NRC NEWS

U.S. NUCLEAR REGULATORY COMMISSION

OFFICE OF PUBLIC AFFAIRS, REGION IV

611 Ryan Plaza Drive - Suite 400

Arlington TX 76011-8064

No. IV-00-34 August 10, 2000

CONTACT: Breck Henderson

Phone: 817-860-8128

Cellular: 817-917-1227

e-mail: bwh@nrc.gov

NOTE TO EDITORS: ANO UNIT 2 INSPECTION REPORT

The U.S. Nuclear Regulatory Commission’s inspection report covering recent steam generator

tube inspections conducted by Entergy Operations Inc. at its ANO Unit 2 nuclear power plant in

Russellville, Arkansas, is complete and available to the public. Entergy performed the inspections

between July 27 and August 8 to locate any defective tubes for repair or plugging. NRC inspectors

concluded that the inspections were properly conducted and evaluated, and that repairs were completed

as necessary. A total of 189 tubes were plugged in both steam generators. The number of tube defects

found and tubes requiring plugging were consistent with expectations. NRC approval is not required

for plant restart.

The full inspection report is attached.

###
August 10, 2000

Craig Anderson, Vice President

Operations

Arkansas Nuclear One

Entergy Operations, Inc.

1448 S.R. 333

Russellville, Arkansas 72801-0967

SUBJECT: NRC's ARKANSAS NUCLEAR ONE INSPECTION REPORT NO. 50-313/00-14;

50-368/00-14

Dear Mr. Anderson:

This refers to the inspection conducted on July 27 to August 8, 2000, at the Arkansas Nuclear One,

Unit 2 facility. The enclosed report presents the results of this inspection. A technical debrief was

discussed on July 28, 2000, with Mr. M. Smith and other members of your staff. The results of this

inspection were discussed on August 8, 2000, with Mr. J. Vandergrift and other members of your staff.

The inspection was an examination of activities conducted under your licenses as they relate to safety

and to compliance with the Commission's rules and regulations and with the conditions of your

licenses. Within these areas, the inspection consisted of a selective examination of procedures and

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

inspection focused on reactor safety.

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

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

from the Publicly Available Records (PARS) component of NRC's document system (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/NRC/ADAMS/index.html (the

Public Electronic Reading Room).

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

you.

Sincerely,

/RA/

P. Harrell, Chief

Project Branch D

Division of Reactor Projects
SUMMARY OF FINDINGS

Arkansas Nuclear One

NRC Inspection Report 50-313/00-14; 50-368/2000-14

This report covers onsite inspection and in-office review of Unit 2 steam generator

inservice inspection surveillance activities. In the Reactor Safety area, the

cornerstones inspected included Mitigating Systems and Barrier Integrity.

There were no inspection findings identified in these areas.

Report Details

Summary of Plant Status

Unit 2 was shutdown for steam generator inspection (Outage 2P00-1) during this

inspection.

1. REACTOR SAFETY

Cornerstones: Mitigating Systems, Barrier Integrity

1R22 Surveillance Testing

a. Inspection Scope

The inspectors reviewed the implementation of the licensee's program for

monitoring steam generator tube degradation. As part of this review the

inspectors:

• Reviewed documents defining the scope of eddy current testing planned for

steam generator tube examinations, including number of tubes, locations,

scope expansion criteria, and specific eddy current probes.

• Reviewed the eddy current data collection, management and analysis

methods.

• Observed a sample of in-progress eddy current data collection.

• Reviewed the licensee's selection criteria for determining which steam

generator tubes were to be in-situ pressure tested.

• Reviewed the eddy current inspection results and the licensee's selection of

tubes to be in-situ pressure tested.
-2-

• Reviewed the licensee's criteria for determining which steam generator tubes

were to be plugged.

b. Findings

The licensee performed eddy current inspections of the Unit 2 steam generator tubes to

determine which tubes were defective and required repair or plugging. In addition, the

licensee performed in-situ pressure testing of a small number of defective tubes to

evaluate their structural and leakage integrity.

The eddy current inspection scope consisted of bobbin coil probe inspections of all

inservice tubes, and rotating pancake coil probe inspections of all indications identified

by the bobbin coil inspection. The inspectors determined that the planned inspection

scope was in accordance with the Technical Specifications and was being appropriately

controlled by the eddy current data management group. In addition, the observed data

collection, management and analysis methods were in accordance with plant

procedures. The licensee indicated that the inspection equipment, data collection

procedures, and data analysis techniques being used during this inspection were

essentially the same as those used during the previous steam generator inspection

(November 1999). The inspectors did not identify any activities contrary to this

assertion, and therefore would expect the sensitivity of the Outage 2P00-1 inspection to

be similar to the November 1999 inspection.

All flaws identified during the Outage 2P00-1 inspection were axial outside diameter

stress corrosion cracking at the eggcrate supports. The licensee identified 64 flaws in

58 tubes in Steam Generator A and 148 flaws in 131 tubes in Steam Generator B.

These results were within the range projected by the licensee.

The licensee performed in-situ pressure testing on a small number of defective steam

generator tubes to evaluate their structural and leakage integrity. They developed

selection criteria for determining which steam generator tubes were to be in-situ

pressure tested. The selection criteria were documented in Revision 1 of Engineering

Report ER-974855-E205, and consisted of a combination of estimated flaw length,

maximum depth, and average depth. The inspectors reviewed the in-situ selection

criteria as well as the licensee's selection of tubes to be in-situ pressure tested.

One tube in Steam Generator A and seven tubes in Steam Generator B were selected

to be pressure tested for leakage integrity at main steam line break (MSLB) pressure

differentials. One Steam Generator A tube and three Steam Generator B tubes leaked

a minimal amount at MSLB conditions. Five tubes were selected to be pressure tested

for structural integrity at three times the normal operating pressure differential (3dp),

four of which had leaked at MSLB conditions. Four of the five tubes were tested to

approximately 500 psi above 3dp with no failure. The licensee indicated that the fifth

tube (Tube 40/108) passed the 3dp pressure, but estimated that it burst at less than
-3-

100 psi above 3dp. Tube 40/108 had not met the selection criteria for pressure testing

at 3dp (the flaw was estimated to be too short to burst), but the licensee elected to

perform this test because the tube leaked when tested at MSLB pressure differentials.

Based on the 3dp pressure test results, the inspectors concluded that it appeared

possible that this flaw might have burst at less than 3dp if the flaw had been deeper.

This was a concern, because a flaw of this length would not have met the licensee's

selection criteria for pressure testing regardless of its depth. Based on the inspector's

concerns, the licensee reevaluated the selection criteria and determined that

nondestructive evaluation uncertainties had not been appropriately considered when

calculating the selection criteria. The selection criteria were modified and the licensee

evaluated the inspection results to determine whether any additional tubes required insitu pressure testing. The licensee concluded that no additional tubes required testing.

The safety significance of this finding was considered very low based on the absence of

adverse consequences. The inspectors determined that the appropriate tubes were insitu pressure tested.

There were no significant findings identified during this inspection.

4. OTHER ACTIVITIES (OA)

4OA6 Management Meetings

.1 Exit Meeting Summary

A technical debrief was discussed on July 28, 2000, with Mr. M. Smith and other

members of your staff. The results of this inspection were discussed on August

8, 2000, with Mr. J. Vandergrift and other members of your staff. The managers

acknowledged the findings presented and also informed the inspectors that no

proprietary material was examined during the inspection.
ATTACHMENT 1

PARTIAL LIST OF PERSONS CONTACTED

Licensee

S. Bennet, Licensing Specialist

M. Cooper, Licensing Specialist

M. Smith, Engineering Programs and Component Manager

D. Harrison, Engineering Programs Supervisor

J. Vandergrift, Director, Nuclear Safety

DOCUMENTS REVIEWED

Unit 2 Engineering Report

ER-974855-E205

Steam Generator Pre-Outage

Degradation Assessment and Repair

Criteria for 2P00

Revisions 1

Unit 2 Training Manual

ANO-2-OTH-ESP-SGMAN

Steam Generator Eddy Current Training

Manual

Revision 4

Engineering Standard

HES-28

ANO-2 Steam Generator Eddy Current

Examination Guidelines

Revision 12

Procedure/Work Plan

5120.500

Steam Generator Integrity Program

Implementation

Change

008-03-0

Procedure/Work Plan

5120.509

Steam Generator Inservice Inspection

Implementation Plan

Change

001-00-0

LIST OF ACRONYMS AND INITIALS USED

MSLB - main steam line break

3dp - three times normal operating pressure differentials