PI

data against the guidance contained in NEI 99-02.

b. Findings

No findings of significance were identified.

.4 Scrams With Loss of Normal Heat Removal

a. Inspection Scope

The inspector reviewed the PI for scrams with loss of normal heat removal (LNHR) for

the period from January 2002 - March 2003 (the inspector notes that Entergys PI data

for Unplanned Scrams Per 7,000 Critical Hours was reviewed during the previous

quarterly inspection, reference report No. 50-247/2003-003). The scrams with LNHR PI

monitors the number of unplanned scrams while critical, during the previous 12 quarters,

that involved a loss of the normal heat removal path through the main condenser. The

inspector reviewed operator logs, licensee event reports, and monthly operating reports

to compare PI data reported by the licensee. The inspector compared the PI data

against the guidance contained in NEI 99-02.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems

.1 Annual Sample Associated with the Failure of No. 22 Component Cooling Water (CCW)

Pump

a. Inspection Scope (71152

The inspector reviewed the corrective actions associated with the failure of No. 22 CCW

pump which occurred on December 5, 2002. This failure was documented in condition

report No. IP2-2002-11242. The pump failure involved inadequate lubrication of the

inboard bearing. This condition was caused by the oiler, which maintains the bearing

housing oil level, being bent in the downward direction thus lowering the oil level at the

bearing. The corrective actions were completed and the condition report closed out on

May 9, 2003. The inspector reviewed the condition report, and its associated apparent

cause evaluation and corrective actions, to verify that the cause(s) of the pump failure

Enclosure

19

was properly identified and evaluated, the corrective actions were appropriate to resolve

the problem, and that actions were properly implemented in a timely manner,

commensurate with the safety significance. The inspector discussed this event with the

system engineer and the Corrective Actions Department staff. The inspector also

reviewed the vendor technical manual and the design change package (MSAP-91-

00014-PGI).

b. Findings

Introduction. A Green non-cited violation was identified by the inspectors. This finding is

based on the failure of Entergy to properly evaluate and take effective corrective actions

associated with the failure of the No. 22 CCW pump on December 5, 2002, resulting

from an improper material substitution on the inboard bearing oil level indication system

performed in March 2000. This finding was determined to be a violation of 10 CFR 50,

Appendix B, Criterion XVI.

Description. On March 15, 2000, the licensee performed corrective maintenance to

resolve excessive oil leaks on the No. 22 CCW pump piping between the bearing

housing and oiler. This work was performed under work order No. NP-99-1111. The

original equipment piping was 1/4-inch schedule 80 carbon steel and consisted of two

lengths of threaded pipe connected by a 90 degree elbow. Due to recurrent oil leaks

from the threaded connections, the original piping was replaced with 3/8-inch stainless

steel tubing. The need for an elbow was negated by placing a 90-degree bend in the

tubing. This change in configuration effectively resolved the problem with oil leaks. No

engineering analysis was performed for this change since it was determined to be a

below the level of detail change in system configuration.

On December 5, 2002, No. 22 CCW pump failed due to inadequate lubrication of the

inboard bearing. Upon licensee investigation it was determined that the oiler was at a

lower position than required due to the tubing being bent in a downward direction at the

90-degree bend. The licensee noted that the tubing was not as rigid as the original

configuration and returned the system back to its original carbon steel piping design.

Analysis. The inspector-identified performance deficiency is an ineffective evaluation

and corrective actions for safety-related equipment, as documented in the December

2002 condition report No. IP2-2002-11242. While the unapproved 2000 modification did

not directly cause the pump to fail, it did introduce a new failure mode because the

stainless steel tubing was less rigid and could be more easily bent (a causal factor of the

pump failure). While the licensee stated that the bearing lubrication piping configuration

was thought to have been outside the scope of a modification, inspector review of the

vendor technical manual drawings identified a clearly depicted threaded pipe

configuration. In addition, design change MSAP-91-00014-PGI credited the 1/4-inch

schedule 80 carbon steel pipe in its seismic analysis assumptions. The inspectors

review of condition report (CR) No. IP2-2002-11242 identified that no evaluation was

conducted to determine how the replacement of the piping occurred without an

engineering evaluation or how the configuration change bypassed the modification

process. The CR evaluation also overlooked the fact that the seismic analysis had been

invalidated by the substitution of stainless steel tubing for threaded carbon steel pipe. No

Enclosure

20

action was taken by Entergy to confirm that the modified configuration could have

performed its safety function in a design basis event between March 2000 and December

2002.

This finding is more than minor since it is associated with the design control attribute of

the mitigating systems cornerstone and affected the cornerstone objective. The

inspectors conducted a Phase 1 SDP screening and determined that the failure to

adequately maintain design control on No. 22 CCW pump was of a very low safety

significance since the redundant train components were operable and unaffected by this

unauthorized tubing modification.

Enforcement. 10 CFR 50, Appendix B, Criterion XVI, states that measures shall be

established to assure that nonconformances are promptly identified and corrected.

Contrary to the above, Entergys evaluation and corrective actions associated with CR

IP2-2002-11242 did not address appropriate corrective actions associated with how an

unapproved modification to the No. 22 CCW pump bearing housing and oiler piping

configuration was made in March 2000. Because this failure to implement adequate

corrective actions is of very low safety significance and has been entered into the

licensees corrective actions program (CR-IP2-2003-03652), this violation is being treated

as a non-cited violation, consistent with Section VI.A of the NRC Enforcement Policy.

(NCV 50-247/2003-007-02). Entergy has proposed corrective actions to reinforce with

maintenance and engineering personnel that configuration control changes require

proper review and approval.

4OA6 Meetings, Including Exit

The inspectors met with Indian Point 2 representatives at the conclusion of the inspection

on July 9, 2003. At that time, the purpose and scope of the inspection were reviewed,

and the preliminary findings were presented. The licensee acknowledged the preliminary

inspection findings.

The inspector asked the licensee whether any materials examined during the inspection

should be considered proprietary. No proprietary information was reviewed during this

inspection.

Enclosure

A-1

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Entergy:

W. Axelson Support Supervisor

S. Baer HP Supervisor

J. Barry Radiological Engineer

R. Cranker Radiation Protection Technician

M. Dampf Health Physics Manager

R. Deschamps Radiological Protection Superintendent

R. Decensi Technical Support Manager

J. Dirset Radiation Protection Technician

R. Fucheck HP Supervisor

D. Gately Radiation Protection Coordinator

B. Kyler HP Supervisor

R. LaVera Radiological Engineer

R. Majes Radiological Engineer

R. Richards HP Supervisor

R. Rodino Radiological Engineer

E. Salisbury Radiological Engineer

W. Scholtens Waste Management Contractor

R. Solanto HP Supervisor

J. Stewart HP Supervisor

R. Tagliomonte Waste Management Supervisor

G. Tiberi Dosimetry Technician

N. Azevedo ISI Supervisor

W. Axelson Support Supervisor

T. Burns Environmental Supervisor

J. Comiotes Director, Nuclear Safety Assessment

L. Cortopassi IP3 Training Manager

F. Dacimo Vice-President

S. Davis IP2 Licence Operator Requalification Training Supervisor

J. DeRoy General Manager Plant Operations, IP3

K. Finucan Emergency Planning Staff

K. Finvean Reactor Vessel Head Inspection, Assistant Project Manager

M. Gillman IP3 Operations Manager

L. Glander Dosimetry Supervisor

J. Goebel Reactor Vessel Head Inspection, Project Manager

F. Inzirillo Manager Emergency Planning

T. Jones Nuclear Safety and Licensing

J. McCann Nuclear Safety and Licensing Manager

F. Mitchell HP Supervisor

D. Pace VP - Engineering - ENN

J. Perrotta Quality Assurance Manager

R. Penny Manager, Engineering Programs

K. Richett HP Technician

Attachment

A-2

R. Sachatello Radiological Consultant

C. Schwarz General Manager, IP2 Plant Operations

G. Schwartz Chief Engineer

H. Salmon Quality Assurance Director

M. Smith Director of IP3 Engineering

R. Solanto HP Supervisor

S. Stevens HP Technician

J. Stewart HP Supervisor

D. Sullivan-Weaver Emergency Planning Staff

J. Tuohy Design Engineering Manager

J. Wheeler Site Training Manager

F. Wilson Superintendent, Operations Training

M. Wilson Emergency Planning Staff

J. Bonner Entergy Northeast Offsite power coordinator

R. Burroni Supr. I&C Maintenance

P. Gropp Manager, DBI Program

T. Jones Licensing

T. Klein Senior Design Engineer

R. Milici Supr. Electrical Design Engineer

T. McCaffrey Supr. Electrical System Engineer

S. Petrosi Manager, Design Engineering

J. Quirk DBD Coordinator

J. Raffaele Sr. Lead Electrical Engineer, DBI Project

J. Tuohy Manager, Engineering Support

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Open/Closed

50-247/03-07-01 FIN Inadequate operability evaluation for the 13.8 KV system

50-247/03-07-02 NCV Ineffective corrective actions associated with an unauthorized

modification to the 22 component cooling water pump

Attachment

A-3

LIST OF CONDITION REPORTS GENERATED

IP2-200303443 Small hole in the mortar joint between concrete blocks in the south wall of

the 23 battery room

IP2-2003-3464 Safety evaluation 95-058-EV reclassification of Fire Zone 25 barriers as

unclassified from Type 1" was not approved by on-site review committee

IP2-2003-2877 Valve 1843B, 22 Safety Injection Pump thrust balance line as a slight

packing leak

IP2-2003-2876 Scaffolding exists in the area of the Refueling Water Storage Tank without

scaffolding tags in place

IP2-2003-2878 22 Safety Injection Pump has evidence of seal leakage at the pump

inboard seal. Dried boron, not active.

IP2-2003-2959 PI-M2 Vapor containment monthly inspection was performed and

information on the conoseal location and schematic drawings were not

incorporated into the procedure.

IP2-2003-03940 Problems with valve labeling and abnormal operating instruction (AOI)

11.1

IP2-2003-2597 NRC observed during plant startup that both PORV block valves were

open contrary to check off list COL 01-01 during plant startup

IP2-2003-04245 Programmatic failure on closure of condition reports to PCNs

IP2-2003-3652 Failure of the 22 Component Cooling Water Bearing

IP2-2003-2595 Seal Injection Flow Momentarily to zero flow during charging pump speed

adjustments

Attachment

A-4

LIST OF DOCUMENTS REVIEWED

Indian Point Energy Center Emergency Plan, Rev 03-01

IP-EP-115, Emergency Plan Forms, Rev 1, 2

IP-EP-130, Emergency Notifications and Mobilization, Rev 0

IP-EP-250, Emergency Operations Facility, Rev 0

IP-EP-251, Alternate Emergency Operations Facility, Rev 1

IP-EP-255, Emergency Operations Facility Management and Liaisons, Void

IP-EP-260, Joint News Center, Rev 0

IP-EP-310, Dose Assessment, Rev 1

IP-EP-410, Protective Action Recommendations, Rev 1

IP-EP-510, Meteorological, Radiological & Plant Data Acquisition System, Rev 1

IP-EP-520, Modular Emergency Assessment & Notification System (MEANS), Rev 1

IP-EP-610, Emergency Termination and Recovery, Rev 1

IP-EP-620, Estimation of Total Population Exposure, Rev 1

IP-1002, Emergency Notification and Communication, Rev 29, Void (IP2)

IP-1010, Central Control Room, Rev 9, 10 (IP2)

IP-1011, Joint News, Void (IP2)

IP-1015, Radiological Monitoring Outside the Protected Area, Rev 11 (IP2)

IP-1019, Coordination of Corporate Response, Void (IP2)

IP-1030, Emergency Operations Facility, Void (IP2)

IP-1038, Offsite Emergency Notifications, Void (IP3)

IP-2001, ED, POM, Shift Managers Procedure, Rev 18 (IP3)

IP-2005, CR Offsite Communicator, Void (IP3)

IP-2300, Emergency Activation of the Emergency Operations Facility, Void (IP3)

IP-2302, EOF Technical Advisor & Information Liaison, Void (IP3)

IP-2303, EOF Radiological Assessment Team Leader, Void (IP3)

IP-2304, EOF Dose Assessment Health Physicist, Void (IP3)

IP-2305, EOF Midas Operator, Void, (IP3)

IP-2306, EOF Security Officer, Void (IP3)

IP-2307, EOF Clerk, Void (IP3)

IP-2308, EOF Direct Line Communicator, Void (IP3)

IP-2309, EOF Offsite Communicator, Void (IP3)

IP-2310, EOF Onsite Radiological Communicator, Void (IP3)

IP-2311, EOF Offsite Radiological Communicator, Void (IP3)

Section 2OS1, Access Control to Radiologically Significant Areas

- Self-assessment, HP access control system, April 6, 2003

- Self-assessment, Remote HP monitoring/3-year period, January 16, 2003

Section 2OS2, ALARA Planning and Controls

- IP1 daily ALARA information for April 6, 2003

- IP2 daily ALARA information for April 6, 2003

- ALARA suggestion program inputs (March/April 2003)

- IPEC ALARA committee meeting presentation handout for March 27, 2003

- Indian Point Energy Center/Radiation protection/Strategic plan for exposure

Attachment

A-5

reduction, 2003 - 2008

Section 2OS3, Radiation Monitoring Instrumentation and Protective Equipment

- Self-assessment, Review of dosimetry records, May 9, 2002 to February 14,

2003

Sections 1EP4, Emergency Action Level and Emergency Plan Changes

Emergency Plan for Indian Point Unit Nos 1 and 2, Rev 01-02a

Indian Point 3 Emergency Plan, Rev 46

Indian Point Energy Center Emergency Plan, Rev 02-01

IP-EP-251, Alternate Emergency Operations Facility, Rev 0

IP-EP-255, Emergency Operations Facility Management and Liaisons, Rev 1, 2

IP-EP-310, Dose Assessment, Rev 0

IP-EP-410, Protective Action Recommendations, Rev 0

IP-EP-510, Meteorological, Radiological & Plant Data Acquisition System, Rev 0

IP-EP-520, Modular Emergency Assessment & Notification System (MEANS), Rev 0

IP-EP-610, Emergency Termination and Recovery, Rev 0

IP-EP-620, Estimation of Total Population Exposure, Rev 0

IC/EALs, Initiating Conditions & Emergency Action Levels, Rev 9 (IP3)

IP-1001, Determining the Magnitude of Release, Rev 17, Void (IP3)

IP-1002, Emergency Notification and Communication, Rev 27, 28 (IP2)

IP-1003, Obtaining Meteorological Data, Rev 18, Void (IP3)

IP-1004, MIDAS Computer System, Rev 16, Void (IP3)

IP-1010, Central Control Room, Rev 6, 7, 8 (IP2)

IP-1013, Protective Action Recommendations, Rev 8, (IP2)

IP-1015, Radiological Monitoring Outside the Protected Area, Rev 10 (IP2)

Section 2OS1, Access Control to Radiologically Significant Areas

- SAO-302, Rev. 19, Radiation work permit (RWP) program

- HP-SQ-3.005, Rev. 24, HP work scheduling

- HP-SQ-3.013, Rev. 12, Routine surveys outside normal RCA

- HP-SQ-3.109, Rev. 28, Control of high radiation, locked high radiation, special

locked high radiation, and very high radiation areas

- RWP 032023, Rev. 0, Charging pump maintenance, repair, and test; Task 22-

Work in 22 charging pump cubicle

- Self-assessment of Technical Support Integration

Attachment

A-6

Section 2OS2, ALARA Planning and Controls

- RS-S-8.005, Rev. 5, ALARA cost-benefit analysis tracking log

- RS-SQ-8.006, Rev. 6, Radiological support ALARA design review

- RS-SQ-8.101, Rev. 5, Temporary shielding program

- IP1 daily ALARA information for May 18, 2003

- IP2 daily ALARA information for May 18, 2003

Condition Reports

199807874

199906643

200000978

200009088

200009884

200009892

200100559

200100750

200205411

200205377

200205379

200205384

200206578

200207521

200207918

200208627

200303284

Design Basis Documents

Emergency Diesel Generators

480 Volt AC System

Procedures

ENN-LI-102, Rev. 2, Corrective Action Process

SAO-112, Rev. 6, Condition Reporting Process (superceded by ENN-LI-102)

Section 2OS3, Radiation Monitoring Instrumentation and Protective Equipment

- HP-9.012, Rev. 4, Operation of Eberline RM-14 and RM-20

- HP-9.067, Rev. 8, Calibration of the Eberline PCM-1A/1B

- HP-9.082, Rev. 1, Efficiency check of G-M pancake type detectors

- HP-9.512, Rev. 5, Calibration of the Eberline RM-14/RM-20 radiation monitors

- HP-9.590, Rev. 1, Calibration of Eberline tool contamination monitor

- HP-9.591, Rev. 4, Calibration procedure for National Nuclear Model Gamma-

40/60 portal monitor

- HP-9.593, Rev. 2, Calibration and operation of Eberline gamma tool monitor

Attachment

A-7

(GTM)

- HP-SQ-3.002, Rev. 17, Equipment and materials release requirements

- HP-SQ-3.011, Rev. 17, Radiation and contamination survey techniques

- Health physics continuing training lesson plan, LP No. HCT0301.01,

Instrumentation sensitivity and contamination control, Rev. 0

- Assessment of the sensitivity to internal contamination for the PCM-1B and

Gamma 60 personnel contamination monitors at IP2, March 2002, by

Cabrera Services

LIST OF BASELINE INSPECTIONS PERFORMED

71111.01 Adverse Weather 1R01

71111.04 Equipment Alignment 1R04

71111.05 Fire Protection 1R05

71111.07 Heat Sink Performance 1R07

71111.11 Operator Requalification 1R11

71111.12 Maintenance Effectiveness 1R12

71111.13 Maintenance Risk Assessment and Emergent Work Activities 1R13

71111.14 Personnel Performance During Non-Routine Plant Evolutions 1R14

71111.15 Operability Evaluations 1R15

71111.16 Operator Workarounds 1R16

71111.19 Post Maintenance Testing 1R19

71111.22 Surveillance Testing 1R22

71111.23 Temporary Plant Modifications 1R23

71114-04 Emergency Action Level and Emergency Plan Changes 1EP4

71121.01 Access Control to Radiologically Significant Areas 2OS1

71121.02 ALARA Planning and Controls 2OS2

71121.03 Radiation Monitoring Instrumentation and Protective Equipment 2OS3

71151 Performance Indicator Verification 4OA1

71152 Problem Identification and Resolution Sample 4OA2

71153 Event Followup 4OA3

LIST OF ACRONYMS

ABFW auxiliary boiler feedwater

AFW auxiliary feedwater

AFWP auxiliary feedwater pump

ALARA as low as reasonably achievable

AOI abnormal operating instruction

BATP boric acid transfer pump

CAP corrective action program

CCR central control room

CCW component cooling water

CFR Code of Federal Regulations

COL check off list

CR condition report

EDG emergency diesel generator

EFPY effective full power years

Attachment

A-8

OAD operations administrative directive

EOF emergency operations facility

EP emergency preparedness

HRA high radiation area

IP Indian Point

IP2 Indian Point Unit 2

IPEC Indian Point Energy Center

IPEEE individual plant examination for external events

kV kilo-volt

LNHR loss of normal heat removal

NCV non-cited violation

NEI Nuclear Energy Institute

NRC Nuclear Regulatory Commission

NRR Nuclear Reactor Regulation

OS occupational safety

PI performance indicator

PM post maintenance

PS public radiation safety

PT penetrant testing

PWT post-work test

RCA radiologically controlled area

RCS reactor coolant system

RMS radiation monitoring system

RPM radiation protection manager

RPS reactor protection system

RSPS risk significant planning standard

RTP rated thermal power

RV reactor vessel

RWP radiation work permit

SAO station administrative order

SDP significance determination process

SI safety injection

SOP system operating procedure

TA temporary alteration

TS technical specifications

UFSAR Updated Final Safety Analysis Report

V volt

WO work order

Attachment

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