Text

IR 05000313-03-002 & IR 05000368-03-002, on 12/29/02 - 03/22/03; Entergy Operations, Inc.; Arkansas Nuclear One, Units 1 and 2; Evaluations of Changes, Tests, or Experiments; Temporary Plant Modifications; ALARA Planning and Controls
	ML031120115
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/21/2003
From:	Laura Smith; NRC/RGN-IV/DRP
To:	Anderson C; Entergy Operations
References
	IR-03-002
	Download: ML031120115 (37)
	v • d • e

April 21, 2003

Craig G. Anderson, Vice President,

Operations

Arkansas Nuclear One

Entergy Operations, Inc.

1448 S.R. 333

Russellville, Arkansas 72801-0967

SUBJECT: ARKANSAS NUCLEAR ONE, UNITS 1 AND 2 - NRC INTEGRATED INSPECTION

REPORT 50-313/03-02; 50-368/03-02

Dear Mr. Anderson:

On March 22, 2003, the NRC completed an inspection at your Arkansas Nuclear One, Units 1

and 2, facility. The enclosed report documents the inspection findings, which were discussed

with you and other members of your staff on April 3, 2003, and as described in Section 4OA6.

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

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

licenses. Within these areas, the inspection consisted of selected examination of procedures

and representative records, observations of activities, and interviews with personnel.

Based on the results of this inspection, the NRC has identified issues that were evaluated under

the risk significance determination process as having very low safety significance (Green) as

well as one issue that required evaluation using our traditional enforcement process. The NRC

has also determined that violations are associated with these issues. These violations are

being treated as noncited violations (NCVs), consistent with Section VI.A of the Enforcement

Policy. These NCVs are described in the subject inspection report. If you contest the violations

or significance of these NCVs, you should provide a response within 30 days of the date of this

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

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

Administrator, U.S. Nuclear Regulatory Commission, Region IV, 611 Ryan Plaza Drive,

Suite 400, Arlington, Texas 76011; the Director, Office of Enforcement, U.S. Nuclear

Regulatory Commission, Washington DC 20555-0001; and the NRC Resident Inspector at the

Arkansas Nuclear One facility.

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

enclosure and your response (if any) will be made available electronically for public inspection

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

of NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at

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

Entergy Operations, Inc.

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Should you have any questions concerning this inspection, we will be pleased to discuss them

with you.

Sincerely,

/RA/

Linda Joy Smith, Chief

Project Branch D

Division of Reactor Projects

Dockets:

50-313

50-368

Licenses:

DPR-51

NPF-6

Enclosure:

NRC Inspection Report

50-313/03-02; 50-368/03-02

cc w/enclosure:

Executive Vice President

& Chief Operating Officer

Entergy Operations, Inc.

P.O. Box 31995

Jackson, Mississippi 39286-1995

Vice President

Operations Support

Entergy Operations, Inc.

P.O. Box 31995

Jackson, Mississippi 39286-1995

Manager, Washington Nuclear Operations

ABB Combustion Engineering Nuclear

Power

12300 Twinbrook Parkway, Suite 330

Rockville, Maryland 20852

County Judge of Pope County

Pope County Courthouse

100 West Main Street

Russellville, Arkansas 72801

Winston & Strawn

1400 L Street, N.W.

Washington, DC 20005-3502

Entergy Operations, Inc.

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Bernard Bevill

Radiation Control Team Leader

Division of Radiation Control and

Emergency Management

Arkansas Department of Health

4815 West Markham Street, Mail Slot 30

Little Rock, Arkansas 72205-3867

Mike Schoppman

Framatome ANP, Inc.

Suite 705

1911 North Fort Myer Drive

Rosslyn, Virginia 22209

Technological Services Branch

Chief

FEMA Region VI

800 North Loop 288

Federal Regional Center

Denton, Texas 76201-3698

Entergy Operations, Inc.

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Electronic distribution by RIV:

Regional Administrator (EWM)

DRP Director (ATH)

DRS Director (DDC)

Senior Resident Inspector (RLB3)

Branch Chief, DRP/D (LJS)

Senior Project Engineer, DRP/D (JAC)

Staff Chief, DRP/TSS (PHH)

RITS Coordinator (NBH)

B. McDermott (BJM)

ANO Site Secretary (VLH)

Dale Thatcher (DFT)

Regional State Liaison Officer (WAM)

ADAMS: Yes

No Initials: ______

Publicly Available Non-Publicly Available

Sensitive

Non-Sensitive

R:\\_ANO\\2003\\AN2003-02RP-RLB.wpd

RIV:RI:DRP/D

SRI:DRP/D

PE:DRP/D

SPE:DRP/D

C:DRS/PSB

KDWeaver

RLBywater

ELCrowe

JAClark

TWPruett

T - LJSmith

E - LJSmith

MPShannon for

4/21/03

C:DRP/D

LJSmith

/RA/

4/21/03

OFFICIAL RECORD COPY

T=Telephone E=E-mail F=Fax

ENCLOSURE

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Dockets:

50-313, 50-368

Licenses:

DPR-51, NPF-6

Report No:

50-313/03-02; 50-368/03-02

Licensee:

Entergy Operations, Inc.

Facility:

Arkansas Nuclear One, Units 1 and 2

Location:

Junction of Hwy. 64W and Hwy. 333 South

Russellville, Arkansas

Dates:

December 29, 2002, through March 22, 2003

Inspectors:

R. Bywater, P.E., Senior Resident Inspector

J. Clark, Senior Project Engineer, Project Branch D

R. Lantz, Senior Emergency Preparedness Inspector

L. Ricketson, P.E., Senior Health Physicist

K. Weaver, Resident Inspector

Approved By:

Linda Joy Smith, Chief, Project Branch D

Division of Reactor Projects

Attachment 1:

Attachment 2:

Supplemental Information

ANO Paper, "Secondary System Pressure Boundary as an Extension of

Containment Liner"

SUMMARY OF FINDINGS

Arkansas Nuclear One, Units 1 and 2

NRC Inspection Report 50-313/03-02; 50-368/03-02

IR05000313-03-02, IR05000368-03-02; Entergy Operations, Inc.; 12/29/02 - 03/22/03; Arkansas

Nuclear One, Units 1 and 2; Evaluations of Changes, Tests, or Experiments; Temporary Plant

Modifications; ALARA Planning and Controls.

The inspection was conducted by the resident inspectors, one senior project engineer, a senior

health physicist, and a senior emergency preparedness inspector. The inspection identified three

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

using 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 Findings

Cornerstone: Mitigating Systems

Green. The licensee did not properly evaluate a temporary alteration that was performed

when a door separating a safety-related switchgear room from the turbine building was

removed for maintenance. As a result, the impact of a potential high energy line break on

equipment needed to mitigate the event was not identified or evaluated by an engineering

evaluation. Failure to perform an engineering evaluation to support this temporary alteration

was a violation of Unit 1 Technical Specification 5.4.1.a. This violation is being treated as a

noncited violation (NCV) consistent with Section VI.A in the Enforcement Policy.

The safety significance of this issue was determined to be very low since this issue screened

as Green during a Phase 1 SDP assessment, because the finding did not result in

equipment becoming incapable of performing its function in the case of a design basis

accident. The issue was considered to be more than minor because it affected the mitigating

systems cornerstone objective for design control and modifications because the ability to

mitigate the consequences of a high energy line break would have been affected if the

finding had affected more than one train of equipment (Section 1R23).

Cornerstone: Barrier Integrity

Severity Level IV. The inspectors identified a noncited violation of 10 CFR 50.59 because

the licensee failed to identify that changes made to the Units 1 and 2 Updated Safety

Analysis Reports required a license amendment request. These changes removed

containment isolation valve controls for secondary system containment penetrations. The

licensee initiated corrective action on March 28, 2003, to prepare a license amendment

request to obtain NRC approval of the changes to the Updated Safety Analysis Reports.

This is an item for traditional enforcement because it involves an issue not appropriate for

evaluation using the SDP. It involves a violation of 10 CFR 50.59, an issue which impacts

NRC oversight ability. The issue is more than minor because it involves a programmatic

issue affecting containment controls for all secondary system penetrations. It was

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considered to be a noncited Severity Level IV violation. Management review determined

it was greater than minor because the change should have received NRC review prior to

implementation. Redundant containment barrier (system piping) existed and the

licensee entered this issue into its corrective action program (Section 1R02).

Cornerstone: Occupational Radiation Safety

Green. The inspectors identified a noncited violation of Units 1 and 2 Technical Specifications 5.4.1.a and 6.8.1.a, respectively, because the licensee failed to follow

procedural requirements. Specifically, the licensee failed to provide the reason radiation

work permits and work activity dose estimates were revised as required by

Procedure NMM RP-105, Revision 1, Section 5.8.

The inspectors determined that this finding was associated with the Occupational

Radiation Safety Cornerstone program and process attributes (ALARA

planning/projected dose) and affected the objective of the cornerstone, which is to

protect the worker from exposure to radiation. Therefore, the finding was greater than

minor. The occurrence involved a failure to maintain or implement, to the extent

practical, procedures needed to achieve occupational doses that were ALARA, which

resulted in unplanned, unintended occupational collective dose for a work activity.

Therefore, the safety significance of the finding was evaluated using the Occupational

Radiation Safety SDP. However, because the licensees 3-year rolling average

collective dose was not greater than 135 person-rem/unit, the finding had no more than

very low safety significance.

B.

Licensee-Identified Findings

Violations of very low safety significance, which were identified by the licensee, have

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

have been entered into the licensee's corrective action program. These violations and

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

Report Details

Summary of Plant Status

Unit 1 began the inspection period at approximately 100 percent power and remained at or near

100 percent throughout the inspection period.

Unit 2 began the inspection period at 100 percent power. On February 9, 2003, Unit 2

operators reduced reactor power to approximately 70 percent in support of engineering data

collection for feedwater system testing. Unit 2 operators returned the unit to 100 percent power

the same day.

1.

REACTOR SAFETY

Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity

[REACTOR-R]

1R01

Adverse Weather Protection (71111.01)

a.

Inspection Scope

On January 21-23, 2003, the inspectors walked down the Unit 2 safety-related battery

rooms to verify that compensatory measures were in place for the cold weather

conditions and to verify that the battery room temperature was sufficient to maintain the

electrolyte temperature above the Technical Specification 4.8.3.b.3 limit of 60°F. The

inspectors reviewed Unit 2 Procedure 2106.032, "Unit Two Freeze Protection Guide,"

Revision 9, to determine all applicable cold weather protection requirements for the

Unit 2 station batteries and verified that these measures were in place to protect the

equipment.

b.

Findings

No findings of significance were identified.

1R02

Evaluations of Changes, Tests, or Experiments (71111.02)

.1

Unit 2 Containment Isolation Valve Designation Change

a.

Inspection Scope

During a routine review of plant status for Unit 2, on or about December 21, 2001, to

January 16, 2002, the inspectors noted that Unit 2 Steam Generator B Sample

Valve 2CV-5859-2 had been de-energized in the open position and backseated due to a

packing leak, which resulted in a Unit 2 control room panel alarm of "CIAS inop." This

valve is located directly outside containment Penetration 2P7 and receives an

engineered safety feature actuation signal to close on a containment isolation actuation

signal (CIAS) on high containment pressure. With the valve de-energized open, it would

not perform its function to close on a CIAS.

The inspectors found that Technical Specification 3.6.3.1, "Containment Isolation

Valves," limiting condition for operation was not entered for this valve, while it was in the

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open position and de-energized for 26 days (December 21, 2001, to January 16, 2002).

This exceeded the action time requirements of the Technical Specifications. The

licensee stated that the valve was no longer considered a containment isolation valve.

The inspectors performed a historic review of the licensing basis for this valve to

determine why Valve 2CV-5859-2 was no longer considered a containment isolation

valve within the scope of the action requirements of Technical Specification 3.6.3.1.

b.

Findings

(1)

Introduction

As a result of an inadequate 10 CFR 50.59 screening, the licensee failed to perform a

safety evaluation for:

Changing the description of containment documented in the Updated Final

Safety Analysis Report (UFSAR)

Changing the description of General Design Criterion (GDC) 57, "Closed System

Isolation Valves," applicability documented in the UFSAR

Changing Procedure 1015.034, "Containment Penetration Administrative

Controls," Revision 1, to indicate that Unit 2 Steam Generator B Sample

Valve 2CV-5859-2 was not a containment isolation valve within the scope of

Technical Specification 3.6.3.1

Several other secondary system containment penetration valves were also included in

these changes. The failure to perform a safety evaluation for these changes to the

Unit 2 UFSAR and Procedure 1015.034 was determined to be an example of a violation

of 10 CFR 50.59.

(2)

Description

Background

As part of this historical review, the inspectors reviewed the Unit 2 Preliminary Safety

Analysis Report and the Unit 2 Final Safety Analysis Report. These documents were

used by the Commission to perform the radiological safety review with respect to a

decision concerning issuance of an operating license for Unit 2. The staffs review was

documented in NUREG-0308, "Safety Evaluation Report, Arkansas Nuclear One,

Unit 2," Supplement 1, dated November 1977. NUREG-0308, Supplement 1, states

that, "the radiological safety review with respect to a decision concerning issuance of an

operating license for ANO-2 has been based on the applicants Final Safety Analysis

Report (Amendment 20) and subsequent Amendments 21-43, all of which are available

for review at the Nuclear Regulatory Commissions Public Document Room . . . ." The

Safety Evaluation Report (SER) summarizes the results of the radiological safety review

of Unit 2 performed by the staff. The SER further states that "The conclusions reached

as a result of the evaluation of the applicants application to operate Unit 2 are presented

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in Section 22.0 of the SER." NUREG-0308, "Containment Isolation Systems,"

Supplement 1, Section 6.2.4, states that "The containment isolation system is designed

to isolate the containment atmosphere from the outside environment under accident

conditions. Double barrier protection, in the form of closed systems and isolation

valves, is provided so that no single valve or piping failure can result in loss of

containment integrity. Containment building penetration piping up to and including the

external isolation valve is designed as seismic Category I equipment and is protected

against missiles which could be generated under accident conditions. Containment

isolation will occur automatically upon receipt of a containment isolation actuation signal

of high containment pressure (5 pounds per square inch gauge.) All fluid penetrations

not required for operation of the engineered safety features equipment will be isolated.

Remotely operated isolation valves have been provided for those safety-related systems

which will not be automatically isolated . . . ."

NUREG-0308, Supplement 1, further states that "we have reviewed the containment

isolation system for conformance to General Design Criteria 54, 55, 56, and 57. We

conclude that the system meets the General Design Criteria and is, therefore,

acceptable."

As stated earlier, the staff documented in NUREG -0308, Supplement 1, that "the

radiological safety review with respect to a decision concerning issuance of an operating

license for ANO-2 has been based on the applicants Final Safety Analysis

Report (FSAR) (Amendment 20) and subsequent Amendments 21 through 43." FSAR

Section 6.2.4 through Amendment 32, dated October 31, 1975, stated that "the

containment isolation systems provide the means of isolating fluid systems that pass

through containment penetrations so as to confine to the containment any radioactivity

that may be released following a postulated accident. Unit-2 does not have a particular

system for containment isolation; however, isolation is achieved by applying common

criteria to penetrations in the various fluid systems, and by using a single parameter,

containment pressure, to actuate the appropriate valves." FSAR, "Design Bases,"

Section 6.2.4.1, states that "the Design basis for the CIS is to minimize the release of

radioactive material from the Containment by closing all fluid penetrations not serving

accident consequence limiting systems, so that the site boundary thyroid and whole

body doses from radioactive material escaping through the containment penetrations

plus the doses from other sources during any postulated accident are within the limits of

10 CFR Part 100. The containment isolation systems are designed in accordance with

10 CFR Part 50, Appendix A, General Design Criteria 54, 55, 56 and 57, and meet the

leak testing criteria of 10 CFR Part 50, Appendix J. The applicable criterion for each

penetration is shown in Table 6.2-26." FSAR Table 6.2-26 and amendments used by

the staff for the radiological safety review concerning issuance of an operating license

for Unit 2 identified Containment Penetrations 2P1, 2P2, 2P3, 2P4, 2P7(SGA),

2P7(SGB), 2P32, 2P35, 2P64, and 2P65 as meeting the GDC 57 criterion. The above

listed penetrations are all associated with the containment building secondary system

piping penetrations.

Based on review of the Unit 2 FSAR, amendments documented in NUREG-0308, and

the current UFSAR through Amendment 16, the inspectors found that Unit 2 Steam

Generator B Sample Valve 2CV-5859-2 had always been previously identified and

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documented in the FSAR as a containment isolation valve, which met 10 CFR Part 50,

Appendix A, GDC 57, until a change was made in the UFSAR in Amendment 13. In

addition, this valve had previously been identified as a containment isolation valve

required to be operable in Unit 2 Technical Specification Table 3.6-1 (referenced by

Technical Specification 3.6.3.1) with an isolation time of <20 seconds on a CIAS. The

requirement that this valve be considered a containment isolation valve remained

identified in Technical Specification Table 3.6-1 until License Amendment 154 was

implemented.

The licensee requested and was granted approval in License Amendment 154 to

remove Technical Specification Table 3.6-1 and replace Technical Specification 3.6.3.1

with the following: "Each containment isolation valve shall be OPERABLE." This was

done in accordance with the guidance of Generic Letter 91-08, "Removal of Component

Lists from Technical Specifications." The SER for Amendment 154 states that "the

licensee confirmed that with the proposed changes, the new Technical Specifications

applied to all valves classified as containment isolation valves by the plant licensing

basis." The NRC staff found that the proposed changes to the Technical Specification

were primarily an administrative change that did not alter the requirements set forth in

the Technical Specifications. The NRC staff concluded that the changes would allow

the licensee to make corrections and updates to the list of components for which the

Technical Specification requirements applied, under the provision that these changes to

plant procedures be controlled as specified in the Administrative Controls Section of the

Technical Specifications. The SER also states that the Technical Specification Bases

were amended to include reference to a new list of containment isolation valves that

would be included in Procedure 2203.005, "Loss of Containment Integrity," and that the

new list would be subject to the administrative control requirements of the Technical

Specifications and 10 CFR 50.59. The inspectors noted that the new list of containment

isolation valves was actually included in Procedure 1015.034, "Containment Penetration

Administrative Control," Revision 0.

During review of NRC Generic Letter 91-08, the inspectors noted that the generic letter

states that "Generally, the UFSAR identifies those valves that are classified as

containment isolation valves. With this Technical Specification change, the limiting

condition for operation (LCO), remedial actions, and surveillance requirements will apply

for all valves that are classified as containment isolation valves by the plant licensing

basis." The inspectors noted that Valve 2CV-5859-2 was identified as a GDC 57

containment isolation valve in the Unit 2 UFSAR, Table 6.2-26, "Containment

Penetrations," and in Procedure 1015.034 at the time when License Amendment 154

was approved and incorporated into the Unit 2 Technical Specifications.

Description of Change to the UFSAR

During review of their containment penetrations, the licensee had identified that the

licensing documents that specified containment isolation valves were not consistent and

determined that this was caused by various personnel incorrectly interpreting the

definition of the containment boundary. In 1995 and 1996, Procedure 1015.034,

"Containment Penetration Administrative Control," Revision 0, and Unit 2 UFSAR,

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Table 6.2-26, were changed with Licensing Document Change Request 2-6-2-002 for

UFSAR, Amendment 13. In this amendment, note 9 was added and the applicability of

GDC 57 was deleted for 38 valves associated with the following Unit 2 secondary

system containment penetrations:

Containment Penetration 2P1

Main Steam from S/G 2E24A - MSIS

Containment Penetration 2P2

Main Steam from S/G 2E24B - MSIS

Containment Penetration 2P3

Feedwater to S/G 2E 24A - MSIS

Containment Penetration 2P4

Feedwater to S/G 2E 24B - MSIS

Containment Penetration 2P7

S/G 2E24A Sample - CIAS

Containment Penetration 2P7

S/G 2E24B Sample - CIAS

Containment Penetration 2P32

S/G 2E24A Blow down - MSIS

Containment Penetration 2P35

EFW to S/G 2E24A - MSIS

Containment Penetration 2P64

S/G 2E4B Blow down - MSIS

Containment Penetration 2P65

EFW to S/G 2E24B - MSIS

Note 9 stated "These penetrations are associated with the secondary side of the steam

generators and are not subject to GDC 57 since the containment barrier integrity is not

breached during DBA LOCA conditions. The containment boundary or barrier against

fission product leakage to the environment is the inside surface of the steam generator

tubes, the outer surface of the lines emanating from the steam generator, and the outer

surface of the steam generator above the bottom tube sheets. Valves associated with

these penetrations are not containment isolation valves."

During discussions with the licensees staff, the inspectors were informed that this

treatment of the secondary system as part of the containment boundary had always

been the licensee's view of containment and that historical inconsistencies in the

designation of secondary system containment valves could be explained, if this was

understood. They also noted that their view was consistent with a 1972 Westinghouse

document describing the containment boundary. See Attachment 2 for additional details

regarding the licensee's perspective. The inspectors could not identify that the 1972

Westinghouse document was ever referenced in any licensing correspondence with the

NRC that was applicable to Unit 2. The inspectors were also not able to find where the

NRC staff had reviewed and accepted this alternative view of the containment boundary.

Screening Review

Approval for UFSAR, Amendment 13, was based on Engineering Report 93-R-0007-01,

"Containment Penetration Design Summary Document," Revision 0. The inspectors

reviewed Engineering Report 93-R-0007-01 and the associated 10 CFR 50.59 screening

review. The following note was included in the Basis for Change for all of the secondary

system containment penetration valves:

"ADMINISTRATIVE/EDITORIAL - Information is confusing in that it leaves the

impression that these valves are Reactor Building Isolation valves when they are not.

Deletion is consistent with note 9" (Unit 2 UFSAR, Table 6.2-26).

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Because the changes made for all of the secondary system containment penetration

valves were viewed as administrative/editorial, these changes met the licensees criteria

for changes specifically exempted from further 10 CFR 50.59 evaluation. This decision

was in accordance with Procedure 1000.131, "10 CFR 50.59 Review Program,"

Attachment 1, which was, at that time, the licensees procedure governing

10 CFR 50.59 determinations/evaluations.

The inspectors reviewed containment descriptions in other sections of the initial

FSAR and the NRC Safety Evaluation Report and did not find that the licensees view of

containment had been previously documented. The inspectors determined that the

change in definition of containment was not an administrative/editorial change because

changes in the potential for increased fission product release during a steam generator

tube rupture accident or other postulated accidents were not evaluated.

While the inspectors acknowledged that the original licensing basis for the containment

boundary was in some cases silent and sometimes inconsistent, the inspectors noted

the Unit 2 FSAR and amendments documented in NUREG-0308 included FSAR,

Table 6.2-26, Amendment 23, which identified all the above secondary system Unit 2

containment penetrations and listed all of the secondary system valves as being

GDC 57 containment isolation valves.

Based on the historical review, the inspectors concluded that on March 31, 1995, the

licensee performed an inadequate 10 CFR 50.59 screening review documented in

Evaluation FFN-95-166, which was approved by the Plant Safety Review Committee on

August 31, 1995, for proposed changes to the Unit 2 UFSAR. As a result of this

inadequate screening, the licensee had not performed a required safety evaluation.

10 CFR 50.59 was changed via publication in the "Federal Register" on October 4, 1999

(64 FR 53582). It is NRC policy to exercise enforcement discretion pursuant to

Section VII.B.6 of the Enforcement Policy and not issue citations or document noncited

violations against the old rule if the revised rule requirements were met.

Safety Evaluation Using the Revised Rule

At the inspectors' request, the licensee performed a safety evaluation of the change

using criteria of the revised 10 CFR 50.59 rule for the Unit 2 Steam Generator B sample

valve.

The licensee identified that the change would result in a very small increase in the

consequences of an accident previously evaluated. The inspectors reviewed the

guidance in Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59,

Changes, Tests, and Experiments," and in NEI 96-07, "Guidelines for 10 CFR 50.59

Implementation," Revision 1, Section 4.3.3, and concluded that this increase in

consequences would be minimal and, therefore, did not require NRC review for that

reason.

The licensee did not identify that the change resulted in a small increase in the

probability of occurrence of a malfunction of equipment important to safety previously

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evaluated in the Safety Analysis Report, which should have received the NRCs approval

prior to implementation. Using the guidance in NEI 96-07, Revision 1, Section 4.3.2, by

definition, departure from GDC 57 results in more than a minimal increase in the

likelihood of occurrence of a malfunction of a component important to safety. In

addition, the probability of malfunction did increase when Valve 2CV-5859-2 was

de-energized in the open position for 26 days (December 21, 2001, through January 16,

2002).

(3)

Analysis

This finding is not suitable for evaluation using the Significance Determination

Process (SDP). The finding is determined by management to be of very low safety

significance. The inspectors reviewed an analysis performed by the licensee and found

that the expected increase in consequences will meet the minimal standard described in

NEI 96-07, Revision 1, Section 4.3.3. This finding is more than minor because the

change should have received NRCs review prior to implementation.

(4)

Enforcement

The inspectors evaluated the change in the UFSAR documented description of

containment, the change in the UFSAR list of containment isolation valves, and the

change in Procedure 1015.034 for compliance with 10 CFR 50.59. At the time that the

10 CFR 50.59 screening was performed in 1995, the rule required the licensee maintain

records of changes in procedures described in the Safety Analysis Report and these

records include a written safety evaluation which provides the basis for the

determination that the change does not involve an unreviewed safety question. The

licensee had not initially performed a safety evaluation for the change in the UFSAR

documented containment description, for the deletion of multiple valves from the UFSAR

list of containment isolation valves, and for the resultant deletion of Technical

Specification applicability for these containment isolation valves in Procedure 1015.034.

The failure to perform the safety evaluation was a violation of 10 CFR 50.59.

Under the previous 10 CFR 50.59 rule, the inadequate screening evaluation resulted in

the licensee not identifying that the change would result in a small increase in the

consequences of an accident previously evaluated and a small increase in the

probability of occurrence of a malfunction of equipment important to safety previously

evaluated in the Safety Analysis Report. This constituted an unreviewed safety question

which should have received NRC approval prior to implementation.

To evaluate conformance with the revised 10 CFR 50.59 rule, the inspectors reviewed

the change in the UFSAR documented containment description, the deletion of multiple

valves from the UFSAR list of containment isolation valves, and the resultant deletion of

the Technical Specification applicability for these containment isolation valves in

Procedure 1015.034 against the requirements of the new rule, using the guidance in

Regulatory Guide 1.187, and NEI 96-07, Revision 1.

Under the revised rule, the increase in consequences was minimal. However, departure

from GDC 57 does result in more than a minimal increase in the likelihood of occurrence

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of a malfunction of a component important to safety. Specifically, the probability of

malfunction increased more than minimally when Unit 2 Steam Generator B Sample

Valve 2CV-5859-2 was de-energized in the open position for 26 days (December 21,

2001, through January 16, 2002). Based on these findings, the inspectors concluded

that a license amendment was also required by the revised rule.

Therefore, the NRC will not exercise enforcement discretion pursuant to Section VII.B.6

of the Enforcement Policy for the 1995 failure to perform a safety evaluation as required

by 10 CFR 50.59.

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

management and is determined to be a Green finding of very low significance. The

licensee entered this finding into its corrective action program as Condition

Report ANO-2003-0242. This violation is being treated as an example of a noncited

violation (50-313/0302-01; 50-368/0302-01) consistent with Section VI.A of the

NRC Enforcement Policy.

.2

Unit 1 Containment Isolation Valve Designation Change

a.

Inspection Scope

During the inspectors review of the Unit 2 issue of deletion of GDC 57 requirements

from certain Unit 2 containment isolation valves, the inspectors also found that multiple

secondary system reactor building penetration valves which were previously identified

as GDC 57 reactor building isolation valves were removed from the Unit 1 UFSAR in the

same manner. The inspectors reviewed Engineering Report 93-R-0007-01, its

associated 10 CFR 50.59 determination, and the Unit 1 UFSAR, Table 5-1, to determine

if the licensee had appropriately performed these reviews.

b.

Findings

(1)

Introduction

The inspectors found that Engineering Report 93-R-007-01 and the associated

10 CFR 50.59 determination was used for changes to both the Units 1 and 2 UFSARs

and containment penetration administrative control procedures. Therefore, the issue of

the licensees departure from GDC 57 applicability for the steam generator secondary

side reactor building/containment building penetration valves (which were previously

identified as GDC 57 valves) from the Units 1 and 2 UFSARs and the Units 1 and 2

Technical Specification requirements as implemented by the respective Unit 1 and 2

containment penetration control procedures, is applicable to both Units 1 and 2.

As a result of an inadequate 10 CFR 50.59 screening, the licensee failed to perform a

safety evaluation for:

Changing the description of containment documented in the Unit 1 UFSAR

-11-

Changing the description of GDC 57 applicability documented in the Unit 1

UFSAR

Several secondary system containment penetration valves were included in these

changes. The failure to perform a safety evaluation for these changes to the Unit 1

UFSAR was determined to be an example of a violation of 10 CFR 50.59.

(2)

Description

Background

As part of the inspectors' review for the Unit 1 issue, the inspectors found that a

previous violation had been cited for Unit 1 concerning the failure to lock closed two

reactor building isolation valves between the feedwater isolation valves and the

containment in accordance with Technical Specification 3.6.5. This violation was

documented in 1994 in NRC Inspection Report 50-313/94-07; 50-368/94-07. The

subject valves were secondary system reactor building penetration feedwater line vent

and drain Valves FW-1038 and FW-1049. They are located in secondary system

penetrations that were affected by the licensee's UFSAR change. The inspectors

reviewed the licensee response and corrective actions associated with the violation to

understand the licensee's interpretation of the definition of containment at that time, just

prior to the deletion of GDC 57 applicability for secondary system penetrations.

In the licensee response to the violation, dated December 16, 1994, the licensee stated:

"The identified concern is that main feedwater vent valves FW-1038 and FW-1049 were

not being maintained as locked valves. Implied in the violation is that General Design

Criterion (GDC) 57, Closed System Isolation Valves, is applicable to the manual test

connection, vent and drain valves located at the given penetrations. GDC 57 requires

manual containment isolation valves be closed and locked to maintain containment

integrity. Locking the test connection, vent and drain valves represents a level of

administrative control not previously considered to be required for these and other test

connection, vent or drain valves.

"The general design basis governing containment building isolation requirements is

leakage through all fluid penetrations, not serving accident-consequence limiting

systems, must be minimized by a double barrier so that no single, credible failure or

malfunction of an active component can result in loss of isolation or intolerable leakage.

The installed double barriers take the form of closed piping systems, both inside and

outside the containment building, and various types of isolation valves.

"Based on ANOs licensing basis and NRC regulations, no specific requirement to lock

these valves previously existed. Industry guidance indicates that test connection, vent

and drain valves should not be classified as GDC isolation valves, but as

administratively con trolled containment system barriers. ANO has procedurally

controlled these valves to ensure they are closed during power operations.

-12-

"The general design basis governing containment isolation requirements is given in

Section 5.2.5 of the current Arkansas Nuclear One- Unit 1 (ANO-1) Safety Analysis

Report. The ANO-1 Safety Analysis Report Table 5.1, Reactor Building Isolation

Valves, lists GDC 55, 56, and 57 penetrations and the applicable containment isolation

valves. There are no test connection, vent and drain valves identified in the table. The

definition of ANSI/ANS-56.2-1984, Containment Isolation Provisions for Fluid Systems,

clearly distinguish between containment isolation valves and the containment

penetration test connection, vent and drain valves by defining test connections or vents

as being provided so that containment isolation valves can be tested. Figures 1 and 2

of ANSI/ANS-56.2-1984 demonstrate the criteria for GDC 55, 56, and 57 and includes

only the main process line valves (main feedwater block valves for P-3 and P-4), and do

not demonstrate provision for test connection, vent or drain valves."

In Section 6.2.3 of the ANO-1 Safety Evaluation Report (SER), it states, in part, "the

reactor building isolation system is designed to isolate the containment atmosphere from

the outside environment under accident conditions. Double barrier protection, in the

form of closed systems and isolation valves, is provided so that no single valve or piping

failure can result loss of containment integrity." The SER further states that "the NRC

staff has reviewed the containment isolation system for conformance to GDC 55, 56,

and 57." The NRC concluded that, the containment isolation system meets the intent of

the general design criteria. From the Final Safety Analysis Report (FSAR) description of

the containment building isolation system it is evident that only the process line valves

were included as containment building isolation valves. Test connection, vent and drain

valves on these penetrations are neither depicted nor described."

The inspectors noted that the process valves that the licensee discussed in their

response to the violation are some of the valves for which the licensee now has deleted

GDC 57 applicability.

The inspectors also found during the review that, during the initial licensing process for

Unit 1, the Commission Staffs Request for Additional Information 5.83 requested

information regarding compliance with GDC 57 for the Unit 1 main feedwater lines to

each steam generator and was specifically evaluated by the Commissions Staff to

determine if the piping configuration used complied with GDC 57. In addition, Request

for Additional Information 10.1 for Unit 1 requested information concerning the turbine

stop valves serving as containment isolation valves for isolation of the unaffected steam

generator in the event of a steamline rupture accident. The licensee responded to

Question 10.1 that the main steam block valves, not the turbine stop valves, were

designed to serve as the containment isolation valves and that these valves met

Criterion 57 of 10 CFR Part 50, Appendix A. The staff also requested information

regarding how the licensee was implementing GDC 57 for penetrations containing the

steam generator sample valves for Unit 1.

-13-

Description of Change to the Unit 1 UFSAR

During review of this issue, the inspectors questioned the licensee concerning the

licensing basis for the Unit 1 steam generator secondary side reactor building

penetrations and the subsequent deletion of the GDC applicability for the following

penetrations:

Containment Penetration P1

Main Steam from Steam Generator E-24A to EFW

and Main Turbine

Containment Penetration P2

Main Steam from Steam Generator E-24B to EFW

and Main Turbine

Containment Penetration P3

Main Feedwater to Steam Generator E24-A

Containment Penetration P4

Main Feedwater to Steam Generator E-24B

Containment Penetration P10

Steam Generator Sampling

Containment Penetration P17

EFW to Steam Generator E-24A

Containment Penetration P65

EFW to Steam Generator E-24B

Based on the inspectors questions, the licensee performed a review of the Unit 1

design and licensing bases and developed a position paper, which is Attachment 2 to

this inspection report. In the licensees position paper, the licensee stated "The

containment boundary or barrier against fission product leakage to the environment is

the inside surface of the steam generator tubes, the outer surface of the lines

emanating from the steam generator, and the outer surface of the steam generator

between the tube sheets. This position is based on the concept of treating the

secondary system pressure boundary as an extension of the containment liner." The

licensee further stated that, although this position was not well documented, this position

has always been the understanding of plant personnel familiar with the original design

and licensing basis. The licensee further stated that this concept was based on

Westinghouse Document WCAP-7451, dated September 1972. The inspectors could

not identify that this document describing the containment boundary was ever

referenced in any licensing correspondence with the Commission that was applicable to

Unit 1.

The licensee stated that, ultimately, the concept of the secondary pressure boundary

being an extension of containment boundary was explicitly approved by the NRC in the

Safety Evaluation Report approving renewal of the Unit 1 operating license. This was

based on the NRC's Request for Additional Information 2.3.2.7-2.

The Request for Additional Information stated "In P&ID LRA-M-237, Sheet 1, the

redundant isolation valves (SS-1017B and SS-1018B) for the test connections of the

sampling system are not highlighted as being within the scope of license renewal.

However, containment isolation provisions require double isolation at the test

-14-

connections for greater assurance of containment integrity. Provide a justification as to

why the second isolation valve on each test connection is not in-scope." The licensees

response, dated August 30, 2000, stated "Please see ANO-1 Safety Analysis Report,

Table 5-1, Penetration 10, and note 8. This penetration is associated with the

secondary side of the steam generator and is not subject to GDC 57. In accordance

with the current licensing basis, the containment boundary or barrier against fission

leakage to the environment is the inside surface of the steam generator tubes, the outer

surface of the lines emanating from the steam generator, and the outer surface of the

steam generator between the tube sheets. Therefore, Valves SS-1017B and SS-1018B

do not meet the scoping criteria for license renewal."

The NRC found the licensees response acceptable. However, the inspectors noted that

the GDC 57 applicability for the secondary system penetrations had already been

deleted without prior Commission approval with the administrative change that is the

subject of this violation for the change to UFSAR, Table 5-1.

Based on the inspectors review, the inspectors concluded that on March 31, 1995, the

licensee performed an inadequate 10 CFR 50.59 screening and, as a result, the

licensee had not performed a safety evaluation as required by 10 CFR 50.59.

Screening Review for Unit 1 changes

Approval for UFSAR, Amendment 14, was based on Engineering Report 93-R-0007-01,

"Containment Penetration Design Summary Document," Revision 0. As stated in the

previous section of this inspection report, the inspectors reviewed Engineering

Report 93-R-0007-01 and the associated 10 CFR 50.59 screening review. The following

note was included in the Basis for Change for all of the secondary system containment

penetration valves:

"ADMINISTRATIVE/EDITORIAL - Information is confusing in that it leaves the

impression that these valves are Reactor Building Isolation valves when they are not.

Deletion is consistent with note 8" (Unit 1 UFSAR, Table 5-1).

Because the changes made for all of the secondary system containment penetration

valves were viewed as administrative/editorial, these changes met the licensees criteria

for changes specifically exempted from further 10 CFR 50.59 evaluation. This decision

was in accordance with Procedure 1000.131, Attachment 1, which was, at that time, the

licensees procedure governing 10 CFR 50.59 determinations/evaluations.

The inspectors determined that the change in definition of containment was not an

administrative change because changes in the potential for increased fission product

release during a steam generator tube rupture accident or other postulated accidents

were not evaluated.

-15-

(3)

Analysis

This finding is not suitable for evaluation using the SDP. The finding is determined by

management to be of very low safety significance. This finding is more than minor,

because the change should have received NRCs review prior to implementation.

(4)

Enforcement

The inspectors evaluated the change in the UFSAR documented description of

containment and the change in the updated UFSAR list of containment isolation valves.

At the time that the 10 CFR 50.59 screening was performed in 1995, the licensee did

not perform a safety evaluation for the change in the UFSAR documented containment

description or for the deletion of multiple valves from the UFSAR list of containment

isolation valves and resultant deletion of Technical Specification applicability for these

containment isolation valves. The failure to perform a safety evaluation was a violation

of 10 CFR 50.59.

Under the previous 10 CFR 50.59 rule, an inadequate screening evaluation resulted in

the licensee not identifying whether the change would result in an increase in the

consequences of an accident previously evaluated or whether there was an increase in

the probability of occurrence of a malfunction of equipment important to safety

previously evaluated in the Safety Analysis Report change should have received NRC

approval prior to implementation, because the change constituted an unreviewed safety

question.

To evaluate conformance with the current 10 CFR 50.59 rule, the inspectors reviewed

the change in the UFSAR documented containment description, deletion of multiple

valves from the UFSAR list of containment isolation valves, and the resultant deletion of

the Technical Specification applicability for these containment isolation valves against

the requirements of the new rule, using the guidance in Regulatory Guide 1.187 and

NEI 96-07, Revision 1.

An evaluation under the new rule was not performed by the licensee. However,

departure from GDC 57 does result in more than a minimal increase in the likelihood of

occurrence of a malfunction of a component important to safety. Based on these

findings, the inspectors concluded that a license amendment was also required by the

current 10 CFR 50.59 rule.

Therefore, NRC will not exercise enforcement discretion pursuant to Section VII.B.6 of

the Enforcement Policy for the 1995 failure to perform a safety evaluation as required by

10 CFR 50.59.

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

management and is determined to be a Green finding of very low significance. The

licensee entered this finding into its corrective action program as Condition

Report (CR) ANO-C-2003-0242. This violation is being treated as an example of a

noncited violation (50-313/0302-01; 50-368/0302-01) consistent with Section VI.A of the

NRC Enforcement Policy.

-16-

1R04

Equipment Alignment

.1

Partial System Walkdown (71111.04)

a.

Inspection Scope

The inspectors performed a partial system walkdown of the Unit 2 service water system

while maintenance was being performed on Pump 2P-4C on February 3-5, 2003. The

inspectors verified proper component alignment and operation in accordance with

Procedure 2104.029, "Service Water System Operations," Revision 53, and system

piping and instrumentation diagrams to verify that the system was in a proper standby

lineup. The inspectors also examined component material condition.

The inspectors performed a partial system walkdown of the Unit 1 service water system

while maintenance was being performed on Pump P-4A on February 18-20, 2003. The

inspectors verified proper component alignment and operation in accordance with

Procedure 1104.029, "Service Water and Auxiliary Cooling System," Revision 55, and

system piping and instrumentation diagrams to verify that the system was in a proper

standby lineup. The inspectors also examined component material condition.

The inspectors performed a partial system walkdown of the Unit 2 vital dc electrical

system during maintenance on Battery 2D-12 on March 14-15, 2003. The inspectors

verified proper system alignment and operation in accordance with Procedure 2107.004,

"DC Electrical System Operation," Revision 20, to verify that the system was in a proper

standby lineup.

b.

Findings

No findings of significance were identified

1R05

Fire Protection

.1

Routine Inspection (71111.05Q)

a.

Inspection Scope

The inspectors referenced the Fire Hazards Analysis Report, Revision 7, during the

following inspections to ensure that conditions were consistent with the requirements of

the licensees fire protection program for fire protection systems design, control of

transient combustibles and ignition sources, fire detection and suppression capability,

fire barriers, and any related compensatory measures. Additional documents reviewed

included Procedure 2203.014, "Alternate Shutdown," Revision 14; Technical

Guideline 2203.014, "Alternate Shutdown," Revision 14; "ANO Pre-Fire Plan,"

Revision 1; Procedure 1000.047, "Control of Combustibles," Revision 15;

Procedure 2203.034, "Fire or Explosion," Revision 5; and CR ANO-2-2003-0229.

Unit 2 Battery Room 2D11, Fire Zone 2103V, conducted on February 21, 2003

-17-

Unit 1 Switchgear Room A3, Fire Zone 99M, conducted on February 19-21, 2003

Unit 1 Cable Spreading Room, Fire Zone 97R, conducted on March 5-7, 2003

Unit 1 Integrated Control System Relay Room, Fire Zone 97R on March 5-7,

2003

Unit 1 North Battery Room, Fire Zone 95-O, on March 5-7, 2003

Unit 2 Battery Room 2D12, Fire Zone 2102Y, conducted on March 11-13, 2003

Unit 2 Core Protection Calculator Room, Fire Zone 2098C, conducted on

February 10, 2003

b.

Findings

No findings of significance were identified.

1R13

Maintenance Risk Assessments and Emergent Work Control (71111.13)

a.

Inspection Scope

The inspectors evaluated and discussed with the licensee the risk assessments listed

below to verify that assessments were performed when required and appropriate

compensatory actions were taken. The inspectors reviewed these assessed risk

configurations against actual plant conditions and any inprogress evolutions or external

events to verify that the assessments were accurate, complete, and appropriate for the

conditions. In addition, the inspectors walked down the control room and plant areas to

verify that compensatory measures identified by the risk assessments were

appropriately performed.

Unit 1 Emergency Diesel Generator K4B risk assessment associated with the

system outage conducted on January 17, 2003

Unit 1 Emergency Diesel Generator K4A risk assessment associated with the

system outage conducted on January 21-22, 2003

Unit 2 Service Water Pump 2P-4C electrical power supply cable replacement

conducted on February 3-4, 2003

Unit 1 Service Water Pump P-4A electrical power supply cable replacement

conducted on February 17-20, 2003

Unit 2 Battery Cell 2D12 replacement and testing conducted January 10, 2003

Unit 2 Battery Cell 2D12 replacement and testing conducted March 14-16, 2003

-18-

Unit 1 Emergency Diesel Generator K4A maintenance activities conducted on

February 14-19, 2003

b.

Findings

No findings of significance were identified.

1R14

Personnel Performance Related to Nonroutine Plant Evolutions and

Events (71111.14, 71153)

a.

Inspection Scope

The inspectors observed the following nonroutine evolutions to verify that they were

conducted in accordance with licensee procedures and Technical Specification

requirements:

On February 2, 2003, Unit 2 experienced a failure of the operating Loop 2 Service

Water Pump 2P-4C. Operators entered Abnormal Operating Procedure 2203.022,

"Loss of Service Water," Revision 8, and promptly started Standby Service Water

Pump 2P-4B to provide service water to Loop 2. The licensee initiated

CR ANO-2-2003-0178 to document the event and corrective actions. As part of an

extent of condition review, the licensee also initiated CR ANO-C-2003-0067 to review

the condition of medium-voltage underground cables at the facility. From this review,

the licensee identified that the cables for Unit 1 Service Water Pump P-4A should be

replaced as a prudent measure. The cables for Pump P-4A were replaced during the

week of February 17, 2003.

b.

Findings

No findings of significance were identified.

1R15

Operability Evaluations (71111.15)

a.

Inspection Scope

The inspectors reviewed operability determinations to assess the correctness of

evaluations, the use of compensatory measures, if needed, and compliance with the

Technical Specifications. The inspectors' review included a verification that operability

determinations were made as specified by the licensee's Procedure LI-102, "Corrective

Action Process," Revision 2, and Procedure 1000.104, "Condition Reporting and

Immediate Reportability Determinations," Revision 17. The technical adequacy of the

determinations was reviewed and compared to the Technical Specifications, Technical

Requirements Manual, UFSAR, and associated licensing-basis documentation, as

appropriate. The operability determinations that were reviewed were documented in the

following CRs:

ANO-C-2003-00067

Service water pump power supply cable evaluation

-19-

ANO-2-2002-00644

Installation of temporary alteration for installation of

heaters to supplement local heating due to Battery

Room 2D11 temperature below 65°F

ANO-2-2003-00237

Various abnormal indications and alarms which occurred

at roughly the same time on Unit 2

ANO-2-2003-00383

Failed support rod from Hanger 2EBD-15-H2 which

provides support from main steam header to stop

Valve 2CV-0250

ANO-1-2003-00225

EDG K-4A fuel oil sight glass leakdown problem

ANO-1-2003-00193

Air entrained in the EDG 1 fuel oil return system

ANO-2-2002-02084

Low voltage on Unit 2 Battery 2D12 Cell 40

ANO-2-2003-00412

Low voltage on Unit 2 Battery 2D12 Cell 40

ANO-2-2003-00193

Unit 2 low spent fuel pool temperature

ANO-1-2003-00378

Emergency Feedwater Valve SV-2613 nameplate and

documentation errors

b.

Findings

No findings of significance were identified.

1R17

Permanent Plant Modifications

.1

Annual Review (71111.17A)

a.

Inspection Scope

The inspectors reviewed the plant modification described in Engineering

Request (ER) ANO-2000-2688-002, "Uprate L-3 Spent Fuel Crane to Handle

Capacity of 122 Tons." This modification replaced the previous Crane L-3

nonsingle-failure proof trolley, rated at 100 tons, with a new trolley of single-failure

proof design rated at 130 tons. This modification was necessary to allow the licensee to

lift spent fuel storage casks of a new, heavier design. The review included the safety

evaluation prepared to determine if the modification required a license amendment per

the requirements of 10 CFR 50.59.

b.

Findings

The licensee's 10 CFR 50.59 evaluation concluded that the proposed modification did

not require a license amendment. The licensee concluded that the upgraded crane

design met the requirements of NUREG-0554, "Single-Failure-Proof Cranes for Nuclear

-20-

Power Plants," and NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."

Therefore, the licensee concluded that the upgraded crane design was acceptable for

implementation without the need for an NRC license amendment.

The inspectors disagreed with this conclusion. The inspectors acknowledged the new

crane was intended to meet single-failure-proof design standards and utilized a trolley

design documented in a vendor topical report that was previously approved by the NRC.

However, the inspectors noted that Generic Letter 85-11, "Completion of Phase II of

Control of Heavy Loads at Nuclear Power Plants NUREG-0612," identified that

installation of a single-failure-proof crane design may reasonably be expected to

eliminate most, perhaps 90 percent, of load drop probability, meaning the failure

probability was not zero. The inspectors concluded that the increase in the maximum

critical load rating of the crane (from 100-130 tons), combined with a required load path

that would carry a loaded spent fuel storage cask over the control rooms, would require

a license amendment.

The inspectors, managers from the NRC Region IV office, and representatives of the

NRC Office of Nuclear Reactor Regulation (NRR) informed the licensee of this

conclusion in a telephone call on February 13, 2003. The inspectors also informed the

licensee that failure to submit a license amendment request for this modification was a

potential violation of 10 CFR 50.59. The licensee entered this issue into its corrective

action program as CR ANO-C-2003-0092. The licensee subsequently submitted a

license amendment request for the Crane L-3 modification to the NRC on February 24,

2003. The license amendment request was still under review by the NRR staff at the

conclusion of the inspection period. Although a potential violation of 10 CFR 50.59 was

determined to exist, the inspectors considered this issue to be an unresolved item,

pending completion of review of the license amendment request by the NRR staff and a

determination of the significance of the violation (50-313/0302-02; 50-368/0302-02).

The upgraded crane had not been used to transport a loaded spent fuel storage cask

and was under administrative controls preventing its use in this manner, pending

resolution of this issue. Therefore, this issue had no immediate safety significance.

1R19

Postmaintenance Testing (71111.19)

a.

Inspection Scope

For the maintenance activities identified below, the inspectors observed the

postmaintenance testing activities in the control room or locally and/or reviewed the test

data obtained from the field. The inspectors observed whether the tests were

performed in accordance with procedures, acceptance criteria were consistent with

Technical Specifications, and the results recorded met the test acceptance criteria. In

addition, the inspectors verified that any deficiencies were recorded and resolved.

These activities included:

Unit 1 Emergency Diesel Generator 1 testing in accordance with

Procedure 1104.036, "Emergency diesel Generator Operation," Revision 41,

conducted on January 24, 2003

-21-

Unit 1 Emergency Diesel Generator 1 testing in accordance with

Procedure 1104.036, "Emergency Diesel Generator Operation," Revision 41,

conducted on February 19, 2003

Unit 2 2D12 battery testing for Cell 40 replacement conducted on March 10-15,

2003, as described in the maintenance action plan associated with

CR ANO-2-2003-0412

b.

Findings

No findings of significance were identified.

1R22

Surveillance Testing (71111.22)

a.

Inspection Scope

The inspectors observed from either the control room or locally the performance of

and/or reviewed the documentation for the following surveillance tests. This was done

to verify that the surveillance tests were performed in accordance with approved

licensee procedures and met Technical Specification requirements. In addition, the

applicable test data was also reviewed to verify whether they met Technical

Specifications, UFSAR, and licensee procedure requirements.

Procedure 2104.036, "2DG1 Monthly Test (Slow Start)," Supplement 1B,

Revision 46, performed on January 29, 2003

Procedure 2105.009, "CEA Exercise Test," Supplement 2, Revision 21,

performed on March 21, 2003

Procedure 2403.023, "2D12 Quarterly Surveillance," Revision 14, performed on

March 17, 2003

Procedure 2403.023, "2D12 Quarterly Surveillance," Revision 14, performed on

March 19, 2003

b.

Findings

No findings of significance were identified.

1R23

Temporary Plant Modifications (71111.23)

a.

Inspection Scope

On January 28-30, 2003, the inspectors reviewed the implementation of

Procedure 2106.032, "Unit 2 Freeze Protection Guide," Revision 9, Section 8, "Battery

Room Low Temperature." This procedurally controlled temporary alteration installed a

temporary portable 15 Kw electric heater in Unit 2 Corridor 2104 to provide

supplemental heating for the safety-related battery rooms in severe wintertime

-22-

conditions. The justification for the temporary non-Q battery room heating was

evaluated and documented in ER ANO 2002-0145-000. The inspectors confirmed that

this temporary alteration was properly installed as authorized by the procedure and the

ER.

On February 3-4, 2003, the inspectors reviewed the implementation of Maintenance

Action Item (MAI) 79434, which installed a temporary alteration evaluated by

ER ANO-2003-0098-00. This temporary alteration installed a temporary power supply to

the Unit 2 Service Water Pump 2P-4C motor while its permanent power supply cables

were being replaced. The inspectors confirmed that this temporary alteration was

implemented and installed as authorized by MAI 79434 and Procedure 1000.028,

"Control of Temporary Alterations," Revision 25.

On February 19-21, 2003, the inspectors observed that Fire Door 48 had been removed

in support of the power supply cable replacement for the Unit 1 Service Water

Pump P-4A motor. The door had been removed per the instructions of MAI 80036.

During the week of March 3, 2003, the inspectors reviewed this maintenance activity to

determine whether an unevaluated temporary alteration to the facility had been

performed.

b.

Findings

A Green noncited violation was identified for performing a maintenance activity that

resulted in the implementation of an unevaluated temporary alteration. Removal of

Door 48 was not evaluated for its impact on safety-related equipment following a

potential high energy line break or for its impact on room cooling capability performance.

Door 48 is a 3-hour rated fire door separating safety-related switchgear Room A3 from

the turbine building. Located outside switchgear Room A3 are two high-pressure

feedwater heaters. The inspectors reviewed the Unit 1 UFSAR, "An Evaluation of Pipe

Breaks Outside the Reactor Building," Section A.7, Amendment 17, and noted that the

main feedwater pipe break analysis stated that steam and water from a postulated

feedwater pipe break are prevented by walls from entering into the switchgear rooms.

The inspectors concluded that this statement implied that doors in the subject walls

were also credited as barriers. The inspectors also reviewed Procedure 1104.027,

"Battery and Switchgear Emergency Cooling System," Revision 20. This procedure

identified that the emergency cooling system was designed to limit maximum room

temperature to 104°F in switchgear Room A3 with turbine building ambient air

temperatures at 122°F. The inspectors also reviewed the component database for

Door 48 and found that this door was not identified as having a design function as a

high-energy line break barrier or room temperature maintenance barrier.

Door 48 was removed per MAI 80036 on February 17, 2003, in order to support cable

replacement activities for Service Water Pump P-4A and was returned to service on

February 20, 2003. The MAI did not include an engineering evaluation for the

acceptability of removal of Door 48 nor was there one performed.

-23-

The inspector concluded that removal of Door 48 resulted in a change to the design

function of the barrier separating switchgear Room A3 and the turbine building and was

a temporary alteration. Procedure 1000.028, "Control of Temporary Alterations,"

Revision 23, requires that all temporary alterations receive an engineering review

documented in an ER.

The licensee entered this issue into its corrective action program as

CR ANO-1-2003-0258. Per UFSAR, Section A.7.3, a postulated main feedwater pipe

break outside of Door 48 was a credible event. With Door 48 removed, the resultant

harsh environment from a break would have potentially incapacitated one of the

redundant trains of equipment necessary to achieve safe shutdown conditions.

However, the remaining train of equipment would have remained unaffected. The

condition did not result in equipment being inoperable because equipment in switchgear

Room A3 was capable of performing its intended function during a design basis event

(e.g., a loss-of-coolant accident) during the time Door 48 was removed.

As part of its corrective action for this finding, the licensee surveyed all doors that

separate the turbine building from safety-related equipment areas. All doors that were

discovered to also have a design function to provide a high energy line break barrier

were identified and were provided administrative controls requiring an engineering

evaluation prior to their removal.

Unit 1 Technical Specification 5.4.1.a requires that written procedures shall be

established, implemented, and maintained, covering activities identified in Regulatory

Guide 1.33, Revision 2, Appendix A, which includes modification work.

Procedure 1000.028 required an engineering evaluation be performed to support a

temporary alteration. Failure to perform an engineering evaluation for removal of

Door 48 was a violation. The safety significance of removal of Door 48 was determined

to be very low since this issue screened as Green during a Phase 1 SDP evaluation.

The issue was considered more than minor because it affected the mitigating systems

cornerstone objective for design control and modifications, because the ability to

mitigate the consequences of a high energy line break would have been affected if the

finding had affected more than one train of equipment. This violation is being treated as

noncited violation, consistent with Section VI.A of the NRC Enforcement Policy

(NCV 50-313/0302-03).

Cornerstone: Emergency Preparedness

1EP4

Emergency Action Level and Emergency Plan Changes (71114.04)

a.

Inspection Scope

The inspectors performed an in-office review of Revision 27 to the Arkansas Nuclear

One Emergency Response Plan, received January 25, 2002, and Procedure 1903.010,

"Emergency Action Level Classification," Revision 37. The inspector compared the

current revisions with their previous revisions and 10 CFR 50.54(q) to determine if the

current revision decreased the effectiveness of the emergency plan.

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b.

Findings

No findings of significance were identified.

2.

Radiation Safety

Cornerstone: Occupational Radiation Safety

2OS2 ALARA Planning and Controls (71121.02)

a.

Inspection Scope

The inspectors reviewed radiation work permits (RWPs) from Refueling Outages 2R15

and 1R17 that both accrued more than 5 rem and that had actual doses that exceeded

the initial work activity dose estimates by more than 50 percent. No high exposure jobs

or work activities in high radiation areas were performed during the inspection; however,

the inspectors observed the movement of Cooling Coil VUC1B (in accordance with

RWP 2003-1040) and the replacement of Spent Fuel Filter F4 (in accordance with

RWP 2003-1017). During the work activities, the inspectors made independent

radiation measurements and observed ALARA practices and contamination control

practices. The following specific items were reviewed and compared with regulatory

requirements:

Plant collective exposure history for the past 3 years, current exposure trends,

and 3-year rolling average dose information

ALARA program procedures

RWP 2002-1420, "Scaffolding and Insulation Activities During 1R17";

RWP 2002-1452 and RWP 2002-1515, "Reactor Head CRD Nozzle Inspection";

RWP 2002-2442, "Steam Generator Inspection and Maintenance"; and

RWP 2002-2507, "Pressurizer Heater Repair"

Radiological work planning

Use of engineering controls to achieve dose reductions

Processes used to estimate and track exposures

Selected corrective action documentation involving higher than planned

exposure levels and radiation worker practice deficiencies since the last

inspection in this area (CRs ANO-C-2001-00297, ANO-2-2002-00882,

ANO-C-2003-00034, and ANO-C-2003-00118)

b.

Findings

Introduction: The inspectors identified a Green noncited violation because the licensee

failed to provide an adequate reason for revising RWPs, as required by a Technical

Specification required procedure.

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Description: During a review of RWP packages, the inspectors noted multiple examples

of the licensee adjusting work activity dose estimates without providing adequate bases.

The dose estimates were adjusted to account for failures to control work activity doses

and failed to implement additional means to control dose.

For example, RWP 2002-1420, "Scaffolding and Insulation Activities During 1R17,"

Revision 4, increased the work activity dose estimate from 15.578-16.578 rem. The

documented reason stated, "The dose estimate would have been exceeded in the next

shift. Additional dose was added to complete work." The licensee provided no other

justification for the work activity dose estimate adjustment. RWP 2002-1452,

Revision 2, increased the work activity dose estimate from 4.2-7.9 rem. The

documented reason stated, "Made corrections to EAD setpoints in the special instruction

to match the ERIMS trigger on the RWP." While this was a legitimate reason for a RWP

revision, it did not address the basis for the adjustment of the work activity dose

estimate. In the most significant example with regard to actual dose exceeding the

legitimate work activity estimate, RWP 2002-2507, Revision 2, increased the work

activity dose estimate from 4.6-6.893 rem. The reasons cited for the revision were:

(1) vendor equipment did not machine correctly, (2) workers were less efficient because

of heat stress conditions, and (3) mockup facilities used for training were inconsistent

with actual plant components. RWP 2002-2507, Revision 3, increased the work activity

dose estimate from 6.893-7.0 rem. The reasons cited for the revision were: (1) the

current projection was insufficient to cover the work on two remaining mechanical nozzle

sealing assembly clamps and (2) additional management was in the field to provide

oversight for inexperienced workers.

With respect to RWP 2002-2507, the inspectors found that equipment ineffectiveness

was a legitimate reason in some cases for revising dose estimates. However, the

licensee had previously cited equipment ineffectiveness as a basis for Revision 1 of the

RWP, in which the work activity dose estimate was increased from 3.660-4.600 rem. No

additional justification was provided to support the use of equipment ineffectiveness a

second time to justify an increase to the dose estimate. Licensee representatives stated

that similar environmental conditions existed in previous refueling outages. Therefore,

containment environmental conditions should have been anticipated. The inconsistent

mockup facility resulted from plant drawing discrepancies and, according to licensee

personnel, was not a recent development. Therefore, the licensee had sufficient

opportunity to identify and correct the situation before the refueling outage. The

inspectors determined through interviews with licensee personnel that the work on all

mechanical nozzle sealing assembly clamps was considered in the original dose

estimate and the number of clamps did not increase. The use of inexperienced workers

did not begin at this stage of the work and should have been considered from the start

and addressed through additional training. The inspectors, after consultation with NRR,

concluded that the bases for the work activity dose estimate adjustments in

RWP 2002-2507, Revisions 2 and 3, were invalid because there were inadequate

justifications for the adjustments. Consequently, the actual dose for this work activity

(RWP 2002-2507) exceeded 5 rem (7.165) and exceeded the legitimate dose estimate

(4.6 rem) by more than 50 percent.

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Analysis: The inspectors determined that this finding was associated with the

Occupational Radiation Safety Cornerstone program and process attributes (ALARA