ML100271264

IR 05000254-09-005 and 05000265-09-005 on 10/01/09 - 12/31/09 for Quad Cities Nuclear Power Station, Units 1 & 2

ML100271264
Person / Time
Site:	Quad Cities
Issue date:	01/27/2010
From:	Ring M NRC/RGN-III/DRP/B1
To:	Pardee C Exelon Generation Co, Exelon Nuclear
References
FOIA/PA-2010-0209 IR-09-005
Download: ML100271264 (45)

See also: IR 05000254/2009005

Text

UNITED STATES

NUCLEAR REGULATORY COMMISSION

REGION III

2443 WARRENVILLE ROAD, SUITE 210

LISLE, IL 60532-4352

January 27, 2010

Mr. Charles G. Pardee

Senior Vice President, Exelon Generation Company, LLC

President and Chief Nuclear Officer (CNO), Exelon Nuclear

4300 Winfield Road

Warrenville, IL 60555

SUBJECT: QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2

NRC INTEGRATED INSPECTION REPORT 05000254/2009005;

05000265/2009005

Dear Mr. Pardee:

On December 31, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an

integrated inspection at your Quad Cities Nuclear Power Station, Units 1 and 2. The enclosed

report documents the inspection findings, which were discussed on January 5, 2010, with

Mr. T. Tulon and other members of your staff.

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

compliance with the Commissions rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed

personnel.

Based on the results of this inspection, three self-revealed findings of very low safety

significance were identified. Two of the findings involved a violation of NRC requirements.

However, because of their very low safety significance, and because the issues were entered

into your corrective action program, the NRC is treating the issues as non-cited violations

(NCVs) in accordance with Section VI.A.1 of the NRC Enforcement Policy. Additionally,

a licensee-identified violation is listed in Section 4OA7 of this report.

If you contest the subject or severity of an NCV, 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 Quad Cities Nuclear Power Station. In addition, if you disagree with the

characterization of any finding in this report, you should provide a response within 30 days of

the date of this inspection report, with the basis for your disagreement, to the Regional

Administrator, Region III, and the NRC Resident Inspector at the Quad Cities Nuclear Power

Station. The information that you provide will be considered in accordance with Inspection

Manual Chapter 0305.

C. Pardee -2-

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

NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Mark A. Ring, Chief

Branch 1

Division of Reactor Projects

Docket Nos. 50-254; 50-265

License Nos. DPR-29; DPR-30

Enclosure: Inspection Report 05000254/2009005; 05000265/2009005

w/Attachment: Supplemental Information

cc w/encl: Distribution via ListServ

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket Nos: 50-254, 50-265

License Nos: DPR-29, DPR-30

Report No: 05000254/2009005 and 05000265/2009005

Licensee: Exelon Nuclear

Facility: Quad Cities Nuclear Power Station, Units 1 and 2

Location: Cordova, IL

Dates: October 1 through December 31, 2009

Inspectors: J. McGhee, Senior Resident Inspector

B. Cushman, Resident Inspector

R. Orlikowski, Senior Resident Inspector - Duane Arnold

M. Bielby, Senior Operations Engineer

C. Moore, Operations Engineer

M. Mitchell, Senior Radiation Protection Inspector

R. Jickling, Senior Emergency Preparedness Inspector

C. Mathews, Illinois Emergency Management Agency

Approved by: M. Ring, Chief

Branch 1

Division of Reactor Projects

Enclosure

TABLE OF CONTENTS

SUMMARY OF FINDINGS ...........................................................................................................1

REPORT DETAILS .......................................................................................................................4

Summary of Plant Status...........................................................................................................4

1. REACTOR SAFETY .......................................................................................................4

1R01 Adverse Weather Protection (71111.01)..............................................................4

1R04 Equipment Alignment (71111.04) ........................................................................5

1R05 Fire Protection (71111.05) ...................................................................................6

1R11 Licensed Operator Requalification Program (71111.11)......................................7

1R12 Maintenance Effectiveness (71111.12)..............................................................11

1R13 Maintenance Risk Assessments and Emergent Work Control (71111.13) ........12

1R15 Operability Evaluations (71111.15) ....................................................................12

1R19 Post-Maintenance Testing (71111.19) ...............................................................13

1R22 Surveillance Testing (71111.22) ........................................................................14

1EP4 Emergency Action Level and Emergency Plan Changes (71114.04) ................15

1EP6 Drill Evaluation (71114.06).................................................................................17

4. OTHER ACTIVITIES.....................................................................................................18

4OA1 Performance Indicator Verification (71151) .......................................................18

4OA2 Identification and Resolution of Problems (71152) ............................................21

4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153) ...............27

4OA5 Other Activities ...................................................................................................30

4OA6 Management Meetings ......................................................................................30

4OA7 Licensee-Identified Violations ............................................................................31

SUPPLEMENTAL INFORMATION ...............................................................................................1

Key Points of Contact ................................................................................................................1

List of Items Opened, Closed and Discussed............................................................................1

List of Documents Reviewed .....................................................................................................2

List of Acronyms Used ..............................................................................................................8

Enclosure

SUMMARY OF FINDINGS

IR 05000254/2009005, 05000265/2009005; 10/01/09 - 12/31/09; Quad Cities Nuclear Power

Station, Units 1 & 2; Other Activities.

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

baseline inspections by regional inspectors. Three Green findings were identified by the

inspectors. Two of the findings were considered Non-Cited Violations (NCVs) of NRC

regulations. The significance of most findings is indicated by their color (Green, White, Yellow,

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

(SDP). Findings for which the SDP does not apply may be Green or be assigned a severity

level after NRC management review. The NRCs program for overseeing the safe operation of

commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process,

Revision 4, dated December 2006.

A. NRC-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

• Green. A finding of very low safety significance and a NCV of 10 CFR 50 Appendix B,

Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the

installation of an inappropriate component into the Unit 2 emergency diesel generator

coolant system. Specifically, the licensee failed to properly perform a part evaluation for

a replacement temperature indicator (TI) designated as augmented quality. This

resulted in the TI probe shearing off in the coolant flow stream and causing foreign

material to enter the coolant system. Immediate corrective actions included the

installation of an appropriately approved TI and recovery of foreign material from the

system.

The same part evaluation process was used for risk-significant components independent

of the system being worked. Therefore, this finding was more than minor because, if left

uncorrected, this performance deficiency could lead to unplanned unavailability of

safety-related or risk-significant equipment and would become a more significant safety

concern. The inspectors performed a Phase 1 SDP screening and concluded that the

issue was of very low safety significance (Green) because the failure of the TI did not

result in unplanned inoperability or loss of function of the diesel generator. The

inspectors determined that this finding did not have a cross-cutting aspect. This

performance deficiency is not indicative of current licensee performance. The decision

to install this type of TI was made in October 2007. The process which allowed this

performance deficiency was identified and corrected through procedure and policy

revisions in February 2008. (Section 4OA2)

• Green: A finding of very low safety significance and a NCV of TS 3.6.2.4,

Residual Heat Removal (RHR) Suppression Pool Spray, was self-revealed for the

licensees failure to meet the Technical Specification (TS) limiting conditions of operation

(LCO) requirement prior to transitioning into an operating mode where the LCO was

required to be satisfied. Specifically, Motor Operator (MO) 1-1001-37B for the Unit 1

RHR torus (suppression pool) spray isolation valve was found to have been inoperable

when the operating crew transitioned Unit 1 from Mode 4 to Mode 2 on May 30, 2009.

The valve actuator had been inadvertently declutched (i.e., motor disengaged) and the

valve was not demonstrated operable by stroking the valve electrically after the actuator

1 Enclosure

motor was declutched. Inspectors determined that the finding was cross-cutting in the

area of Problem Identification and Resolution - Corrective Action (P.1(a)) because plant

personnel failed to identify the physical contact with the valve actuator that resulted in

the valve being declutched; therefore, operators incorrectly assessed the system

condition as in compliance with TS 3.6.2.4. Immediate licensee corrective actions

included engagement of the motor and stroke testing of the valve.

The finding is more than minor because it was associated with the equipment

performance quality attribute of the Mitigating Systems Cornerstone and affected the

objective of ensuring availability, reliability, and capability of systems that respond to

initiating events to prevent undesirable consequences. Specifically, failure to verify

system availability and capability prior to entering the required modes resulted in fewer

available mitigating systems than assumed in the operating risk evaluations. The

inspectors determined the finding could be evaluated using the SDP in accordance with

IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial

Screening and Characterization of findings, Table 4a. Inspectors answered all of the

questions for the Mitigating Systems Cornerstone No. Therefore, the finding screened

as Green or very low safety significance. (Section 4OA3)

Cornerstone: Barrier Integrity

• Green. A finding of very low safety significance was self-revealed for the failure to

perform maintenance that would ensure the portable emergency flooding pump (Darley

pump) was in a standby condition and readily available to accomplish the requirements

of QCOA 0010-16, Flood Emergency Procedure. Specifically, the failure to perform

adequate maintenance resulted in the need to replace the battery and gasoline for the

pump and, upon pump start, fuel sprayed out of the fuel pump. Although the staged

portable pump would not have supported the external flooding emergency response

procedure, no violation of regulatory requirements occurred. The inspectors did not

identify a cross-cutting aspect associated with this finding because the issue is not

reflective of current licensee performance. Immediate corrective actions included

replacement of the degraded battery and overhaul of the pumps fuel pump. Other

actions included identification of preventative maintenance tasks and establishing a

program owner of the pump and support equipment.

This issue was more than minor because it was associated with the Structures,

Systems, and Components (SSC) Performance attribute of the Barrier Integrity

Cornerstone objective of maintaining the functionality of spent fuel pool cooling.

The finding affected the cornerstone objective of providing assurance that physical

design barriers protect the public from radionuclide releases caused by events including

external flooding. Specifically, the pump could fail due to maintenance preventable

component failure resulting in inadequate or degraded makeup to the spent fuel pool

during an external flooding event. The inspectors determined the finding could be

evaluated using the SDP in accordance with IMC 0609, Significance Determination

Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of

findings, Tables 4a and 4b. The inspectors determined that even though this equipment

is assumed to completely fail, the licensee could provide an alternate portable pump

already located on site and capable of performing the safety function during this slow

developing event. Since alternate equipment was available and the delay in mobilizing

the alternate equipment would not have resulted in loss of capability to mitigate the

2 Enclosure

impact of the flooding event, the issue is of very low safety significance or Green.

(Section 4OA2)

B. Licensee-Identified Violations

A violation of very low safety significance that was identified by the licensee was

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

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

corrective action tracking number are listed in Section 4OA7 of this report.

3 Enclosure

REPORT DETAILS

Summary of Plant Status

Unit 1

Unit 1 operated at 100 percent thermal power throughout the evaluated period from October 1

until December 31, 2009, with the exception of planned power reductions for routine

surveillances, planned equipment repair, and control rod maneuvers.

Unit 2

Unit 2 operated at or near 100 percent thermal power from October 1 until December 16 with

the exception of planned power reductions for routine surveillances and control rod maneuvers.

On December 16, 2009, operators attempted to replace a light bulb in the indication circuit for

the extraction steam check valve A on the 2D feedwater heaters. The light bulb separated with

the base remaining in the socket. During the evolution the D heaters tripped, resulting in a

partial loss of feedwater heating and a resulting change in reactor power. Operators lowered

power about 150 MWth (50 MWe) by inserting one high reactivity-worth control rod. Power

increased by 0.59 percent during the loss of feedwater heating transient. By 10:45 a.m. that

same morning, feedwater heaters had been restored and the control rod was withdrawn to

restore the unit to 100 percent thermal power. The unit remained at 100 percent power for the

duration of the evaluated period.

1. REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01)

.1 Winter Seasonal Readiness Preparations

a. Inspection Scope

The inspectors conducted a review of the licensees preparations for winter conditions to

verify that the plants design features and implementation of procedures were sufficient

to protect mitigating systems from the effects of adverse weather. Documentation for

selected risk-significant systems was reviewed to ensure that these systems would

remain functional when challenged by inclement weather. During the inspection, the

inspectors focused on plant-specific design features and the licensees procedures used

to mitigate or respond to adverse weather conditions. Additionally, the inspectors

reviewed the Updated Final Safety Analysis Report (UFSAR) and performance

requirements for systems selected for inspection, and verified that operator actions were

appropriate as specified by plant-specific procedures. Cold weather protection, such as

heat tracing and area heaters, was verified to be in operation where applicable. The

inspectors also reviewed corrective action program (CAP) items to verify that the

licensee was identifying adverse weather issues at an appropriate threshold and

entering them into the CAP in accordance with station corrective action procedures.

Specific documents reviewed during this inspection are listed in the Attachment to this

report. The inspectors reviews focused specifically on the following plant systems due

to their risk significance or susceptibility to cold weather issues:

4 Enclosure

• heating steam, and

• circulating water/de-icing valve.

This inspection constituted one winter seasonal readiness preparations sample as

defined in Inspection Procedure (IP) 71111.01-05.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignment (71111.04)

.1 Quarterly Partial System Walkdowns

a. Inspection Scope

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

systems:

• 1/2 B diesel driven fire pump; and

• Unit 1 emergency diesel generator and diesel generator cooling water pump.

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

Reactor Safety Cornerstone at the time they were inspected. The inspectors attempted

to identify any discrepancies that could impact the function of the system, and, therefore,

potentially increase risk. The inspectors reviewed applicable operating procedures,

system diagrams, UFSAR, Technical Specification (TS) requirements, outstanding work

orders (WOs), condition reports, 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 and support equipment

were aligned correctly and operable. The inspectors examined the material condition of

the components and observed operating parameters of equipment to verify that there

were no obvious deficiencies. The inspectors also verified that the licensee had properly

identified and resolved equipment alignment problems that could cause initiating events

or impact the capability of mitigating systems or barriers and entered them into the CAP

with the appropriate significance characterization. Documents reviewed are listed in the

Attachment to this report.

These activities constituted two partial system walkdown samples as defined in

IP 71111.04-05.

b. Findings

No findings of significance were identified.

5 Enclosure

.2 Semi-Annual Complete System Walkdown

a. Inspection Scope

On November 5, 2009, the inspectors performed a complete system alignment

inspection of the Unit 2 emergency diesel generator to verify the functional capability of

the system. This system was selected because it was considered both safety significant

and risk significant in the licensees probabilistic risk assessment. The inspectors

walked down the system to review mechanical and electrical equipment lineups;

electrical power availability; system pressure and temperature indications, as

appropriate; component labeling; component lubrication; component and equipment

cooling; hangers and supports; operability of support systems; and to ensure that

ancillary equipment or debris did not interfere with equipment operation. A review of a

sample of past and outstanding work orders was performed to determine whether any

deficiencies significantly affected the system function. In addition, the inspectors

reviewed the CAP database to ensure that system equipment alignment problems were

being identified and appropriately resolved. Documents reviewed are listed in the

Attachment to this report.

These activities constituted one complete system walkdown sample as defined in

IP 71111.04-05.

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 walkdowns which were focused on availability,

accessibility, and the condition of firefighting equipment in the following risk-significant

plant areas:

• Unit 2 Reactor Bldg. El. 5540, NW Corner Room - 2A Core Spray, Fire Zone

11.3.3;

• Unit 1 Turbine Bldg. El. 5950, Diesel Generator, Fire Zone 9.1;

• Unit 1 Turbine Bldg. El. 5950, Reactor Feed Pumps, Fire Zone 8.2.6.A;

• Crib House Bldg. El. 5598, Basement, Fire Zone 11.4.A; and

• Crib House Bldg. El. 5950, Ground Floor/Service Water Pumps, Fire Zone

11.4.B.

The inspectors reviewed areas to assess if the licensee had implemented a fire

protection program that adequately controlled combustibles and ignition sources within

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

passive fire protection features in good material condition, and implemented adequate

compensatory measures for out-of-service, degraded or inoperable fire protection

equipment, systems, or features in accordance with the licensees fire plan.

The inspectors selected fire areas based on their overall contribution to internal fire risk

6 Enclosure

as documented in the plants Individual Plant Examination of External Events with later

additional insights, their potential to impact equipment which could initiate or mitigate a

plant transient, or their impact on the plants ability to respond to a security event.

Using the documents listed in the Attachment to this report, the inspectors verified that

fire hoses and extinguishers were in their designated locations and available for

immediate use; that fire detectors and sprinklers were unobstructed; that transient

material loading was within the analyzed limits; and fire doors, dampers, and penetration

seals appeared to be in satisfactory condition. The inspectors also verified that minor

issues identified during the inspection were entered into the licensees CAP.

Documents reviewed are listed in the Attachment to this report.

These activities constituted five quarterly fire protection inspection samples as defined in

IP 71111.05-05.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification Program (71111.11)

.1 Resident Inspector Quarterly Review (71111.11Q)

a. Inspection Scope

On November 4, 2009, the inspectors observed licensed operator continuing training to

verify that operator performance was adequate, evaluators were identifying and

documenting crew performance problems, and training was being conducted in

accordance with licensee procedures. The inspectors evaluated the following areas:

• licensed operator performance;

• crews communications and accuracy of documentation;

• ability to take timely actions in the conservative direction;

• correct use and implementation of abnormal and emergency procedures;

• control board manipulations;

• oversight and direction from supervisors; and

• ability to identify and implement Emergency Plan actions and notifications.

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

expectations and lesson objectives. Documents reviewed are listed in the Attachment to

this report.

This inspection constituted one quarterly licensed operator requalification program

sample as defined in IP 71111.11.

b. Findings

No findings of significance were identified.

7 Enclosure

.2 Facility Operating History (71111.11B)

a. Inspection Scope

The inspectors reviewed the plants operating history from January 2007 through

September 2009 to identify operating experience that was expected to be addressed by

the Licensed Operator Requalification Training (LORT) program. The inspectors verified

that the identified operating experience had been addressed by the facility licensee in

accordance with the stations approved Systems Approach to Training (SAT) program to

satisfy the requirements of 10 CFR 55.59(c). The documents reviewed during this

inspection are listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

.3 Licensee Requalification Examinations

a. Inspection Scope

The inspectors performed an inspection of the licensees LORT test/examination

program for compliance with the stations SAT program which would satisfy the

requirements of 10 CFR 55.59(c)(4). The reviewed operating examination material

consisted of two operating tests, each containing two dynamic simulator scenarios and

five job performance measures (JPMs). The two biennial written examinations reviewed

consisted of two parts. Each written examination contained 30 questions consisting of

15 written exam questions and 15 static exam questions. The inspectors reviewed the

annual requalification operating test and biennial written examination material to

evaluate general quality, construction, and difficulty level. The inspectors assessed the

level of examination material duplication from week to week during the current year

operating test. The examiners assessed the amount of written examination material

duplication from week to week for the biennial written examination administered in

calendar year 2009. The inspectors reviewed the methodology for developing the

examinations, including the LORT program 2-year sample plan, probabilistic risk

assessment insights, previously identified operator performance deficiencies, and plant

modifications. The documents reviewed during this inspection are listed in the

Attachment to this report.

b. Findings

No findings of significance were identified.

.4 Licensee Administration of Requalification Examinations

a. Inspection Scope

The inspectors observed the administration of a requalification operating test to

assess the licensees effectiveness in conducting the test to ensure compliance with

10 CRF 55.59(c)(4). The inspectors evaluated the performance of one operating crew in

parallel with the facility evaluators during four dynamic simulator scenarios and

evaluated various licensed crew members concurrently with facility evaluators during the

8 Enclosure

administration of several JPMs. The inspectors assessed the facility evaluators ability

to determine adequate crew and individual performance using objective, measurable

standards. The inspectors observed the training staff personnel administer the operating

test, including conducting pre-examination briefings, evaluations of operator

performance, and individual and crew evaluations upon completion of the operating test.

The inspectors evaluated the ability of the simulator to support the examinations.

b. Findings

No findings of significance were identified.

.5 Examination Security

a. Inspection Scope

The inspectors observed and reviewed the licensees overall licensed operator

requalification examination security program related to examination physical security

(e.g., access restrictions and simulator considerations) and integrity (e.g., predictability

and bias) to verify compliance with 10 CFR 55.49, Integrity of Examinations and Tests.

The inspectors also reviewed the facility licensees examination security procedure and

the implementation of security and integrity measures (e.g., security agreements,

sampling criteria, bank use, and test item repetition) throughout the examination

process. No examination security compromises occurred during these observations.

The documents reviewed during this inspection are listed in the Attachment to this

report.

b. Findings

No findings of significance were identified.

.6 Licensee Training Feedback System

a. Inspection Scope

The inspectors assessed the methods and effectiveness of the licensees processes for

revising and maintaining its LORT program up-to-date, including the use of feedback

from plant events and industry experience information. The inspectors reviewed the

licensees quality assurance oversight activities, including licensee training department

self-assessment reports. The inspectors evaluated the licensees ability to assess the

effectiveness of its LORT program and their ability to implement appropriate corrective

actions. This evaluation was performed to verify compliance with 10 CFR 55.59(c) and

the licensees SAT based program. The documents reviewed during this inspection are

listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

9 Enclosure

.7 Licensee Remedial Training Program

a. Inspection Scope

The inspectors assessed the adequacy and effectiveness of the remedial training

conducted since the previous biennial requalification examinations and the training from

the current examination cycle to ensure that they addressed weaknesses in licensed

operator or crew performance identified during training and plant operations. The

inspectors reviewed remedial training procedures and individual remedial training plans.

This evaluation was performed in accordance with 10 CFR 55.59(c) and with respect to

the licensees SAT based program. The documents reviewed during this inspection are

listed in the Attachment to this report.

b. Findings

No findings of significance were identified.

.8 Conformance With Operator License Conditions

a. Inspection Scope

The inspectors reviewed the facility and individual operator licensees' conformance with

the requirements of 10 CFR Part 55. The inspectors reviewed the facility licensee's

program for maintaining active operator licenses and to assess compliance with

10 CFR 55.53(e) and (f). The inspectors reviewed the procedural guidance and the

process for tracking on-shift hours for licensed operators and which control room

positions were granted watch-standing credit for maintaining active operator licenses.

The inspectors reviewed the facility licensee's LORT program to assess compliance with

the requalification program requirements as described by 10 CFR 55.59(c). Additionally,

medical records for 10 licensed operators were reviewed for compliance with

10 CFR 55.53(I). The documents reviewed during this inspection are listed in the

Attachment to this report.

b. Findings

No findings of significance were identified.

.9 Annual Operating Test Results and Biennial Written Examination Results (71111.11B)

a. Inspection Scope

The inspectors reviewed the overall pass/fail results of the individual JPM operating

tests, the simulator operating tests, and the biennial written examination (required to be

given per 10 CFR 55.59(a)(2)) administered by the licensee from September 2009

through November 2009 as part of the licensees operator licensing requalification cycle.

These results were compared to the thresholds established in Inspection Manual

Chapter 0609, Appendix I, Licensed Operator Requalification Significance

Determination Process (SDP)." The evaluations were also performed to determine if the

licensee effectively implemented operator requalification guidelines established in

NUREG 1021, Operator Licensing Examination Standards for Power Reactors, and

10 Enclosure

IP 71111.11, Licensed Operator Requalification Program. The documents reviewed

during this inspection are listed in the Attachment to this report.

This inspection constituted one inspection sample as defined in IP 71111.11.

b. Findings

No findings of significance were identified.

1R12 Maintenance Effectiveness (71111.12)

.1 Routine Quarterly Evaluations (71111.12Q)

a. Inspection Scope

The inspectors evaluated degraded performance issues involving the following

risk-significant systems:

• Z2900; Safe Shutdown Makeup Pump, and

• Z4700; Instrument Air.

The inspectors reviewed events such as where ineffective equipment maintenance had

resulted in valid or invalid automatic actuations of engineered safeguards systems and

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

problems in terms of the following:

• implementing appropriate work practices;

• identifying and addressing common cause failures;

• scoping of systems in accordance with 10 CFR 50.65(b) of the maintenance rule;

• characterizing system reliability issues for performance;

• charging unavailability for performance;

• trending key parameters for condition monitoring;

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

• verifying appropriate performance criteria for SSCs/functions classified as (a)(2)

or appropriate and adequate goals and corrective actions for systems classified

as (a)(1).

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

and condition monitoring of the system. In addition, the inspectors verified maintenance

effectiveness issues were entered into the CAP with the appropriate significance

characterization. Documents reviewed are listed in the Attachment to this report.

This inspection constituted two quarterly maintenance effectiveness samples as defined

in IP 71111.12-05.

b. Findings

No findings of significance were identified.

11 Enclosure

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

.1 Maintenance Risk Assessments and Emergent Work Control

a. Inspection Scope

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

maintenance and emergent work activities affecting risk-significant and safety-related

equipment listed below to verify that the appropriate risk assessments were performed

prior to removing equipment for work:

• Work Week 45 - 1A residual heat removal (RHR) room cooler, 1A RHR service

water (RHRSW) loop, 1B RHR seal cooler, 1-1001-16A boroscope and Votes

testing, 1-1001-37A MOV equipment qualification inspection; and

• Work Week 51 - Unit 1 250 Vdc battery reconfiguration using Unit 1 125 Vdc

alternate battery with emergent Unit 2 125 Vdc battery low specific gravity

problems, 2A RHR loop and 2B RHRSW pump unavailability.

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

Reactor Safety Cornerstones. As applicable for each activity, the inspectors verified that

risk assessments were performed as required by 10 CFR 50.65(a)(4) and were accurate

and complete. When emergent work was performed, the inspectors verified that the

plant risk was promptly reassessed and managed. The inspectors reviewed the scope

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

probabilistic risk analyst 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 maintenance risk assessments and emergent work control activities constituted

two samples as defined in IP 71111.13-05.

b. Findings

No findings of significance were identified.

1R15 Operability Evaluations (71111.15)

.1 Operability Evaluations

a. Inspection Scope

The inspectors reviewed the following issues:

• IR 987904: 1A RHR Room Cooler Tube Sheet Has Pitting, and

• IR 994823: TS SR 3.8.4.8 Frequency Not Met.

The inspectors selected these potential operability issues based on the risk significance

of the associated components and systems. The inspectors evaluated the technical

adequacy of the evaluations to ensure that TS operability was properly justified and the

subject component or system remained available such that no unrecognized increase in

12 Enclosure

risk occurred. The inspectors compared the operability and design criteria in the

appropriate sections of the TS and UFSAR to the licensees evaluations to determine

whether the components or systems were operable. Where compensatory measures

were required to maintain operability, the inspectors determined whether the measures

in place would function as intended and were properly controlled. The inspectors

determined, where appropriate, compliance with bounding limitations associated with the

evaluations. Additionally, the inspectors also reviewed a sampling of corrective action

documents to verify that the licensee was identifying and correcting any deficiencies

associated with operability evaluations. Documents reviewed are listed in the

Attachment to this report.

This operability inspection constituted two samples as defined in IP 71111.15-05.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19)

.1 Post-Maintenance Testing

a. Inspection Scope

The inspectors reviewed the following post-maintenance activities to verify that

procedures and test activities were adequate to ensure system operability and functional

capability:

• WO 1121775, 250 Vdc Battery Charger #2 4-Hour Load Test;

• WO 1261246, Replace Battery Changeover Relay R12 EC 376690;

• QCMMS 4100-33, 1/2-4101B Diesel Driven Fire Pump Annual Capacity Test;

• WO 1130535, OP PMT Filter B Train Control Room HVAC; and

• WO1107582, Replace Unit 2 DGCWP Alternate Feed Contactor.

These activities were selected based upon the structure, system, or component's ability

to impact risk. The inspectors evaluated these activities for the following (as applicable):

the effect of testing on the plant had been adequately addressed; testing was adequate

for the maintenance performed; acceptance criteria were clear and demonstrated

operational readiness; test instrumentation was appropriate; tests were performed as

written in accordance with properly reviewed and approved procedures; equipment was

returned to its operational status following testing (temporary modifications or jumpers

required for test performance were properly removed after test completion); and test

documentation was properly evaluated. The inspectors evaluated the activities against

TS, the UFSAR, 10 CFR Part 50 requirements, licensee procedures, and various

NRC generic communications to ensure that the test results adequately ensured that the

equipment met the licensing basis and design requirements. In addition, the inspectors

reviewed corrective action documents associated with post-maintenance tests to

determine whether the licensee was identifying problems and entering them in the CAP

and that the problems were being corrected commensurate with their importance to

safety. Documents reviewed are listed in the Attachment to this report.

13 Enclosure

This inspection constituted five post-maintenance testing samples as defined in

IP 71111.19-05.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22)

.1 Surveillance Testing

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:

• QCOS 1400-01, 2A Core Spray Pump Performance Test (IST);

• QCIS 0300-02, Unit 1 Division 1 Scram Discharge Volume Calibration and

Functional Test (Routine);

• QCOS 7500-05, 1/2 B Standby Gas Treatment Operability Test (Routine);

• QCOS 1600-07, Reactor Coolant Leakage in the Drywell (RCS);

• QCEMS 0230-11, Modified Performance Test of Unit 1(2) 125 Vdc Normal or

Alternate Battery (Routine); and

• QCOS 6900-14, Station Battery Allowable Value Verification Surveillance

(Routine).

The inspectors observed in plant activities and reviewed procedures and associated

records to determine the following:

• did preconditioning occur;

• were the effects of the testing adequately addressed by control room personnel

or engineers prior to the commencement of the testing;

• were acceptance criteria clearly stated, demonstrated operational readiness, and

consistent with the system design basis;

• plant equipment calibration was correct, accurate, and properly documented;

• as-left setpoints were within required ranges, and the calibration frequency were

in accordance with TS, the UFSAR, procedures, and applicable commitments;

• measuring and test equipment calibration was current;

• test equipment was used within the required range and accuracy, applicable

prerequisites described in the test procedures were satisfied;

• test frequencies met TS requirements to demonstrate operability and reliability;

tests were performed in accordance with the test procedures and other

applicable procedures, jumpers and lifted leads were controlled and restored

where used;

• test data and results were accurate, complete, within limits, and valid;

• test equipment was removed after testing;

• where applicable for inservice testing activities, testing was performed in

accordance with the applicable version of Section XI, American Society of

14 Enclosure

Mechanical Engineers code, and reference values were consistent with the

system design basis;

• where applicable, test results not meeting acceptance criteria were addressed

with an adequate operability evaluation or the system or component was

declared inoperable;

• where applicable for safety-related instrument control surveillance tests,

reference setting data were accurately incorporated in the test procedure;

• where applicable, actual conditions encountering high resistance electrical

contacts were such that the intended safety function could still be accomplished;

• prior procedure changes had not provided an opportunity to identify problems

encountered during the performance of the surveillance or calibration test;

• equipment was returned to a position or status required to support the

performance of its safety functions; and

• all problems identified during the testing were appropriately documented and

dispositioned in the CAP.

Documents reviewed are listed in the Attachment to this report.

This inspection constituted four routine surveillance testing samples, one inservice

testing sample, and one reactor coolant system leak detection inspection samples as

defined in IP 71111.22, Sections -02 and -05.

b. Findings

No findings of significance were identified.

1EP4 Emergency Action Level and Emergency Plan Changes (71114.04)

.1 Emergency Action Level and Emergency Plan Changes

a. Inspection Scope

Since the last NRC inspection of this program area, Emergency Plan Annex,

Revisions 26 and 27 were implemented based on the licensees determination, in

accordance with 10 CFR 50.54(q), that the changes resulted in no decrease in

effectiveness of the Plan, and that the revised Plan as changed continues to meet the

requirements of 10 CFR 50.47(b) and Appendix E to 10 CFR Part 50. The inspectors

conducted a sampling review of the Emergency Plan changes and a review of the

Emergency Action Level (EAL) changes to evaluate for potential decreases in

effectiveness of the Plan. However, this review does not constitute formal NRC approval

of the changes. Therefore, these changes remain subject to future NRC inspection in

their entirety.

This emergency action level and emergency plan changes inspection constituted one

sample as defined in IP 71114.04-05.

15 Enclosure

b. Findings

(1) Unresolved Item (URI) 05000254/2009005-01: Changes to EAL HU6 Potentially

Decrease the Effectiveness of the Plans without Prior NRC Approval

Introduction: The inspectors reviewed changes implemented to the Quad Cities Station

Radiological Emergency Plan Annex EALs and EAL Basis. In Revision 26, the licensee

changed the basis of EAL HU6, "Fire not extinguished within 15 minutes of detection

within the protected area boundary, by adding two statements. The two changes added

to the EAL basis stated that if the alarm could not be verified by redundant control room

or nearby fire panel indications, notification from the field that a fire exists starts the

15-minute classification and fire extinguishment clocks. The second change stated the

15-minute period to extinguish the fire does not start until either the fire alarm is verified

to be valid by additional control room or nearby fire panel instrumentation, or upon

notification of a fire from the field. These statements conflict with the previous

Quad Cities Station Annex, Revision 25, basis statements and potentially decrease the

effectiveness of the Plans.

Description: Quad Cities Station Radiological Emergency Plan Annex, Revision 25,

EAL HU6, initiating condition stated, "Fire not extinguished within 15 minutes of

detection, or explosion, within the protected area boundary." The threshold values for

HU6 were, in part: 1) Fire in any Table H2 area not extinguished within 15 minutes of

control room notification or verification of a control room alarm; or 2) Fire outside any

Table H2 area with the potential to damage safety systems in any Table H2 area not

extinguished within 15 minutes of control room notification or verification of a control

room alarm. Table H2, Vital Areas, were identified as main control room, reactor

building, diesel generator rooms, 4 kilovolt switchgear area, battery rooms, B train

control room heating-ventilation and air conditioning, service water pumps, and turbine

building cable tunnel. The basis defined fire as "combustion characterized by heat and

light. Sources of smoke such as slipping drive belts or overheated electrical equipment

do not constitute fires. Observation of flame is preferred but is not required if large

quantities of smoke and heat are observed."

The basis for Revision 25, EAL HU6 thresholds 1 and 2 stated, in part, the purpose of

this threshold is to address the magnitude and extent of fires that may be potentially

significant precursors to damage to safety systems. As used here, notification is visual

observation and report by plant personnel or sensor alarm indication. The 15-minute

period begins with a credible notification that a fire is occurring or indication of a valid fire

detection system alarm. A verified alarm is assumed to be an indication of a fire unless

personnel dispatched to the scene disprove the alarm within the 15-minute period.

The report, however, shall not be required to verify the alarm. The intent of the

15-minute period is to size the fire and discriminate against small fires that are readily

extinguished (e.g., smoldering waste paper basket, etc.).

Revision 26 of the Quad Cities Station Radiological Emergency Plan Annex, changed

the threshold basis for EAL HU6 by adding the following two statements: 1)"If the alarm

cannot be verified by redundant control room or nearby fire panel indications, notification

from the field that a fire exists starts the 15-minute classification and fire extinguishment

clocks," and 2) "The 15-minute period to extinguish the fire does not start until either the

fire alarm is verified to be valid by utilization of additional control room or nearby fire

panel instrumentation, or upon notification of a fire from the field."

16 Enclosure

The two statements added to the basis in Revision 26 conflict with the Revision 25

threshold basis and initiating condition. The changed threshold basis in Revision 26

could add an indeterminate amount of time to declaring an actual emergency until a

person responded to the area of the fire and made a notification to the control room of a

fire in the event that redundant control room or nearby fire panel indications were not

available.

Pending further review and verification by the NRC to determine if the changes to EAL

HU6 threshold basis potentially decreased the effectiveness of the Plans, this issue was

considered an unresolved item (URI 05000254/2009005-01; 05000265/2009005-01).

1EP6 Drill Evaluation (71114.06)

.1 Emergency Preparedness Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of an after-hours licensee emergency drill on

November 11, 2009, to identify any weaknesses and deficiencies in classification,

notification, and protective action recommendation development activities. The

after-hours drill was preceded by an unannounced, after-hours drive-in drill.

The inspectors observed emergency response operations in the Technical Support

Center to determine whether the event classification, notifications, and protective action

recommendations were performed in accordance with procedures. The inspectors also

attended the licensee drill critique to compare any inspector-observed weakness with

those identified by the licensee staff in order to evaluate the critique and to verify

whether the licensee staff was properly identifying weaknesses and entering them into

the corrective action program. As part of the inspection, the inspectors reviewed the drill

package and other documents listed in the Attachment to this report.

This emergency preparedness drill inspection constituted one sample as defined in

IP 71114.06-05.

b. Findings

No findings of significance were identified.

.2 Emergency Preparedness Termination and Recovery Drill Observation

a. Inspection Scope

The inspectors evaluated the conduct of an emergency preparedness termination and

recovery drill on December 2, 2009, to identify any weaknesses and deficiencies in the

conduct of the drill and to assess the licensees ability to assess performance via a

formal critique process in order to identify and correct Emergency Preparedness

weaknesses. The inspectors observed emergency response operations in the Technical

Support Center to determine whether the recovery and termination activities associated

with the drill were performed in accordance with procedures. The inspectors also

attended the licensee drill critique to compare any inspector-observed weakness with

those identified by the licensee staff in order to evaluate the critique and to verify

whether the licensee staff was properly identifying weaknesses and entering them into

17 Enclosure

the corrective action program. As part of the inspection, the inspectors reviewed the drill

package and other documents listed in the Attachment to this report.

This emergency preparedness drill inspection constituted one sample as defined in

IP 71114.06-05.

b. Findings

No findings of significance were identified.

4. OTHER ACTIVITIES

4OA1 Performance Indicator Verification (71151)

.1 Mitigating Systems Performance Index - Emergency Alternating Current Power System

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance

Index (MSPI) - Emergency Alternating Current (AC) Power System performance

indicator for Quad Cities Units 1 and 2 for the period from the 4th quarter 2008 through

the 3rd quarter 2009. To determine the accuracy of the performance indicator (PI) data

reported during those periods, PI definitions and guidance contained in the Nuclear

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

Guideline, Revision 6, were used. The inspectors reviewed the licensees operator

narrative logs, MSPI derivation reports, issue reports, event reports and NRC integrated

inspection reports for the period of October 1, 2008, through September 30, 2009, to

validate the accuracy of the submittals. The inspectors reviewed the MSPI component

risk coefficient to determine if it had changed by more than 25 percent in value since the

previous inspection, and if so, that the change was in accordance with applicable

guidance. The inspectors also reviewed the licensees issue report database to

determine if any problems had been identified with the PI data collected or transmitted

for this indicator, and none were identified. Documents reviewed are listed in the

Attachment to this report.

This inspection constituted two MSPI emergency AC power system samples as defined

in IP 71151-05.

b. Findings

No findings of significance were identified.

.2 Mitigating Systems Performance Index - High Pressure Injection Systems

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance

Index - High Pressure Injection Systems performance indicator for Quad Cities Units 1

and 2 for the period from the 4th quarter 2008 through the 3rd quarter 2009. To

determine the accuracy of the PI data reported during those periods, PI definitions and

guidance contained in the NEI Document 99-02, Regulatory Assessment Performance

18 Enclosure

Indicator Guideline, Revision 6, were used. The inspectors reviewed the licensees

operator narrative logs, issue reports, MSPI derivation reports, event reports and

NRC integrated inspection reports for the period of October 1, 2008, through

September 30, 2009, to validate the accuracy of the submittals. The inspectors

reviewed the MSPI component risk coefficient to determine if it had changed by more

than 25 percent in value since the previous inspection, and if so, that the change was in

accordance with applicable guidance. The inspectors also reviewed the licensees issue

report database to determine if any problems had been identified with the PI data

collected or transmitted for this indicator, and none were identified. Documents

reviewed are listed in the Attachment to this report.

This inspection constituted two MSPI high pressure injection system samples as defined

in IP 71151-05.

b. Findings

No findings of significance were identified.

.3 Mitigating Systems Performance Index - Heat Removal System

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance

Index - Heat Removal System performance indicator for Quad Cities Units 1 and 2 for

the period from the 4th quarter 2008 through the 3rd quarter 2009. To determine the

accuracy of the PI data reported during those periods, PI definitions and guidance

contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator

Guideline, Revision 6, were used. The inspectors reviewed the licensees operator

narrative logs, issue reports, event reports, MSPI derivation reports, and NRC integrated

inspection reports for the period of October 1, 2008, through September 30, 2009, to

validate the accuracy of the submittals. The inspectors reviewed the MSPI component

risk coefficient to determine if it had changed by more than 25 percent in value since the

previous inspection, and if so, that the change was in accordance with applicable

guidance. The inspectors also reviewed the licensees issue report database to

determine if any problems had been identified with the PI data collected or transmitted

for this indicator, and none were identified. Documents reviewed are listed in the

Attachment to this report.

This inspection constituted two MSPI heat removal system samples as defined in

IP 71151-05.

b. Findings

No findings of significance were identified.

.4 Mitigating Systems Performance Index - Residual Heat Removal System

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance

Index - Residual Heat Removal System performance indicator for Quad Cities Units 1

19 Enclosure

and 2 for the period from the 4th quarter 2008 through the 3rd quarter 2009. To

determine the accuracy of the PI data reported during those periods, the PI definitions

and guidance contained in the NEI Document 99-02, Regulatory Assessment

Performance Indicator Guideline, Revision 6, were used. The inspectors reviewed the

licensees operator narrative logs, issue reports, MSPI derivation reports, event reports

and NRC integrated inspection reports for the period of October 1, 2008, through

September 30, 2009, to validate the accuracy of the submittals. The inspectors

reviewed the MSPI component risk coefficient to determine if it had changed by more

than 25 percent in value since the previous inspection, and if so, that the change was in

accordance with applicable guidance. The inspectors also reviewed the licensees issue

report database to determine if any problems had been identified with the PI data

collected or transmitted for this indicator, and none were identified. Documents

reviewed are listed in the Attachment to this report.

This inspection constituted two MSPI residual heat removal system samples as defined

in IP 71151-05.

b. Findings

No findings of significance were identified.

.5 Mitigating Systems Performance Index - Cooling Water Systems

a. Inspection Scope

The inspectors sampled licensee submittals for the Mitigating Systems Performance

Index - Cooling Water Systems performance indicator for Quad Cities Units 1 and 2 for

the period from the 4th quarter 2008 through the 3rd quarter 2009. To determine the

accuracy of the PI data reported during those periods, PI definitions and guidance

contained in the NEI Document 99-02, Regulatory Assessment Performance Indicator

Guideline, Revision 6, were used. The inspectors reviewed the licensees operator

narrative logs, issue reports, MSPI derivation reports, event reports and NRC integrated

inspection reports for the period of October 1, 2008, through September 30, 2009, to

validate the accuracy of the submittals. The inspectors reviewed the MSPI component

risk coefficient to determine if it had changed by more than 25 percent in value since the

previous inspection, and if so, that the change was in accordance with applicable

guidance. The inspectors also reviewed the licensees issue report database to

determine if any problems had been identified with the PI data collected or transmitted

for this indicator, and none were identified. Documents reviewed are listed in the

Attachment to this report.

This inspection constituted two MSPI cooling water system samples as defined in

IP 71151-05.

b. Findings

No findings of significance were identified.

20 Enclosure

.6 Radiological Effluent Technical Specification/Offsite Dose Calculation Manual

Radiological Effluent Occurrences

a. Inspection Scope

The inspectors sampled licensee submittals for the Radiological Effluent Technical

Specifications (RETS)/Offsite Dose Calculation Manual (ODCM) Radiological Effluent

Occurrences performance indicator for the period of December 2008 through

November 2009. The inspectors used PI definitions and guidance contained in the

NEI Document 99-02, Regulatory Assessment Performance Indicator Guideline,

Revision 6 to determine the accuracy of the PI data reported during those periods.

The inspectors reviewed the licensees issue report database and selected individual

reports generated since this indicator was last reviewed to identify any potential

occurrences such as unmonitored, uncontrolled, or improperly calculated effluent

releases that may have impacted offsite dose. The inspectors reviewed gaseous

effluent summary data and the results of associated offsite dose calculations for selected

dates between December 2008 and November 2009 to determine if indicator results

were accurately reported. The inspectors also reviewed the licensees methods for

quantifying gaseous and liquid effluents and determining effluent dose. Documents

reviewed are listed in the Attachment to this report.

This inspection constituted one RETS/ODCM radiological effluent occurrences sample

as defined in IP 71151-05.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152)

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

Preparedness, Public Radiation Safety, Occupational Radiation Safety, and

Physical Protection

.1 Routine Review of Items Entered into the Corrective Action Program (CAP)

a. Inspection Scope

As part of the various baseline inspection procedures discussed in previous sections of

this report, the inspectors routinely reviewed issues during baseline inspection activities

and plant status reviews to verify that they were being entered into the licensees CAP at

an appropriate threshold, that adequate attention was being given to timely corrective

actions, and that adverse trends were identified and addressed. Attributes reviewed

included: the complete and accurate identification of the problem; that timeliness was

commensurate with the safety significance; that evaluation and disposition of

performance issues, generic implications, common causes, contributing factors, root

causes, extent of condition reviews, and previous occurrences reviews were proper and

adequate; and that the classification, prioritization, focus, and timeliness of corrective

actions were commensurate with safety and sufficient to prevent recurrence of the issue.

Minor issues entered into the licensees CAP as a result of the inspectors observations

are included in the attached List of Documents Reviewed.

21 Enclosure

These routine reviews for the identification and resolution of problems did not constitute

any additional inspection samples. Instead, by procedure they were considered an

integral part of the inspections performed during the quarter and documented in

Section 1 of this report.

b. Findings

No findings of significance were identified.

.2 Daily Corrective Action Program Reviews

a. Inspection Scope

In order to assist with the identification of repetitive equipment failures and specific

human performance issues for followup, the inspectors performed a daily screening of

items entered into the licensees CAP. This review was accomplished through

inspection of the stations daily condition report packages.

These daily reviews were performed by procedure as part of the inspectors daily plant

status monitoring activities and, as such, did not constitute any separate inspection

samples.

b. Findings

No findings of significance were identified.

.3 Semi-Annual Trend Review

a. Inspection Scope

The inspectors performed a review of the licensees CAP and associated documents to

identify trends that could indicate the existence of a more significant safety issue. The

inspectors review was focused on repetitive equipment issues and associated corrective

actions, but also considered the results of daily inspector CAP item screening discussed

in Section 4OA2.2 above, licensee trending efforts, and licensee human performance

results. The inspectors review nominally considered the 6-month period of

January 1, 2009, through June 30, 2009, although some examples expanded beyond

those dates where the scope of the trend warranted.

The review also included issues documented outside the normal CAP in major

equipment problem lists, repetitive and/or rework maintenance lists, departmental

problem/challenges lists, system health reports, quality assurance audit/surveillance

reports, self assessment reports, and Maintenance Rule assessments. The inspectors

compared and contrasted their results with the results contained in the licensees

CAP trending reports. Corrective actions associated with a sample of the issues

identified in the licensees trending reports were reviewed for adequacy. Additionally,

the inspectors reviewed CAP open priority 1, 2, and 3 corrective actions for timeliness.

In addition, all open priority 4 action tracking items (ACITs) were reviewed to ensure they

were properly categorized and that the justifications for extension were appropriate and

properly documented.

22 Enclosure

This review constituted a single semi-annual trend inspection sample as defined in

IP 71152-05.

b. Findings

No findings of significance were identified.

.4 Selected Issue Followup Inspection: Issue Report 966501, Darley Pump Leaking

Gasoline from the Fuel Pump

a. Inspection Scope

During a review of items entered in the licensees CAP, the inspectors followed up on a

corrective action item documenting gasoline leaking from the fuel pump of the portable

emergency flooding pump (Darley pump) on September 17, 2009, during preparations

for a pump capacity demonstration run. The pump capacity demonstration was a new

procedure developed in response to a non-cited violation (NCV) documented in

Inspection Report 05000254/2007005.

This review constituted one in-depth problem identification and resolution sample as

defined in IP 71152-05.

b. Findings

Introduction: A finding of very low safety significance was self-revealed for the failure to

maintain the portable emergency flooding pump and supporting equipment in a condition

required to support implementation of QCOA 0010-16, Flood Emergency Procedure.

Description: In Inspection Report 05000254/2007005, inspectors documented a NCV of

TS 5.4.1 for the licensees failure to develop adequate surveillance procedures for

equipment used during an external flooding event. Corrective action for this issue

included revising the external flooding procedure and developing and implementing a

procedure to test a portable pump used as the sole source of makeup water to the spent

fuel pool following an external flooding event. The action to develop and implement the

pump test procedure was issued in May and stated, Develop test procedure and

conduct test to confirm flow of greater than or equal to 200 gpm by mid-July. Brief

NRC Resident as appropriate. The action was closed to an Engineering Change (EC) 366481, on July 18, 2007, with no actual test performed. The documented justification

for this closure stated that discussions with the NRC resident clarified the intent of the

action and no physical testing needed to be performed. Followup discussions with the

resident inspectors stationed at Quad Cities in July 2007 had no recollection of the

conversation and their understanding of the intended action remained unchanged from

the original report.

Licensee staff generated Issue Report (IR) 738335 in February 2008 to document the

review of the NCV response and generate a closure package of all related IRs. The lack

of preventative maintenance (PM) testing was identified and an action tracking item was

generated to Develop PM/testing requirements for the Darley pump associated with the

external flooding event. The original corrective action due date was July 16, 2008.

The action was extended several times, and on May 18, 2009, during a review of

corrective actions for NRC-identified issues, the licensee staff identified that a CAP

23 Enclosure

action item (ACIT 624645-03) had been inappropriately closed. In addition, the licensee

determined that ACIT 624645-03 was inappropriately tagged as an Action Tracking Item

(ACIT) and should have been a corrective action. Issue Report 921197 was generated

and ACIT 624645 was upgraded to a corrective action with a July 31, 2009, due date.

The test procedure was developed and the pump was scheduled to run on

September 17, 2009.

The capacity test was implemented with WO 01247374. When mechanics pulled the

pump and support components from the storage location, they found that the engine

battery had to be replaced and the gasoline stored with the motor had to be replaced.

Since the mechanics performing the test had never operated the pump, they decided to

run it in the weld shop before taking it down to the river. When the mechanics started

the pump, fuel was spraying out of the fuel pump. They immediately shut down the

pump and contained the fuel leak (IR 966501).

The Darley pump fuel system was repaired and the capacity test was completed

satisfactorily on September 25, 2009. Review of recent pump operating history and

PM tasks revealed that the pump had not been operated since the NCV was identified in

2007. The annual maintenance performed under PM 164250 in July of 2009 changed

the oil and inspected the filters and spark plugs with no post-maintenance operation

required. The PM also failed to identify that the battery was beyond the expected life

and did not determine that the battery would maintain its charge.

Analysis: The inspectors determined that the failure to perform maintenance that would

ensure the pump was in a standby condition and readily available to accomplish the

requirements of QCOA 0010-16 was a performance deficiency fully within the licensees

ability to control, and therefore a finding. This issue was more than minor because it

was associated with the SSC Performance attribute of the Barrier Integrity Cornerstone

element of maintaining the functionality of spent fuel pool cooling. The finding affected

the cornerstone objective of providing assurance that physical design barriers protect the

public from radionuclide releases caused by events including external flooding.

Specifically, the pump could fail due to a maintenance preventable component failure

resulting in inadequate or degraded makeup to the spent fuel pool during an external

flooding event. The inspectors did not identify a cross-cutting aspect associated with

this finding because the maintenance issue is a legacy issue and not reflective of current

licensee performance. The pump and PM tasks had been in place for several years.

Inspectors reviewed maintenance requirements for other temporary equipment staged in

support of external events and emergency operating procedures, some of which was put

in place after the Darley pump was staged, and did not identify any similar issues.

The inspectors determined the finding could be evaluated using the SDP in accordance

with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 -

Initial Screening and Characterization of findings, Tables 4a and 4b. The inspectors

determined that even though this equipment is assumed to completely fail, the licensee

could provide an alternate portable pump already located on site and capable of

performing the safety function during this slow developing event. The alternate pump

had maintenance and test procedures in place to provide a basis for reliability. Since

alternate equipment was available and the delay in mobilizing the alternate equipment

would not have resulted in loss of capability to mitigate the impact of the flooding event,

the issue is of very low safety significance or Green.

24 Enclosure

Enforcement: Technical Specification 5.4.1 required that written procedures be

established, implemented, and maintained for the items specified in Regulatory

Guide 1.33, Quality Assurance Program Requirements. QCOA 0010-16,

Flood Emergency Procedure, was the licensee procedure used to meet the

Regulatory Guide 1.33 requirement for an emergency flooding event. The procedure

specified that the portable pump staged in the protected area warehouse is to be used to

respond to the event. Although the regulatory guide did not specifically require

maintenance procedures for portable equipment, failure to maintain the staged

equipment in a condition to be used to mitigate the event does not support timely

implementation of the procedure to provide spent fuel pool makeup and is a finding.

Enforcement action does not apply because the performance deficiency did not involve a

violation of a regulatory requirement. Because the finding does not involve a violation of

regulatory requirements and has a very low safety significance, it is identified as

(FIN 05000254/2009005-02; 05000265/2009005-02). The issue was added to the

licensees CAP program as IR 966501 and IR 968809. Immediate corrective actions

included replacement of the degraded battery and overhaul of the pumps fuel pump.

Other actions included identification of preventative maintenance tasks and establishing

a program owner of the pump.

.5 Selected Issue Followup Inspection: Incident Report 984769, Temperature Indicating

Probe Found Broken in the Unit 2 Diesel Generator Coolant System

a. Inspection Scope

During a review of items entered in the licensees CAP, the inspectors followed up on a

corrective action item documenting a failed temperature indicating probe (TI) in the

Unit 2 diesel generator coolant system on October 27, 2009, during planned

maintenance on the Unit 2 emergency diesel generator (EDG).

This review constituted one in-depth problem identification and resolution sample as

defined in IP 71152-05.

b. Findings

Introduction: A finding of very low safety significance and associated NCV were

self-revealed when a TI failed in the Unit 2 diesel generator coolant system.

Description: On October 27, 2009, while performing corrective maintenance on

TI 2-6641-8205, technicians noted that the tip had broken off the probe when comparing

it to the length of the new TI. This TI provides local indication of the jacket coolant water

temperature at the inlet to the diesel engine and provides no alarm function.

The TI was scheduled for replacement in October 2008 when Operations identified the

TI reading abnormally at zero degrees. A work order was written and scheduled for

October 2009. During the performance of the maintenance, it was noted that the new TI

was longer than the one recently removed. A new work order was written to retrieve any

foreign material from the system. The broken tip was recovered from the diesel

generator coolant system.

25 Enclosure

The licensee investigation discovered that the installation analysis for this TI was

approved under the non-safety below level of design detail (NSBLD) process in October

2007 under Revision 3 of SM-AA-300, Procurement Engineering Support Activities.

Using this provision, NSBLD changes must be documented and shall identify the

change with justification of the changes technical acceptability. The length of the probe

was the only difference to the previously installed TI. The TI was installed with a

3.25 inch probe, which was longer than the previous 2 inch probe. The added length

increased the shear force from the coolant flow and caused the probe to break off.

An operating experience (OPEX) review would have revealed an event at another

nuclear facility where the same make and model TI experienced the same failure

mechanism in a diesel generator coolant system. Under Revision 3 of SM-AA-300,

OPEX reviews for NSBLD were not required, nor were additional peer reviews required.

The lack of an OPEX review was an identified vulnerability by the licensees corporate

supply organization in a common cause analysis which was performed for a lack of

technical rigor issued in February 2008. A corrective action from this common cause

analysis was to implement Revision 4 of SM-AA-300 which limited NSBLD reviews to

non-safety host component applications. Revision 4 was implemented at Quad Cities in

February 2008. Since this specific TI is classified as augmented quality, Revision 4

would prevent use of the NSBLD process of a non-identical replacement. A full item

equivalency evaluation would be required for any non-identical replacement.

An extent of condition review is scheduled to be performed at Quad Cities by

Procurement Engineering for all NSBLD reviews that were performed under Revision 3

of SM-AA-300 from August 2007 through February 2008.

Analysis: The inspectors determined that the approval of an inappropriate component

designated as augmented quality was a performance deficiency and a finding. The

same parts evaluation process was used for risk-significant components independent of

the system being worked. Therefore, this finding was more than minor because, if left

uncorrected, this performance deficiency could lead to unplanned unavailability of

safety-related or risk-significant equipment and would become a more significant safety

concern. This performance deficiency challenged the Mitigating Systems Cornerstone

attribute of Equipment Performance by challenging equipment availability and reliability.

The inspectors performed a Phase 1 SDP screening and concluded that the issue was

of very low safety significance (Green) because the failure of the TI did not result in

unplanned inoperability or loss of function of the diesel generator. The inspectors

determined that this finding did not have a cross-cutting aspect. This performance

deficiency is not indicative of current licensee performance. The decision to install this

type of TI was made in October 2007. The process which allowed this performance

deficiency was identified and corrected through procedure and policy revisions to

SM-AA-300 in February 2008.

Enforcement: The TI was designated augmented quality in the licensees quality

assurance program. The licensees quality assurance program applied controls

equivalent to safety-related components for Class 1E equipment qualification to

augmented quality equipment and systems. This correlation is applicable to several

Appendix B criteria included in the program such as both Section 3 - Design Control,

and Section 5 - Instructions Procedures and Drawings, of the licensees Quality

Assurance program for augmented quality.

26 Enclosure

Title 10 CFR 50, Appendix B, Criterion V states in part that activities affecting quality

shall be prescribed by instructions and procedures of a type appropriate to the

circumstances and shall be accomplished in accordance with these instructions or

procedures.

Contrary to the above, on October 30, 2007, SM-AA-300 was not appropriate to the

circumstances in that it did not require an approval process with technical rigor

equivalent to the process used for safety-related components when a non-identical

temperature indicating probe designated augmented quality was approved for use.

That part was approved for use through a NSBLD review per Revision 3 of SM-AA-300

instead of undergoing a full item equivalency evaluation, and the part subsequently

failed resulting in foreign material in the diesel generator coolant system. The foreign

material did not cause any adverse consequences in this instance.

Because this issue is of very low safety significance, and this issue has been entered

into the licensees corrective action program as Issue Report 984769, this issue is being

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

(NCV 05000265/2009005-03).

Corrective actions for this event included replacement of the TI with an appropriately

approved TI. The licensee has also scheduled to perform an extent of condition review

of NSBLD reviews performed under Revision 3 of SM-AA-300 from August 2007 through

February 2008.

4OA3 Follow-Up of Events and Notices of Enforcement Discretion (71153)

.1 (Closed) Licensee Event Report 05000254/2009-003-00: Failure of RHR Torus Spray

Isolation Valve to Open Due to Declutch Mechanism Problems

a. Inspection Scope

Inspectors reviewed the event, evaluation, and corrective actions for the motor operated

valve failure reported in Licensee Event Report (LER) 05000254/2009-003. Documents

reviewed as part of this inspection are listed in the Attachment to this report. This LER is

closed.

This event follow-up review constituted one sample as defined in IP 71153-05.

b. Findings

Introduction: A finding of very low safety significance and an NCV of Technical

Specification (TS) 3.6.2.4, Residual Heat Removal (RHR) Suppression Pool Spray,

was self-revealed for the licensees failure to meet the TS limiting condition for operation

(LCO) requirements prior to transitioning into an operating mode where the LCO was

required to be satisfied. Specifically, MO 1-1001-37B, motor operator for the Unit 1 RHR

torus (suppression pool) spray isolation valve, was found to have been inoperable when

the operating crew transitioned Unit 1 from Mode 4 to Mode 2 on May 30, 2009. The

valve actuator had been inadvertently declutched (i.e., motor disengaged) and the valve

was not demonstrated operable by stroking the valve electrically after the actuator motor

was declutched.

27 Enclosure

Discussion: On June 4, 2009, with Unit 1 in Mode 1 at 100 percent power following

startup from a forced outage, MO 1-1001-37B, torus spray shutoff valve, was determined

to be inoperable because it would not open remotely using the control switch during

performance of the residual heat removal power operated valve test surveillance.

The torus spray valve had been closed using the motor and a clearance order had been

placed on the valve during the outage. Another motor operated valve in the residual

heat removal system on that same clearance, MO 1-1001-7C, RHR C torus suction line

isolation valve, had failed to open on May 28, 2009, when the clearance tag was

removed and valve stroking was being performed to restore the component to a standby

configuration. Operators reported manually declutching (disengaging the actuator

motor) the 7C valve while placing the clearance tag in order to verify the valve was

closed. Inspectors identified that the action of manually verifying valve position was not

a normal practice as supported by OP-AA-103-105, Limitorque Motor-Operated Valve

Operations, and Operations department management. Investigation into the 7C failure

revealed that the actuator lubricant was degraded in the area of the clutch return spring

preventing the motor from engaging when called upon from the control circuit. The

RHR C valve actuator was rebuilt using MOV Long Life grease, new tripper cams, new

trip lever assembly, and a new outer declutch arm snap ring. The rebuilt actuator was

verified to operate correctly in all modes and returned to service prior to unit restart on

May 30, 2009.

Inspectors interviewed operating personnel regarding the positioning of MO 1-1001-37B

torus spray valve. Operators stated that they did not manually declutch the 37B valve

since the valve was already closed (normal position) when they hung the tag. The

licensees investigation attempted to identify both how the motor on the 37B valve was

declutched and why the actuator did not return to the motor mode of operation

automatically as designed. The licensee verified that the actuator was not able to

transition from the motor mode to the manual mode without external (human)

intervention.

Although the licensee could not identify how or when the valve actuator motor was

declutched, the licensees investigators concluded that the declutch lever was most likely

bumped during work activities on top of the Torus during the recent outage with the unit

in Mode 4. Investigation further determined that with the valve motor disengaged,

increased friction in the actuator caused by degraded lubricant in the area of the clutch

return spring prevented the engagement of the motor to open the valve. The actuator

motor was engaged by manually manipulating the declutch lever and stroke testing the

valve.

Inspectors reviewed the grease sampling methodology and the preventative

maintenance frequency for the SMP-00 type actuators and determined that both were

conducted in accordance with the industry standards for these type valves.

Analysis: The failure of plant personnel to demonstrate operability of MO 1-1001-37B by

stroking the valve electrically prior to changing modes was a performance deficiency.

The finding is more than minor because it was associated with the equipment

performance quality attribute of the Mitigating Systems Cornerstone and affected the

objective of ensuring availability, reliability and capability of systems that respond to

initiating events to prevent undesirable consequences. Specifically, failure to verify

system availability and capability prior to entering the required modes resulted in fewer

28 Enclosure

available mitigating systems than assumed in the operating risk evaluations. Inspectors

determined that the finding was cross-cutting in the area of Problem Identification and

Resolution - Corrective Action because plant personnel failed to identify the valve

actuator contact that resulted in the valve being declutched; therefore, operators

incorrectly assessed the system condition as in compliance with TS 3.6.2.4 (P.1(a)).

The inspectors determined the finding could be evaluated using the SDP in accordance

with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 -

Initial Screening and Characterization of Findings, Table 4a. Inspectors answered all of

the questions for the Mitigating Systems Cornerstone No. Therefore, the finding

screened as Green or very low safety significance.

Enforcement: Technical Specification 3.0, Limiting Condition for Operation (LCO)

Applicability, LCO 3.0.4 stated in part that when an LCO is not met, entry into a mode in

the Applicability shall only be made:

• when the associated actions to be entered permit continued operation while in

the mode or other specified condition in the Applicability for an unlimited time;

• after performance of a risk assessment addressing inoperable systems and

components, and acceptability of entering the mode; or

• when an allowance is stated in the specification.

Technical Specification 3.6.2.4, Residual Heat Removal (RHR) Suppression Pool

Spray, required two RHR suppression pool spray subsystems to be operable in

Modes 1, 2 and 3.

Contrary to the above, on May 30, 2009, the licensee changed operating modes from

Mode 4 to Mode 2 with the MO 1-1001-37B valve inoperable in violation of TS 3.6.2.4

LCO conditions since only one RHR suppression pool (Torus) spray subsystem was

operable. Specifically, TS 3.6.2.4 had no allowance provided to permit mode change

with less than two subsystems operable, no prior risk assessment was performed, and

the specification did not permit operation for an unlimited time, the mode change

resulted in non-compliance with TS LCO 3.6.2.4.

Because this finding is of very low safety significance, and this issue has been entered

into the licensees corrective action program as IR 928048, this violation is being treated

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

(NCV 05000254/2009005-04).

Immediate corrective actions for this event included engagement of the actuator motor

by manually manipulating the declutch lever and stroke testing the valve. Since the

hardened grease in this area of the actuator assembly was only an issue if the actuator

was manually declutched, the valve was left in standby, and overhaul of the valve

actuator was scheduled for the next refueling outage.

29 Enclosure

4OA5 Other Activities

.1 World Association of Nuclear Operators Plant Assessment Report Review

a. Inspection Scope

The inspectors reviewed the final report for the World Association of Nuclear Operators

plant assessment conducted in February 2009. The inspectors reviewed the report to

ensure that issues identified were consistent with the NRC perspectives of licensee

performance and to verify if any significant safety issues were identified that required

further NRC followup.

b. Findings

No findings of significance were identified.

.2 Quarterly Resident Inspector Observations of Security Personnel and Activities

a. Inspection Scope

During the inspection period, the inspectors conducted observations of security force

personnel and activities to ensure that the activities were consistent with licensee

security procedures and regulatory requirements relating to nuclear plant security.

These observations took place during both normal and off-normal plant working hours.

These quarterly resident inspector observations of security force personnel and activities

did not constitute any additional inspection samples. Rather, they were considered an

integral part of the inspectors' normal plant status review and inspection activities.

b. Findings

No findings of significance were identified.

4OA6 Management Meetings

.1 Exit Meeting Summary

On January 5, 2010, the inspectors presented the inspection results to T. Tulon and

other members of the licensee staff. The licensee acknowledged the issues presented.

The inspectors confirmed that none of the potential report input discussed was

considered proprietary.

.2 Interim Exit Meetings

Interim exits were conducted for:

• The results of the licensed operator requalification training program inspection

and with the site vice president, Mr. T. Tulon, on October 2, 2009.

• The licensed operator requalification training biennial written examination and

annual operating test examination materials were discussed with the training

manager, Mr. K. Moser, on November 12, 2009.

30 Enclosure

• The licensed operator requalification training program annual inspection results

with operations training manager, Mr. D. Snook, on November 20, 2009, via

telephone.

• The results of the Radiological Effluent TS/Offsite Dose Calculation Manual

Radiological Effluent Occurrences performance indicator verification program

inspection with the plant manager, Mr. R. Gideon, on December 16, 2009.

• The annual review of Emergency Action Level and Emergency Plan changes

with the licensee's emergency preparedness coordinator, Mr. F. Swan, via

telephone on December 21, 2009.

The inspectors confirmed that none of the potential report input discussed was

considered proprietary. Proprietary material received during the inspection was returned

to the licensee.

4OA7 Licensee-Identified Violations

The following violation of very low significance (Green) was identified by the licensee

and is a violation of NRC requirements which meets the criteria of Section VI of the

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

• Technical Specification 5.5.1 requires implementation of the Offsite Dose

Calculation Manual. Offsite Dose Calculation Manual, Revision 8, Part 12.2.1,

Radioactive Liquid Effluent Monitoring Instrumentation, Section C requires that

when the service water effluent gross activity monitor is operated with less than

the minimum number of operable channels, the licensee shall collect and analyze

grab samples for beta or gamma activity once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Contrary to the

above, grab samples were not collected while the Unit 1 service water effluent

gross activity monitor was inoperable from June 2-20, 2009. Specifically,

following fuse replacement, the licensee failed to recognize that the instrument

remained uninitialized; therefore, that compensatory samples were required. The

finding was documented in the licensees corrective action program as

IR 933472. Corrective actions included returning the monitor to service and

reviewing captured monitor data from June 2-20, 2009, to ensure that no release

events occurred during the monitor outage, revising the monitor repair and

maintenance procedures to clear direct communication with the Chemistry

Department subject matter experts during work on the system, and reinforcing

the expectation that control room operators turn over all abnormal indications to

supervisors each shift. The finding was determined to be of very low safety

significance because, although the finding related to the effluent release

program, it was not a failure to implement the effluent program or an event that

resulted in a dose to the public in excess of Appendix I criterion or

10 CFR 20.1301(e).

ATTACHMENT: SUPPLEMENTAL INFORMATION

31 Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. Tulon, Site Vice President

R. Gideon, Plant Manager

D. Kimler, Shift Operations Superintendent

S. Darin, Engineering Manager

W. Beck, Regulatory Assurance Manager

J. Burkhead, Nuclear Oversight Manager

J. Garrity, Work Control Manager

K. Moser, Training Manager

V. Neels, Chemistry/Environ/Radwaste Manager

D. Collins, Radiation Protection Manager

D. Thompson, Security Manager

Nuclear Regulatory Commission

M. Ring, Chief, Reactor Projects Branch 1

Illinois Emergency Management Agency

R. Zuffa, Unit Supervisor, Resident Inspector Section

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Opened

05000254/2009005-01; URI Changes to EAL HU6 Potentially Decrease the Effectiveness05000265/2009005-01 of the Plans without Prior NRC Approval

05000254/2009005-02; FIN Darley Pump Leaking Gasoline from the Fuel Pump

05000265/2009005-02

05000265/2009005-03 NCV Temperature Indicating Probe Found Broken in the Unit 2

Diesel Generator Coolant System

05000254/2009005-04 NCV Failure of RHR Torus Spray Isolation Valve to Open Due to

Declutch Mechanism Problems

Closed

05000254/2009005-02; FIN Darley Pump Leaking Gasoline from the Fuel Pump

05000265/2009005-02

05000265/2009005-03 NCV Temperature Indicating Probe Found Broken in the Unit 2

Diesel Generator Coolant System

05000254/2009005-04 NCV Failure of RHR Torus Spray Isolation Valve to Open Due to

Declutch Mechanism Problems05000254/2009003-00 LER Failure of RHR Torus Spray Isolation Valve to Open Due to

Declutch Mechanism Problems

1 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.

Section 1R01

- QCOP 0010-01; Winterizing Checklist; Revision 48

- QCOP 0010-02; Required Cold Weather Routines; Revision 28

- WC-AA-107; Seasonal Readiness; Revision 06

- IR 99493; U-2 FW Heater LCV Response to Lowering Circulating Water Inlet Temp

- WO 1183498; Cycle CW De-Ice Valve

- WO 1282535; Ice Melt Valve Stuck Shut

- QCOP 4400-06; Circulating Water System De-icing; Revision 14

- ECR 59777; Design Alternate Method for Operation of Ice Melt Valve

- IR 993018; Wire Rope Rating on Ice Melt Valve

- IR 986355; Ice Melt Valve Stuck Shut

- WO 01194645; MM Union Leaking Inside U1 Cond Demin Vault (HTG STM)

- WO 01215488; MM Repair Piping Leak Underground Next to Cribhouse

- WO 01242820; MM Seal Cracks in Ceiling Above Bus 23-1

Section 1R04

- QCOP 4100-01; Firewater System Lineup for Standby Operation; Revision 4

- QCOP 6600-01: Diesel Generator 1(2) Preparation For Standby Operation; Revision 38

- WO #01272234; EM Change RMS-9 Setting at SWGR 19 CUB 5D Per EC 377092

- WO #01107582; EM Replace U2 DGCWP Alternate Feed Contactor

- WO #920850; IM CAL DG HX 2-6661B Cooling Water Inlet PI 2-3941-67A

- WO #945963; IM CAL DG HX 2-6661B Cooling Water Outlet PI 2-3941-67B

- WO #01107581; EM Replace U2 DGCWP Normal Feed Contactor

- WO #01245102; EM Support OP QCOS 6600-17 U2 DGCW Pump Alternate Feed Test

- QCOS 6600-17; Operating Cycle Diesel Cooling Water Pump Alternate Power Feed Test for

Appendix R; Revision 15

- EC 360507; Unit 2 EDG Voltage Regulator (Place EDG in Droop Mode Prior to

Synchronization to the Grid)

- EC 377665; TMOD to Bypass Faulty SSES Switch Local/Remote Contact at U2 Diesel

Generator

Section 1R05

- OP-AA-201-008; Pre-fire Plan Manual Index - Pre-Plan RB-16; Revision 2

- Pre-plan TB-74; Fire Zone 9.1, Unit 1 Turbine Bldg. El. 595-0, Diesel Generator; Revision 24

- Pre-plan TB-73; Fire Zone 8.2.6.A, Unit 1 Turbine Bldg. El. 595-0, Reactor Feed Pumps;

Revision 24

- Pre-plan CH-44; Fire Zone 11.4.A, Crib House Bldg. El. 559-8 Basement; Revision 0

- Pre-plan CH-45; Fire Zone 11.4.B, Ground Floor/Service Water Pumps; Revision 22

2 Attachment

Section 1R11

- SY-AA-101-132; Assessment and Response to Suspicious Activity and Security Threats;

Revision 14

- QCOA 0010-20; Security Event; Revision 25

- EP-AA-1006; Radiological Emergency Plan Annex for Quad Cities Station; Revision 27

- Requalification Examination Results/Calendar Year 2009

- Quad Cities, Units 1 and 2 NRC Integrated Inspection Reports; dated various from

January 2007 through September 2009

- OP-AA-105-102; Attachment 1; Active License Tracking Log (for 1st & 2nd Quarters of 2009);

Revision 9

- OP-AA-105-102; Attachment 2, Reactivation of License Log (2 for LSRO, 2 RO); Revision 9

- Quad Cities Classroom Sample Plan for Training Years 2008 and 2009; 6/18/2009

- Quad Cities Simulator Sample Plan for Training Years 2008 and 2009

- 71111.11 Appendix C Responses/Justifications; 9/28/2009

- TQ-AA-224-F070; Evaluation Feedback Summary, LORT Cycle 08-1 through 08-5; LORT

Cycle 09-1 through 09-4

- TQ-AA-1002; Attachment 3; LORT Quarterly Curriculum Review Committee Meeting Minutes;

all of 2008 and first two quarters of 2009

- Special LORT CRC Meeting Minutes; 1/23/2009

- TQ-AA-150; Operator Training Programs; Revision 2

- TQ-AA-150-F07; Simulator Evaluation Form - STA or IA

- TQ-AA-150-F08; Simulator Evaluation Form - Individual

- TQ-AA-150-F09; Simulator Evaluation Form - Crew

- TQ-AA-210-5101; Training Observation Forms; dated various

- TQ-AA-306; Simulator Management

- TQ-AA-306-F06; BWR Critical Condition for Cold Startup; Revision 0

- TQ-AA-306-F07; BWR Power Coefficient of Reactivity and Control Rod Worth; Revision 0

- TQ-AA-306-F08; BWR Xenon Worth; Revision 0

- TQ-AA-306-F06; BWR Site Specific Shutdown Margin and Reactivity Anomaly Tests

- TQ-AA-306-JA-02; Simulator Testing Report Update

- Differences between the Quad Cities Simulator and Quad Cities U-1 & U-2; Revision 14;

7/17/09

- Differences between the Quad Cities Simulator and Quad Cities U-1 & U-2; Revision 15;

9/29/09

- LS-AA-126-1005; Attachment 2; Check-In Self-Assessment Report Template

- LS-AA-126-1001; Attachment 2; FASA Self-Assessment Report

- Simulator Malfunction Test Procedure, Grid Frequency Disturbance (ED16)

- Simulator Malfunction Test Procedure, Reactor Building Instrument Air System (IA02)

- Simulator Malfunction Test Procedure, Main Steam Isolation Valve Closure (MS01)

- Quad Cities Simulator Malfunction Testing Schedule; Revision 8; 5/5/2008

- Simulator Transient Tests; dated various

- Safety System Functional Failure, Rolling Twelve Months Unit 1 and Unit 2; 9/28/09

- Action Request Reports; various dates for LORT 2009

- LORT Attendance Sheets; 2009

3 Attachment

Section 1R12

- Enterprise Maintenance Rule Production Database for the following systems:

• Z2900; Safe Shutdown Makeup Pump

• Z4700; Instrument Air

- System Engineer Notebook and Accountability Logs for the following systems:

• Safe Shutdown Makeup Pump

• Instrument Air

- IR 712670; Safe shutdown makeup pump failed surveillance; 12/17/07

- IR 713041; Broken SSMP part not found during repairs; 12/18/07

- IR 711934; SSMP Suction line did not fill during fill; 12/14/07

- IR 712059; SSMP fails to sustain flow and pressure; 12/15/07

- IR 731013; SSMP Sparking on Startup; 2/4/08

- IR 729984; SSMP failed operability test per TIC-1982; 2/1/08

- IR 729951; SSMP Local FIC failed PMT; 1/31/08

- IR 734472; MRULE A-1 determination for SSMP required; 2/11/08

- IR 741838; SSMP feed breaker problems during system restoration; 2/27/08

- IR 787063; Local SSMP flow controller not reading correctly; 6/16/08

- IR 890904; SSMP controller connector degraded; 3/10/09

- IR 930013; Historical FME identified in SSMP piping inspection; 6/10/09

- IR 956294; SSMP FIC Valve position discrepancy with local valve indication; 8/21/09

- IR 947201; FPI - SSMP Breaker and fuse coordination for CT-2; 7/29/08

- IR 1003024; SSMP Draws a vacuum when starting for PMT; 12/09/09

- IR 1002036; Drain valve for SSMP room cooler may be blocked; 12/06/09

- IR 991490; NCV 09-006-02 Closure package - SSMP Breaker coordination 11/10/09

- IR 673268; 1B Instrument Air Compressor Excessive Leakage; 9/20/07

- IR 762652; 1A Instrument Air Compressor Trip; 04/12/08

- IR 856509; Red Trend Code for 1/2B Instrument Air Compressor - EC 364602

- IR 871161; 1A Instrument Air Compressor Trip; 01/24/09

- IR 871939; 1A Instrument Air Compressor Trip; 01/26/09

- IR 977823; 1A Instrument Air Compressor Tripped Due to Low Oil Pressure; 2/7/09

- IR 936122; Compressor does not auto start; 6/27/09

Section 1R13

- WO #01075655; EM Perform Boroscope INSP of MO 1-1001-16A MOV

- WO #01120751; EM MOV 1-1001-37A MOV EQ Inspection

- WO #01123089; MM Inspect/Clean 1B RHR Pump Seal Cooler

- WO #01131318; EM Votes Test MOV 1-1001-16A

- WO #01190642; MM U-1A RHR HX Room Cooler Air/Water Side Clean Inspect

Section 1R15

- IR 849245; 1B RHR Room Cooler Heat Exchanger Has Tube Sheet Pitting

- WO 862709; 1B RHR Air/Water Side Room CLR CLN/INSP

- IR 987904; 1A RHR Room Cooler Heat Exchanger Tube Sheet Has Pitting

- WO 01190642; MM U-1A RHR HX Room Cooler Air/Water Side Clean Inspect

4 Attachment

- IR 849681; 1B RHR Room Cooler Reassembled at Risk

- EC 373177; Determination of Minimum Wall Thickness of Tubesheet for RHR Room

Cooler 1-574B

- IR 994823; TS SR 3.8.4.8 Frequency Not Met

- QC-SURV-01; Risk Assessment for Missed Surveillance for U2 125 Vdc Battery

Section 1R19

- QCMMS 4100-32; 1/2 -4101A Diesel Driven Fire Pump Annual Capacity Test; Revision 24

- WO 1261246; Replace Battery Changeover Relay R12 EC 376690

- EC 376690; 1/2 A Fire Pump Controller Replace Battery Changeover Relay R12; Revision 1

- QCOS 4100-01; Monthly Diesel Fire Pump Test; Revision 28

- QCOP 4100-03; Diesel Fire Pump Operation; Revision 17

- QCMMS 4100-33; 1/2 - 4101B Diesel Driven Fire Pump Annual Capacity Test; Revision 24

- WO 1121775; 250 Vdc Battery Charger #2 4 Hour Load Test

- WO 1130534; Control RM HVAC Air Filter Unit In Place DOP LK Test

- QCOS 5750-02; Control Room Emergency Filter System Test; Revision 45

- QCIS 5700-04; Main Control Room Air Filter Unit DOP-Freon Test; Revision 0

- QCOS 6600-17; Operating Cycle Diesel Cooling Water Pump Alternate Power Feed Test for

Appendix R; Revision 15

- QCEPM 0400-15; Emergency Diesel Generator Transfer Panel Inspection; Revision 9

- WO 01107582; Replace Unit 2 DGCWP Alternate Feed Contactor

Section 1R22

- QCOS 1400-01; Quarterly Core Spray System Flow Rate Test; Revision 38

- QCOS 1400-07; Core Spray Pump Performance Test; Revision 10

- QCOS 7500-05; Standby Gas Treatment System Monthly Operability Test; Revision 30

- QCIS 0300-02; Unit 1 Division 1 Scram Discharge Volume Rochester Instruments Calibration

and Functional Test; Revision 09

- QCOS 1600-07, Revision 027; Reactor Coolant Leakage in the Drywell

- QCEMS 0230-11; Modified Performance Test of Unit 1(2) 125 Vdc Normal or Alternate

Battery; Revision 0

- QCOS 6900-02; Station Safety Related Battery Quarterly Surveillance; Revision 33

- QCOP 6900-24; Transfer of Unit 2 125 Vdc Battery Bus Between Normal and Alternate

Battery; Revision 12

- QCOS 6900-14; Station Battery Allowable Value Verification Surveillance; Revision 13

Section 1EP4

- Quad Cities Station Radiological Emergency Plan Annex; Revisions 25, 26, and 27

Section 1EP6

- EP-AA-1006; Radiological Emergency Plan Annex for Quad Cities Station; Revision 27

- Quad Cities Generating Station 2009 Termination and Recovery Drill Briefing Package;

December 2, 2009

- EP-AA-115; Termination and Recovery; Revision 7

- EP-AA-111-F-01; Termination/Recovery Checklist; Revision A

5 Attachment

Section 4OA1

- CY-QC-120-724; Continuous Liquid Effluent Analysis; Revision 1

- CY-QC-120, 723; Allocation of Radioactive Liquid Discharges; Revision 0

- CY-QC-120-720; Plant Effluent Dose Calculations; Revision 4

- CY-QC-120-725; Gaseous Release of Tritium Calculation; Revision 1

- Cy-QC-120-726; Fe-55, Sr-89, Sr-90 and Gaseous Alpha Release; Revision 3

- NEI 99-02; Regulatory Assessment Performance Indicator Guideline, Revision 6

- Enterprise Maintenance Rule Production Database for the following systems:

• Z2300; High Pressure Coolant Injection System

• Z1000; Residual Heat Removal System

• Z6600; Diesel Generator System

• Z1300; Reactor Core Isolation Cooling System

• Z9700; 345 kV Switchyard

- System Engineer Notebook and Accountability Logs for the following systems:

• Residual Heat Removal

• RHR Service Water

• Reactor Core Isolation Cooling

• HPCI

• Emergency Diesel Generators

Section 4OA2Q

- IR 984769; Well Broke Off TI in Diesel Generator Coolant System

- WO 1198663; U-2 EDG Eng Temp Indicator TI-2-6641-8205 Not Working

- WO 1280197; Well Broke Off TI In U2 Diesel Generator Coolant System

- SM-AA-300; Procurement Engineering Support Activities; Revision 5

- IR 624645; Flood Emergency Pump Testing Documentation; 05/02/07

- IR 638004; Clarify UFSAR 3.4.1.1 Required Flow Rate to SFP During Flood; 06/07/07

- IR 738335; NCV 07-005-02 GR NCV & X-cutting WRT External Flooding Event; 02/19/08

- IR 921197; Inappropriate ACIT Closure of Darley Pump NCV; 05/18/09

- IR 927463; Request For Darley Pump Testing in On-line Schedule; 06/03/09

- IR 966501; Darley Pump Leaking Gasoline from the Fuel Pump; 09/17/09

- IR 968809; Adequacy of Preventative Maintenance on Darley Pump; 09/22/09

- WO 01247374; Darley Pump Baseline Testing; 9/17/09

- QCOA 0010-16; Flood Emergency Procedure; Revision 12

- QCMMS 1500-12; Portable Emergency Flood Pump Capacity Test; Revision 0

- QCOP 4100-19; Emergency Portable Pump Operations; Revision 7

- PMID/RQ 164250; Perform Maintenance on the External Portable Pump; 09/17/09

Section 4OA3

- 10 Medical Files for Licensed Operators; Various Dates

- Licensee Event Report 254/09-003; Failure of RHR Torus Spray Isolation Valve Due to

Declutch Mechanism Problems; 8/3/09

- IR 928048; MO 1-1001-37B Failed to Open During QCOS 1000-09; 6/4/09

- IR 924666; 1-1001-7C Will Not Open; 5/28/09

- OP-AA-103-105; Limitorque Motor-operated Valve Operations; Revision 1

6 Attachment

Section 4OA7

- AR 933472933472 Service Water Effluent Radiation Monitor Inoperable; 6/20/09

7 Attachment

LIST OF ACRONYMS USED

AC Alternating Current

ADAMS Agencywide Document Access Management System

ACIT Action Tracking Item

CAP Corrective Action Program

CFR Code of Federal Regulations

DGCWP Diesel Generator Cooling Water Pump

EAL Emergency Action Level

EC Engineering Change

EDG Emergency Diesel Generator

IMC Inspection Manual Chapter

IP Inspection Procedure

IR Issue Report

IST Inservice Test

JPM Job Performance Measure

LCO Limiting Condition for Operation

LER Licensee Event Report

LORT Licensed Operator Requalification Training

MO Motor Operator

MOV Motor Operated Valve

MSPI Mitigating System Performance Index

NCV Non-Cited Violation

NEI Nuclear Energy Institute

NRC U.S. Nuclear Regulatory Commission

NSBLD Non-Safety Below Level of Design Detail

OP Operations

OPEX Operating Experience

ODCM Offsite Dose Calculation Manual

PARS Publicly Available Records

PI Performance Indicator

PM Planned or Preventative Maintenance

PMT Post Maintenance Test

RCS Reactor Coolant System

RETS Radiological Effluent Technical Specification

RHR Residual Heat Removal

RHRSW Residual Heat Removal Service Water

SAT Systems Approach to Training

SDP Significance Determination Process

SSC Systems, Structures, and Components

TI Temperature Indicator

TS Technical Specification

UFSAR Updated Final Safety Analysis Report

URI Unresolved Item

Vdc Volt direct current

WO Work Order

8 Attachment

C. Pardee -2-

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

NRCs document system (ADAMS). ADAMS is accessible from the NRC Website at

http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Mark A. Ring, Chief

Branch 1

Division of Reactor Projects

Docket Nos. 50-254; 50-265

License Nos. DPR-29; DPR-30

Enclosure: Inspection Report 05000254/2009005; 05000265/2009005

w/Attachment: Supplemental Information

cc w/encl: Distribution via ListServ

DOCUMENT NAME: G:\1-Secy\1-Work In Progress\QUA 2009005.doc

Publicly Available Non-Publicly Available Sensitive 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 E RIII

NAME MRing:cms

DATE 01/27/2010

OFFICIAL RECORD COPY

Letter to C. Pardee from M. Ring dated January 27, 2010

SUBJECT: QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 INTEGRATED

INSPECTION REPORT 05000254/2009005; 05000265/2009005

DISTRIBUTION:

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