Text

IR 05000461-07-004; on 07/01/2007 - 09/30/2007; Amergen Energy Company, LLC; Clinton Power Station, Integrated Inspection Report and Event Follow-up
	ML073100335
Person / Time
Site:	Clinton
Issue date:	11/06/2007
From:	Ring M; NRC/RGN-III/DRP/B1
To:	Crane C; Exelon Generation Co, Exelon Nuclear
References
	IR-07-004
	Download: ML073100335 (45)
	v • d • e

November 6, 2007

Mr. Christopher M. Crane

President and Chief Nuclear Officer

Exelon Nuclear

Exelon Generation Company, LLC

4300 Winfield Road

Warrenville, IL 60555

SUBJECT: CLINTON POWER STATION

NRC INTEGRATED INSPECTION REPORT 05000461/2007004

Dear Mr. Crane:

On September 30, 2007, the U.S. Nuclear Regulatory Commission (NRC) completed an

integrated inspection at your Clinton Power Station. The enclosed report documents the

inspection results, which were discussed on October 4, 2007, with Mr. F.A. Kearney and other

members of your staff.

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

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

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

personnel.

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

significance (Green) were identified. All of these issues involved violations of NRC

requirements. However, because of the very low safety significance and because they were

entered into your corrective action program, the NRC is treating these violations as Non-Cited

Violations consistent with Section VI.A.1. of the NRC Enforcement Policy.

If you contest the subject or severity of a Non-Cited Violation, you should provide a response

within 30 days of the date of this inspection report, with the basis for your denial, to the U.S.

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

20555-0001; with copies to the Regional Administrator, Region III, 2443 Warrenville Road,

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

Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Clinton

Power Station.

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

enclosure, and your response (if any) will be available electronically for public inspection in the

C. Crane -2-

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

NRCs document system (ADAMS), accessible from the NRC Web site 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 No. 50-461

License No. NPF-62

Enclosure: Inspection Report No. 05000461/2007004

w/Attachment: Supplemental Information

cc w/encl: Site Vice President - Clinton Power Station

Plant Manager - Clinton Power Station

Regulatory Assurance Manager - Clinton Power Station

Chief Operating Officer

Senior Vice President - Nuclear Services

Vice President - Operations Support

Vice President - Licensing and Regulatory Affairs

Manager Licensing - Clinton Power Station

Senior Counsel, Nuclear, Mid-West Regional Operating Group

Document Control Desk - Licensing

Assistant Attorney General

Illinois Emergency Management Agency

State Liaison Officer, State of Illinois

Chairman, Illinois Commerce Commission

C. Crane -2-

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

NRCs document system (ADAMS), accessible from the NRC Web site 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 No. 50-461

License No. NPF-62

Enclosure: Inspection Report No. 05000461/2007004

w/Attachment: Supplemental Information

cc w/encl: Site Vice President - Clinton Power Station

Plant Manager - Clinton Power Station

Regulatory Assurance Manager - Clinton Power Station

Chief Operating Officer

Senior Vice President - Nuclear Services

Vice President - Operations Support

Vice President - Licensing and Regulatory Affairs

Manager Licensing - Clinton Power Station

Senior Counsel, Nuclear, Mid-West Regional Operating Group

Document Control Desk - Licensing

Assistant Attorney General

Illinois Emergency Management Agency

State Liaison Officer, State of Illinois

Chairman, Illinois Commerce Commission

DOCUMENT NAME:C:\FileNet\ML073100335.wpd

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

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

OFFICE RIII RIII RIII RIII

NAME M. Ring for M. Ring

B. Dickson

DATE 11/05/2007 11/06/2007

OFFICIAL RECORD COPY

Letter to C. Crane from M. Ring dated November 6, 2007

SUBJECT: CLINTON POWER STATION

NRC INTEGRATED INSPECTION REPORT 05000461/2007004

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U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No: 50-461

License No: NPF-62

Report No: 05000461/2007004

Licensee: AmerGen Energy Company, LLC

Facility: Clinton Power Station

Location: Route 54 West

Clinton, IL 61727

Dates: July 1 through September 30, 2007

Inspectors: B. C. Dickson, Senior Resident Inspector

D. Tharp, Resident Inspector

A. Barker, Senior Project Engineer

J. McGhee, Reactor Engineer

A. Koonce, Reactor Engineer

D. Melendez, Reactor Engineer

M. Mitchell, Health Physicist

Approved by: Mark Ring, Chief

Branch 1

Division of Reactor Projects

Enclosure

SUMMARY OF FINDINGS

IR 05000461/2007004, AmerGen Energy Company LLC, on 07/01/07 - 09/30/2007 Clinton

Power Station, Event Follow-up.

This report covers a three month period of baseline resident inspection and announced

baseline inspection on radiation protection. The inspection was conducted by Region III

inspectors and the resident inspectors. Three Green findings, involving Non-Cited Violations,

were identified. The significance of most findings is indicated by their color (Green, White,

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

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

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

operation of commercial nuclear power reactors is described in NUREG-1649, Reactor

Oversight Process, Revision 3, dated July 2000.

A. Inspector-Identified and Self Revealing Findings

Cornerstone: Barrier Integrity

Green. A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50,

Appendix B, Criterion IV, Procurement Document Control, was self revealed following

receipt of laboratory results that showed that Division 1 control room ventilation system

charcoal filter penetration values were higher than allowed by Clintons Technical

Specifications. This issue occurred because the licensee failed to establish proper

purchase specifications for charcoal used in the control room ventilation system.

Additionally, this issue led to Division 1 control room ventilation subsystem being

inoperable from May 9 through May 16, 2005, concurrent with the Division 2 control

room ventilation subsystem being inoperable due to planned maintenance from

May 9 through May 14, 2005. Licensee corrective actions included entering the issue

into the corrective action program, revising the charcoal purchase specifications, and

adding limitations to work orders to prevent scheduling work that could impact the

operability of redundant systems.

This issue was more than minor because it affected the objective of the Barrier Integrity

cornerstone of assuring that physical design barriers protect the public from radionuclide

releases caused by accidents or events. Additionally, this issue is associated with the

barrier performance attribute of maintaining Radiological Barrier functionality of the

control room. Failure to ensure adequate purchase specifications resulted in there

being a period where both trains of control room ventilation were inoperable without the

knowledge of the operators. The issue was of very low safety significance because it

only represented a degradation of the radiological barrier function provided for the

control room. (Section 4OA3.1)

Green. The inspectors identified a performance deficiency involving a Non-Cited

Violation of Technical Specifications when the licensee failed to meet the required

completion time for an action statement in Technical Specification 3.4.5. Specifically,

Technical Specification 3.4.5 does not allow reactor coolant system pressure boundary

leakage and requires a shutdown to Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if pressure boundary

1 Enclosure

leakage is discovered. Upon entry into the drywell following a shutdown of the reactor

on June 19, 2007, the licensee discovered the existence of reactor coolant system

pressure boundary leakage. Indications of the leakage had been discovered at 0433 on

June 18, 2007, but the plant was not placed in Mode 3 until approximately 31 hours3.587963e-4 days 0.00861 hours 5.125661e-5 weeks 1.17955e-5 months later

at 1125 on June 19, 2007. Licensee corrective actions included replacing the leaking

flexible hose, scheduling replacement of other flexible hoses, and establishing a

preventive maintenance replacement frequency for the flexible hoses.

This issue was more than minor because operating with a degraded pressure boundary

affected the reactor coolant system equipment and barrier performance attribute of the

Barrier Integrity cornerstone, in that, reactor coolant system pressure boundary leakage

results in a reduction in the reasonable assurance that physical design barriers protect

the public from radionuclide releases caused by accidents or events. The issue was of

very low safety significance because the potential maximum size of the leak was well

within the capability of the available mitigating equipment. The finding is related to the

cross-cutting area of Human Performance (Decision Making) in that operators had

initially entered TS 3.4.5 for pressure boundary leakage, but later chose not to treat the

leakage as pressure boundary leakage, and treat it as unidentified leakage until the

actual location could be determined (H.1(b)). (Section 4OA3.2)

Green. A performance deficiency involving a Non-Cited Violation of 10 CFR Part 50

Appendix B, Criteria V, Instructions, Procedures, and Drawings, was self-revealed

following an event on August 17, 2007, where a spent fuel bundle being moved to a

temporary storage location came in contact with and rested upon another fuel bundle

seated in its storage location. The licensee procedure that governs spent fuel pool

movement failed to provide adequate guidance on how high to lift the fuel bundle prior to

traversing across the spent fuel pool. Licensee corrective actions included revising the

fuel handling procedure to provide specific instructions regarding how high to lift a fuel

bundle during spent fuel pool movements.

This issue was more than minor because it affected the barrier integrity objective of

assuring that physical design barriers protect the public from radionuclide releases

caused by accidents or events. The inspectors determined that this issue only degraded

the Fuel Cladding Barrier and its associated cornerstone, therefore, this issue was of

very low safety significance. This finding is related to the cross-cutting area of Human

Performance (Resources) because the licensee did not provide complete and accurate

procedures. Specifically, the procedure relied on the skills of the operator, did not

provided specific values on how high to lift a fuel bundle, and did not require

independent verification (H.2(c)). (Section 4OA3.3)

B. Licensee-Identified Violations

No findings of significance were identified.

2 Enclosure

REPORT DETAILS

Summary of Plant Status

The plant was operated at approximately 96 to 97 percent rated thermal power (maintaining

100 percent electrical output) throughout the inspection period with several derates of

approximately 2-3 percent at the requests of the grid operator. The grid operator made those

requests due to concerns regarding grid stability.

Exceptions to relatively steady state operation occurred on July 1, 2007, when operators

lowered reactor power to approximately 82 percent to make repairs to an electrohydraulic

control system leak on the #4 main turbine control valve and on September 9, 2007, when

operators lowered reactor power of approximately 75 percent to perform control rod pattern

adjustments, quarterly main turbine valve testing and main steam isolation valve testing.

1. REACTOR SAFETY

Cornerstone: Initiating Events, Mitigating Systems, and Barrier Integrity

1R01 Adverse Weather Protection (71111.01)

a. Inspection Scope

The inspectors evaluated site readiness for severe thunderstorms prior to arrival of

weather fronts associated with severe thunderstorms, high winds and rain that resulted

in severe thunderstorm and tornado watches being identified. On each occasion

operations shift manning was verified to be adequate and in accordance with site

procedures. Site walkdowns were performed to evaluate potential vulnerabilities for

missile generation during high winds or tornados and to assess the implementation of

the site procedures. The communications protocol between the control room and the

transmission system operator was also reviewed during the inspections and examples of

the quality of communication were observed due to severe weather conditions during

the inspection.

This review represented two inspection samples of review prior to impending weather

conditions.

b. Findings

No findings of significance were identified.

1R04 Equipment Alignments (71111.04Q)

a. Inspection Scope

The inspectors performed partial walkdowns of accessible portions of divisions of

risk-significant mitigating systems equipment during times when the divisions were of

increased importance due to redundant divisions or other related equipment being

3 Enclosure

unavailable. The inspectors utilized the valve and electric breaker checklists listed at

the end of this report to verify that the components were properly positioned and that

support systems were lined up as needed. The inspectors also examined the material

condition of the components and observed operating parameters of equipment to verify

that there were no obvious deficiencies. The inspectors reviewed outstanding work

orders and issue reports (IRs) associated with the divisions to verify that those

documents did not reveal issues that could affect division function. The inspectors used

the information in the appropriate sections of the Updated Safety Analysis Report

(USAR) to determine the functional requirements of the systems. The documents listed

at the end of this report were also used by the inspectors to evaluate this area.

The inspectors performed two samples by verifying the alignment of the reactor core

isolation cooling system and the Division 1 standby gas treatment system while

Division 2 was out of service.

b. Findings

No findings of significance were identified.

1R05 Fire Protection (71111.05Q)

a. Inspection Scope

The inspectors conducted fire protection walkdowns which were focused on availability,

accessibility, and the condition of fire fighting equipment, the control of transient

combustibles and ignition sources, and on the condition and operating status of installed

fire barriers. The inspectors selected fire areas for inspection based on their overall

contribution to internal fire risk, as documented in the individual plant examination of

external events with later additional insights, and their potential to impact equipment

which could cause a plant transient, to verify that fire hoses and extinguishers were in

their designated locations and available for immediate use, that fire detectors and

sprinklers were not obstructed, that transient material loading was within the analyzed

limits, and that fire doors, dampers, and penetration seals appeared to be in satisfactory

condition. The inspectors verified that minor issues identified during the inspection were

entered into the licensees corrective action program.

The inspectors reviewed portions of the licensees fire protection evaluation report and

the USAR to verify consistency in the documented analysis with installed fire protection

equipment at the station.

The inspectors completed seven samples by inspection of the following areas:

Fire Area D-5d, Division 1 diesel generator room;

Fire Area A-2b, Division 1 residual heat removal pump and heat exchanger

room;

Fire Area A-1, 707' auxiliary building, general access area;

Fire Area CB-1c, 719' control building, heating ventilation and air conditioning

equipment area;

Fire loading and associated fire hazard permits in approved storage area;

4 Enclosure

Fire Area A-2 and Fire Zone A-2a, reactor core isolation cooling pump room; and

Fire Area D-6a, b, Division 2 diesel generator and day tank room.

b. Findings

No findings of significance were identified.

1R06 Flood Protection Measures (71111.06)

a. Inspection Scope

The inspectors observed the flood protection equipment in the B and C residual heat

removal pump rooms. The inspectors verified that flooding mitigation plans and

equipment were consistent with the design requirements and risk analysis assumptions.

The inspectors reviewed USAR Section 3.4.1 for internal flooding protection measures,

reviewed the licensees flooding mitigation procedures, and reviewed issue reports

related to possible flood protection issues. Additionally, plant walkdowns were

performed to verify design barriers were properly maintained. Penetrations between

rooms, watertight doors, electrical conduit seals and covers, and room drains were

inspected to verify material condition met design assumptions. The inspectors

performed a review of the stations maintenance database to verify preventative

maintenance was current and equipment deficiencies were being appropriately reported

and resolved. Additionally, the inspectors reviewed the maintenance rule scoping and

performance criteria and determined that the function was being tracked appropriately.

The corrective action program was also reviewed for the past 12 months for issues

related to internal flood protection. The inspectors completed a one inspection sample

by completing the internal flooding review of the B residual heat removal pump and

heat exchanger room and the C residual heat removal pump room.

b. Findings

No findings of significance were identified.

1R07 Heat Sink Performance (71111.07A)

a. Inspection Scope

The inspectors verified readiness and availability of the Division 3 emergency diesel

generator heat exchanger by performing the following activities:

Observed in-progress eddy current testing and reviewed post-eddy current

testing results;

Observed the condition of heat exchanger end bell and tube sheet during

walkdown;

Reviewed the results of tube inspections and outage work packages to

determine whether maintenance was performed in accordance with the

licensees maintenance program for heat exchangers and reviewed issue

reports to verify that deficiencies were identified and incorporated into the

licensees corrective action program;

5 Enclosure

Reviewed the evaluation and corrective actions for Action Request 678934,

Div III DG HX (1DG13A) Chemistry Results,

Reviewed the evaluation and corrective actions for Action Request 675035,

Excessive Erosion Discovered on the Flange Faces 1VH03A, and

Verified the heat exchanger was properly classified under the Maintenance

Rule and identified issues received appropriate program reviews.

This inspection represented the completion of one annual sample.

b. Findings

No findings of significance were identified.

1R11 Licensed Operator Requalification (71111.11)

a. Inspection Scope

The inspectors observed an evaluation of an operating crew on September 19, 2007.

The scenario (ESG LOR Exam 87d) consisted of a failure of the A reactor recirculation

flow control valve, a failure to scram using the manual pushbuttons, unisolable flooding

in emergency core cooling system pump room and an emergency depressurization.

The inspectors reviewed licensed-operator requalification training to evaluate operator

performance in mitigating the consequences of a simulated event, particularly in the

areas of Human Performance. The inspectors evaluated operator performance

attributes which included communication clarity and formality, timely performance of

appropriate operator actions, appropriate alarm response, proper procedure use and

adherence, and senior reactor operator oversight and command and control.

Additionally, simulator physical fidelity and training department actions to incorporate

current plant deficiencies and annunciators were evaluated.

The inspectors also assessed the performance of the training staff evaluations involved

in the requalification process. For any weaknesses identified during the session, the

inspectors observed that the licensee evaluators also noted the issues and discussed

them in the critique at the end of the session. Discrepancies were reviewed with the

training staff. The inspectors verified that all issues were captured in the training

program and licensee corrective action program.

These activities completed one inspection sample.

b. Findings

No findings of significance were identified.

6 Enclosure

1R12 Maintenance Effectiveness (71111.12)

a. Inspection Scope

The inspectors reviewed the effectiveness of the licensees maintenance efforts in

implementing 10 CFR Part 50.65 (the maintenance rule (MR)) requirements, including a

review of scoping, goal-setting, performance monitoring, short and long-term corrective

actions, and current equipment performance problems. These systems were selected

based on their designation as risk-significant under the maintenance rule. The

inspectors also reviewed issue reports and associated documents for appropriate

identification of problems, entry into the corrective action system, and appropriateness

of planned or completed actions. The inspectors completed three samples by reviewing

the following:

Division 1 DC Battery and DC distribution;

Containment and reactor vessel isolation system and

Reactor core isolation cooling system primary containment isolation valves.

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

a. Inspection Scope

The inspectors observed the licensees risk assessment processes and considerations

used to plan and schedule maintenance activities on safety-related structures, systems,

and components particularly to ensure that maintenance risk and emergent work

contingencies had been identified and resolved. The inspectors completed five

samples by assessing the effectiveness of risk management activities for the following

work activities or work weeks:

Reviewed licensee risk assessment and subsequent area operator staging

associated with standby liquid control system pump and valve quarterly

operability run;

Reviewed licensees risk assessment and detailed work plan for reserve auxiliary

transformer static VAR compensator prior to the removal of test equipment;

Reviewed licensee risk assessment for planned surveillance on high pressure

core spray system;

Reviewed licensee risk assessment associated with six-year preventative

maintenance of the Division 3 diesel generator (including walkdown of risk

sensitive area to ensure proper flagging);

Reviewed the licensees risk assessment of 138 kv line work by Ameren IP and

subsequent emergency reserve auxiliary transformer outage.

b. Findings

No findings of significance were identified.

7 Enclosure

1R15 Operability Evaluations (71111.15)

a. Inspection Scope

The inspectors reviewed the following operability determinations and evaluations

affecting mitigating systems to determine whether operability was properly justified and

the component or system remained available such that no unrecognized risk increase

had occurred. The inspectors completed four samples of operability determinations and

evaluations by reviewing the following:

Operability Evaluation 655836: Residual heat removal water leg pump supply

check valves (1E12F084A and 1E12F085A) failure to close;

Operability Evaluation 630815-03: Division 1, 2, and 3 shutdown service water

valves (1SX025A, 1SX025B, and 1SX025C) exceed the vendor allowable

deflection;

Action Request 667663: Through body leak of shutdown service water valve

(1SX019B);

Action Request 660299: Type C Integrated leakage rate testing for test lines

going through containment penetration 1MC-152 not being performed;

Operability Evaluation 664280: Through wall leak on shutdown service water line

(1SX20AB); and

Operability Evaluation 671001: Division 3 essential switchgear/shutdown service

water piping wall thickness below minimum screening criteria.

b. Findings

No findings of significance were identified.

1R19 Post-Maintenance Testing (71111.19)

a. Inspection Scope

The inspectors reviewed the post maintenance testing activities associated with

maintenance or modification of important mitigating, barrier integrity, and support

systems that were identified as risk significant in the licensees risk analysis. The

inspectors reviewed these activities to verify that the post maintenance testing was

performed adequately, demonstrated that the maintenance was successful, and that

operability was restored. During this inspection activity, the inspectors interviewed

maintenance and engineering department personnel and reviewed the completed post

maintenance testing documentation. The inspectors used the appropriate sections of

the Technical Specifications (TS) and USAR, as well as the documents listed at the end

of this report, to evaluate this area.

Testing subsequent to the following activities was observed and evaluated:

Work Order 00318466: Troubleshooting and repair of Division 2, 250 Volt

battery charger (1DC07E);

Work Order 01048964-03: Post maintenance testing for optical isolator pair

(manual scram push button) P664-A-A81-A313/A305;

8 Enclosure

Work Order 00914954-03: Replacement of 1SX-303A;

Work Order 00909977-04: Replacement of standby gas treatment system

pressure switches;

Work Order 00909977-02: Replacement of combustible gas control system

pressure switch;

Work Order 00969911: Replacement of Division 1 main control room ventilation

supply fan; and

Reviewed post maintenance testing of Division 3 emergency diesel generator

following six year maintenance.

Performance of this review comprised seven samples.

b. Findings

No findings of significance were identified.

1R22 Surveillance Testing (71111.22)

a. Inspection Scope

The inspectors witnessed selected surveillance testing and/or reviewed test data to

verify that the equipment tested using the surveillance procedures met the TS, the

Technical requirements Manual (TRM), the USAR, and licensee procedural

requirements, and demonstrated that the equipment was capable of performing its

intended safety functions. The activities were selected based on their importance in

verifying mitigating systems capability and barrier integrity. The inspectors used the

documents listed at the end of this report to verify that the testing met the frequency

requirements; that the tests were conducted in accordance with the procedures,

including establishing the proper plant conditions and prerequisites; that the test

acceptance criteria were met; and that the results of the tests were properly reviewed

and recorded. In addition, the inspectors interviewed operations, maintenance, and

engineering department personnel regarding the tests and test results.

Seven samples were completed.

The inspectors evaluated the following surveillance tests:

CPS 9069.01, Shutdown service water operability test pump B;

CPS 9000.01, Control room surveillance log and CPS 3315.02 Leak detection;

CPS 9051.01, High pressure core spray and water leg pump operability;

CPS 9052.01, Low pressure core spray and residual heat removal A pumps

water leg pump operability;

CPS 9052.04, Low pressure core spray and residual heat removal A vent and

fill;

CPS 9054.01, Reactor core isolation cooling operability; and

CPS 9080.02, Diesel generator 1B operability - manual and quick start

operability.

9 Enclosure

These tests included three in-service testing (IST) surveillance samples and one reactor

coolant system leakage detection surveillance sample.

b. Findings

No findings of significance were identified.

1R23 Temporary Plant Modifications (7111.23)

The inspectors reviewed and evaluated the following temporary plant modification on

risk significant equipment to verify that the instructions were consistent with applicable

design modification documents and that the modifications did not adversely impact

system operability or availability. The inspectors interviewed operations, engineering

and maintenance personnel, as appropriate, and reviewed the design modification

documents and the 10 CFR 50.59 evaluations against the applicable portions of the

USAR. The documents listed at the end of the report were also used by the inspectors

to evaluate this area. The inspectors reviewed the issues that the licensee entered into

its corrective action program to verify that identified temporary modification problems

were being entered into the program with the appropriate characterization and

significance. The inspectors also reviewed the licensees corrective actions for

temporary modification related issues documented in selected condition reports. The

condition reports are specified in the List of Documents Reviewed.

The inspectors completed one inspection sample by reviewing the following temporary

modification:

Defeating Inclined fuel transfer system upender/fuel handling platform Interlock,

Revision 0.

b. Findings

No findings of significance were identified.

2. EMERGENCY PREPAREDNESS

1EP6 Drill Evaluation (71114.06)

a. Inspection Scope

The inspectors observed the emergency response activities associated with the drill

conducted on August 9, 2007. Specifically, the inspectors verified that the emergency

classification and simulated notifications were properly completed, and that the licensee

adequately critiqued the training. Additionally, the inspectors observed licensee

activities during the drill in the simulated control room and the Technical Support Center

(TSC).

The drill included an explosion in the Division 2 emergency diesel generator and a

subsequent anticipated transient without a scram.

10 Enclosure

b. Findings

No findings of significance were identified.

2. RADIATION SAFETY

Cornerstone: Occupational Radiation Safety

2OS1 Access Control to Radiologically Significant Areas (71121.01)

.1 Review of Licensee Performance Indicators for the Occupational Exposure Cornerstone

a. Inspection Scope

The inspectors reviewed the licensees Occupational Exposure Control cornerstone

performance indicators (PIs) to determine whether or not the conditions surrounding the

PIs had been evaluated, and identified problems had been entered into the corrective

action program for resolution. These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

.2 Plant Walkdowns and Radiation Work Permit Reviews

a. Inspection Scope

The inspectors reviewed licensee controls and surveys in the following two radiologically

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

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

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

including surveys, postings and barricades were acceptable:

Spent fuel pool re-rack project; and

Phase separator resin transfer.

These reviews represented one inspection sample.

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

access these two areas and other high radiation work areas to identify the work control

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

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

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

the actions required when their electronic dosimeters noticeably malfunctioned or

alarmed. These reviews represented one inspection sample.

The inspectors walked down and surveyed (using an NRC survey meter) these two

areas to verify that the prescribed radiation work permit, procedure, and engineering

11 Enclosure

controls were in place, that licensee surveys and postings were complete and accurate,

and that air samplers were properly located. These reviews represented one inspection

sample.

The inspectors reviewed RWPs for airborne radioactivity areas to verify barrier integrity

and engineering controls performance (e.g., HEPA ventilation system operation) and to

determine if there was a potential for individual worker internal exposures of greater

than 50 millirem committed effective dose equivalent. No areas in the plant were

airborne areas with a potential of greater than 5 millirem committed effective dose

equivalent. Work areas having a history of, or the potential for, airborne transuranics

were evaluated to verify that the licensee had considered the potential for transuranic

isotopes and provided appropriate worker protection. The license has not had a history

of transuranic contamination. These reviews represented one inspection sample.

The adequacy of the licensees internal dose assessment process for internal exposures

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

reviews represented one inspection sample.

The inspectors also reviewed the licensees physical and programmatic controls for

highly activated and/or contaminated materials (non-fuel) stored within spent fuel or

other storage pools. These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

.3 Problem Identification and Resolution

a. Inspection Scope

The inspectors reviewed the licensees self-assessments, audits, Licensee Event

Reports, and Special Reports related to the access control program to verify that

identified problems were entered into the corrective action program for resolution.

These reviews represented one inspection sample.

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

high radiation area radiological incidents when available (non-PIs identified by the

licensee in high radiation areas greater than 1R/hr). Staff members were interviewed

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

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

to safety and risk based on the following:

Initial problem identification, characterization, and tracking;

Disposition of operability/reportability issues;

Evaluation of safety significance/risk and priority for resolution;

Identification of repetitive problems;

Identification of contributing causes;

Identification and implementation of effective corrective actions;

12 Enclosure

Resolution of Non-Cited Violations (NCVs) tracked in the corrective action

system; and

Implementation/consideration of risk significant operational experience feedback.

These reviews represented one inspection sample.

The inspectors evaluated the licensees process for problem identification,

characterization, and prioritization and verified that problems were entered into the

corrective action program and resolved. For repetitive deficiencies and/or significant

individual deficiencies in problem identification and resolution, the inspectors verified

that the licensees self-assessment activities were capable of identifying and addressing

these deficiencies. These reviews represented one inspection sample.

The inspectors reviewed licensee documentation packages for all PI events occurring

since the last inspection to determine if any of these PI events involved dose rates

greater than 25 R/hr at 30 centimeters or greater than 500 R/hr at 1 meter. Barriers

were evaluated for failure and to determine if there were any barriers left to prevent

personnel access. Unintended exposures greater than 100 millirem total effective dose

equivalent (or greater than 5 rem shallow dose equivalent or greater than 1.5 rem lens

dose equivalent), were evaluated to determine if there were any regulatory

overexposures or if there was a substantial potential for an overexposure. There were

no unintended exposures approaching 100 millirem. These reviews represented one

inspection sample.

b. Findings

No findings of significance were identified.

.4 Job-In-Progress Reviews

a. Inspection Scope

The inspectors observed the following two work activities that were being performed in

radiation areas, airborne radioactivity areas, or high radiation areas that presented the

greatest radiological risk to workers:

Spent fuel pool re-rack project; and

Phase separator resin transfer.

The inspectors reviewed radiological job requirements for these two activities including

RWP requirements and work procedure requirements, and attended

As-Low-As-Reasonably-Achievable (ALARA) job briefings. These reviews represented

one inspection sample.

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

radiological conditions in the work area were adequately communicated to workers

through pre-job briefings and postings. The inspectors also verified the adequacy of

radiological controls including required radiation and contamination surveys; radiation

13 Enclosure

protection job coverage; and contamination controls. These reviews represented one

inspection sample.

Radiological work in high radiation work areas having significant dose rate gradients

was reviewed to evaluate the application of dosimetry to effectively monitor exposure to

personnel and to verify that licensee controls were adequate. These work areas

involved areas where the dose rate gradients were severe (diving activities in the fuel

pool re-rack project) which increased the necessity of providing multiple dosimeters.

These reviews represented one-inspection sample.

b. Findings

No findings of significance were identified.

.5 High Risk Significant, High Dose Rate (HRA), and Very High Radiation Area Controls

a. Inspection Scope

The inspectors held discussions with the acting Radiation Protection Manager

concerning high dose rate/high radiation area and very high radiation area controls and

procedures, including procedural changes that had occurred since the last inspection, in

order to verify that any procedure modifications did not substantially reduce the

effectiveness and level of worker protection. These reviews represented one inspection

sample.

The inspectors discussed with RP supervisors the controls that were in place for special

areas that had the potential to become very high radiation areas during certain plant

operations, to determine if these plant operations required communication beforehand

with the RP group, so as to allow corresponding timely actions to properly post and

control the radiation hazards. These reviews represented one inspection sample.

The inspectors conducted plant walkdowns to verify the posting and locking of

entrances to high dose rate HRAs, and very high radiation. These reviews represented

one inspection sample.

b. Findings

No findings of significance were identified.

.6 Radiation Worker Performance

a. Inspection Scope

During job performance observations, the inspectors evaluated radiation worker

performance with respect to stated radiation protection work requirements and

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

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

accounted for the level of radiological hazards present. These reviews represented one

inspection sample.

14 Enclosure

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

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

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

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

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

acting Radiation Protection Manager. These reviews represented one inspection

sample.

b. Findings

No findings of significance were identified.

.7 Radiation Protection Technician Proficiency

a. Inspection Scope

During job performance observations, the inspectors evaluated Radiation Protection

Technician (RPT) performance with respect to radiation protection work requirements

and evaluated whether they were aware of the radiological conditions in their workplace,

the RWP controls and limits in place, and if their performance was consistent with their

training and qualifications with respect to the radiological hazards and work activities.

These reviews represented one inspection sample.

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

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

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

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

These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

Cornerstone: Public Radiation Safety

2PS2 Radioactive Material Processing and Transportation (71122.02)

.1 Radioactive Waste System

a. Inspection Scope

The inspectors reviewed the liquid and solid radioactive waste system descriptions in

the Updated Final Safety Analysis Report (UFSAR), and the 2006 Annual Radioactive

Effluent Release Report for information on the types and amounts of radioactive waste

(radwaste) generated and disposed. The inspectors reviewed the scope of the

licensees audit/self-assessment activities, with regard to radioactive material processing

and transportation programs to determine if those activities satisfied the requirements of

15 Enclosure

10 CFR 20.1101(c) and the quality assurance audit requirements of Appendix G to

10 CFR Part 20 and of 10 CFR 71.137, as applicable.

These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

.2 Radioactive Waste System Walkdowns

a. Inspection Scope

The inspectors walked down portions of the liquid and solid radwaste processing

systems to verify that these systems were consistent with the descriptions in the UFSAR

and in the Process Control Program and to assess the material condition and operability

of those systems. The inspectors reviewed the status of radioactive waste process

equipment that was not operational and/or was abandoned in place. The inspectors

discussed with the licensee the administrative and/or physical controls preventing the

inadvertent use of this equipment to ensure that the equipment would not contribute to

an unmonitored release path or be a source of unnecessary personnel exposure.

The inspectors reviewed changes to the waste processing system to verify the changes

were reviewed and documented in accordance with 10 CFR 50.59 and to assess the

impact of the changes on radiation dose to members of the public. The inspectors

reviewed the licensees processes for transferring waste resin into shipping containers

to determine if appropriate waste stream mixing and sampling was performed so as to

obtain representative waste stream samples for analysis. The inspector also reviewed

the methodologies for waste concentration averaging to determine if representative

samples of the waste product were provided for the purposes of waste classification in

accordance with 10 CFR 61.55.

These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

.3 Waste Characterization and Classification

a. Inspection Scope

The inspectors reviewed the licensees methods and procedures for determining the

classification of radioactive waste shipments including the use of scaling factors to

quantify difficult-to-measure radionuclides. The inspectors reviewed the licensees most

recent radiochemical sample analysis results for each of the licensees waste streams,

and the associated calculations used to account for difficult-to-measure radionuclides.

These waste streams consisted of radwaste demineralizer resins, various filter media,

and dry active waste (DAW). The inspectors also reviewed the licensees use of scaling

16 Enclosure

factors to quantify difficult-to-measure radionuclides (e.g., pure alpha or beta emitting

radionuclides). The reviews were conducted to verify that the licensees program

assured compliance with 10 CFR 61.55 and 10 CFR 61.56, as required by Appendix G

of 10 CFR Part 20. The inspectors also reviewed the licensees waste characterization

and classification program to determine if reactor coolant chemistry data was

periodically evaluated to account for changing operational parameters that could

potentially affect waste stream classification and thus validate the continued use of

existing scaling factors between sample analysis updates.

These reviews represented one inspection sample.

b. Findings

No findings of significance were identified.

.4 Shipment Preparation and Records

a. Inspection Scope

The inspectors reviewed the documentation of shipment packaging, surveying, package

labeling and marking, vehicle inspections and placarding, emergency instructions, and

licensee verification of shipment readiness for six selected non-excepted radioactive

material and radwaste shipments, made between March 2006 and January 2007. The

shipment documentation reviewed included:

Equipment in a B-25 Box Shipped as Low Specific Activity (LSA);

Two Spent Resins Shipped as Type A;

Two Phase Separator Resins Shipped as Type A; and

Part 61 Resin Samples Shipped as LSA.

For each shipment, the inspectors determined if the requirements of 10 CFR Parts 20

and 61 and those of the Department of Transportation (DOT) in 49 CFR Parts 170-189

were met. Specifically, records were reviewed, and staff involved in shipment activities

were interviewed to determine if packages were labeled and marked properly, if

packages and transport vehicle surveys were performed with appropriate

instrumentation, whether survey results satisfied DOT requirements, and if the quantity

and type of radionuclides in each shipment were determined accurately. The inspectors

also determined whether shipment manifests were completed in accordance with DOT

and NRC requirements, if they included the required emergency response information, if

the recipient was authorized to receive the shipment, and if shipments were tracked as

required by 10 CFR Part 20.

Selected staff involved in shipment activities were interviewed by the inspectors to

determine if they had adequate skills to accomplish shipment related tasks and to

determine if the shippers were knowledgeable of the applicable regulations to satisfy

package preparation requirements for public transport with respect to NRC

Bulletin 79-19, Packaging of Low-Level Radioactive Waste for Transport and Burial.

Also, the inspectors reviewed the transportation specific training for the authorized

shippers to assure that they met the requirements of 49 CFR Part 172, Subpart H.

17 Enclosure

These reviews represented two inspection samples.

b. Findings

(1) Shipment Total Quantity Re-characterized After Shipping

A shipment of phase separator resins was shipped from Clinton Power Station

September 30, 2005, and delivered to a vendor on October 1, 2005. The total curie

quantity in the shipment was in excess of the vendors Agreement State license limits.

The vendor communicated this discrepancy to shipping personnel at Clinton Power

Station on October 3, 2005. The shipper then re-characterized the total quantity of the

shipment by reviewing dose rate survey data and applying a dose to curie

methodology. Contrary to Clinton procedure, the re-characterization was not reviewed

by other Clinton personnel and new paperwork for the shipment, including a new NRC

Form 541 was generated and transferred to the recipient.

This event remains under review by the NRC and is categorized as an Unresolved Item

(URI) (URI 05000461/2007004-01).

.5 Identification and Resolution of Problems for Radwaste Processing and Transportation

a. Inspection Scope

The inspectors reviewed selected condition reports, self-assessment and audit reports,

along with field observation reports that addressed the radioactive waste and radioactive

materials shipping program, since the last inspection to determine if the licensee had

effectively implemented the corrective action program and if problems were identified,

characterized, prioritized, and corrected. The inspectors also determined whether the

licensee's self-assessment program was capable of identifying repetitive deficiencies, or

significant individual deficiencies in problem identification and resolution.

The inspectors also selectively reviewed other corrective action program reports

generated since the previous inspection that dealt with the radioactive material or

radwaste shipping program, interviewed staff, and reviewed documents to determine if

the following activities were being conducted in an effective and timely manner,

commensurate with their importance to safety and risk:

Initial problem identification, characterization, and tracking;

Disposition of operability/reportability issues;

Evaluation of safety significance/risk and priority for resolution;

Identification of repetitive problems;

Identification of contributing causes;

Identification and implementation of effective corrective actions;

Resolution of Non-Cited Violations tracked in corrective action system(s); and

Implementation/consideration of risk significant operational experience feedback.

These reviews represented one inspection sample.

18 Enclosure

b. Findings

No findings of significance were identified.

4. OTHER ACTIVITIES (OA)

4OA1 Performance Indicator Verification (71151)

To perform a periodic review of performance indicator (PI) data to determine its

accuracy and completeness.

Cornerstones: Mitigation Systems

1. Mitigating Systems Performance Indicators (MSPI)

a. Inspection Scope

The inspectors sampled the licensees submittals for performance indicators for the

period of July 2006 through June 2007. The inspectors used performance indicator

definitions and guidance contained in revision 5 of Nuclear Energy Institute (NEI)

document 99-02, Regulatory Assessment Performance Indicator Guideline, to verify

the accuracy of the performance indicator data. The inspectors performed three

samples by reviewing the following:

MSPI Emergency AC;

MSPI Residual heat removal and

MSPI Shutdown service water.

b. Issues and Findings

No findings of significance were identified.

2. Safety System Functional Failures Performance Indicator

a. Inspection Scope

The inspectors reviewed, at a minimum, the most recent 24 months of Licensee Event

Reports, licensee data reported to the NRC, plant logs, issue reports, and NRC

inspection reports to verify the following performance indicators reported by the

licensee for the 2nd Quarter of 2007 for Safety System Functional Failures.

The inspectors verified that the licensee accurately reported performance as defined by

the applicable revision of Nuclear Energy Institute Document 99-02, Regulatory

Assessment Performance Indicator Guideline.

This performance indicator review constituted one inspection sample.

19 Enclosure

b. Findings

No findings of significance were identified.

Cornerstones: Public Radiation Safety and Barrier Integrity

a. Inspection Scope

The inspectors reviewed the licensees determination of PIs for the Public Radiation

Safety Performance Indicator (Radiological Environmental Technical

Specification/Offsite Dose Calculation Manual Radiological Effluent Occurrences) and

Barrier Integrity Performance Indicator (Reactor Coolant System Specific Activity) to

determine if the licensee accurately determined these performance indicators and had

identified all occurrences. Specifically, the inspectors reviewed the licensees

corrective action program documents for the 4th quarter of CY 2006 and the 1st, 2nd, and

3rd quarters of CY 2007 and Public Radiation Safety and Barrier Integrity Performance

Indicator data to ensure that there were no PI occurrences that were not identified by

the licensee. In addition to record reviews, the inspectors observed a chemistry

technician obtain and analyze a reactor coolant system sample. The inspectors

interviewed members of the licensees staff who were responsible for performance

indicator data acquisition, verification and reporting, to determine if their review and

assessment of the data was adequate.

These reviews represented two inspection samples.

b. Findings

No findings of significance were identified.

4OA2 Identification and Resolution of Problems (71152)

a. Inspection Scope

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

issues during baseline inspection activities and plant status reviews to verify that they

were being entered into the licensees corrective action system at an appropriate

threshold, that adequate attention was being given to timely corrective actions, and that

adverse trends were identified and addressed. In addition, the inspectors reviewed the

following issue:

b. Findings

There were no findings of significance identified. The inspectors reviewed procedure

OP-AA-102-103, Operator Work-Around Program, and the issues being tracked for

program resolution. As of July 23, 2007, there were two operator workarounds that

were being tracked for resolution. The inspectors review determined that the issues

were appropriately characterized as operator workarounds.

20 Enclosure

The inspectors also reviewed selected operations department concerns and out of

tolerance items that were identified in operations narrative logs to identify potential

operator workarounds that were not in the program. The inspectors concluded that

there were no additional workarounds identified from the sources reviewed.

On August 2, 2007, the inspector attended a WorkAround Board (WAB) meeting. One

of the items under review that was identified during forced outage C1F049 and

documented by IR 642141, was the operation of 1E12-F009, residual heat removal

shutdown cooling inboard isolation valve, from the main control room. The valve would

not open or close electrically (i.e., mechanically bound). An operator workaround

program review was not requested by IR 642141. The resident inspectors questioned

the licensee on the appropriateness of not conducting an operator workaround program

review. This resulted in the August 2 review of this condition. In addition, the licensee

generated IR 656515 on the missed opportunity to conduct an operator workaround

program review of 1E12-F009 valve operation. On August 2, 2007, the WAB approved

1E12-F009 valve operation as an operator workaround. The WAB member dialogue

on the issues that were reviewed on August 2, 2007, was at the appropriate level of

detail, and considered operator compensatory actions required to comply with plant

procedures, design requirements and technical specifications.

4OA3 Event Follow-up (71153)

.1 (Closed) LER 05000461/2005-001-00. Inadequate Procurement Specification for

Charcoal Results in Inoperable Control Room Ventilation Subsystem.

Introduction: A performance deficiency involving a Non-Cited Violation of

10 CFR Part 50, Appendix B, Criterion IV, Procurement Document Control, was self

revealed following receipt of results that showed that Division 1 control room ventilation

(VC) system charcoal filter penetration values were higher than allowed by Clintons

Technical Specifications. This issue occurred because the licensee failed to establish

proper purchase specifications for charcoal used in the VC system. Additionally, this

issue led to Division 1 VC subsystems being inoperable from May 9 through

May 16, 2005, concurrent with Division 2 VC subsystems being inoperable due to

planned maintenance from May 9 through May 14, 2005.

Description: On May 4, 2005, the licensee took charcoal absorber samples from the

Division 1 VC recirculation charcoal bed filter in order to perform charcoal penetration

testing as required by VC system Technical Specification surveillance requirement 3.7.3.3. On May 9, 2005, planned maintenance was performed on the Division 2

emergency diesel generator resulting in it becoming inoperable. The licensee also

declared the Division 2 VC subsystem inoperable because of the diesel inoperability.

The Division 2 emergency diesel generator and the Division 2 VC system were

declared operable on May 14, 2005, following completion of planned maintenance on

the emergency diesel generator. On May 16, 2005, the licensee received the

Division 1 charcoal sample analysis results which showed that after applying a

correction factor, the charcoal penetration value was 8.168. This charcoal penetration

was higher than the charcoal penetration operability value of 6.0 contained in Clintons

Technical Specifications.

21 Enclosure

Because of the unsatisfactory results, the licensee concluded that the seven day

required action completion time to restore an inoperable VC subsystem was not met,

nor was the required action to be in Mode 3 and Mode 4 per Clintons Technical

Specifications. Additionally, the Division 2 VC subsystem was inoperable from

May 9 to May 14, 2005, due to the Division 2 emergency diesel generator maintenance

outage. The licensee concluded that the station was in Technical Specification (TS) 3.7.3 required action D.1, requiring entry into TS limiting condition for operation

(LCO) 3.0.3. The licensee did not enter the TS 3.7.3 required action and LCO 3.0.3,

since the sample results were not known during the emergency diesel generator

outage.

A licensee investigation concluded that the cause of this event was that charcoal

purchasing requirements were inadequate for the unique application at Clinton. The

licensee purchased the charcoal installed in the Division 1 VC system to

ANSI/ASME N509-1980, Nuclear Power Plant Air Cleaning Units and Components.

According to the licensee, this industry standard had no limit on as-manufactured

moisture levels. The licensee concluded that the Clinton VC systems unique design

attributes, which included higher than normal airflow velocity, thin charcoal beds, and

no airflow heaters coupled with high as-manufactured moisture levels significantly

reduced the charcoal residence time and increased the penetration levels. The

licensee investigation also concluded that 10 percent of the charcoal contained in the

failed filter was caked after 17 months of operations. The licensee stated that this

caking was again proof of high as-manufactured moisture content.

The failed charcoal bed filter contained coal from lot 55, batch 68. The licensees extent

of condition review determined that this event was limited to the VC A recirculation

charcoal bed filter. The bases of this conclusion were that penetration tests results

were satisfactory for the other beds in the VC system and no lot 55, batch 68, charcoal

was in storage. The licensee investigation also determined that the charcoal in lot 55,

batch 68, had as-manufactured high moisture content of at least 12 percent by weight.

All charcoal currently in storage was confirmed to have a moisture content of less than

eight percent.

The inspectors were concerned that the licensee maintenance scheduling process

would allow for the redundant VC subsystem to be made inoperable without knowing

the results of the Technical Specification required sample analysis. The inspectors

concluded poor scheduling of work contributed to both VC subsystems becoming

inoperable per the licensees Technical Specifications. The licensee did not address

this issue in the root cause report. In response to the inspectors questions in this area,

the licensee provided issue report 340314, Enhancement to Manage Risk of Charcoal

Samples. This issue report addressed the inspectors concern, in that, it incorporated

corrective actions that implemented an administrative trigger using model work orders

to ensure that no work would be scheduled for the next work week that would cause an

inoperability of the opposite train or its support systems. This administrative trigger

was placed into model work orders for all VC system and standby gas treatment

system predefined activities.

Analysis: The inspectors determined that the licensees failure to provide adequate

purchase specifications to satisfy the requirements for the Clinton unique VC system

22 Enclosure

was a performance deficiency warranting a significance evaluation. The inspectors

concluded that the finding was more than minor in accordance with IMC 0612, Power

Reactor Inspection Reports, Appendix B, because the finding affected the objective of

the Barrier Integrity cornerstone of assuring that physical design barriers protect the

public from radionuclide releases caused by accidents or events. Additionally, this

issue is associated with the barrier performance attribute of maintaining Radiological

Barrier functionality of the control room. Failure to ensure adequate purchase

specifications resulted in there being a period where both trains of control room

ventilation were inoperable without the knowledge of the operators.

The inspectors completed a Phase 1 significance determination using IMC 0609,

Significance Determination Process, Appendix A, Attachment 1, dated

March 23, 2007. Using IMC 0609 Appendix A, SDP Phase 1 screening worksheet, the

inspectors determined that this issue degraded the Containment Barrier. The

inspectors answered Yes, to whether the finding only represents a degradation of the

radiological barrier function provided for the control room, or auxiliary building, or spent

fuel pool, or SBGT system (BWR). As a result, the Phase 1 Worksheets screened as

Green.

The finding is also related to the cross-cutting area of Human Performance as defined

in IMC 0305, Operating Reactor Assessment Program, specifically, the finding is

related to the resources and work control component because the licensee did not

provide an adequate purchase specification reflective of the requirements for the VC

system. However, this issue was determined not to have a cross-cutting aspect

because it was over two years old and not reflective of current performance in this area

based on the inspectors review of the licensees effectiveness review of corrective

actions established to prevent this issue from occurring again.

Enforcement: 10 CFR Part 50, Appendix B, Criteria IV, Procurement Document

Control states that measures shall be established to assure that applicable regulatory

requirements, design bases, and other requirements which are necessary to assure

adequate quality are suitably included or referenced in the documents for procurement

of material, equipment, and services, whether purchased by the applicant or by its

contractors or subcontractors.

Contrary to the above, Clinton Power Station failed to specify in a procurement

document the moisture content for charcoal filters to be used in the VC system. This

issue resulted in charcoal being used in the division 1 VC system that was outside the

penetration limits established by Technical Specification 3.7.3, from May 4 to

May 16, 2007. This issue also resulted in both control room ventilation systems being

inoperable. The licensee entered this issue into the corrective action program (CAP)

as issue reports 335698 and 340314. Corrective action for this issue included revising

the charcoal purchase specifications to limit as-manufactured content of eight percent

by weight. Additionally, the licensee added limitations to model work orders to prevent

or limit the scheduling of work that could impact the operability of redundant systems

that contained charcoal. Because the licensee entered the issue into the CAP and the

finding is of very low safety significance, this violation is being treated as an NCV,

23 Enclosure

consistent with Section VI.A of the NRC Enforcement policy.

(NCV 05000451/2007-004-01)

.2 (Closed) LER 05000461/2007-003-00, IGSCC Causes RCS Pressure Boundary Leak

and Reactor Shutdown

a. Inspection Scope

The inspectors observed the station response to a steam leak in the drywell on

June 18, 2007. The inspectors arrived in the control room approximately two hours

after the initial alarms and indications of a leak were received and discussed the

actions that had been taken with control room operators. The inspectors then

proceeded to the outage control center, which had been manned for troubleshooting

and to support the control room in decision making, to discuss proposed actions with

licensee management. The inspectors reviewed the licensees troubleshooting

documents, equipment prompt investigation, root cause report, and LER 2007-003.

b. Findings

Introduction: A Non-Cited Violation of Technical Specifications (TS) having very low

safety significance (Green), was self-revealed when the licensee failed to meet the

required completion time for an action statement in TS 3.4.5. Specifically, TS 3.4.5

does not allow reactor coolant system pressure boundary leakage and requires a

shutdown to Mode 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months if pressure boundary leakage is discovered. Upon

entry into the drywell following a shutdown of the reactor on June 19, 2007, the

licensee discovered the existence of reactor coolant system (RCS) pressure boundary

leakage. Indications of the leakage had been discovered at 0433 on June 18, 2007,

but the plant was not placed in Mode 3 until approximately 31 hours3.587963e-4 days 0.00861 hours 5.125661e-5 weeks 1.17955e-5 months later at 1125 on

June 19, 2007.

Description: At 0433 on June 18, 2007, plant operators responded to main control

room alarms and other indications, and determined the presence of a steam leak in the

drywell. The initial alarms included: Transient Test Alarm, Fission Product Particulate

High Radiation, and Fission Product Iodine High. Other indications included transient

test system trouble, channel 119, main steam line C elbow tap differential pressure, an

increased trend in drywell pressure rise, and the fission product monitor particulate and

iodine channels had taken a step change. The operators took appropriate actions in

accordance with station procedures for reactor coolant leakage and abnormal release

of airborne radioactivity. The operators noted that a spike in main steam line C elbow

tap differential pressure was an indication of a steam leak somewhere in the area of

the elbow tap. The C main steam line elbow taps were located within the area

considered reactor coolant system pressure boundary. The operators entered TS 3.4.5

actions C.1, be in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and C.2, be in Mode 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The licensee manned the outage control center at 0641. After further deliberations, the

licensee noted that there were other components (mechanical joints for flexi-hose) that

are associated with the C main steam line flow instrumentation that are not considered

pressure boundaries and, therefore, the increase in drywall leakage was not

24 Enclosure

necessarily pressure boundary leakage. Based on this, the licensee then decided to

consider the steam leak as unidentified leakage. With the change in classification to

unidentified leakage the leak rates did not meet any of the requirements for remaining

in TS 3.4.5. The licensee exited action statements C.1 and C.2. However, although

small, the leakage continued to increase. Due to the steady increase in unidentified

leakage, operators commenced a normal plant shutdown at 2011 on June 18, 2007, to

enter the drywell and identify the actual location of the steam leak. At 0635 on

June 19, 2007, maintenance personnel entered the drywell and found pressure

boundary leakage on a one-inch diameter ASME Section III Class II stainless steel

braided flexible hose assembly on the C main steam line flow elbow tap. Operators

once again entered the actions of TS 3.4.5 due to reactor coolant pressure boundary

leakage. The plant entered Mode 3 at 1125 on June 19, 2007, (approximately 31 hours3.587963e-4 days 0.00861 hours 5.125661e-5 weeks 1.17955e-5 months

after indication of pressure boundary leakage), and Mode 4 at 2300 on June 19, 2007.

The licensee replaced the leaking flexible hose and initiated a root cause investigation.

The root cause determined that the leak was caused by intergranular stress corrosion

cracking of the flexible hose.

Analysis: The inspectors determined that failure to meet the required completion times

for TS actions was a performance deficiency warranting a significance evaluation. The

inspectors determined that the finding was more than minor by using Inspection Manual

Chapter 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening,

because operating with a degraded pressure boundary affected the reactor coolant

system (RCS) equipment and barrier performance attribute of the Barrier Integrity

cornerstone, in that, RCS pressure boundary leakage results in a reduction in the

reasonable assurance that physical design barriers protect the public from radionuclide

releases caused by accidents or events. The finding also affected the cross-cutting

component of Decision Making in the area of Human Performance in that operators

had initially entered TS 3.4.5 for pressure boundary leakage, but subsequently, the

organization decided not to treat the leak as pressure boundary leakage, but rather as

unidentified leakage until the actual location could be determined.

The inspectors completed a significance determination using IMC 0609, Significance

Determination Process. The Phase 1 worksheet directed the inspectors to consider

the reactor coolant system leakage under the Initiating Events Cornerstone column.

The inspectors answered Yes to question one under the Initiating Events cornerstone

column stating that assuming worst case degradation, the finding would result in

exceeding the TS limit for identified RCS leakage. The Phase 1 worksheet directed the

inspector to Phase II. For the Phase II screening, the inspectors used the table for a

small break loss of coolant accident, and conservatively set the initiating event

frequency to 1 (X=0), because the event was actually occurring in the form of the

steam leak. Since all of the mitigating equipment was available, the finding was

determined to be of very low safety significance (Green). The licensee established the

following corrective actions to address this issue: 1) replaced the leaking flexible hose,

2) scheduled replacement of all other flexible hoses used on the main steam lines

during the next refueling outage and 3) established a 16-year preventive maintenance

replacement frequency for the flexible hoses used on the main steam line.

Enforcement: Technical Specification 3.4.5 a. limits RCS operational leakage to

No pressure boundary leakage, and states that if pressure boundary leakage is

25 Enclosure

present, the plant is required to be in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and Mode 4 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Contrary to the above, on June 18, 2007, pressure boundary leakage was present and

the operators did not place the plant in Mode 3 for nearly 31 hours3.587963e-4 days 0.00861 hours 5.125661e-5 weeks 1.17955e-5 months and Mode 4 for

approximately 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months . Because this violation was of very low safety significance and

it was entered into the licensees corrective action program as IR 641375, this violation

was treated as a NCV, consistent with Section VI.A of the NRC Enforcement Policy.

The LER is closed. (NCV 05000461/2007004-02)

.3 Spent Fuel Bundle Incident During Rerack Project

Introduction: A performance deficiency involving a Non-Cited Violation of 10 CFR

Part 50 Appendix B, Criteria V, Instructions, Procedures, and Drawings, was

self-revealed following an event on August 17, 2007, where a spent fuel bundle being

moved to a temporary storage location came in contact with and rested upon another

fuel bundle seated in its storage location. The licensee procedure that governs spent

fuel pool movement failed to provide adequate guidance on how high to lift the fuel

bundle prior to traversing across the spent fuel pool.

Description: During spent fuel moves on August 17, 2007, in support of the spent fuel

pool re-rack project, a spent fuel bundle being moved by a spent fuel bridge operator

came in contact with the bail handle of a spent fuel bundle seated in its storage

location. Upon recognition of this situation the spent fuel bridge operator immediately

stopped. Immediately following the stoppage of the spent fuel bridge the licensee

observed that the fuel bundle was leaning slightly and the grapple-engaged light was

no longer lit. The grapple-engage light being no longer lit was an indication that the

grapple switch was no longer engaged and that an interlock was in place that

prevented the fuel bundle from being lifted vertically. Prior to resuming fuel movement,

the licensee verified that the grapple was indeed engaged by use of an underwater

camera. After this verification, the interlock associated with the grapple-engage light

was bypassed. The spent fuel bundle was then raised and placed back in its original

location.

A situation where a fuel bundle comes in contact with and rests upon another fuel

bundle is a concern because it challenges the integrity of the fuel cladding. The fuel

cladding is considered one of three physical barriers designed to separate the fuel from

the public and the environment.

Upon review of procedure CPS 3703.02,Fuel Handling Platform Operations, the

inspectors concluded that the procedure failed to provide adequate guidance on how

high the bundle should be lifted prior to moving the bundle to a new location. This

procedure was the administrative procedure which governs the operation of the bridge

and is used by the bridge operators to perform spent fuel pool movement. For example

step 8.2.2.15 requires lifting the bundle or blade until the grapple normal up light is lit

if moving the fuel bundle into the inclined fuel transfer system upender. The normal

up indication occurs at a reference vertical position (Z coordinate) of 0 inches. As

the fuel bundle is lowered from the normal up position, the Z coordinate increases.

Appendix A of the CPS 3703.02, showed several minimum fuel grapple Z coordinates

26 Enclosure

for various movements. The appendix showed that the fuel bundle must be at or above

Z position of 66 inches to enter the spent fuel pool racks.

The inspectors noted that a caution prior to step 8.2.2.15 states, it is not required to

raise the main hoist to normal up prior to platform or trolley movement, however

adequate clearance shall exist. The inspectors concluded that this step does not

contain adequate guidance on how high to lift the fuel bundle prior to moving it to

another location in the spent fuel pool. The inspectors concluded that the lack of

specific guidance caused this event to occur. Additionally, the procedure did not

require any form of verification or peer checks prior to fuel movement.

Analysis: The inspectors determined that the licensees failure to provided adequate

details in the procedure was a performance deficiency warranting a significance

evaluation. The inspectors concluded that the finding was more than minor in

accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, because

the finding affects the barrier integrity objective of assuring that physical design barriers

protect the public from radionuclide releases caused by accidents or events. The

finding is associated with the procedure quality attribute of the Barrier Integrity

cornerstone because it challenged the functionality of the fuel cladding.

The inspectors completed a Phase 1 significance determination using IMC 0609,

Significance Determination Process, Appendix A, Attachment 1, dated

March 23,2007. Using IMC 0609 Appendix A, SDP Phase 1 screening worksheet, the

inspectors determined that this issue degraded only the Fuel Cladding Barrier and its

associated cornerstone. Therefore, in accordance with the screening worksheet, this

issue screens directly as Green. This finding is related to a cross-cutting component

in the area of Human Performance associated with Resources (H.2(c)) because the

licensee did not provide complete and accurate procedures. Specifically, the

procedure relied on the skill of the operator and did not require independent

verification.

Enforcement: 10 CFR Part 50, Appendix B, Criteria V, Instructions, Procedures, and

Drawings states that activities affecting quality shall be prescribed by documented

instructions, procedures, or drawings, of a type appropriate to the circumstances and

shall be accomplished in accordance with these instructions, procedures, or drawings.

Contrary to the above, Clinton Power Station (CPS) procedure 3703.02 failed to

provide adequate instructions for handling spent fuel bundles. This resulted in a spent

fuel bundle that was being moved to a temporary storage location coming in contact

with another fuel bundle seated in its storage location on August 17, 2007. This issue

was entered into the licensees corrective action program as issue report 661918.

Corrective action included revising the fuel handling procedure. Because this violation

was of very low safety significance and was entered into the licensees CA program,

this violation is being treated as an NCV, consistent with Section VI.A.1 of the NRC

Enforcement Policy. (NCV 05000451/2007-004-03)

27 Enclosure

4OA4 Cross-Cutting Aspects of Findings

.1 A finding described in section 4OA3.2 of this report had, as its primary cause, a Human

Performance deficiency, in that, licensee decision making in determining the source of

a steam leak in the drywell resulted in exceeding the TS allowed time for placing the

plant in Modes 3 and 4 when RCS pressure boundary leakage was present.

.2 A finding described in section 4OA3.3 of this report had, as its primary cause, a Human

Performance deficiency, in that, the licensee failed to provide adequate procedural

guidance in regard to moving spent fuel in the spent fuel pool.

4OA6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to Mr. B Hanson and other members of

licensee management at the conclusion of the inspection on October 4, 2007. The

inspectors asked the licensee whether any materials examined during the inspection

should be considered proprietary. No proprietary information was identified.

.2 Interim Exit Meetings

Interim exit meetings were conducted for:

Occupational Radiation Safety program for radiation monitoring

instrumentation and protective equipment with Mr. F. Kearney, Plant Manager on

August 10, 2007.

Radioactive Material Processing and Transportation program with Mr. B. Hanson

on September 21, 2007.

ATTACHMENT: SUPPLEMENTAL INFORMATION

28 Enclosure

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

B. Hanson, Site Vice President

R. Kearney, Plant Manager

R. Schenck, Work Management Director

G. Vickers, Radiation Protection Director

J. Gackstetter, Regulatory Assurance Manager

R. Frantz, Regulatory Assurance Representative

M. Hiter, Access Control Supervisor

M. Friedmann, Acting Regulatory Assurance Director

C. VanDerburgh, Nuclear Oversight Manager

J. Domitrovich, Maintenance Director

D. Schavey, Operations Director

J. Rappeport, Acting Chemistry Manager

J. Lindsay, Training Manager

C. Williamson, Security Manager

R. Peak, Site Engineering Director

T. Chalmers, Shift Operations Superintendent

LIST OF ITEMS OPENED, CLOSED AND DISCUSSED

Opened

05000461/2007004-01 URI Shipment Total Quantity Re-characterized After

Shipping (Section 2PS2.3)

05000451/2007-004-01 NCV Inadequate Procurement Specification for

Charcoal Results in Inoperable Control Room

Ventilation Subsystem

05000451/2007-004-02 NCV Failure to Comply with Technical Specification 3.4.5 for RCS Pressure Boundary Leak

05000451/2007-004-03 NCV Inadequate Procedure Results in Spent Fuel

Bundle Incident

Closed

05000461/2007-003-00 LER IGSCC Causes RCS Pressure Boundary Leak

and Reactor Shutdown

05000461/2005-001-00 LER Inadequate Procurement Specification for

Charcoal Results in Inoperable Control Room

Ventilation Subsystem

1 Attachment

05000451/2007-004-01 NCV Inadequate Procurement Specification for

Charcoal Results in Inoperable Control Room

Ventilation Subsystem

05000451/2007-004-02 NCV Failure to Comply with Technical Specification 3.4.5 for RCS Pressure Boundary Leak

05000451/2007-004-03 NCV Inadequate Procedure Results in Spent Fuel

Bundle Incident

Discussed

None

2 Attachment

LIST OF DOCUMENTS REVIEWED

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

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

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

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

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

1R01 Adverse Weather

OP-AA-108-111-1001, Severe Weather and Natural Disaster Guidelines; Revision 2

OP-AA-102-102, General Area Checks and Operator Field Rounds; Revision 5

WC-AA-107, Seasonal Readiness; Revision 3

CPS 1019.05, Transient Equipment/Materials; Revision 11

CPS 4302.01, Tornado/High Winds; Revision 18e

1R04 Equipment Alignments

CPS 3319.01V001, Standby Gas Treatment Valve Lineup; Revision 8

CPS 3319.01E001, Standby Gas Treatment Electrical Lineup; Revision 10c

1R05 Fire Protection

CPS 1893.04M310, 719 Control: HVAC Equipment Area Pre-fire Plan; Revision 5a

Section 3.2.1.2, Fire Area A-2: Zone A-2a, Fire Protection Evaluation Report,

CPS 1893.01, Fire Protection Impairment Reporting, Revision 16a

CPS Updated Final Safety Analysis Report, Fire Hazard Analysis

CPS 1892.01, Fire Protection Impairment Reporting, 08/02/07, Revision 16b.

Section 6.2.6.3, Updated Safety Analysis Report

ANSI 56.8, 1987, Containment System Leakage Testing Requirements

Reg. Guide 1.163, Performanced - Based Containment Leak - Test Program

10 CFR 50 App. J., Primary Reactor Containment Leakage Testing for Water - Cooled Power

Reactors

CPS - ITS

CPS Risk Evaluation #1631, Risk Analysis for Missed Surveillance, failure to complete LLRT

on Containment Monitoring Test Lines

IR 660299, Type C Testing for LLRT Test Lines

1R06 Flood Protection

CPS 4304.01, Flooding, Revision 4e

IR 670686, Setpoints For Maximum Normal Operating Water Level Switches

IR 670693, Flood Water Level & Secondary Containment Temperature Instrumentation

IR 670429, Max Safe Flood Alarm Out of Cal in RH [residual heat removal] A Room

IR 670415, Max Safe Flood Alarms Out of Cal in RI Pump Room

1R07 Heat Sink Performance

WO 1005225, Open, Inspect, Boroscope, Eddy Current 100%, and Clean 1DG13A

ER-AA-340-1002, Service water heat exchanger and component inspection guide; Revision 3

IR 678934, Division three diesel generator heat exchanger chemistry results; October 2, 2007

IR 675035, Diesel generator jacket water cooler zinc modification execution impact;

September 24, 2007

3 Attachment

IR 675091, Excessive erosion discovered on the flange faces of 1VH03A; September 25, 2007

WO 01005225, Open, Inspect, Boroscope, Eddy Current 100%, and Clean

AR 00678934, Division II Diesel Generator HX (1DG13A) Chemistry Results

AR 00675035, Diesel Generator Jacket Water Cooler Zinc Mod Execution Impact

ER-AA-340-1002, Service Water Heat Exchanger and Component Inspection Guide,

Revision 3

1R12 Maintenance Effectiveness

IR 518197, NOS ID MSPI Basis Document Requires Revision

IR 519897, Potential Air Leak on 1VQ002 or 1VQ005

IR 520828, CPS 4001.02C001 Setpoint Update Needed

IR 524365, Automatic Reactor Scram on Reactor High Water Level

IR 524768, RCIC Isolation Troubleshooting Results

IR 552156, Condition Monitoring Failure - LD Inst 1E31N085A - RCIC Isolation

IR 550868, 1SX01PA: Low Margin for Division I SX Design Flows

IR 548493, 1C91P633: STS Failure, 5004-3H, RCIC Division 1 Card C-A14-A125

IR 530970, Invalid Assumption in Calculation

IR 556556, Unexpected Annunciators during STS Summary Card B-A11-A111

IR 557804, Evaluate Transmitter 1E31N085B Time Delay Settings

IR 557809, Received Annunciator 5004-3H, STS Failure

IR 648094, 1E51-F054 RCIC Drain Pot Bypass Valve has Small Packing Leak

IR 639727, RCIC Pump Outboard Bearing Bubbler Crosshairs not @ 90 Degrees

IR 631320, NRC Question: Impact of EC 365827 Changes on Close Time Test

IR 589210, RCIC Speed > 4600 RPM during Surveillance Testing

IR 589461, Found RCIC Turbine & Governor Oil Below the Standby Band

IR 627033, 1E51-N590: New RGSC Card Cat ID 1148300-01 Failed Testing

IR 587482, 1E51F046 Potential MOV Over Thrust Risk

IR 584757, 1E51N590: RCIC RGSC Module Replacement & Procedure Enhance

IR 562400, 1 Foot Steam Plum Leak at Base of 1E51N010

IR 555469, 1E51N010: 5063-2C Turbine Stm Line Wtr Drain Trap Level High

IR 555312, Crack Found in Housing for Level Switch during WO 924362

IR 535878, Minor Steam Leak under Insulation of RCIC Turbine 1E51-C002

IR 531654, RCIC Quick Start Surv CPS 9054.01C004

IR 531065, RCIC Turbine Stayed at Incorrect Speed during 9054.01C004

IR 529655, NOS IDD RCIC Restoration with out of Cal Pressure Transmi

IR 528901, IMD Lessons Learned form 1E31N085A Replacement

IR 524345, RCIC Division 1 Isolated Following Scram

IR 517311, Revise ARPs 5062-3D & 5063-2E and 3310.01

IR 517453, New Fittings for RCIC Vortex Mod does not Match Design

IR 517558, Mini-PM for RCIC Turbine 1E51-C002 - C1R11

IR 517768, Feasibility Study for RPV Head Piping

IR 517775, Security Required for RCIC Tank Control Sooner than Planned

IR 517874, TS 3.3.3.2, Remote Shutdown Instrumentation, Enhancement

IR 518522, E51-N655F Failed Initial +/- .25% Input/Output Comparison

IR 518849, Heater Support Stands not secured to Tank

IR 518932, Free Release of Transmitters

IR 519207, Transposition Error for RPM Values - RCIC Turbine Speed

IR 519208, 1TR-CM018 Points above Alarm Setpoint with no Alarm Leds Lit

IR 519566, Items Discovered on Carts not Properly RAM Tagged/Controlled

4 Attachment

IR 519776, Red Hose Found by RCIC Tank

IR 520032, NOS ID - Unavailability Baseline Data Errors in MSPI Basis Doc

IR 520069, Work Week Preps Did Not Identify Issues

IR 520293, NRC Question on MSPI Implementation

IR 520803, Contamination Found on Floor in RCIC Tank Valve Room

IR 520906, TS LCO 3.3.5.1/3.3.5.2 Logical Connections Need Indenting

IR 522049, NSRB Issue - Unsecured Cables at Contaminated Area Boundary

IR 522119, NOS ID - Deficiencies Found During NSRB Tour

IR 523330, NRC Identified MSPI Basis Document Error for HP

IR 524547, Possible Training Process Problem

IR 524893, Poor Decision Making in PORC Meeting

IR 525078, Actions Required RCIC High Steam Flow Time Delay EC 362261

IR 526991, Correction Needed for IR 518932 Response

IR 528013, New Card Received Damaged from Stores

IR 528655, NOS ID Line Not Properly Controlled

IR 528787, RP Work Practices Need Improvement

IR 529288, Enhancements to RCIC Restoration Procedure

IR 529423, NOS ID Technical Rigor Weakness for RCIC OpEval 524768-02

IR 529475, Revise Tornado Missile Hazard Analysis

IR 529411, Change to 9054.01C004, RCIC Quick Start Surveillance

IR 529718, Process Improvements in Status Control

IR 531090, Enhancements: Clarify Tech Spec Bases for LCO 3.5.3 RCIC

IR 531215, RCIC Procedure Lacks Steam Line Draining Steps

IR 531392, Ramp Generator Card Received with Jumper not used at CPS

IR 531410, FLS Provided Inaccurate Estimate for First Time Performance

IR 531412, RCIC Ramp Generator Card in Repair Program

IR 537945, Support for Estimating Flows for Mitigating Strategies

IR 539389, Trng: Simulator: Simulator Responses to Evaluated Scenario

IR 541955, RCIC Isolations Incorrectly Referenced in CPS 9532.16

IR 543371, Potential Trend in Equipment Failure Cause Identification

IR 543520, Evaluate Common Causes of RCIC System Issues

IR 543638, NOS-ID Appropriate Cause Evaluation not Assigned

IR 543661, Enhancement IR to Revise 9080.23

IR 546682, Trng: Drawing Reference is Incorrect

IR 553475, 1LL59BP18E: Electrolyte Level below Plates on Two Cells

IR 554669, B.5.B Phase 2 and 3 Closure Actions

IR 555735, Expectation not Met as Required by CPS Policy #54

IR 556474, Ownership of 3209.01 Raw Water Treatment

IR 557274, E-2 Ready to Work Exceptions not in Goal for WW645

IR 557833, RCIC Vent Pipe Leaks at Sampling Point with Sample Equipment

IR 558269, Fix RCIC Sampling

IR 559377, NOS ID Separation Criteria not Maintained During EC Install

IR 560786, NOS IDd Gaps in Technical Rigor Used in OP Determinations

IR 561779, 9030.01C034 & C035 Missing an Expected Alarm

IR 562215, Security Tour Improvement Opportunities

IR 562813, Procedural Conflict for RCIC Instrument Valve Position

IR 563344, 1E51F378 has Minor Packing Leak

IR 563348, Minor Oil Leaks on Oil Sightglass & Govenor Assembly

IR 563632, Replace LD Transmitters to Prevent RCIC Isolation

5 Attachment

IR 563745, Generate WO for LD Transmitter in Relation to CR-563632

IR 565444, NRC Non-Cited Violation 2006-11-02, RCIC Suction Vortexing

IR 567532, RCIC Valve 1E51-337 with Lockwire

IR 569112, ORM NTSP Value Incorrect for RCIC Storage Tank Level Low

IR 570095, NOS ID Project Plan for PI&R Inspection not Developed

IR 570783, Cause Analysis for HPCS Vortex Issue

IR 574514, Inappropriate Assignment of Corrective Actions

IR 579912, C1R11 Surv Opt HIT - 9438.05 Enhancements

IR 582944, Non-Oily Sheen on Top of Suppression Pool

IR 583348, Clean Up of Cask Wash Down Pit

IR 585490, Black Marker Found Floating in Suppression Pool

IR 588801, Procedure Enhancement to 9030.01C34

IR 588802, Procedure Enhancement to 9030.01C035 RI MS Sup Press

IR 588886, Enhance to 9054.02

IR 589228, 1RI01T: Notice of Violation (White Finding) Related to HPCS

IR 590313, Division 1 ADS Backup Air Bottles Losing 25-50 # per Day

IR 596320, Perform ISI/NDE and Provide Support Work Pre C1R11

IR 597183, Trng - Revision Needed to Operations Strategies Document

IR 599370, Install Walkway across Berm at the RCIC Tank

IR 599388, Install Sidewalk to RCIC Tank

IR 599404, Install Gate on Existing Walkway

IR 599942, Procedure Review/Enhancement for Remote Shutdown Actions

IR 599989, RCIC TTV Gimpel Valve Discontinuation Future Material

IR 600927, RCR Identifies Weakness in OP-AA-108-115/OP-AA-106-1006

IR 600993, INPO CDE Website, MSPI PRA Data

IR 603347, NOS ID Post August 2006 Organization Weakness not Addressed

IR 603659, NOS ID Inadequate Appl of Standards during Work Activities

IR 604716, 1E51-N636A as Found DAC Values OOT

IR 608092, Trng - Changes to Transient Mitigation Needed

IR 608292, 2007 NRC PI&R Observation on Use of OPEX

IR 610762, Work Orders Removed from Week 0719 Parts

IR 610943, PMRQ for RV 1E12-F036 Specifies Incorrect Testing and Freque

IR 613782, 9027.01C007 Section 8.7.7 Challenges SX Loads

IR 613800, Work Orders Removed from Week 0719

IR 615740, NOS ID Stated Apparent Cause was not addressed in EOC

IR 617461, RCIC Isolation Bypass Switch Operations in not Crisp

IR 618123, E-4 Clearance Indicator not at Goal of 100%

IR 602553, RCIC Transmitters not Available for SOW

IR 602760, Perform Inspection of Component Supports and Support

IR 602867, Procedure Enhancement for Filling RCIC Storage Tank

IR 618251, Fuse Missing form Interlock for Doors 329 and 330

IR 618423, Trng - Enhancement to 3310.01 RCIC

IR 619749, Tornado Missile Found in the RCIC Berm

IR 621632, NOS ID Trend Codes for CAP Products not Consistency Entered

IR 622636, Enhancement to ECCS Availability Due to WL Pump Availability

IR 622674, Missed Opportunity: Late Cancellation of a Work Order

IR 623199, RAT SVC Tripped Resulting in Unplnd Entry into 72 hr SD LCO

IR 623608, RCIC System Outage Questions

IR 623611, CDBI FASA Â Configured Design Specification not Updated

6 Attachment

IR 624129, Use of Admin Controls during RCIC Surveillance

IR 624937, WO 19406 Removed from Week 0719 due to Parts

IR 625517, MSPI Related Work Windows are not Consistently Scheduled

IR 626291, A TDRFP Oil Leak at 1PSFW105

IR 626496, RHR Unprojected Unavailability SSPI/MSPI Impact

IR 626577, Clearance Program Tag Font Remains Ambiguous/Unclear

IR 626578, Safety - Cannot Access Scaffold #5465 Safely

IR 627067, 1C88N2403: TT Point 30 Failed Calibration

IR 627527, 1E51N003: Computer Point E51DA001 Out of Spec Low (OOT)

IR 627710, Enhancement to ARP 5063-2D

IR 628094, 1C88K607: Received Spurious Trip of TTHMI Server Channel 119

IR 628259, CDBI FASA Enhancement to Loss of AC Off Normal 4200.01

IR 629191, CDBI FASA - Check Valve 1E22-F016 Not in ATLAS

IR 630647, 1RIX-AR013 Alert Alarm due to Spike

IR 632033, PMS Requires Valve Cycling in D/W and STM Tunnel

IR 632043, LL on LLRT Type C Test

IR 632321, Enhancement to CPS 3315.02, Leak Detection (LD)

IR 632785, CDBI FASA: Leakage to RCIC Tank not Updated in Alt. ST Docs

IR 634385, Placement of Red/Green Dots in the MCR

IR 634633, INPO Walk Downs List of Discrepancies

IR 635283, 1VX14S Air Gap Between Housing and Cooling Coils 1VX14AA/AB

IR 636156, NRC Performance Indicator for SSFFS in Action Region

IR 636753, Chemistry Sample Results for 1VY04A RCIC Room Cooler

IR 636859, 9030.01C007 Enhancement

IR 637549, Trng - Procedure Enhancement 3310.01 RCIC

IR 638917, 1C71S001C: Unexpected Alarms Division 3 NSPS Inverter

IR 641088, No Engineered Test Points for Jumper/Test Leads in 9532.13

IR 644916, Set Test Condition for RI Min Flow Valve Therm. Overl. Test

IR 647267, 1C91P633: Received 5004-3H STS Failure

IR 648508, 3220.01 Requires Revision to Support C1R11

IR 649191, Procedure Step Needs Revision Regarding Bypass Switch

IR 652248, CAPR not Implemented as Worded

IR 654201, NOS ID: Trending of PAR Data for Common Causes

IR 616603, 1E51N052: Unexpected MCR Alarm During 9054.06

IR 654764, NRC Questions RCIC Turbine Inboard Bearing Oil Level

IR 627562, 1E51F064: Abnormal Voltage Indicated During Surveillance

IR 627546, 9861.02D015 Test Set A had Leakage in Excess of 20,000 SCCM

ER-AA-310, Implementation of the Maintenance Rule, Revision 6

ER-AA-310-1001, Maintenance Rule - Scoping, Revision 3

ER-AA-310-1002, Maintenance Rule - SSC Risk Significance Determination, Revision 2

ER-AA-310-1003, Maintenance Rule - Performance Criteria Selection, Revision 3

ER-AA-310-1004, Maintenance Rule - Performance Monitoring, Revision 5

ER-AA-310-1005, Maintenance Rule - Dispositioning Between (a) (1) and (a) (2), Revision 5

ER-AA-310-1006, Maintenance Rule - Expert Panel Roles and Responsibilities, Revision 3

ER-AA-310-1007, Maintenance Rule - Periodic (a) (3) Assessment, Revision 4

Performance Criteria Report for RI System

Failure Report for RI System

Scoping/Risk Significance Detailed Report for RI System

7 Attachment

Assessment of Maintenance Effectiveness; 10CFR50.65 (a) (3) Assessment; Clinton Power

Station; 3/1/2004 to 3/1/2006.

CPS Technical Specifications: TS 3.8.4, 3.8.5, and 3.8.6 and associated bases

USAR Section 8.3.2

Vendor Manual K2989-0001, Power Conversion Products Battery Chargers

IR 546815, Division 2 DC Bus Voltage Fluctuations

IR 651865, Potential Adverse Trend on Division 4 DC Bus Voltage

CPS System DC Equipment Failure Report for July 2005 to July 2007

1R15 Operability Evaluations

IR 671001, Essential switchgear cooling C shutdown service water line thickness below

minimum screening criteria; September 13, 2007

EC 367363, Minimum wall calculation for reducer small ends for 3"X1.5" swg at valve

1SX025C on line 1SX23AC-3", Revision 0

IR 667633, Through body leak 1SX019B; September 4, 2007

IR 667865, Issues identified during disassembly of 1SX019B valve; September 5, 2007

IR 667974, LL valve has through body leak after decision to defer repair; September 5, 2007

IR 668216, ASME code case N-513-1, Augmented volumetric exam of 1SX20AA;

September 5, 2007

IR 668217, ASME code case N-513-1, augmented volumetric exam of 1SX23BA;

September 5, 2007

IR 668220, ASME code case N-513-1, augmented volumetric exam of 1SX23AB;

September 5, 2007

IR 668221, ASME code case N-513-1, augmented volumetric exam of 1SX23AC;

September 5, 2007

IR 668223, ASME code case N-513-1, augmented volumetric exam of 1SX04AC;

September 5, 2007

IR 668304, Division 3 control room ventilation chiller shutdown service water supply/return

vacuum breakers leaking; September 6, 2007

IR 670094, Division 3 shutdown service water ultrasonic testing results below screening

criteria; September 11, 2007

IR 670193, Essential switchgear cooling A shutdown service water piping ultrasonic testing

results below screening criteria; September 11, 2007

IR 670479, Isolated points on shutdown service water A piping less than minimum evaluation

criteria; September 12, 2007

EC 367242, Temporary acceptance of pitting of the VC-B head pressure controller valve body,

1SX019B; Revision 0

IR 666899, 1SX019B: Cavitation damage found in valve body; August 31, 2007

IR 658234, NRC Questions on OPEVAL 655836-02; August 07, 2007

1R19 Post Maintenance Testing

IR 671554, 0VC03CA fan replacement post job review - WW 0736; September 14, 2007

EC 366835, Modification to replace supply fan 0VC03CA; Revision 0

WO 969991, 0VC03CA -VC A Supply Fan Noise; September 9, 2007

IR 675847, Cut O-ings; September 26, 2007

IR 675914, Division 3 emergency diesel generator turbocharger drain tubing missing;

September 26, 2007

IR 676864, Out of specification parameters on Division 3 diesel generator during surveillance

run; September 28, 2007

8 Attachment

IR 676905, K1 Lockout relay failed to trip emergency diesel generator; September 28, 2007

WO 795653, Replace Normally deenergized relay 1E22S001B-8; September 28, 2007

WO 1005265, Division 3 diesel generator six year maintenance: September 28, 2007

CPS 9080.03, Diesel generator 1C operability - Manual quick start operability; Revision 28e

CPS 3316.01,

CPS Operations Narrative Logs for August 14, 2007

WO 909977, Replace Rosemount Transmitter

WO 318466, EM Troubleshoot/Repair 1DC07E

1R22 Surveillance Testing

CPS 9054.01D002, RCIC (1E51 - C001) High Pressure Operability Checks Checklist,

Revision 23a

CPS 9054.01C002, RCIC (1E51 - C001) High Pressure Operability Check, Revision 2b

CPS 9054.01, RCIC System Operability Check, Revision 42e

IR 657976, Unexpected Alarm 5063 -1D RCIC Pump Suction Press

IR 658065, 1E51-F059 Limit switch Didnt Operate Properly During 9054.01

IR 627527, 1E51N003: Computer Point E51DA001 Out of Spec Low (OOT)

IR 658055, RCIC Turbine Steam Disch Press Gauge Reads Zero #

CPS 9052.01, LPCS/RHR A Pumps & LPCS/RHR A Water Leg Pump Operability, 09/13/06;

Revision 43d,

IR 655830, 1E21F349A Unquantifiable Seat Leakage

IR 655831, 1E21F350A Unquantifiable Seat Leakage

IR 655833, 1E21R501, Gauge Overanged During 9052.01

IR 655836, 1E12F084A, WLP Check To RHR A, failed to close

IR 655838, 1E12F085A, WLP to RHR a, failed to close

IR 658200, Senior Resident Has Question on RCIC Alarm; August 07, 2007

Clinton Power Station Updated Final Safety Analysis Report

CPS Technical Specification Surveillance Requirement 3.8.1.3 and associated bases

CPS 9080.02, Diesel generator 1B Operability - Manual and Quick Start Operability;

Revision 47d

IR 660893, NRC Questioned KVAR Loading on 9080.02 Division 2 Diesel; August 15, 2007

CPS 9069.01, Shutdown Service Water Operability Test; Revision 45

CPS Technical Specification 3.7.1/3.7.2 and associated bases

CPS 9861.09D007, Leakage Test on Valve 1CC075B and 1CC076B, Revision 1a

2PS2 Radioactive Material Processing and Transportation

USAR Chapter 11 Solid and Liquid Radwaste Management System, Revision 11

2006 Annual Radioactive Effluent Release Report; dated April 27, 2007

AR 435702; Chemistry, Radwaste, Effluent and Environmental Monitoring Audit Report Audit

NOSA-CPS-06-04; dated April 19, 2006

AR 302506; Contamination Found in a Twenty Foot Sea Van; dated February 17, 2005

AR 309873; Incorrect Dose Rate Listed in Waste Shipment Notification; dated March 8, 2005

AR 345179; Legacy Filter Waste Stores Without an Inventory; dated June17, 2005

AR 354609; Laundry Trailer Found with Small Hole in Flooring; dated July 19, 2005

AR 430473; FlatBed Trailer Separated from Semi-Tractor While Moving; dated

December 5, 2005

AR 451440; Incomplete Departure Survey Documentation; dated February 8, 2006

AR 476188; Radioactive Shipping Documentation Errors; dated April 7, 2006

AR 539967; Inability to Transfer Waste Sludge to Vendor; dated October 4, 2006

9 Attachment

AR 569198; Radiation Protection Audit Report; dated September 5, 2007

AR 577885; 0WX02TA:Lessons Learned On 10 CFR 61 Analysis; dated January 11, 2007

AR 578765; Invalid Data initially Generated; dated January 12, 2007

AR 585131; Deficiencies in Shipping Documents Found; dated January 30, 2007

AR 617139; Radwaste Vendor Processing Creates Elevated Dose Rates in Walkway; dated

April 14, 2007

AR 626907; Water on Floor of Sealand Number 59; dated May 8, 2007

AR 643431; Fluid Found on Incoming Radioactive Material Shipment; dated June 22, 2007

AR 661628; Rotor Cribbing Found Deteriorating More Than Last Inspection; dated August 17,

2007

AR 672300; Approximately 50 Milliliters Leak From Sealand Onto Trailer; dated

September 17, 2007

ASSA 563448; Transportation and Radwaste Self-assessment; dated July 20, 2007

ASSA 581780-04; detailed Review of 2004 Through 2006 Shipping Records; dated

January 31, 2007

RP-AA-100; Process Control Program For Radioactive Wastes; Revision 4

RP-AA-600; Radioactive Material/Waste Shipments; Revision 10

RP-AA-602; Packaging of Radioactive Material Shipments; Revision 12

RP-AA-603; Inspection and Loading of Radioactive Material Shipments; Revision 3

RP-CL-605-1001; CPS 10 CFR Part 61 Program; Revision 1

CPS 3909.01; Operating Spent Resin System; Revision 21

CPS 3909.02; Operating Phase Separators; Revision 20e

CPS 3909.03; Operating Waste Sludge System; Revision 20d

CPS 3909.04; Operating Concentrate Waste System; Revision 15b

CPS 3909.05; Operating Fuel Pool Filter Demineralizer Sludge System; Revision 12b

CPS 6418.04; Analysis of Radwaste Samples for Solid Waste Processing; Revision 9b

CY-AA-110-200; Sampling; Revision 5

CPS 3222.10; Reactor Sample Station; Revision 10

2OS1 Access Control to Radiologically Significant Areas

AR289955; NOSA-CPS-05-06; Health Physics Functional Area Nuclear Oversight Audit; dated

July 27, 2005

AR 537844; Purple Painted tool Found Outside the Radiologically Restricted Area; dated

August 29, 2006

AR 575308; Building wall Damaged in Rotor Storage Building; dated January 4, 2007

AR 582546; 1DR1-142 Door Hanging-up; dated January 24, 2007

AR 601273; High Radiation Area Near Miss Event; dated March 8, 2007

AR 633042; Emergent High Radiation Area Access; dated May 23, 2007

AR 630925; Radiography Inspection Terminated Due To Boundary Dose Rates; dated

May 16, 2007

RP-AA-210; Dosimetry Issue; Usage and Control; Revision 10

RP-AA-460; Controls for High and Very High Radiation Areas; Revision 12

RP-AA-460-101; Radiological Key Control and Area Access Requirements; Revision 3a

RP-AA-460-1001; Additional High Radiation Exposure Control; Revision 2

RP-AA-460-1002; High Radiation Area and Locked High Radiation Areas Briefing Form

RWP 10002868; Spent Fuel Pool Re-Rack Diving; Revision 0

RWP 10007956; Spent Fuel Pool Re-Rack; Revision 3

10 Attachment

OA2 Identification and Resolution of Problems

OP-AA-102-103; Operator Work-Around Program; Revision 1

CPS 9000.10; Accident Monitoring And Remote Shutdown Instrumentation Log; Revision 31a

IR 610835; Eval. Division 1 & 2 H2O2 Monitor INOP Printers: OPS Workaround; March 30,

2007

IR 624538; NOS ID Operations Workaround Leads To Fire Door Impairment; May 2, 2007

IR 575009; Permanent Access Solution Needed For 1FW004 Valve; January 3, 2007

IR 617958; Off-Normal Response Enhancement; April 16, 2007

IR 572918; Eval. Main Turbine BRG #9 Vibe Problem As Operator Challenge; December 25,

2006

IR 567237; 1KYCP105H Repeat Timer Did Not Advance - OP Challenge; December 9, 2006

IR 642744; TDRFP B Turning Gear Failed To Disengage Automatically; June 21, 2007

IR 652502; 1HG02CA Greater Than 3 Hours Run Time In July 2007; July 22, 2007

IR 656515; Missed Opportunity For Operator Work Around/Challenge Review; August 2, 2007

WO 843352; Task 01; Contingent Task To Disassemble, Inspect And Repair Valve 1E12F009

Due Seat Leakage

OA3 Event Followup

LER 05-461/2007-003-00, IGSCC Causes Pressure Boundary Leak and Reactor Shutdown;

August 16, 2007

IR 641375, 4001.01 Reactor Coolant Leakage; June 18, 2007

RCR 641375, Flexible hose failure on Main Steam Line C results in manual reactor shutdown

for replacement

11 Attachment

LIST OF ACRONYMS USED

ADAMS Agency wide Documents Access and Management System

ALARA As-Low-As-Reasonably-Achievable

CPS Clinton Power Station

CR Condition Report

DAW Dry Active Waste

DOT Department of Transportation

EP Emergence Preparedness

ERAT Emergency Reserve Auxilary Transformer

FPER Fire Protection Evaluation Report

HPCS High Pressure Core Spray

HRA High Radiation Area

HVAC Heating Ventilation and Cooling

IMC Inspection Manual Chapter

IR Issue Report

LCO Limited Condition for Operation

LPCS Low Pressure Core Spray

LSA Low Specific Activity

MR Maintenance Rule

MSPI Mitigating System Performance Indicator

NCV Non-Cited Violation

NEI Nuclear Energy Institute

NRC Nuclear Regulatory Commission

OPC Operational Support Center

OSC Operational Support Center

PARS Publicly Available Records

PI Performance Indicator

PMT Post Maintenance Testing

Radwaste Radioactive Waste

RCS Reactor Coolant System

RHR Residual Heat Removal

RP Radiation Protection

RPT Radiation Protection Technician

RWP Radiation Work Permit

SDP Significant Determination Process

SVC Static VAR Compensator

TS Technical Specification

URI Unresolved Item

USAR Updated Safety Analysis Report

WAB Work Around Board

WO Work Order 12 Attachment

ML073100335

Text

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