Text

IR 05000341-06-003; 04/01/2006-06/30/2006; Fermi Power Plant, Unit 2; Fire Protection, Maintenance Risk Assessment, Operability Evaluations, Refueling and Outage Activities
	ML062160540
Person / Time
Site:	Fermi
Issue date:	08/02/2006
From:	Christine Lipa; NRC/RGN-III/DRP/RPB4
To:	Cobb D; Detroit Edison
References
	IR-06-003
	Download: ML062160540 (48)
	v • d • e

August 2, 2006

Mr. Donald K. Cobb

Assistant Vice President

Nuclear Generation

Detroit Edison Company

6400 North Dixie Highway

Newport, MI 48166

SUBJECT: FERMI POWER PLANT, UNIT 2, NRC INTEGRATED

INSPECTION REPORT 05000341/2006003

Dear Mr. Cobb:

On June 30, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed an integrated

inspection at your Fermi Power Plant, Unit 2. The enclosed report documents the inspection

findings which were discussed on July 11, 2006, with you and other members of your staff.

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

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, five findings of very low safety significance were

identified, all of which involved violations of NRC requirements. However, because these

findings were of very low safety significance and because the issues were entered into your

corrective program, the NRC is treating these findings as Non-Cited Violations in accordance

with Section VI.A.1 of the NRCs Enforcement Policy. If you contest the subject or severity of

a Non-Cited Violation, you should provide a response within 30 days of the date of this

inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission,

ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional

Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road,

Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory

Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Fermi 2

facility.

In accordance with 10 CFR 2.390 of the NRCs Rules of Practice, a copy of this letter and

its enclosure will be made available electronically for public inspection in the NRC Public

D. Cobb -2-

Document Room or from the Publicly Available Records (PARS) component of 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).

Sincerely,

/RA/

Christine A. Lipa, Chief

Branch 4

Division of Reactor Projects

Docket No. 50-341

License No. NPF-43

Enclosure: Inspection Report 05000341/2006003

w/Attachment: Supplemental Information

cc w/encl: K. Hlavaty, Plant Manager

R. Gaston, Manager, Nuclear Licensing

D. Pettinari, Legal Department

Michigan Department of Environmental Quality

Waste and Hazardous Materials Division

M. Yudasz, Jr., Director, Monroe County

Emergency Management Division

Supervisor - Electric Operators

State Liaison Officer, State of Michigan

Wayne County Emergency Management Division

D. Cobb -2-

Document Room or from the Publicly Available Records (PARS) component of 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).

Sincerely,

Christine A. Lipa, Chief

Branch 4

Division of Reactor Projects

Docket No. 50-341

License No. NPF-43

Enclosure: Inspection Report 05000341/2006003

w/Attachment: Supplemental Information

cc w/encl: K. Hlavaty, Plant Manager

R. Gaston, Manager, Nuclear Licensing

D. Pettinari, Legal Department

Michigan Department of Environmental Quality

Waste and Hazardous Materials Division

M. Yudasz, Jr., Director, Monroe County

Emergency Management Division

Supervisor - Electric Operators

State Liaison Officer, State of Michigan

Wayne County Emergency Management Division

DOCUMENT NAME:E:\Filenet\ML062160540.wpd

G Publicly Available G Non-Publicly Available G Sensitive G Non-Sensitive

To receive a copy of this document, indicate in the concurrence box "C" = Copy without attach/encl "E" = Copy with attach/encl "N" = No copy

OFFICE RIII RIII

NAME RLerch:dtp CLipa

DATE 08/02/06 08/02/06

OFFICIAL RECORD COPY

Donald K. Cobb -3-

ADAMS Distribution:

LXR1

DHJ

RidsNrrDirsIrib

GEG

KGO

RMM3

CAA1

LSL (electronic IRs only)

C. Pederson, DRS (hard copy - IRs only)

DRPIII

DRSIII

PLB1

TXN

ROPreports@nrc.gov (inspection reports, final SDP letters, any letter with an IR number)

U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No: 50-341

License No: NPF-43

Report No: 05000341/2006003

Licensee: Detroit Edison Company

Facility: Fermi Power Plant, Unit 2

Location: Newport, Michigan

Dates: April 1 through June 30, 2006

Inspectors: R. Michael Morris, Senior Resident Inspector

T. Steadham, Resident Inspector

A. Wilson, NRC Headquarters

M. Franke, Senior Resident Inspector, Perry

M. Jordan, NRC Consultant

R. Langstaff, Senior Reactor Inspector

M. Mitchell, Radiation Specialist

Approved by: C. Lipa, Chief

Branch 4

Division of Reactor Projects

Enclosure

SUMMARY OF FINDINGS

IR 05000341/2006003; 04/01/2006-06/30/2006; Fermi Power Plant, Unit 2; Fire Protection,

Maintenance Risk Assessment, Operability Evaluations, Refueling and Outage Activities.

This report covers a 3-month period of inspection by resident inspectors and announced

baseline inspections by a regional radiation specialist inspector. Five Green findings, all of

which were associated with non-cited violations (NCVs) were identified. The significance of

most findings is indicated by their color (Green, White, Yellow, Red) using Inspection

Manual Chapter 0609, Significance Determination Process (SDP). Findings for which the

SDP does not apply may be Green after NRC management review. 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-Revealed Findings

Cornerstone: Initiating Events

Green. The inspectors identified an NCV of 10 CFR 50.65(a)(4) for the failure to

perform an adequate risk assessment for the Division I battery load test. The licensee

failed to consider the effect the test would have on the temperature in the reactor

protection system motor generator set rooms. Consequently, the load bank used for the

test caused the room temperature to increase which necessitated the unanticipated

installation of a temporary fan to cool the room. The licensee entered this issue into

their corrective action program to evaluate any programmatic or procedural deficiencies

that may have contributed to this event.

This finding is more than minor because the licensees risk assessment failed to

consider maintenance activities that could increase the likelihood of an initiating event,

specifically a loss of shutdown cooling from a reactor protection system motor generator

set trip on high temperature. The finding is of very low safety significance because it did

not affect the ability of operators to recover from a loss of shutdown cooling if it had

occurred. The cause of the finding is related to the cross-cutting element of Human

Performance. (Section 1R13.2)

Cornerstone: Mitigating Systems

Green. The inspectors identified an NCV of license condition 2.C(9) due to the

presence of unauthorized transient combustible materials in the residual heat removal

complex. An office chair and a plastic trash bin half filled with paper were secured next

to the electrical panel and associated cable raceway for emergency diesel generator 12

ventilation in the emergency diesel generator 12 switchgear room. The licensee entered

this issue into their corrective action program and removed the unauthorized transient

combustible materials from the residual heat removal complex.

This finding is more than minor because it affected the Mitigating Systems Cornerstone

attribute for protection against external factors. Specifically, a fire involving the

unauthorized transient combustibles could have affected a nearby electrical panel and

associated cable raceway containing mitigating system equipment important to safety.

2 Enclosure

The finding is of very low safety significance because the unauthorized transient

combustible materials would not have ignited from existing sources of heat or electrical

energy. The cause of the finding is related to the cross-cutting element of Problem

Identification and Resolution. (Section 1R05.2)

Green. The inspectors identified an NCV of Technical Specification 3.1.5.a.2,

Amendment 38, for the standby liquid control (SLC) system being inoperable for longer

than the allowed time without the plant being placed in hot shutdown. The licensee

failed to properly evaluate the operability of SLC during sparging activities when the

issue was raised in 1999. As a result, the licensee initiated a 21-hour sparge on the

SLC tank on August 24, 1999, and failed to take actions in accordance with the

Technical Specifications. After the deficient evaluation was identified on June 1, 2006,

the licensee revised the applicable procedures to declare the SLC system inoperable

during sparging the SLC tank. The licensee entered this issue into their corrective

action program.

This finding is more than minor because it represented a programmatic deficiency in the

licensees chemical control program which affected the ability of the fire brigade to

respond to and mitigate the effects of a fire. Upon management review, the finding is of

very low safety significance because the quantities of the relevant chemicals were low

and the storage location was sufficiently remote from mitigating equipment.

(Section 1R05.3)

Cornerstone: Emergency Preparedness

Green. The inspectors identified an NCV of license condition 2.C(9), for the failure to

appropriately store chemicals in accordance with the fire hazards analysis. The licensee

failed to evaluate the fire fighting response guidelines in NFPA-49 for various chemicals

brought into the protective area and, therefore, failed to appropriately store them as

required by the licensees fire hazards analysis. As a result, five normally stored

chemicals in the building have recommended fire fighting strategies that are inconsistent

with the licensees approved fire protection pre-plan. The licensee entered this issue

into their correction action program.

This finding is more than minor because it affected the equipment performance attribute

of the reactor safety cornerstone objective of ensuring the availability, reliability, and

capability of mitigating equipment to respond to initiating events to prevent undesirable

consequences. The finding is of very low safety significance because the total time of

sparging activities was short. (Section 1R15.2)

Cornerstone: Occupational Radiation Safety

Green. A self-revealed NCV was identified for the licensees failure to comply with

Technical Specification 5.4.1.a, written procedures shall be established, implemented,

and maintained covering applicable procedures recommended in Regulatory

Guide 1.33. The licensee did not adequately control the modification of the ventilation

equipment used to vent airborne radioactive particulate to the refuel floor during reactor

vessel floodup. Consequently, while raising reactor vessel water level, the improper

venting led to personnel contaminations, uptakes of radioactive material, and the

3 Enclosure

evacuation of the Reactor Building. The licensee entered this issue into their corrective

action program and conducted an investigation into the event. The corrective actions

recommended the development and implementation of an acceptable methodology for

raising reactor water level.

This finding is more than minor because it affected the Occupational Radiation Safety

Cornerstone of Radiation Safety due to individual worker unplanned, unintended dose.

The finding was evaluated using the SDP and was determined to be a finding of very

low safety significance because there was not a substantial potential for overexposure

and the licensees ability to assess dose was not compromised. (Section 1R20)

B. Licensee-Identified Violations

A violation of very low safety significance, which was identified by the licensee, has been

reviewed by the inspectors. Corrective actions taken or planned by the licensee have

been entered into the licensees corrective action program. This violation and corrective

actions are listed in Section 4OA7 of this report.

4 Enclosure

REPORT DETAILS

Summary of Plant Status

Unit 2 began this inspection period shutdown for refueling outage 11 (RF11). Reactor

startup began on May 3 but was halted at 95 percent power on May 9 due to indications

of a fuel leak. Suppression testing commenced that same day, reducing reactor power

to approximately 63 percent until May 12 when the operators began increasing reactor

power after suppressing the fuel leak. On May 14, the reactor was at full power where it

remained at or near until May 19 when the operators began a reactor shutdown to

replace the leaking fuel assembly. After completing the work, operators began a reactor

startup on May 28. The reactor reached full power on May 30 where it remained at or

near until June 15 when an automatic reactor scram occurred as a result of a failure of

main unit transformer 2B. The failed transformer was disconnected and reactor startup

began on June 16. On June 21, the reactor reached 63 percent power (maximum

planned with transformer 2B out of service) and remained there for the remainder of the

inspection period.

1. REACTOR SAFETY

Cornerstone: Mitigating Systems, Barrier Integrity, Initiating Events, Emergency

Preparedness

1R01 Adverse Weather (71111.01A)

a. Inspection Scope

The inspectors reviewed licensee procedures for mitigating the effects of hot weather.

The inspectors reviewed severe weather procedures, emergency plan implementing

procedures related to severe weather, and annunciator response procedures, and

performed walkdowns. This included the reactor building and turbine building ventilation

preparations. Additionally, the inspectors reviewed condition assessment resolution

documents (CARD) and verified problems associated with adverse weather were

entered into the corrective action program with the appropriate significance

characterization.

These activities represented one hot weather systems preparation inspection sample.

b. Findings

No findings of significance were identified.

5 Enclosure

1R04 Equipment Alignments (71111.04)

.1 Partial System Walkdowns (71111.04Q)

a. Inspection Scope

The inspectors performed partial system walkdowns of the following risk-significant

systems:

C Safety Relief Valves performed the week of April 3, 2006;

C SLC A performed the week of April 24, 2006;

C Division II Residual Heal Removal (RHR) and Residual Heat Removal Service

Water (RHRSW) Lineup performed the week of May 14, 2006;

C Standby Electrical Power (emergency diesel generator [EDG]) lineup performed

the week of May 14, 2006; and

C Division I RHR and RHRSW Shutdown Cooling performed the week of

May 21, 2006.

The inspectors selected these systems based on their risk significance relative to the

reactor safety cornerstones. The inspectors reviewed operating procedures, system

diagrams, Technical Specification (TS) requirements, Administrative TS, and the impact

of ongoing work activities on redundant trains of equipment in order to identify

conditions that could have rendered the systems incapable of performing their intended

functions. The inspectors also walked down accessible portions of the systems to verify

system components were aligned correctly.

In addition, the inspectors verified equipment alignment problems were entered into the

corrective action program with the appropriate significance characterization.

These activities represented five quarterly partial system walkdown inspection samples.

b. Findings

No findings of significance were identified.

.2 Complete System Walkdown (71111.04S)

a. Inspection Scope

The inspectors performed a complete system walkdown of the following risk-significant

system:

General Service Water (GSW) performed the week of April 24, 2006

The inspectors reviewed operating procedures, system diagrams, TS requirements, and

applicable sections of the Updated Final Safety Analysis Report (UFSAR) to ensure the

correct system lineup. The inspectors verified acceptable material condition of system

components, availability of electrical power to system components, and that ancillary

equipment or debris did not interfere with system performance.

6 Enclosure

These activities represented one semi-annual complete system walkdown inspection

sample.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05)

.1 Routine Resident Inspector Tours (71111.05Q)

a. Inspection Scope

The inspectors conducted fire protection tours of the following risk-significant plant

areas:

Condensate Pump Room;

Standby Gas Treatment, Pipe Room;

Top of Torus;

RHR Complex, Division I RHR;

RHR Complex, Division I EDG, Switchgear Rooms, Ventilation Rooms;

RHR Complex, Division I RHRSW Pump Room;

Hemyc Wrap for the Fire Barrier;

Division I, RHR Pump Room;

Division II, RHR Pump Room;

Main Unit Transformer 2B; and

Division II Electrical Switchgear Room.

The inspectors verified fire zone conditions were consistent with assumptions in the

licensee's fire hazards analysis. The inspectors walked down fire detection and

suppression equipment, assessed the material condition of fire fighting equipment, and

evaluated the control of transient combustible materials. In addition, the inspectors

verified fire protection related problems were entered into the corrective action program

with the appropriate significance characterization.

These activities represented eleven routine quarterly fire protection inspection samples.

b. Findings

No findings of significance were identified.

.2 RHR Complex, Division II EDG, Switchgear and Ventilation Rooms

a. Inspection Scope

The inspectors also conducted fire protection tours of the RHR complex, Division II

EDG, switchgear and ventilation rooms which are risk-significant plant areas.

7 Enclosure

The inspectors verified fire zone conditions were consistent with assumptions in the

licensee's fire hazards analysis. The inspectors walked down fire detection and

suppression equipment, assessed the material condition of fire fighting equipment, and

evaluated the control of transient combustible materials. In addition, the inspectors

verified fire protection related problems were entered into the corrective action program

with the appropriate significance characterization.

These activities represented one routine quarterly fire protection inspection sample.

b. Findings

Introduction: The inspectors identified an NCV of license condition 2.C(9) having very

low safety significance (Green) for the presence of unauthorized transient combustible

materials in the RHR complex.

Description: On May 15, 2006, the inspectors identified unauthorized transient

combustible materials in the RHR complex EDG 12 switchgear room. Specifically, the

inspectors identified an office chair and a plastic trash bin approximately half full of

paper secured approximately one foot from panel H21-P351, a safety-related electrical

panel for EDG 12 room ventilation, and associated cable raceway.

Section 9A.1.3.2.e of the UFSAR stated that the fire protection program had a

component to minimize the amount of combustibles to which safety-related areas may

be exposed. Procedure MOP11 implemented the fire protection program by prescribing

methods for controlling transient combustible material and the location of plant support

equipment. Step 3.5.1 of procedure MOP11 required a Plant Support Equipment

Approval form be obtained before placing any support equipment in the RHR complex.

The procedural requirement existed to ensure the introduction of transient combustible

materials was reviewed by fire protection personnel. However, no Plant Support

Equipment Approval form was submitted for the chair and trash bin identified in the

EDG 12 switchgear room within the RHR complex.

After the inspectors informed the fire protection supervisor of the issue, the fire

protection supervisor initiated CARD 06-23388 to initiate corrective actions. Licensee

personnel performed a walkdown of the RHR complex and identified three additional

trash bins and two chairs in other switchgear rooms within the RHR complex. The trash

bins and chairs were removed from the switchgear rooms.

Analysis: The inspectors determined the licensees failure to properly control transient

combustibles was a performance deficiency because the licensee is expected to comply

with their fire hazards analysis and because it was within the licensees ability to foresee

and prevent. The finding was greater than minor in accordance with Inspection Manual

Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue

Disposition Screening, issued September 30, 2005, because the finding affected the

Mitigating Systems Cornerstone attribute for protection against external factors, i.e., fire.

Specifically, a fire involving the unauthorized transient combustibles could have affected

a nearby electrical panel and associated cable raceway containing mitigating system

equipment important to safety. The inspectors identified that a credible fire scenario

existed in that equipment important to safety was located within the zone of influence of

8 Enclosure

the unauthorized transient combustible materials as described by Table 2.3.2,

Calculated Values (in feet) for Use in the Ball and Column Zone of Influence Chart for

Fires in an Open Location Away from Walls of IMC 0609, Appendix F, Fire Protection

Significance Determination Process, issued February 28, 2005.

The inspectors completed a significance determination of this issue using IMC 0609,

Appendix F. The inspectors reviewed IMC 0609, Appendix F, Attachment 2,

Degradation Rating Guidance Specific to Various Fire Protection Program Elements,

and determined the unauthorized transient combustible materials represented a low

degradation rating because the materials would not have ignited from existing sources

of heat or electrical energy. As such, the finding screened to Green under Question 1 of

IMC 0609, Appendix F, Task 1.3.1, Qualitative Screening for All Finding Categories,

and was considered a finding of very low safety significance. The primary cause of this

finding was related to the cross-cutting aspect of problem identification and resolution

because the licensees response to several recent instances of unauthorized transient

combustibles was not effective in preventing this instance of unauthorized transient

combustibles.

Enforcement: License condition 2.C.(9) required the licensee to implement and

maintain in effect all provisions of the approved fire protection program as described in

the UFSAR. Section 9A.1.3.2.e of the UFSAR stated the fire protection program had a

component to minimize the amount of combustibles to which safety-related areas may

be exposed. Procedure MOP11 implemented the fire protection program by prescribing

methods for controlling transient combustible material and the location of plant support

equipment. Step 3.5.1 of procedure MOP11 required a Plant Support Equipment

Approval form be obtained before placing any support equipment in the RHR complex, a

safety-related area. Contrary to the above, on May 15, 2006, the inspectors identified

support equipment, i.e., a chair and a trash bin, had been placed in the EDG 12

switchgear room within the RHR complex without a Plant Support Equipment Approval

form having been obtained. Once identified, the licensee initiated CARD 06-23388,

performed a walkdown of the RHR complex, and removed unauthorized chairs and trash

bins from the switchgear rooms in the RHR complex. Because this violation is of very

low safety significance and because it was entered into the licensees corrective action

program as CARD 06-23388, this violation is being treated as an NCV, consistent with

Section VI.A.1 of the NRC Enforcement Policy: NCV 05000341/2006003-01:

Unauthorized Transient Combustibles in Safety-Related Areas.

.3 Fire Protection - Drill Observation (71111.05A)

a. Inspection Scope

The inspectors assessed fire brigade performance and the drill evaluators' critique

during an unannounced fire brigade drill on June 21, 2006. The drill simulated a fire in

the chemical storage room in the radioactive waste building. The inspectors focused on

the command and control of fire brigade activities, fire fighting and communication

practices, material condition and use of fire fighting equipment, and implementation and

adequacy of pre-planned fire fighting strategies.

9 Enclosure

These activities represented one annual fire protection - drill observation inspection

sample.

b. Findings

Introduction: The inspectors identified a Green NCV of license condition 2.C(9) for the

failure to appropriately store chemicals in accordance with the fire hazards analysis.

Description: The inspectors watched Fire Brigade Drill Scenario Number 6 which

involved a simulated fire in the chemical storage room on the first floor of the radioactive

waste building. Firefighters entered the room wearing full protective clothing and

positive-pressure, self-contained breathing apparatus. In accordance with fire protection

Pre-Plan FP-RDWST, Rev. 4, Radioactive Waste Building Zones 22, 23, 24, and 25, the

brigade simulated extinguishing the fire by using a water hose in a fog pattern.

The inspectors later questioned the adequacy of the fire protection pre-plan because it

did not appear to take into account differences in fire fighting strategies with the various

types of chemicals in the room. After reviewing the list of chemicals in the room against

the fire fighting strategies recommended by NFPA-49, Hazardous Chemicals Data 1994

Edition, the inspectors identified four chemicals normally stored in the room where

NFPA-49 recommends using special protective clothing when fighting a fire involving

those chemicals: monoethylamine solution, sodium hydroxide, potassium hydroxide,

and sulfuric acid. Additionally, NFPA-49 recommends against the use of water when

fighting fires involving sulfuric acid.

The inspectors reviewed the storage locations of these chemicals and determined they

were not segregated in such a manner to ensure a fire in that room did not involve any

of those chemicals. Further, the fire protection pre-plan contained no guidance on any

special precautions to be followed when fighting a fire involving any of those four

chemicals. The inspectors determined it was unreasonable to assume the responding

fire brigade would be able to easily determine what chemicals were on fire given the lack

of labeling and amount of smoke that likely would be present in the room during an

actual fire.

The licensees fire hazards analysis, as documented in UFSAR, Section 9A.5.G.3,

required chemicals be stored in accordance with the guidelines of NFPA-49. Although

the literal storage requirements for these chemicals were generally adhered to, e.g.,

stored in a cool, dry, ventilated room in metal cabinets, etc., the inspectors determined

the fire fighting strategies for the four chemicals of interest above were inseparable from

the storage guidelines because the licensee is expected to take all relevant information

into account when determining the appropriate chemical storage requirements. For

example, although NFPA-49 contained no guidance to store sulfuric acid separately

from nitric acid, the fact that water is suitable for fires involving nitric acid but not for fires

involving sulfuric acid logically concludes either, a) water should not be used if the

chemicals are in the same cabinet, b) the sulfuric acid should be stored in a separate

container, or c) the quantity of sulfuric acid is controlled sufficiently low so as to not

require segregation; none of which occurred. Likewise, because chemical suits are

recommended for those four chemicals but standard fire fighter turnout gear is suitable

for all other normally stored chemicals in the room, it is reasonable to expect the

10 Enclosure

licensee will take the emergency response personal protective equipment guidelines into

account when storing chemicals which the licensee also failed to do.

The inspectors questioned the licensee on how chemicals were controlled such that they

did not adversely affect the fire protection strategy and were informed that chemicals

are evaluated based on the effect they would have on the plant but not on the effect

they would have on fire fighting techniques. For example, there were no controls in

place to either ensure the fire brigade did not use water on fires involving sulfuric acid or

to control the amount of sulfuric acid below some threshold to preclude any alteration in

the fire fighting strategy.

Analysis: The inspectors determined the licensees failure to properly store chemicals in

accordance with guidelines contained in NFPA 49 was a performance deficiency

because the licensee is expected to comply with their fire hazards analysis and because

it was within the licensees ability to foresee and prevent. The finding is more than

minor because it represented a programmatic deficiency in the licensees chemical

control program which affected the ability of the fire brigade to respond to and mitigate

the effects of a fire. This finding affected the emergency planning cornerstone because

it affected the ability of the fire brigade to respond to a fire which could potentially affect

the licensees emergency plan.

The finding is not suitable for SDP evaluation, but has been reviewed by NRC

management and is determined to be a finding of very low safety significance (Green)

because the quantities of the relevant chemicals were low and the storage location was

sufficiently remote from mitigating equipment.

Enforcement: Fermi 2 Facility Operating License NPF-43, condition 2.C(9), required,

in part, that the licensee implement and maintain in effect all provisions of the

approved fire protection program as described in Section 9A of the UFSAR as

amended and approved in the Fermi 2 safety evaluation report through supplement 6.

UFSAR 9A.5.G.3 required hazardous chemicals be stored in accordance with the

guidelines of NFPA 49-1994, Hazardous Chemicals Data 1994 Edition. Contrary to

the above, on June 21, 2006, the licensee failed to utilize the guidelines contained in

NFPA 49-1994 when storing chemicals in the radioactive waste building. Because

this violation is of very low safety significance and because it was entered into the

licensees corrective action program as CARD 06-24243, this violation is being

treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy:

NCV 05000341/2006003-02: Improper Storage of Chemicals Affecting Fire Fighting

Response.

1R06 Flood Protection (71111.06)

a. Inspection Scope

The inspectors performed an inspection related to the licensee's precautions to mitigate

the risk from internal flooding events. The inspectors performed a walkdown of the

following plant areas to assess the adequacy of watertight doors and verify drains and

sumps were clear of debris and were operable:

11 Enclosure

High Pressure Coolant Injection Pump Room;

Reactor Core Isolation Cooling Pump Room;

Division I Core Spray Pump Room;

Division II Core Spray Pump Room; and

Auxiliary Building T Room.

The inspectors also reviewed the work activities associated with internal flooding to

verify identified problems were being entered into the corrective action program with the

appropriate characterization and significance.

These activities represented one internal flood protection inspection sample.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance (71111.07A)

a. Inspection Scope

The inspectors reviewed completed test reports and observed the performance of

inspections for the RHR service water heat exchanger.

The inspectors selected this heat exchanger because its associated systems were risk

significant in the licensee's risk assessment and were required to support the operability

of other risk-significant, safety-related equipment. During these inspections, the

inspectors observed the as-found condition of the heat exchanger and verified no

deficiencies existed that would mask degraded performance. The inspectors discussed

the as-found condition as well as the historical performance of the heat exchanger with

engineering department personnel and reviewed applicable documents and procedures.

In addition, the inspectors verified heat sink problems were entered into the corrective

action program with the appropriate significance characterization, and completed

corrective actions were adequate and appropriately implemented.

These activities represented one heat sink performance inspection sample.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification (71111.11Q)

a. Inspection Scope

On June 13, 2006, the inspectors observed an operations support crew during the

annual requalification examination in mitigating the consequences of events in

SS-OP-802-330, Anticipated Transient Without Scram with Small Steam Leak, Rev. 0,

dated January 26, 2006, on the simulator. The inspectors evaluated the following areas:

12 Enclosure

C licensed operator performance;

C crews clarity and formality of communications;

ability to take timely actions in the conservative direction;

prioritization, interpretation, and verification of annunciator alarms;

correct use and implementation of abnormal and emergency procedures;

C control board manipulations;

C oversight and direction from supervisors; and

C ability to identify and implement appropriate TS actions and Emergency Plan

actions and notifications.

The crews performance in these areas was compared to pre-established operator

action expectations and successful critical task completion requirements.

These activities represented one quarterly licensed operator requalification inspection

sample.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12Q)

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following three

risk-significant systems:

C Station Blackout Diesel Generators CTG 11-1, 2, 3, 4, and 120 kV switchyard;

C RHR System A and B; and

C Molded Case Circuit Breakers.

The inspectors assessed performance issues with respect to the reliability, availability,

and condition monitoring of the system. Specifically, the inspectors independently

verified the licensee's actions to address system performance or condition problems in

terms of the following:

C implementing appropriate work practices;

C identifying and addressing common cause failures;

C scoping of systems in accordance with 10 CFR 50.65(b);

C characterizing system reliability issues;

C tracking system unavailability;

C trending key parameters (condition monitoring);

C ensuring 10 CFR 50.65(a)(1) or (a)(2) classification and/or re-classification; and

C verifying appropriate performance criteria for systems classified as (a)(2) and/or

appropriate and adequate goals and corrective actions for systems classified as

(a)(1).

In addition, the inspectors verified maintenance effectiveness issues were entered into

the corrective action program with the appropriate significance characterization.

13 Enclosure

These activities represented three quarterly maintenance effectiveness inspection

samples.

b. Findings

No findings of significance were identified.

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13Q)

.1 Routine Maintenance Risk Assessments

a. Inspection Scope

The inspectors reviewed the licensee's evaluation and management of plant risk for the

maintenance and operational activities affecting risk-significant and safety-related

equipment listed below:

maintenance risk for week of April 2, 2006;

maintenance risk for week of April 17, 2006;

maintenance risk for week of April 30, 2006; and

maintenance risk for week of June 26, 2006.

These activities were selected based on their potential risk significance relative to the

reactor safety cornerstones. As applicable for each activity, the inspectors reviewed the

scope of maintenance work, discussed the results of the assessment with the licensee's

probabilistic risk analyst and/or shift technical advisor, and verified plant conditions were

consistent with the risk assessment. The inspectors also reviewed TS requirements and

walked down portions of redundant safety systems, when applicable, to verify risk

analysis assumptions were valid and applicable requirements were met.

These activities represented four quarterly maintenance risk assessment and

emergency work control inspection samples.

b. Findings

No findings of significance were identified.

.2 Inadequate Maintenance Risk Assessment

a. Inspection Scope

The inspectors reviewed the activities surrounding the Division I battery load test

performed during RF11 to determine if the licensee appropriately considered the risk

impacts of performing the test. The inspectors interviewed licensee staff, reviewed

documents, and performed walkdowns. The inspectors considered ancillary equipment

affected by the test to determine what affect, if any, the test would have on it.

These activities represented one quarterly maintenance risk assessment and

emergency work control inspection sample.

14 Enclosure

b. Findings

Introduction: The inspectors identified a Green NCV of 10 CFR 50.65(a)(4) for the

failure to perform an adequate risk assessment for the Division I battery load test.

Description: On April 3, 2006, the inspectors identified a temporary fan installed in the

blocked-open doorway to the dc motor control center (DC MCC) area. Upon entering

the room, the inspectors noticed that additional doors inside the DC MCC area leading

to each of the reactor protection system motor generator (RPSMG) set rooms were

open. After questioning why the doors were blocked open and a fan installed, the

inspectors learned operators took those actions to provide additional cooling to the

RPS MG sets because the Division I 130/260 VDC battery load surveillance test,

procedure 42.309.05, was in progress.

Because the air-cooled load bank used for the test was temporarily installed in the

DC MCC area, temperatures in the room started to increase after the test commenced.

However, two area room coolers were out of service due to a scheduled outage of the

GSW system which left the DC MCC, vital battery, battery charger, and RPS MG set

rooms with no cooling. In order to help prevent a loss of shutdown cooling, which would

have occurred had the RPS MG sets tripped, the operators had previously blocked

opened the doors to the RPS MG set rooms which further increased the temperature in

the DC MCC area. Operators then blocked open the double doors to the DC MCC area

and installed a large utility fan in the doorway to provide additional cooling to the area.

The test was successfully completed, temperatures dropped, and the fan was removed.

The outage risk associated with this test did not consider the effects it would have on

the key safety function of maintaining decay heat available due to the additional heat

from the load bank with no room cooling. Moreover, the outage risk associated with the

GSW outage assumed that the RPS MG sets would not be running. Consequently,

prudent risk management actions were not developed prior to performing the battery

load test. However, because operators installed a fan in the open doorway early

enough into the test, shutdown cooling remained in operation.

Analysis: The inspectors determined the failure to perform an adequate risk analysis of

maintenance activities prior to performing maintenance was a performance deficiency

because the licensee is expected to comply with the requirements of the maintenance

rule. This finding is more than minor because the licensees risk assessment failed to

consider maintenance activities that could increase the likelihood of an initiating event,

specifically a loss of shutdown cooling. In addition, this finding affected the initiating

event cornerstone because it is associated with an increase in the likelihood of an

initiating event. The inspectors utilized the maintenance risk and shutdown risk SDP to

assess the risk of this finding. The finding is of very low safety significance because the

finding did not affect the ability of operators to recover from a loss of shutdown cooling

had it occurred. The inspectors determined the cause of this finding impacted the

Human Performance cross-cutting area because the cause of the inadequate risk

assessment was due to a personnel error.

Enforcement: 10 CFR 50.65(a)(4) requires, in part, that before performing

maintenance activities, the licensee shall assess and manage the increase in risk

15 Enclosure

that may result from the proposed maintenance activities. Contrary to the above,

beginning on April 1, 2006, and continuing through April 6, 2006, the licensee

performed surveillance procedure 42.309.05 without adequately assessing and

managing the increase in risk prior to performing the activity. Because this violation is

of very low safety significance and because it was entered into the licensees corrective

action program as CARD 06-21892 and 06-24495, this violation is being treated as an

NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy:

NCV 05000341/2006003-03: Inadequate Maintenance Risk Assessment.

1R14 Personnel Performance During Non-Routine Plant Evolutions and Events (71111.14)

a. Inspection Scope

The inspectors reviewed the licensees actions in response to the following non-routine

events to ensure the licensee took appropriate actions in accordance with licensee

procedures:

unplanned reactor building contamination, CARD 06-21534;

control rod position indication malfunction, CARD 06-23491 & 06-23489;

main turbine bypass valves opened at power, CARD 06-24113; and

mode 5 reactor scram during installation of shorting links, CARD 06-23588.

The inspectors reviewed operator logs, procedures, corrective action documents, other

documents, and interviewed personnel. The inspectors also evaluated the licensees

operational decision making involved with these non-routine events.

These activities represented four inspection samples.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15)

.1 Routine Review of Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following documents to ensure the identified condition did

not render the involved equipment inoperable or result in an unrecognized increase in

plant risk, and the licensee appropriately applied TS limitations and appropriately

returned the affected equipment to an operable status:

CARD 06-23114, Motor Operator Valve Motor Replacement for RHR Shutdown

Cooling Inboard Suction Bypass Valve; and

CARD 06-23898, Division II Emergency Equipment Cooling Water Makeup

Pump In-Service Test Flow Unattainable.

These activities represented two operability evaluation inspection samples.

16 Enclosure

b. Findings

No findings of significance were identified.

.2 Standby Liquid Control Operability During Air Sparging Operations

a. Inspection Scope

The inspectors reviewed the licensees practice of placing an air sparge on the SLC tank

to determine if system operability was justified. The inspectors reviewed previous

engineering evaluations to determine the technical adequacy of the conclusions. The

inspectors reviewed operator logs, TS, design basis documents, UFSAR, and other

documents. The inspectors interviewed operators, engineers, and other licensee staff.

These activities represented one operability evaluation inspection sample.

b. Findings

Introduction: The inspectors identified a Green NCV of TS for the SLC system being

inoperable for longer than the action time to be in hot shutdown with both SLC

subsystems inoperable.

Description: In April 1999, the licensee reviewed an operating experience report issued

by another licensee discussing the inoperability of the SLC system during air sparging

activities. Air sparging the SLC tank was done to facilitate mixing of the sodium

pentaborate in the tank and was performed prior to the monthly chemistry analysis and

after any chemical addition to the tank. The air sparge header was located near the

bottom of the tank in proximity to the SLC pump suction line. The concern was that if

the pumps were operating while the air sparge was operating, air could be drawn into

the pumps and compromise their ability to perform their design function. The licensee

initiated CARD 99-13240 to evaluate the applicability of the issue to Fermi.

The licensee determined that although pump performance would be impacted if the

pumps were required while an SLC tank air sparge was in progress, the pumps

remained operable. Based on input from the pump vendor, the licensee concluded that

entrained air in the pumped water would cause increased pump vibration and a

negligible reduction in delivered flow rate. Because the vendor stated the increased

vibration would only affect long-term pump reliability, the licensee concluded that

long-term pump degradation was not a concern because SLC had a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months mission

time.

In an effort to lessen the probability of requiring the SLC pumps during an actual event

while the air sparge was operating, the licensee identified the need to revise the system

operating procedures to minimize duration of sparge operations from hours to minutes.

The action to revise procedure 23.149, Standby Liquid Control System, was originally

due on June 30, 1999.

The inspectors reviewed the licensees evaluation and noted the CARD did not contain

either any documentation from the vendor or any attempt to quantify the reduction in

17 Enclosure

flow rate to ensure the minimum required flow was maintained. The inspectors asked

the licensee for the vendors recommendations in writing but were later told the pump

vendor declined to support their previous conclusion in writing. The inspectors were

concerned that if the pump vendor was unwilling to state in writing that the pumps would

operate for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months during sparging, then pump operability was not justified.

The inspectors brought their concern to the licensee who entered the issue into their

corrective action program as CARD 06-23785 on June 1, 2006. After further review, the

licensee concluded that SLC operability during sparging could not be supported and

revised procedure 23.149 accordingly. Upon review of control room logs, the inspectors

determined that although the licensee significantly reduced the total sparging time since

1999, the tank was air sparged for 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months on August 24, 1999, which exceeded the

time to be in hot shutdown of 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months with both SLC subsystems inoperable while in

Mode 1 or 2. In addition, the inspectors concluded the maximum unavailability in any

1-year period after identification of the issue in 1999 was approximately 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months .

Analysis: The inspectors determined the licensees failure to appropriately evaluate

SLC operability during sparging operations was a performance deficiency because the

licensee is expected to adequately evaluate issues that affect the operability of TS

equipment and because it was within the licensees ability to foresee and prevent. The

finding is more than minor because it affected the equipment performance attribute of

the reactor safety cornerstone objective of ensuring the availability, reliability, and

capability of mitigating equipment to respond to initiating events to prevent undesirable

consequences.

The inspectors assessed the finding using the SDP. Because the inspectors

considered this finding to represent an actual loss of a safety function of SLC, the

inspectors performed a phase 2 SDP analysis. A phase 3 analysis was subsequently

performed by the senior reactor analyst (SRA). The SRA performed the risk evaluation

using the Fermi Standardized Plant Analysis Risk (SPAR) Model, Level 1, Revision 3P,

Change 3.21, created October 2005. The SRA ran the SPAR model assuming common

cause failure of both SLC pumps, with an exposure time of 90 hours0.00104 days 0.025 hours 1.488095e-4 weeks 3.4245e-5 months . Using the above

information the SRA obtained a change in core damage frequency (CDF) of 3.1E-8

(Green) for internal events. The dominant sequences involved a failure of the reactor to

scram after a transient, loss of condenser heat sink, and loss of main feedwater, and

failure of the SLC system.

Anticipated transient without scram events are not assumed to be caused by external

events and, therefore, the risk contribution from external events is insignificant.

Similarly, because the internal events CDF is less than 1E-7, large early release

frequency (LERF) is not significant per IMC 0609, Appendix H. The SRA concluded

the total CDF considering internal events, external events, and LERF is estimated at

3.1E-8 (Green).

Enforcement: Technical Specification 3.1.5.a.2, Amendment 38, required that while in

Modes 1 and 2, with the SLC system otherwise inoperable, the licensee must restore

the system to operable status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least hot shutdown within the

next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and was in effect on August 24 and 25, 1999. Contrary to the above,

beginning on August 24, 1999, and continuing until August 25, 1999, while in Modes 1

18 Enclosure

and 2, the SLC system was inoperable for 21 hours2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months while the SLC tank was being air

sparged; therefore, on August 25, 1999, with the SLC system inoperable for greater

than 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , the plant was not in at least hot shutdown. Because this violation is of

very low safety significance and because it was entered into the licensees corrective

action program as CARD 06-23785, this finding is being treated as an NCV, consistent

with Section VI.A.1 of the NRC Enforcement Policy: NCV 05000341/2006003-04:

Improper Evaluation of Standby Liquid Control Operability During Tank Sparging.

.3 Inappropriate Use of Risk in Operability Evaluations

a. Inspection Scope

The inspectors reviewed CARD 06-23913 to ensure that the identified condition did not

render the involved equipment inoperable or result in an unrecognized increase in plant

risk and that the licensee appropriately applied TS limitations and appropriately returned

the affected equipment to an operable status.

These activities represented one operability evaluation inspection sample.

Introduction: The inspectors identified an Unresolved Item (URI) when the licensee

removed pipe insulation, credited for environmental qualification of nearby equipment,

while at power without an adequate engineering evaluation.

Description: On June 8, 2006, the licensee initiated CARD 06-23913 to request a work

request to replace contaminated insulation on the suction and discharge pipe for the B

RHR pump. As a result, Work Request (WR) 000Z062027 was released and work

began on June 13, 2006. While performing a plant tour on June 15, 2006, the

inspectors identified the insulation was missing from the suction pipe for the B RHR

pump and questioned the licensee if the insulation removal had an approved

engineering evaluation. Because the equipment was in a potentially harsh environment,

the uninsulated pipe would increase the temperature profile of the room during accident

conditions which could affect the environmental qualification of electrical equipment in

the room. The licensee stated the evaluation was documented in CARD 06-23913 and

concluded that removing the insulation while at power was acceptable.

The inspectors reviewed the subject evaluation and became concerned that one of the

assumptions for the evaluation was that an accident was not considered as credible

during the period of time the insulation was to be removed. Upon further review, the

inspectors learned the licensee used non-accident heat loads to determine the

environmental effects of removing the insulation. The licensees justification was that if

the total time the insulation was removed was less than 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months , then the probability

of either a high energy line break or loss of coolant accident was negligible and, hence,

did not need to be assumed to occur.

The definition of operability stated, however, the equipment must be capable of

performing its specified function(s). The inherent assumption was the occurrence,

conditions, or event would exist and the safety function could be performed. Therefore,

19 Enclosure

the inspectors concluded the use of probabilities of the occurrence of accidents while

the insulation was removed was an unacceptable assumption in the subsequent

operability evaluation.

While reviewing this evaluation, the inspectors discovered the licensee used the

same method of evaluating on-line insulation removal since at least September 20,

2001, and found five additional CARDs where the licensee approved removing

insulation from equipment in potentially harsh areas while at power, likewise with

unacceptable evaluations. Because the extent of condition of this issue is potentially

significant and could extend to work other than on-line insulation removal, this item is

unresolved pending the inspectors review of the licensees full extent of condition

review and subsequent risk evaluation and is identified as Unresolved Item

(URI) 05000341/2006003-05: Inappropriate Use of Risk in Operability Evaluations.

1R19 Post-Maintenance Testing (71111.19)

a. Inspection Scope

The inspectors reviewed post-maintenance testing (PMT) activities associated with the

following scheduled maintenance:

Division I Switchyard Modifications, EDP 31880;

Feedwater Check Valve PMT, WR T210040100;

Drywell Cooler Number 4 Replacement, WR 000Z052131;

Reactor Core Isolation Cooling PMT;

Reactor Recirculation Pump Discharge Valve B310SF031A, Stem Replacement,

WR 000Z050487;

Main Steam Isolation Valve (MSIV), Limit Switch Replacement; and

WR 000Z060156, Replace Control Rod Drive Pump Inboard Bearing Oil Level

Sight glass.

The inspectors reviewed the scope of the work performed and evaluated the adequacy

of the specified PMT. The inspectors verified the PMT was performed in accordance

with approved procedures, the procedures clearly stated acceptance criteria, and the

acceptance criteria were met. The inspectors interviewed operations, maintenance, and

engineering department personnel and reviewed the completed PMT documentation.

In addition, the inspectors verified PMT problems were entered into the corrective action

program with the appropriate significance characterization.

These activities represented seven PMT inspection samples.

b. Findings

No findings of significance were identified.

20 Enclosure

1R20 Refueling and Outage Activities (71111.20)

.1 Routine Refueling Outage Inspection Activities

a. Inspection Scope

The inspectors observed the licensees performance during RF11, which was in

progress at the beginning of this inspection and concluded on May 5, 2006.

This inspection consisted of a review of the licensees outage schedule, safe shutdown

plan and administrative procedures governing the outage, periodic observations of

equipment alignment, and plant and control room outage activities. Specifically, the

inspectors determined whether the licensee effectively managed elements of shutdown

risk pertaining to reactivity control, decay heat removal, inventory control, electrical

power control, and containment integrity.

The inspectors performed the following activities daily, during the outage:

attended control room operator and outage management turnover meetings to

verify the current shutdown risk status was well understood and communicated;

performed walkdowns of the main control room to observe the alignment of

systems important to shutdown risk;

observed the operability of reactor coolant system (RCS) instrumentation and

compared channels and trains against one another;

performed walkdowns of the turbine, auxiliary, and reactor buildings and the

drywell to observe ongoing work activities, to ensure work activities were

performed in accordance with plant procedures, and to verify procedural

requirements regarding fire protection, foreign material exclusion, and the

storage of equipment near safety-related structures, systems, and components

were maintained;

verified the licensee maintained secondary containment in accordance with TS

requirements; and

reviewed selected issues the licensee entered into its corrective action program

to verify identified problems were being entered into the program with the

appropriate characterization and significance.

Additionally, the inspectors performed the following specific activities:

monitored refueling activities to verify the licensee adhered to established

procedures and TS requirements for handling of irradiated fuel;

performed drywell closeout;

verified shutdown cooling tagouts;

verified completion of restart restraint items; and

observed control rod withdrawal to criticality and portions of the plant power

ascension.

In particular, the inspectors reviewed the licensees restart restraint process and verified

the closure of selected issues. Documents reviewed during these inspection activities

are listed at the end of this report.

21 Enclosure

Because inspection activities for this refueling outage constituted one inspection sample

in Inspection Report 05000341/2006002, and since only one sample is counted per

outage, the inspection activities for this inspection period do not constitute an additional

refueling and outage inspection sample.

b. Findings

Introduction: The inspectors identified a Green NCV of Technical Specification 5.4.1.a,

for the failure to adequately control the modification of the ventilation equipment used to

vent airborne radioactive particulate to the refuel floor during reactor vessel floodup.

Description: At 1450, on March 26, 2006, operators initiated core spray at approximately

3000 gpm to raise reactor water level, to permit removal of the reactor vessel head. The

reactor vessel head vent pipe had been disconnected, to permit installation of a

ventilation unit for the venting of gases to the refuel floor. The ventilation unit consisted

of a nominal 3600 scfm fan, a charcoal filter, and a HEPA particulate filter. Neither filter

had been recently tested. The ventilation unit was configured with two hoses placed to

take suction close to the reactor vessel head vent. The exhaust of the ventilation unit

ran to a point below an intake for the Standby Gas Treatment System (SGTS), to permit

capture of the exhaust by the SGTS.

At approximately 1500, the ventilation unit intake hoses were observed being pushed

away from the reactor vessel head vent, due to flow from the vent. Core spray was shut

down at 1502 and at 1504 visible moisture was seen being emitted from the vent. A

continuous air monitor on the refuel floor alarmed at 1510 and RP ordered the

evacuation of all but essential personnel from the refuel floor. By 1525 all personnel

were removed from the floor and shortly after this, the entire reactor building was

evacuated due to the spread of contamination. Decontamination of several workers was

required. Twenty-eight workers were whole body counted, with 26 showing uptakes of

varying levels of Co-60, Co-58, and Mn-54.

There were three conditions that contributed to the cause of the event. First of all,

coolant activity levels were higher than expected due to a crud burst during shutdown

and the temporary loss of RWCS allowed Co-60 to enter and remain in the coolant,

possibly plating out on reactor internals. The second condition related to the

temperature of the material vented from the reactor vessel head. Reactor vessel

outside shell temperatures indicated 215 degrees F, which equates to internal metal

temperatures above the atmospheric boiling point of water. This indicates that some of

the coolant flashed to steam as the level in the reactor vessel rose, which could increase

the carryover of coolant activity to the vented gas. The third condition was the

inadequate processing of vented material from the reactor vessel head. The venting of

the airborne radioactive particulate would not have resulted in building contamination

and personnel uptakes if the ventilation unit had effectively removed this material to the

SGTS.

The use of the ventilation system was not in accordance with its design configuration.

The ventilation system for previous outages had suction hoses connected to a hood that

was placed over the reactor vessel head vent, to improve the capture of vented material.

The use of the hood was discontinued due to its impact on water level instrumentation.

22 Enclosure

An engineering evaluation was not performed on the impact of the change in

configuration of the ventilation system. In addition, the exhaust arrangement from the

ventilation unit to the SGTS had not been evaluated for effectiveness. Licensee

Procedure MES12, Performing Temporary Modifications, requires the modification

process be followed and an evaluation be performed.

Furthermore, licensee procedural guidance did not exist for the overall process of taking

the operating reactor to a condition allowing fuel movement. The event could have been

prevented if appropriate acceptance criteria for allowable reactor vessel temperature

and coolant activity levels existed. Thus, the root cause of the event was determined to

be a procedural and programmatic weakness.

The licensee initiated CARD 06-21534, Continuous Air Monitor Alarm on Refueling

Floor, to track the investigation of the event in their CAP. The primary corrective action

recommendation is to develop and implement an acceptable methodology for raising

reactor vessel water level. In addition, the design and configuration of the current

ventilation exhaust capture system will be evaluated and modified, as appropriate to

assure that it is adequate for the expected reactor vessel fill rate and radioactive

material concentrations. The methodology may involve an alternate vent path, such as

using the attached piping to vent the reactor vessel to the drywell.

Analysis: The inspectors determined the licensees lack of control of the Temporary

Modification process constituted a design control issue. The licensees failure to

adequately control the process used to vent airborne radioactive particulate to the refuel

floor during reactor vessel floodup represents a performance deficiency as defined in

NRC Inspection Manual Chapter 0612, Appendix B, Issue Screening. The issue was

determined to be more than minor because if left uncorrected the issue could become a

more significant safety concern if coolant activity levels were higher or if the vessel was

flooded quicker.

The finding was assessed using NRC Inspection Manual Chapter 0609, Appendix C,

Occupational Radiation Safety Significance Determination Process due to individual

worker unplanned, unintended dose. The finding was determined to be of very low

safety significance because the inspectors answered, NO, to all four phase 1

screening questions.

Enforcement: Technical Specification 5.4.1.a requires that procedures recommended in

Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, be established,

implemented and maintained. Section 4.a of that document, in part requires

instructions for filling, venting, and draining the reactor pressure vessel. Contrary to the

above, the initial installation of the ventilation system and the changes made to the

ventilation system that was used as part of the reactor vessel floodup during outages

was not processed through the Temporary Modification Procedure. This finding is being

treated as an NCV, consistent with Section VI.A.1 of the NRC Enforcement Policy and is

identified as NCV 05000341/2006003-06: Inadequate Controls for Venting the Reactor

Pressure Vessel Head. This issue is in the licensees corrective action program as

CARD 06-22667.

23 Enclosure

.2 Forced Outage 06-01

a. Inspection Scope

The inspectors observed the licensees performance during Forced Outage 06-01 from

May 20, 2006, through May 29, 2006, which was scheduled to locate and replace a

failed fuel assembly. During power ascension following RF11, operators identified a

potential fuel leak because off gas radiation levels were slightly elevated from normal.

Operators began suppression testing later that day, which identified the failed assembly.

Operators initiated a manual unit shutdown to replace the failed fuel. While the unit was

shutdown, additional assemblies subjected to similar flux profile histories were also

replaced although fuel sipping operations identified only one fuel assembly with a fuel

cladding defect.

This inspection consisted of a review of the licensees outage schedule, safe shutdown

plan and administrative procedures governing the outage, periodic observations of

equipment alignment, and plant and control room outage activities. Specifically, the

inspectors determined whether the licensee effectively managed elements of shutdown

risk pertaining to reactivity control, decay heat removal, inventory control, electrical

power control, and containment integrity.

The inspectors performed the same daily activities, during the outage as described in

Section 1R20.1 for the refueling outage.

These activities represented one forced outage inspection sample.

b. Findings

No findings of significance were identified.

.3 Forced Outage 06-02

a. Inspection Scope

The inspectors observed the licensees performance during Forced Outage 06-02 from

June 15, 2006, through June 17, 2006. On June 15, a reactor scram occurred due to a

main turbine generator trip which occurred when main unit transformer 2B failed. The

inspectors responded to the control room and to the transformer area to assess the

licensees response to the event.

This inspection consisted of a review of the licensees outage schedule, safe shutdown

plan and administrative procedures governing the outage, and plant and control room

outage activities. Specifically, the inspectors determined whether the licensee

effectively managed elements of shutdown risk pertaining to reactivity control, decay

heat removal, inventory control, and electrical power control.

24 Enclosure

The inspectors performed the following activities during the outage:

attended control room operator and outage management turnover meetings to

verify the current shutdown risk status was well understood and communicated;

performed walkdowns of the main control room to observe the alignment of

systems important to shutdown risk;

observed the operability of RCS instrumentation and compared channels and

trains against one another; and

observed control rod withdrawal to criticality and portions of the plant power

ascension.

These activities represented one forced outage inspection sample.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22Q)

a. Inspection Scope

The inspectors reviewed the test results for the following activities to determine whether

risk-significant systems and equipment were capable of performing their intended safety

function and to verify testing was conducted in accordance with applicable procedural

and TS requirements:

MSIV Limit Switch (routine);

SLC Squib Valve Test (routine);

Integrity Test for Containment Penetrations X-7A, X-7B, X-7C, and X-7D (LLRT);

Scram Nitrogen Accumulator Integrity Test (routine);

Scram Discharge Volume Vent and Drain Valves Operability (routine);

Secondary Containment Integrity Test (LLRT);

MSIV Channel Functional Test (isolation valve);

LOOP/LOCA Test (routine);

Reactor Core Isolation Cooling System Automatic Actuation and Flow Test

(routine); and

High Pressure Coolant Injection Flow Rate Test at 165 psig Reactor Steam

Pressure (routine).

The inspectors reviewed the test methodology and test results to verify equipment

performance was consistent with safety analysis and design basis assumptions. In

addition, the inspectors verified surveillance testing problems were being entered into

the corrective action program with the appropriate significance characterization.

These activities represented seven routine, two local leak rate test (LLRT), and one

containment isolation valve surveillance inspection samples.

25 Enclosure

b. Findings

No findings of significance were identified.

1EP6 Drill Evaluation (71114.06)

a. Inspection Scope

The inspectors observed the licensee perform classifications and protective action

recommendations during licensed operator requalification training on June 20, 2006.

The inspectors observed activities in the control room simulator. The inspectors also

attended the post-drill critique in the simulator. The focus of the inspectors activities

was to note any weaknesses and deficiencies in the shift managers performance as

emergency director and ensure the licensee evaluators noted the same weaknesses

and deficiencies and entered them into the corrective action program. As part of the

inspection, the inspectors reviewed the drill package included in the list of documents

reviewed at the end of this report.

These activities represented one drill evaluation inspection sample.

b. Findings

No findings of significance were identified.

2OS1 Access Control to Radiologically Significant Areas (IP 71121.01)

.1 Plant Walkdowns and Radiation Work Permit Reviews

a. Inspection Scope

The inspectors reviewed licensee controls and surveys in the following three

radiologically significant work areas within radiation areas, high radiation areas and

airborne radioactivity areas in the plant and reviewed work packages which included

associated licensee controls and surveys of these areas to determine if radiological

controls including surveys, postings and barricades were acceptable:

Turbine Building Main Steam Reheater Replacement;

Drywell Activities; and

Refuel Floor Activities.

These activities represented one inspection sample.

The inspectors reviewed the radiation work permits (RWPs) and work packages used to

access the three areas and other high radiation work areas to identify the work control

instructions and control barriers that had been specified. Electronic dosimeter alarm set

points for both integrated dose and dose rate were evaluated for conformity with survey

indications and plant policy. Workers were interviewed to verify they were aware of the

actions required when their electronic dosimeters noticeably malfunctioned or alarmed.

26 Enclosure

These activities represented one inspection sample.

The inspectors walked down and surveyed (using an NRC survey meter) the three areas

to verify the prescribed RWPs, procedure, and engineering controls were in place,

licensee surveys and postings were complete and accurate, and air samplers were

properly located.

These activities represented one inspection sample.

The inspectors reviewed RWPs for the following airborne radioactivity areas to verify

barrier integrity and engineering controls performance, e.g., high efficiency particulate

air filter ventilation system operation, and to determine if there was a potential for

individual worker internal exposures of greater than 50 millirem committed effective

dose equivalent. There were no areas where there was a potential for individual worker

internal exposures of greater than 50 millirem committed effective dose equivalent.

Work areas having a history of, or the potential for, airborne transuranic isotopes were

evaluated to verify the licensee had considered the potential for transuranic isotopes

and provided appropriate worker protection. There where no areas having a history of,

or the potential for, airborne transuranic isotopes.

These activities represented one inspection sample.

The adequacy of the licensees internal dose assessment process for any actual internal

exposures greater than 50 millirem committed effective dose equivalent was assessed.

There were no internal exposures greater than 50 millirem committed effective dose

equivalent.

These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

.2 Problem Identification and Resolution

a. Inspection Scope

The inspectors reviewed three corrective action reports related to access controls and

high radiation area radiological incidents. Staff members were interviewed and

corrective action documents were reviewed to verify that follow-up activities were being

conducted in an effective and timely manner commensurate with their importance to

safety and risk based on the following:

initial problem identification, characterization, and tracking;

disposition of operability/reportability issues;

evaluation of safety significance/risk and priority for resolution;

identification of repetitive problems;

identification of contributing causes;

identification and implementation of effective corrective actions;

27 Enclosure

resolution of NCVs tracked in the corrective action system; and

implementation/consideration of risk-significant operational experience feedback.

These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

.3 Job-In-Progress Reviews

a. Inspection Scope

The inspectors observed the following three jobs that were being performed in radiation

areas, airborne radioactivity areas, or high radiation areas for observation of work

activities that presented the greatest radiological risk to workers:

Drywell Cooler Number Four Removal;

Cutout and Replace Check Valve E1100F031A; and

Perform Refuel Activities.

The inspectors reviewed radiological job requirements for the three activities including

RWP requirements and work procedure requirements, and attended As-Low-As-Is-

Reasonably-Achievable (ALARA) job briefings.

These activities represented one inspection sample.

Job performance was observed with respect to these requirements to verify radiological

conditions in the work area were adequately communicated to workers through pre-job

briefings and postings. The inspectors also verified the adequacy of radiological

controls including required radiation, contamination, and airborne surveys for system

breaches; radiation protection job coverage which included audio and visual surveillance

for remote job coverage; and contamination controls.

These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

.4 Radiation Worker Performance

a. Inspection Scope

During job performance observations, the inspectors evaluated radiation worker

performance with respect to stated radiation protection work requirements and

evaluated whether workers were aware of the significant radiological conditions in their

workplace, the RWP controls and limits in place, and that their performance had

accounted for the level of radiological hazards present.

28 Enclosure

These activities represented one inspection sample.

The inspectors reviewed radiological problem reports which found the cause of the

event was due to radiation worker errors to determine if there was an observable pattern

traceable to a similar cause, and to determine if this perspective matched the corrective

action approach taken by the licensee to resolve the reported problems. These

problems, along with planned and taken corrective actions were discussed with the

Radiation Protection Manager.

These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

.5 Radiation Protection Technician (RPT) Proficiency

a. Inspection Scope

During job performance observations, the inspectors evaluated RPT performance with

respect to radiation protection work requirements and evaluated whether they were

aware of the radiological conditions in their workplace, the RWP controls and limits in

place, and if their performance was consistent with their training and qualifications with

respect to the radiological hazards and work activities.

These activities represented one inspection sample.

The inspectors reviewed two radiological problem reports which found the cause of the

event was radiation protection technician error to determine if there was an observable

pattern traceable to a similar cause, and to determine if this perspective matched the

corrective action approach taken by the licensee to resolve the reported problems.

These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

2OS2 As-Low-As-Is-Reasonably-Achievable Planning And Controls (ALARA) (IP 71121.02)

.1 Inspection Planning

a. Inspection Scope

The inspectors reviewed plant collective exposure history, current exposure trends,

ongoing and planned activities in order to assess current performance and exposure

challenges. This included determining the plants current 3-year rolling average for

collective exposure in order to help establish resource allocations and to provide a

perspective of significance for any resulting inspection finding assessment.

29 Enclosure

These activities represented one inspection sample.

The inspectors reviewed the outage work scheduled during the inspection period and

associated work activity exposure estimates for the following five work activities which

were likely to result in the highest personnel collective exposures:

Drywell Cooler Number Four Removal;

Cutout and Replace Check Valve E1100F031A;

Refuel Floor Activities;

Main Steam Reheater Replacement; and

In-Service Inspections.

These activities represented one inspection sample.

The inspectors determined site specific trends in collective exposures and source-term

measurements. The inspectors reviewed procedures associated with maintaining

occupational exposures ALARA and processes used to estimate and track work activity

specific exposures.

These activities represented two inspection samples.

b. Findings

No findings of significance were identified.

.2 Radiological Work Planning

a. Inspection Scope

The inspectors evaluated the licensees list of planned work activities for RF11 ranked

by estimated exposure that were in progress and reviewed the following three work

activities of exposure significance:

06-1113, CRD Exchange;

06-1205, East/West MSR Replacement; and

06-1124, Drywell Cooler Number Four Removal.

For these three activities, the inspectors reviewed the ALARA work activity evaluations,

exposure estimates, and exposure mitigation requirements in order to verify the licensee

had established procedures and engineering and work controls that were based on

sound radiation protection principles in order to achieve occupational exposures that

were ALARA. This also involved determining that the licensee had reasonably grouped

the radiological work into work activities, based on historical precedence, industry

norms, and/or special circumstances.

The inspectors compared the results achieved including dose rate reductions and

person-rem used with the intended dose established in the licensees ALARA planning

for these three work activities. Reasons for inconsistencies between intended and

actual work activity doses were reviewed.

30 Enclosure

These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

.3 Verification of Dose Estimates and Exposure Tracking Systems

a. Inspection Scope

The licensees process for adjusting exposure estimates or re-planning work, when

unexpected changes in scope, emergent work or higher than anticipated radiation levels

were encountered, was evaluated. This included determining that adjustments to

estimated exposure (intended dose) were based on sound radiation protection and

ALARA principles and not adjusted to account for failures to control the work. The

frequency of these adjustments was reviewed to evaluate the adequacy of the original

ALARA planning process.

These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

.4 Job Site Inspections and ALARA Control

a. Inspection Scope

The inspectors observed the following five jobs that were being performed in radiation

areas, airborne radioactivity areas, or high radiation areas for observation of work

activities that presented the greatest radiological risk to workers.

Drywell Cooler Number Four Removal;

Cutout and Replace Check Valve E1100F031A;

Refuel Floor Activities;

Main Steam Reheater Replacement; and

In-Service Inspections.

The licensees use of engineering controls to achieve dose reductions was evaluated to

verify procedures and controls were consistent with the licensees ALARA reviews,

sufficient shielding of radiation sources was provided for, and the dose expended to

install/remove the shielding did not exceed the dose reduction benefits afforded by the

shielding.

These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

31 Enclosure

.5 Radiation Worker Performance

a. Inspection Scope

Radiation worker and RPT performance was observed during work activities being

performed in radiation areas, airborne radioactivity areas, and high radiation areas that

presented the greatest radiological risk to workers. The inspectors evaluated whether

workers demonstrated the ALARA philosophy in practice by being familiar with the work

activity scope and tools to be used, by utilizing ALARA low dose waiting areas, and that

work activity controls were being complied with. Also, radiation worker training and skill

levels were reviewed to determine if they were sufficient relative to the radiological

hazards and the work involved.

These activities represented one inspection sample.

b. Findings

No findings of significance were identified.

4. OTHER ACTIVITIES (OA)

4OA2 Identification and Resolution of Problems (71152)

.1 Routine Review of Identification and Resolution of Problems

a. Inspection Scope

As discussed in previous sections of this report, the inspectors routinely reviewed issues

during baseline inspection activities and plant status reviews to verify they were being

entered into the licensee's corrective action system at an appropriate threshold,

adequate attention was being given to timely corrective actions, and adverse trends

were identified and addressed.

b. Findings

No findings of significance were identified.

.2 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a screening review of each item entered into the licensees

corrective action program to identify trends that might indicate the existence of a more

significant safety issue. The inspectors considered repetitive or closely related issues

that may have been documented by the licensee outside the normal corrective action

program, such as in:

C trend reports or performance indicators,

C major equipment problem lists,

32 Enclosure

C repetitive and/or rework maintenance lists,

C departmental problem/challenges lists,

C system health reports,

C quality assurance audit/surveillance reports,

C self assessment reports,

C maintenance rule assessments, or

C corrective action backlog lists.

The inspectors verified the licensee was identifying issues at an appropriate threshold

and entering them into their corrective action program by comparing those issues

identified by the NRC during the conduct of the plant status and inspectable area

portions of the program with those issues identified by the licensee.

b. Issues

Unidentified drywell leakage was fluctuating after startup from RF11 but has since

leveled out. From the lowest value, unidentified leakage has increased from about

0.06 gpm to an average daily value of 0.14 gpm. Additionally, the inner seal pressure

for the B reactor recirculation pump has been fluctuating by as much as 40-60 psig;

however, there does not appear to be a correlation between the seal pressure

oscillations and drywell leakage. These issues are in the licensees corrective action

program as CARDs 06-24313 for the unidentified leakage and 06-23791 for the seal

pressure oscillations.

4OA3 Event Followup (71153)

.1 Reactor Scram due to Main Transformer Fault

a. Inspection Scope

As described in Section 1R20.3 of this report, the inspectors responded to the control

room on June 15, 2006, when the reactor automatically shut down as a result of the

failure of main unit transformer 2B. The inspectors observed plant parameters and

status, evaluated the performance of mitigating systems and licensee actions, confirmed

that the licensee properly classified the event in accordance with emergency action level

procedures and made timely notifications to NRC and state/county governments, as

required by 10 CFR 50.72 (Event Number 42643). The inspectors determined and

communicated details regarding the event to NRC management, risk analysts and

others in Region III and Headquarters as input to an evaluation per Management

Directive 8.3 for determining the appropriate level of event response. Based on the

events that occurred, routine resident inspection efforts were deemed appropriate.

b. Findings

No findings of significance were identified.

33 Enclosure

.2 Review of Licensee Event Reports (LER)

a. (Closed) LER 50-341/2006-001: At 0039 hours4.513889e-4 days 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months on April 1, 2006, Fermi 2 feedwater line

check valves B2100F010A and B2100F076A failed their LLRT. The air leakage rate of

the inboard check valve B2100F010A was 324.21 standard cubic feet per hour (SCFH),

and the leakage rate of outboard check valve B2100F076A was above the

measurement capability of the leak rate monitor. The penetration (X-9A) minimum-

pathway air leakage value was determined to be 324.21 SCFH which is greater than the

allowable containment leakage rate (La) value of 296.3 SCFH per TS 5.5.12 and higher

than the allowable secondary containment bypass leakage rate of 0.1 La or 29.63 SCFH

per TS Surveillance Requirement 3.6.1.3.11. The B2100F076A failure was attributed to

soft seat degradation which was primarily caused by extending its service time to three

operating cycles. The B2100F010A valve failure was attributed to soft seat degradation

due to a slight misalignment of the valve disc to the in-body seat, compounded by wear

between the internal shaft and valve disc. The slight misalignment caused the soft seat

along the top portion of the disc to contact the seat first, resulting in a scraping action as

the disc flexed to its full seat position. For both valves, the soft seats were replaced,

and the soft seat service time has been limited to two operating cycles. The internal

shaft for the B2100F010A valve was replaced, and the alignment between the disc and

the valve seat was adjusted. Both valves were retested and met their associated LLRT

acceptance criteria prior to restart of the unit.

The LER was reviewed by the inspectors. No findings of significance were identified

and no violation of NRC requirements occurred. The licensee documented the LLRT

failure in CARD 06-21751. This LER is closed.

4OA6 Exit Meetings

.1 Exit Meeting Summary

On July 11, 2006, the inspectors presented the inspection results to Mr. D. Cobb and

other members of licensee management at the conclusion of the inspection. The

inspectors asked the licensee whether any material examined during the inspection

should be considered proprietary. No proprietary information was identified.

.2 Interim Exit Meetings

On April 7, 2006, an interim exit meeting was conducted for the Access Control to

Radiological Areas and ALARA inspection with Mr. Kevin Hlavaty, Plant Manager, and

other licensee staff.

4OA7 Licensee-Identified Violations

The following violation of very low significance was identified by the licensee and is a

violation of NRC requirements, which meet the criteria of Section VI of the NRC

Enforcement Manual, NUREG-1600, for being dispositioned as an NCV.

34 Enclosure

Cornerstone: Public Radiation Safety

The licensees procedure 67.000.103, Surveying of Outgoing Shipments, directs the

staff to survey outgoing vehicles used to carry an exclusive use shipment of radioactive

material. The procedure relies on the proper identification of the incoming shipment as

an exclusive use shipment. This procedure is used to implement the requirements of

49 CFR 173.443 and 49 CFR 177.843 that require the specific release survey of

vehicles in exclusive use situations. Contrary to the above, and as described in

CARD 06-21389, on March 20, 2006, an exclusive use radioactive material shipment

was received by the licensee. The shipment contained one package of Limited Quantity

radioactive material and four boxes of non-radioactive material and the radiation

protection staff assigned to accept the shipment incorrectly identified the shipment as a

non-exclusive use shipment. After the packages were removed from the conveyance,

the vehicle was released without the required survey. This was identified by licensee

supervision but not before the vehicle had departed the site. The carrier was contacted

and the vehicle returned to the licensees site before further transportation activity had

commenced and a survey was completed. No contamination was found and no dose

rates above background were identified. The finding is of very low safety significance

because it did not result in an unmonitored release nor were any dose limits

approached.

ATTACHMENT: SUPPLEMENTAL INFORMATION

35 Enclosure

KEY POINTS OF CONTACT

Licensee

D. Gipson, Chief Nuclear Officer

D. Cobb, Assistant Vice President, Nuclear Generation

K. Hlavaty, Plant Manager

S. Bartman, Nuclear Production

J. Davis, Manager, Outage Management

R. Gaston, Licensing Manager

S. Hassoun, Principal Engineer, Licensing

H. Higgins, Radiation Protection Manager

J. Korte, Manager, Nuclear Security

J. Plona, Engineering Director

NRC

C. Lipa, Chief, Division of Reactor Projects, Branch 4

1 Attachment

LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

05000341/2006003-05 URI Inappropriate Use of Risk in Operability Evaluations

(Section 1R15.3)

Opened and Closed

05000341/2006003-01 NCV Unauthorized Transient Combustibles in Safety-Related

Areas (Section 1R05.2)05000341/2006003-02 NCV Improper Storage of Chemicals Affecting Fire Fighting

Response (Section 1R05.3)05000341/2006003-03 NCV Inadequate Maintenance Risk Assessment

(Section 1R13.2)05000341/2006003-04 NCV Improper Evaluation of Standby Liquid Control Operability

During Tank Sparging (Section 1R15.2)05000341/2006003-06 NCV Inadequate Controls for Venting the Reactor Pressure

Vessel Head (Section 1R20.1)

Closed

05000341/2006-001 LER Excessive Feedwater Check Valve Leakage at Containment

Penetration

Discussed

None.

2 Attachment

LIST OF DOCUMENTS REVIEWED

The following is a list of documents reviewed during the inspection. Inclusion on this list does

not imply that the NRC inspectors reviewed the documents in their entirety but rather that

selected sections of portions of the documents were evaluated as part of the overall inspection

effort. Inclusion of a document on this list does not imply NRC acceptance of the document or

any part of it, unless this is stated in the body of the inspection report.

1R01: Adverse Weather Protection

CARD 06-23861, 06/05/06; Procedure Enhancement for TBHVAC (NRC Comment)

Procedure 27.000.06, Rev 0, 02/27/03; Performance Evaluation Procedure, Hot Weather

Operations

Open/Closed Work Requests by Related Work Code; 05/04/06

1R04: Equipment Alignment

Drawing 6M721-5706-3, 2/16/00; RHR Service Water Make Up Decant and Overflow Systems

Functional Operating Sketch

Drawing 6M721-5706-1, 3/5/04; Residual Heat Removal (RHR) Division II Functional Operating

Sketch

03-00120, 01/02/03; Pinhole leak in piping

03-13694, 6/17/03; Document the Condition of General Service Water Piping

04-24918, 10/25/04; P4100F402A installed at the bottom of the pipe

06-11615, 05/19/06; EDG Electrical Lineup Load Description Changes

06-21618, 03/28/06; E1100F050A Failed PI Leak Test SR 3.4.5.1

06-22730, 4/24/06; NRC-Identified Concerns in GSW Pumphouse

06-23447, 05/18/06; 23.205 Att 1B enhancement

06-23494, 05/20/06; E1100F050A actuator failed to open valve

23.131, Rev.86; General Service Water System

43.000.005, Rev. 30; Visual Examination of Piping and Components (VT-2)

23.208, Rev. 81; RHR Complex Service Water Systems

23.205, Rev. 94; Residual Heat Removal System

1R05: Fire Protection

CARD 06-23365; Door is not latching; Dated May 15, 2006 (NRC-Identified)

CARD 06-23388; Transient Combustibles in the RHR; Dated May 16, 2006 (NRC-Identified)

MOP11; Fire Protection; Revision 10

Fire Brigade Drill Scenario 6, Rev. 11/29/94; First Floor Radioactive Waste Chemical Lab

Storage Area

Fire protection Pre-Plan FP-RDWST, Rev. 4; Radwaste Building Zones 22, 23, 24, and 25

1R06: Flood Protection Measures

CARD 06-21354; High Pressure Coolant Injection Room Flooded due to Drains backing Up;

3/18/06

CARD 06-22600; Moderate Energy Line Break Evaluation; 4/21/06 (NRC-Identified)

3 Attachment

Nuclear Generation Memo TMPE-94-0308; May 18, 1994; Flood Protection Review

6M721-2223, Rev U, 11/24/06; Diagram Equipment Drains All Floors Auxiliary and Reactor

Buildings

6M721-2224, Rev W, 11/24/04; Diagram Floor Drains All Floors Auxiliary and Reactor Buildings

6M721-2032, Rev BO, 04/19/06; Sump Pump Diagram Radwaste System

6M721-2032-1, Rev AI, 04/19/06; Sump Pump Diagram Radwaste System

1R07: Heat Sink Performance

WR TG25060421; Perform RHR Division II Heat Exchanger Performance Test; 3/22/06

1R11: Licensed Operator Requalification

Scenario SS-OP-802-3300, Rev. 0; Anticipated Transient Without Scram with Small Steam

Leak; 1/26/06

1R12: Maintenance Effectiveness

Design Specification 3071-128-EZ-06; Electrical Design Instructions Molded Case Circuit

Breakers

Maintenance Rule Conduct Manual MMR, Appendix E, Rev 4; Maintenance Rule SSC Specific

Functions

Memo dtd 01/30/06, TMIS-06-0011; Summary of Expert Panel Meeting 184 Conducted

January 24, 2006

CTG11-1 Get Well Plan; July 2003

Deviation Event Report 93-0528, 09/15/93

Vendor Manual VME5-18, Rev 0; Spectrum Technologies, Series 5600, MCC

CARD 97-10182, 10/27/97; Defective Molded Case Circuit Breaker

CARD 03-01098,05/30/03; Reactor Protection System - Function Failure

03-19510,07/20/03; Safety Eval 95-0002 is Used as an Operability Evaluation for LPCI with the

RHR Minimum Flow Valves Open

CARD 04-22685, 07/02/04; Generator Transformer - Function Failure

CARD 04-23307. 08/03/04; Motor Control Centers & Dist. Cabinets

CARD 05-23490, 06/21/05; Auxiliary Electrical - Function Failure

CARD 06-21363, 04-02-06; Auxiliary Electrical - Function Failure

CARD 06-21527, 04/09/06; Residual Heat Removal System - Function Failure

CARD 06-22270, 04/12/06, Maintenance Rule Function T4100-09 should be included in

Maintenance Rule Scope Investigation

Procedure 35.306.008, Rev 46; Maintenance Procedure, ITE Gould Motor Control Center Load

Compartment

Procedure 35.306.018, Rev 5; Maintenance Procedure, Spectrum Technology Motor Control

Center Load Compartment

1R13: Maintenance Risk Assessment and Emergent Work Evaluation

Fermi 2 Daily Plant Status, April 2, 2006

Schedulers Evaluation for Fermi 2, April 2, 2006

Fermi 2 Daily Plant Status, April 17, 2006

Schedulers Evaluation for Fermi 2, April 17, 2006

4 Attachment

Fermi 2 Daily Plant Status, April 30, 2006

Schedulers Evaluation for Fermi 2, April 30, 2006

Fermi 2 Daily Plant Status, June 26, 2006

Schedulers Evaluation for Fermi 2, June 26, 2006

CARD 06-21892, 4/4/06; NRC Questions related to Temporary Cooling Installed for Division I

Battery Test (NRC-Identified)

CARD 06-24495, 7/7/06; Work Risk Assessment and Temporary Equipment Controls

(NRC-Identified)

WR 0219060414; Perform 42.309.05 Division I (5 Year) 130/260 VDC batter Check (2A-1 Only)

WR 1219060414; Perform 42.309.05 Division I (5 Year) 130/260 VDC batter Check (2A-2 Only)

1R14: Non-Routine Events

CARD 06-24113; Main Steam Bypass Valves Opened Unintentionally During Power Ascension;

6/18/06

CARD 06-23588; Mode 5 Reactor SCRAM During Installation of SRM Shorting Links; 5/24/06

CARD 06-23501; Full In Light for Control Rod 50-27 on Full Core Display is Intermittent;

5/20/06

1R15: Operability Evaluations

Drawing - 5744, Rev. BK; Emergency Equipment Cooling Water Division I; 11/23/05

WR 000Z060107; Remove and Reinstall Insulation on E1100F031B; 2/9/06

WR 000Z062027; Replace Contaminated Insulation; 6/12/06

CARD 06-24156; Affects of Accidents not Addressed in Insulation Removal Evaluation for

E1100F031B; 6/20/06 (NRC Identified)

CARD 01-17302; There are no Site Guidelines for Insulation Removal on Operable Equipment;

8/14/01

CARD 06-23898; EECW M/U Pump IST Flow Unattainable; 6/7/06

CARD 02-14782; Engineering Evaluation for On-Line Insulation Removal; 6/27/02

CARD 03-16498; Engineering Evaluation for On-Line Insulation Removal; 9/9/03

CARD 05-21940; Engineering Evaluation for On-Line Insulation Removal; 3/23/05

CARD 06-23913; Replace Contaminated Insulation; 6/8/06

CARD 06-23785; Standby Liquid Control Operability During Sparging Activities; 6/1/06 (NRC-

Identified)

CARD 99-13240; Inoperability of Standby Liquid Control During Air Sparging; 4/15/99

1R19: Post-Maintenance Testing

EDP - 31880 120 KV Switchyard Upgrade

Engineering Change Request 33690-1, Rev A, 03/14/06; Replacement of Drywell Cooler Coils

T470B003 and T4700B004

Equivalent Replacement Evaluation (ERE) 34173, Rev 0, 04/18/06; E1150F608 MOV Motor

Replacement

Oil Sample Analysis Reports for C1106C001A, E Control Rod Drive Pump; 01/01/06 -

06/29/06

CARD 06-22258, 04/12/06; CTG equipment issues discovered during IPTE 04-02

CARD 06-22982, 4/30/06; Inboard MSIV A will not slow close

CARD 06-23031, 5/1/06; MSIV B2103F028B RPS Limit Switch did not actuate when expected

5 Attachment

CARD 06-22634, 4/22/06; A Inboard MSIV limit switch, PIS B21N572A, will not change state

Drawing SD-F-0179, Rev. A, 9/25/05; Diagram Line Breaker Control 120KV, POS GK

Drawing SD-2500-01, Rev. A, 2/28/06; One Line Diagram Plant 4160V & 480V

Drawing SD-2500-02, Rev. A, 2/23/06; One Line Diagram 13.8KV

Drawing SD-F-0001, Rev. A, 2.23/06; One Line Diagram 120 KV Switchyard

IPTE 04-02, 120KV Switchyard Upgrade

Procedure 24.137.01, Rev. 34; Main Steam Isolation Channel Functional Test

24.206.01, Rev 63, 05/08/06; RCIC System Pump and Valve Operability Test

43.000.005, Rev 30, 03/22/06; Visual Examination of Piping and Components (VT-2)

43.401.303, Rev 32, 10/24/05; Local Leakage Rate Testing for Penetration X-9A

WR B203040100, 04/10/06; B2100F010A - Nuclear Boiler Feedwater Supply Inboard Primary

Containment Check Valve

WR 000Z050487, 04/29/06;B3105F031A - Reactor Recirculation Pump A Discharge Valve

WR T210040100, 04/10/06; B2100F076A - Nuclear Boiler Feedwater Supply Check Valve

WR 000Z052131, 04/19/06; Inservice Testing, Drywell Cooler #4

WR 000Z060156, 6/19/06; E Control Rod Drive Inboard Pump Bearing Oil level Bulls eye Dirty

1R20: Refueling and Outage Activities

Apparent Cause Determination for Damage Found in the Main Generator During RF11 Robot

Inspections CARD 06-21922

Inspection Requirement Form, Requisition Number: 9086929; 04/19/2006

Drawing 6M721-3722, Rev A; Flow Diagram & Details of Purging Unit - Reactor Pressure

Vessel - Unit 2

Safety Tagging Record 2006-002089

CARD 06-21534, 03/26/2006; Continuous Air Monitor Alarm on Refueling Floor

CARD 06-22590, 04/21/2006; NRC identified concern with Div 1 Core Spray and Defense in

Depth Investigation

CARD 06-22642, 04/22/2006; Are EDG Surveillances Testing What They Are Setup to Test

CARD 06-22667, 04/23/2006; RPV Venting Unit Configuration Control Discrepancies

CARD 06-23114, 05/03/2006; NRC Concern: Review of ERE-34173 E1150F608 MOV Motor

Replacement

CARD 06-23793, 06/02/2006; NRC Concern - Material Released Without the Requirements of

EED Being Verified

Procedure 24.106.06, Rev 25; Surveillance Procedure, SCRAM Discharge Volume Vent and

Drain Valves SCRAM Operability Test

Procedure 32.000.07, Rev 33; Crane Operation Procedure, Reactor Building Crane Operation

Procedure 35.717.001, Rev 29; Maintenance Procedure, Reactor Building Crane - Quarterly

Preventive Maintenance

Procedure 35.717.003, Rev 3; Maintenance Procedure, Reactor Building Crane - Annual PM

Inspection

Procedure 43.401.303, Rev 32; Surveillance Procedure, Local Leakage Rate Testing for

Penetration X-9A

Work Control Conduct Manual MWC13, Rev 0; Outage Nuclear Safety

Maintenance Conduct Manual MMA07, Rev 14; Hoisting, Rigging and Load Handling

6 Attachment

1R22: Surveillance Testing

Drawing 6M721-5703-1, Rev. Y; Control Rod Drive System Functional Operating Sketch

CARD 06-22739, 4/24/06; Acrid smell from EDG #13 local control cabinet

CARD 06-23031, 5/1/06; MSIV B2103F028B RPS Limit Switch did not actuate when expected

CARD 06-22653, 04/23/06; HCU 46-27 Conduit for wiring to accumulator pressure switch

needs repair

CARD 06-22999, 05/01/06; Enhancements to 24.405.03

CARD 06-21681; Potential Enhancement for MSIV Switch Calibrations; 3/20/06 (NRC-

Identified)

CARD 06-21720; NRC Question Regarding MSIV Switch Testing; 3/31/06 (NRC-Identified)

CARD 06-22046; Loss of SLC Squib Valve A Continuity Light During Performance of

24.139.03; 4/7/06

Procedure 24.106.06, Rev 25; Surveillance Procedure, SCRAM Discharge Volume Vent and

Drain Valves SCRAM Operability Test

Procedure 24.106.08, Rev. 3; CRD Hydraulic Unit Accumulator Integrity Test

Procedure 24.137.01, Rev. 34; Main Steam Isolation Channel Functional Test

Procedure 24.405.03, Rev. 33; Secondary Containment Operability Test

Procedure 24.307.03, Rev. 38; EDG 13 - Loss of Offsite Power and ECCS Start with Loss of

Offsite Power Test

Procedure 24.202.02, Rev. 42; HPCI Flow Rate Test at 165 psig Reactor Steam Pressure

Procedure 24.206.04, Rev. 44; RCIC System Automatic Actuation and Flow Test

Procedure 43.401.500, Rev 34; Surveillance Procedure, Local Leakage Rate Testing for

Penetration X-7A, X-7B, X-7C, and X-7D

Procedure 43.401.500, Rev 35; Surveillance Procedure, Local Leakage Rate Testing for

Penetration X-7A, X-7B, X-7C, and X-7D

Procedure 24.402.06; Rev. 32; Drywell to Suppression Chamber Bypass Leak Test

Procedure 35.139.002; Rev. 27; SLC System Explosive Valve Insert Replacement

Procedure 24.139.03, Rev. 40; SLC Manual Initiation, RWCU Isolation, and Storage tank

Heater Operability Test

WR 0311060425; Perform 24.402.06 Drywell to Suppression Chamber Bypass Leak Test;

3/25/06

WR 0518041022; Perform 44.010.063 RPS MSIV Outboard Valve Limit Switch, Div. I and II,

Calibration

WR 1245060421; Perform 24.139.03 SLC Loop B Pump Flow, Manual Initiate, and Squib

Firing; 4/21/06

1EP6: Drill Evaluation

Scenario SS-OP-802-3300, Rev. 0; Anticipated Transient Without Scram with Small Steam

Leak; 1/26/06

2OS1: Access Control to Radiologically Significant Areas

CARD 05-26818; Primary Containment Atmosphere Sample Pump; T5001-C003 Will Not Start;

dated December 5, 2005

CARD 06-11546; WGI Employee Felt Something in His Left Eye after Removing Protective

Clothing at Drywell Step-off Pad; dated April 5, 2006

CARD 06-21534; High Radiation Alarm on Refuel Floor; dated March 26, 2006

7 Attachment

CARD 06-21640; Unnecessary Contamination of Personnel; dated March 29, 2006

CARD 06-21639; Evaluate Dose an Dose Rate Alarms for Fast Entry Electronic Dosimeters;

dated March 29, 2006

CARD 06-20962; Worker Electronic Dosimeter Alarm on Incorrect Task; dated February 23,

2006

CARD 06-21177; Unexpected Dose Rate Alarm; dated March 8, 2006

CARD 06-21787; Entering High Radiation Area on Incorrect Task; dated April 1, 2006

CARD 06-21807; Evaluate Reactor Building Ventilation Impact on the Spread of Contamination

during RF11 Vessel Fill Up; dated April 2, 2006

CARD 06-21857; Drywell Stepoff Pad Poor Radiation Protection Practices; dated April 3, 2006

CARD 06-21873; Potential Release of Radioactive Contamination; dated April 4, 2006

CARD 06- 21868; Workers Entered Top of Torus High Radiation Area on Wrong Radiation

Work Permit; dated April 4, 2006

CARD 06-21944; Foreign Material Found Inside Main Condenser Upper Steam Space; dated

April 5, 2006

CARD 06-21958; Worker Received Puncture Wound; dated April 6, 2006

EP-225; Radiological Medical Emergencies; Revision 13

RF11 ALARA Self-Assessment; dated April 4, 2006

2OS2 As-Low-As-Is-Reasonably-Achievable Planning And Controls (ALARA)

RWP 06-1110; In-Service Inspections; Revision 0

RWP 06-1103; Install and Remove Temporary Shielding, Install Permanent Shielding;

Revision 0

RWP 06-1105; Job Progress for Scaffold Activities in the Reactor Building Steam Tunnel and

Drywell; Revision 0

RWP 06-1124; Replace Number 4 Drywell Cooler; Revision 4

RWP 06-1164; E1100F031A Cutout and Replace Check Valve; Revision 0

RWP 06-1251; Perform Refuel Activities on Reactor Building 5; Revision 0

CARD 06-21765; Work Initiated Without Radiation Protection Review; dated April 1, 2006

CARD 06-21911; E100F031A Glovebag Not Installed Properly; dated April 5, 2006

CARD 06-21947; E1100F060A Re-pack Extra Time and Dose Above Initial Estimate; dated

April 5, 2006

Procedure 63.000.100; Respirator Evaluation Work Sheet (RWP 06-1117); Revision 0

4OA2: Identification and Resolution of Problems

CARDs initiated between 1/1/06 and 06/30/06

4OA3: Event Followup (71153)

NRC-06-0037 Letter; Licensee Event report No. 2006-001, Excessive Feedwater Check Valve

Leakage At Containment Penetration; dated May 24, 2006

4OA7 Licensee-Identified Violations

CARD 06-21389; Failure to Perform Required Surveys in Accordance with

Procedure 67.000.102; dated March 21, 2006

Procedure 67.000.103; Survey of Outgoing Radioactive Material Shipments; Revision 16

Procedure 67.000.102; Survey of Incoming Radioactive Material Shipments; Revision 0

8 Attachment

LIST OF ACRONYMS USED

ALARA As Low As Reasonable Achievable

CARD Condition Assessment Resolution Document

CDF Core Damage Frequency

CFR Code of Federal Regulations

CTG Combustion Turbine Generator

DRP Division of Reactor Projects

EDG Emergency Diesel Generator

GSW General Service Water

HPCI High Pressure Coolant Injection

IMC Inspection Manual Chapter

LER Licensee Event Report

LERF Large Early Release Frequency

LLRT Local Leak Rate Test

MCC Motor Control Center

MSIV Main Steam Isolation Valve

NCV Non-Cited Violation

NRC Nuclear Regulatory Commission

PI Performance Indicator

PMT Post-Maintenance Testing

RCIC Reactor Core Isolation Cooling

RCS Reactor Coolant System

RHR Residual Heat Removal

RHRSW Residual Heat Removal Service Water

RPS Reactor Protection System

RPSMG Reactor Protection System Motor Generator

RPT Radiation Protection Technician

RWCU Reactor Water Cleanup

RWP Radiation Work Package

SCFH Standard Cubic Feet Per Hour

SCFM Standard Cubic Feet Per Minute

SDP Significance Determination Process

SGTS Standby Gas Treatment System

SLC Standby Liquid Control

SPAR Standardized Plant Analysis Risk

SRA Senior Reactor Analyst

TM Temporary Modifications

TS Technical Specifications

UFSAR Updated Final Safety Analysis Report

WR Work Request

9 Attachment

ML062160540

Text

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