Text

IR 05000346-02-007, on 02/20 - 10/24/2002, Firstenergy Nuclear Operating Company, Davis-Besse Nuclear Power Station. Special Inspection
	ML023370100
Person / Time
Site:	Davis Besse
Issue date:	11/29/2002
From:	Grobe J; NRC/RGN-III
To:	Myers L; FirstEnergy Nuclear Operating Co
References
	SIP Test Sample upto2-6-04 IR-02-007
	Download: ML023370100 (36)
	v • d • e

November 29, 2002

Mr. Lew Myers

Chief Operating Officer

FirstEnergy Nuclear Operating Company

Davis-Besse Nuclear Power Station

5501 North State Route 2

Oak Harbor, OH 43449-9760

SUBJECT: DAVIS-BESSE NUCLEAR POWER STATION

NRC SPECIAL INSPECTION -REACTOR VESSEL HEAD REPLACEMENT -

REPORT NO. 50-346/02-07(DRS)

Dear Mr. Myers:

On October 24, 2002, the US Nuclear Regulatory Commission (NRC) completed a special

inspection at your Davis-Besse Nuclear Power Station. This inspection reviewed your actions

to resolve Restart Checklist Item No. 2.a, associated with the adequacy of the reactor vessel

head replacement and Restart Checklist Item No. 2.b associated with the adequacy of the

containment vessel restoration following head replacement. Specifically, this inspection

focused on review of a sample of activities as described in the Davis-Besse Reactor Head

Resolution Plan. To evaluate the implementation of this plan, our inspection included reviews

and observations in three areas under your plan: (1) non-destructive examinations performed

on the replacement head welds that occurred at the Midland Michigan site; (2) the American

Society of Mechanical Engineers (ASME) Code data packages for the replacement head; and

(3) activities associated with the temporary containment access opening and restoration.

Additionally, we reviewed the examination of the original vessel head penetration nozzles that

your staff conducted in accordance with your commitments to NRC Bulletin 2001-01,

Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles. This report

presents the results of our review.

Based on our inspection, we confirmed that: (1) adequate records were assembled to ensure

that the replacement head was designed and fabricated in conformance with ASME Code

requirements and that the original ASME Code Section III N-stamp remained valid; (2) the

engineering evaluation associated with construction of the temporary containment access

opening considered appropriate loads and demonstrated that stress in the containment shell

materials would not exceed design limits; (3) the temporary containment vessel opening was

restored such that the original ASME Code construction requirements were maintained; (4) the

work activities to construct and restore the temporary containment opening and closure

occurred in a controlled manner and in accordance with procedure requirements; and (5) that

your managers demonstrated an active oversight role for the control of the contractors on the

containment building temporary construction opening. Therefore, we concluded that the

L. Myers -2-

Davis-Besse Reactor Head Resolution Plan was effectively implemented. At the conclusion of

this inspection, your staff had not completed the final acceptance pressure tests for the vessel

head and containment vessel. Therefore, Restart Checklist Item No. 2.a and 2.b will remain

open pending completion of this testing.

In accordance with 10 CFR Part 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/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

John A. Grobe, Chairman

Davis-Besse Oversight Panel

Docket No. 50-346

License No. NPF-3

Enclosure: NRC Special Inspection Report

No. 50-346/02-07(DRS)

See Attached Distribution

L. Myers -2-

Davis-Besse Reactor Head Resolution Plan was effectively implemented. At the conclusion of

this inspection, your staff had not completed the final acceptance pressure tests for the vessel

head and containment vessel. Therefore, Restart Checklist Item No. 2.a and 2.b will remain

open pending completion of this testing.

In accordance with 10 CFR Part 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/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

John A. Grobe, Chairman

Davis-Besse Oversight Panel

Docket No. 50-346

License No. NPF-3

Enclosure: NRC Special Inspection Report

No. 50-346/02-07(DRS)

See Attached Distribution

DOCUMENT NAME: G:DRS\ML023370100.wpd

See Previous Concurrence

To receive a copy of this document, indicate in the box: "C" = Copy without attachment/enclosure "E" = Copy with attachment/enclosure "N" = No copy

OFFICE RIII * RIII * RIII RIII *

NAME MHolmberg:sd DHills CLipa JCreed (4OA3.3)

DATE 11/13/02 11/14/02 11/25/02 11/13/02

OFFICE NRR RIII

NAME TQuay JGrobe

DATE 11/26/02 11/29/02

OFFICIAL RECORD COPY

L. Myers -3-

cc w/encl: B. Saunders, President - FENOC

Plant Manager

Manager - Regulatory Affairs

M. OReilly, FirstEnergy

Ohio State Liaison Officer

R. Owen, Ohio Department of Health

Public Utilities Commission of Ohio

President, Board of County Commissioners

Of Lucas County

President, Ottawa County Board of Commissioners

D. Lochbaum, Union of Concerned Scientists

ADAMS Distribution:

AJM

DFT

SPS1

RidsNrrDipmIipb

GEG

HBC

CST1

C. Ariano (hard copy)

DRPIII

DRSIII

PLB1

JRK1

DB0350

U.S. NUCLEAR REGULATORY COMMISSION

REGION III

Docket No: 50-346

License No: NPF-3

Report No: 50-346/02-07(DRS)

Licensee: FirstEnergy Nuclear Operating Company

Facility: Davis-Besse Nuclear Power Station

Location: 5501 North State Route 2

Oak Harbor, OH 43449

Dates: February 20, 2002 through October 24, 2002.

Inspectors: M. Holmberg, Reactor Inspector, Division of Reactor

Safety, Region III

James Belanger, Senior Physical Security Inspector,

Division of Reactor Safety, Region III

Donald Jones, Reactor Inspector, Division of Reactor

Safety, Region III

Richard McIntyre, Senior Reactor Engineer, Quality and

Maintenance Section, Division of Inspection Program

Management, Office of Nuclear Reactor Regulation

Doug Simpkins, Resident Inspector, Division of Reactor

Projects, Region III.

John Jacobson, Reactor Inspector, Division of Reactor

Safety, Region III

Approved by: David Hills, Chief

Mechanical Engineering Branch

Division of Reactor Safety, Region III

TABLE OF CONTENTS

SUMMARY OF FINDINGS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2

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

4OA3 Event Follow up (IP 93812) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

.1 Davis-Besse Reactor Head Resolution Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

b. Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

c. Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4

.2 Implementation of the Davis-Besse Reactor Head Resolution Plan . . . . . . . . . . 5

b. Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5

b.1 Consumers Energy Part 21 Report Evaluation . . . . . . . . . . . . . . . . . . . . 5

b.2 Radiographic Examination of the Vessel Head Welds Conducted at

Midland Michigan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

b.3 ASME Section III Data Package For The Midland Vessel Head . . . . . . . 6

b.4 ASME Section XI Data Package for the Midland Vessel Head . . . . . . . 7

b.5 ASME Section XI Design Reconciliation for the Replacement Vessel

Head . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

b.6 Containment Access Opening . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

b.7 Containment Access Opening Restoration . . . . . . . . . . . . . . . . . . . . . . . 9

b.8 Oversight of Containment Construction Opening and Restoration . . . . 13

c. Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13

.3 Reactor Head Replacement Project Security Measures . . . . . . . . . . . . . . . . . . 14

b. Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

c. Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14

.4 Conclusions on Reactor Vessel Head Replacement Activities . . . . . . . . . . . . . 14

4OA5 Other . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

.1 Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles

(Temporary Instruction(TI)- 2515/145) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15

b. Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

c. Findings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

.2 Reactor Pressure Vessel Head and Vessel Head Penetration Nozzles (TI

2515/150) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

4OA6 Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

.1 Exit Meeting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

SUMMARY OF FINDINGS

IR 05000346-02-07; FirstEnergy Nuclear Operating Company; on 02/20-10/24/02; Davis-Besse

Nuclear Power Station. Special Inspection.

This report covers an 8-month special inspection of licensee activities associated with the

reactor vessel head examination and replacement. This inspection was conducted by a

resident inspector, inspectors based in the Region III Office, and technical staff from the

Office of Nuclear Reactor Regulation (NRR). The significance of most findings

is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter 0609,

Significance Determination Process (SDP). Findings for which the SDP does not apply may

be Green 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. Inspection Findings

No findings of significance were identified.

B. Licensee Identified Findings

No findings of significance were identified.

2

REPORT DETAILS

Background and Event Overview

On March 6, 2002, Davis-Besse personnel notified the NRC of degradation (corrosion) of the

reactor vessel head material adjacent to a control rod drive mechanism (CRDM) nozzle. This

condition was caused by coolant leakage and boric acid corrosion of the head material induced

by an undetected crack in the adjacent CRDM nozzle. The degraded area covered in excess of

20 square inches where the low-alloy structural steel was corroded away, leaving the thin

stainless steel cladding layer. This condition represented a loss of the reactor vessels

pressure retaining design function, since the cladding was not considered as pressure

boundary material in the structural design of the reactor pressure vessel. While the

cladding did provide a pressure retaining capability during reactor operations, the identified

degradation represented an unacceptable reduction in the margin of safety of one of the

three principal fission product barriers at the Davis-Besse Nuclear Power Station (reference

NRC report 50-346/02-03(DRS)).

At a public meeting held on June 4, 2002, the licensee described the reactor pressure vessel

closure head (RPVCH) replacement program for the Davis-Besse Nuclear Power Station. At

this meeting, the licensee discussed several options for resolving the degraded RPVCH

including replacement with a vessel head from the Midland Michigan plant. The Midland

Michigan plant had previously halted construction and was never completed. In a letter to the

NRC dated August 9, 2002, the licensee described a plan for the replacement of the RPVCH

using the Midland RPVCH. Because the Midland RPVCH was similar in design to the

Davis-Besse RPVCH and was readily available, the licensee chose this option over repairing

the existing RPVCH or fabricating a new RPVCH. The licensee issued the Davis-Besse

Reactor Head Resolution Plan (DBHRP) which described the project management, planning,

and execution of tasks needed to remove the replacement RPVCH from the containment

building at Midland and subsequently install the replacement RPVCH on the reactor vessel

at Davis-Besse. The NRC inspectors reviewed the licensee activities associated with the

DBHRP during this inspection. Additionally, this inspection included review of the

nondestructive examinations (NDE) of the original RPVCH nozzles that the licensee conducted

in accordance with commitments to NRC Bulletin 2001-01, Circumferential Cracking of Reactor

Pressure Vessel Head Penetration Nozzles. Given the high public interest in this subject area

at Davis-Besse, and therefore the need to clearly communicate the rationale for NRC staff

conclusions regarding the licensees RPVCH replacement activities, this report documents the

inspectors observations.

3

4. OTHER ACTIVITIES

4OA3 Event Follow up (IP 93812)

.1 Davis-Besse Reactor Head Resolution Plan (DBHRP)

a. Inspection Scope

On May 24, 2002, the licensee issued the DBHRP, Revision 0 and on July 10, 2002,

issued Revision 1 to the DBHRP. The NRC inspectors reviewed the DBHRP to evaluate

the adequacy of the planned work scope and licensee staffing.

b. Observations

The DBHRP described the project management, planning, and execution of tasks

needed to remove the replacement Reactor Pressure Vessel Closure Head (RPVCH)

from the containment building at Midland and subsequently install the replacement head

on the reactor vessel at Davis Besse. In this plan, the licensee described the activities

necessary to support the RPVCH replacement which included:

C procurement and certification;

C RPVCH modifications;

C temporary fuel removal;

C access through the Davis-Besse containment opening;

C installation of new RPVCH;

C restoration, inspection and testing of the RPVCH and containment;

C storage and disposition of original RPVCH; and

C updating the design and licensing basis.

The inspectors noted that the original (Revision 0) of the DBHRP defined a contractor

project team with clearly designated responsibilities including history of related work

experience. The inspectors reviewed the personnel work histories for contractors and

licensee personnel assigned to this project. Based on this review, the inspectors

concluded that the licensee had assembled head replacement personnel with extensive

experience in the nuclear industry and with extensive experience in similar engineering

projects, such as steam generator replacement. The inspectors considered that the

scope of the DBHRP and qualifications of project personnel were sufficient to

accomplish the head replacement project. The licensee subsequently removed the

specific work experience and education history of the personnel assigned to this project

from the DBHRP (Revision 1), but did not substantively change the actual project

organization or staffing.

c. Findings

No findings of significance were identified.

4

.2 Implementation of the Davis-Besse Reactor Head Resolution Plan

a. Inspection Scope

To evaluate the implementation of the DBHRP the inspectors reviewed activities in three

areas under this plan:

1) From June 12, 2002, through June 14, 2002, the inspectors reviewed NDE performed

on the replacement head welds that occurred at Midland, Michigan;

2) From August 20, 2002, through August 22, 2002, the inspectors reviewed the

American Society of Mechanical Engineers (ASME) Code data packages and

Consumers Energy 10 CFR Part 21 Report for the replacement head; and

3) From August 13, 2002, through October 24, 2002, the inspectors observed the

activities and reviewed records associated with the containment building access opening

and restoration.

b. Observations

The licensee purchased the Midland RPVCH from Framatome ANP, who in turn had

purchased the head from Consumers Energy, the owner of the Midland, Michigan plant.

The Midland, Michigan plant had not been completed (construction permit was issued in

1972 and construction work was suspended in 1984). The Midland RPVCH had not

been placed in service since original fabrication, and had been stored inside the

containment building at the Midland, Michigan site. To confirm that the Midland head

could be used at Davis-Besse, the licensees vendor (Framatome ANP, Inc.) performed

non-destructive examination (NDE) and reviewed head fabrication documentation.

Specifically, Framatome provided the licensee with the required documentation, NDE,

analyses and ASME Code reconciliation necessary to ensure the original ASME Code

N-stamp documentation was valid, and that the Midland RPVCH complied with

applicable NRC and industry requirements. This documentation was assembled by

Framatome ANP in ASME Code Section III and XI Quality Assurance (QA) data

packages. Framatome ANP, Inc. notified the NRC, by letter dated September 9, 2002,

that the Midland head conformed to all ASME Code Section III, Class A requirements,

that the supporting documentation was valid, and that all markings and identification

symbols matched the head configuration.

b.1 Consumers Energy Part 21 Report Evaluation

On May 23, 2002, pursuant to the reporting requirements of 10 CFR Part 21.21(b),

Consumers Energy notified the NRC by letter, that the status of the Midland head was

indeterminate, because the Midland reactor head had been in storage since 1986

without any routine maintenance or any oversight of a formal QA program.

The inspectors reviewed preliminary safety concern PSC 3-02, prepared by the

licensees vendor (Framatome ANP) to address this issue. In this document,

Framatome ANP identified the concern, the cause, and corrective actions. The

5

completed corrective actions included an in-depth review of the original NSS-13 QA

Data Package for the reactor vessel closure head, verification of Code markings, full

visual and non-destructive re-examination of all the vessel head welds and completion

of the ASME Code Section XI pre-service NDE.

On September 12, 2002, Framatome ANP informed the NRC by letter, that the reactor

vessel head intended for use at Midland Unit 2 was manufactured in accordance with

the requirements of ASME Code Section III, Class A, 1968 Edition with Summer 1968

Addenda. In this letter, Framatome ANP asserted that the this head, with a few very

minor variations, was identical to the original reactor vessel head installed at

Davis-Besse, including configuration, materials, pressure and temperature rating, and

vessel interface parameters.

The inspectors reviewed the corrective actions discussed in PSC 3-02 and considered

the completed actions adequate to address the lack of QA controls identified in the

Part 21 notification for the Midland reactor head.

b.2 Radiographic Examination of the Vessel Head Welds Conducted at Midland, Michigan

The licensee repeated radiographic examinations of the Midland head dome-to-flange

weld and the control rod drive nozzle-to-flange weld to replace the original radiographic

examination (RT) records, which could not be located. However, three lifting lugs on the

closure head dome, spaced 120 degrees apart, prevented a complete examination of

the dome-to-flange weld. The licensee subsequently determined that only 95 percent of

the head dome-to-flange weld had been examined. Therefore, on August 1, 2002, the

licensee submitted a letter to the NRC requesting relief (RR-A26 and RR-A27) from the

ASME Code requirements to perform a 100 percent examination of this weld and to

have the original RT records. The inspectors reviewed the RT records completed on

these welds and did not identify any other deviations from Code requirements.

b.3 ASME Section III Data Package For The Midland Vessel Head

The inspectors reviewed the ASME Code Section III portions of the RPVCH

documentation package assembled as Framatome QA Data Package

No. 23-5018698-00. The data package was prepared in accordance with the

Framatome ANP safety-related Quality Manual No. 56-5015885-00, which was

audited and approved by the licensee.

The RPVCH was originally fabricated at the Babcock & Wilcox Mount Vernon Works for

Consumer Power Company and was designed in accordance with the ASME Boiler and

Pressure Vessel Code,Section III, 1968 Edition, Summer 1968 Addenda. The RPVCH

design pressure was 2500 pounds-per-square-inch-gage and design temperature was

650 degrees Fahrenheit. The RPVCH ASME Code edition, design pressure and

temperature were the same as the original Davis Besse vessel head. The data package

included records which demonstrated that the Midland vessel head components were

stress-relieved at 1100 degrees Fahrenheit for sufficient time to meet Code

requirements. The CRDM nozzles and structural J-groove welds received no post weld

6

heat treatment in order to limit distortion, as allowed by the Section III Code

requirements.

The Section III Code QA data package included rubbings of the name plate N-stamp for

the reactor vessel and closure head (the N number is the same for both). The rubbing

included the design pressure and temperature, the hydrostatic test pressure, and the

date of manufacture (1975). The inspectors confirmed that this package contained

records required by the Code including: the design specifications, design analyses,

drawings, NDE records, hydrostatic test records and certified material test reports for

pressure boundary materials.

Based on review of the this package, the inspectors concluded that sufficient records

existed to confirm that the RPVCH was fabricated in accordance with the ASME Code

Section III and construction QA requirements.

b.4 ASME Section XI Data Package for the Midland Vessel Head

The inspectors reviewed the ASME Code Section XI portion of the RPVCH

documentation package assembled as Framatome QA Data Package

No. 23-5019258-00. The data package was prepared in accordance with the

Framatome ANP safety-related Quality Manual No. 56-5015885-00, which was

audited and approved by the licensee. Additionally, the inspectors reviewed the

documentation in process traveler, 50-5018614-00, for modification and preparation

of the Midland RPVCH, which contained supplemental NDE records. The NDE

completed by the licensee exceeded the minimum required by the Code and included:

C visual examination of the entire RPVCH to identify signs of degradation or

evidence of welding while the head was in storage at Midland;

C dye penetrant examination (PT) records of all 69 CRDM J-groove welds;

C ultrasonic examination (UT) Examination of all 69 CRDM nozzles;

C PT of the head cladding at 6 sample areas;

C PT & RT of all 69 CRDM flange-to-nozzle welds;

C RT of closure head-to-flange weld;

C magnetic particle examination (MT) of the head lifting lug attachments;

C UT and MT of the closure head-to-flange weld; and

C eddy current examination of all 69 CRDM nozzle internal surfaces.

Based on review of the ASME Section XI data package and supporting documentation

for supplemental NDE conducted on the RPVCH, the inspectors concluded that

adequate records existed to confirm that the RPVCH was designed and fabricated in

conformance with ASME Code requirements and that the original ASME Code

Section III N-stamp remained valid.

b.5 ASME Section XI Design Reconciliation for the Replacement Vessel Head

The ASME Code Section XI required reconciliation of any differences which may

exist for the replacement Code component in design, fabrication and examination

requirements to ensure that the replacement component is satisfactory for the

7

specified design and operating conditions. In a letter dated August 9, 2002, the

licensee informed the NRC of the ASME Code reconciliation activities to be completed

for the replacement RPVCH. In this letter, the licensee included a summary table for

ASME Section XI, Article IWA-4000 Repair/Replacement Activities. The material

presented in this summary table, along with the supporting vendor documents

(51-5019457-00 & 01, Davis Besse RV Closure Head Replacement Reconciliation,

and 33-5019877-00, Davis Besse Original Closure Head Replacement Design Report,

and RPVCH drawings), provided the record of licensee activities with respect to

performing the reconciliation of Code requirements.

The inspectors confirmed that the Davis Besse Original Closure Head Replacement

Design Report was prepared, reviewed and approved by qualified personnel and was

certified by two registered professional engineers, who specialized in ASME Section III

Code stress analysis. Further, the inspectors confirmed that the registered professional

engineers performed independent design reviews. Therefore, the inspectors concluded

that the Design Report and supporting documents provided an adequate basis for the

ASME Section XI reconciliation of the RPVCH.

b.6 Containment Access Opening

The Davis-Besse containment lacked an access opening of sufficient size to permit

removal of the old RPVCH and reinstallation of the new RPVCH. Therefore, the

licensee cut a temporary access opening in the shield building and containment vessel

of sufficient size to support the RPVCH replacement. The licensee performed a

detailed engineering evaluation of the work activities associated with construction of

the containment access opening as documented in engineering work request (EWR)

02-0146. This package included a Design Report, in which the licensee evaluated the

design requirements applicable to the containment work activities. In this report, the

licensee reviewed design requirements applicable to the construction of containment

vessel and shield building access openings, temporary containment reinforcement, head

rigging and transport, ventilation, and restoration of the construction openings. The

licensee did not perform a 10 CFR 50.59 safety evaluation for activities associated with

this containment access opening because it was considered a maintenance activity, that

did not change the design.

In EWR 02-0146, the licensee identified supporting calculations and applicable

design requirements for the construction openings in the metal containment vessel

and the concrete containment shield building. The inspectors reviewed calculation

12501-C-003, Evaluation of Containment Vessel for Construction Opening, which

confirmed adequate structural integrity for a 20 feet wide by 20 feet high square

opening in the containment vessel. The inspectors noted that this finite element

calculation bounded the size of the actual containment vessel construction opening,

which was 13 feet high and 18 feet wide. This calculation considered appropriate loads

for the de-fueled plant conditions, which included, seismic, tornado induced, dead

weight and rigging loads. The calculation demonstrated that for five load combinations,

which included polar crane dead loads combined with seismic induced loads, stress in

the containment shell materials would not exceed Code design limits.

8

The inspectors observed the following activities associated with construction of the

temporary access opening in containment.

b.6.1 Shield Building Access Opening

The inspectors observed the licensee contractor cutting the construction access opening

in the shield building to support the vessel head replacement. The access cut in the

containment shield building was 16.5 feet high and 21.5 feet wide. The opening was

made by the licensee contractor using a hydro demolition technique (high pressure

water jet process) to remove the concrete. This high pressure water jet process left the

original rebar intact and undamaged. The inspectors noted that even the fine rebar tie

wire remained intact. The licensee contractor followed the work order 02-003545-10

which implemented the contractor Work Plan and Inspection Record (WPIR) C-CRA-02,

Cut Vessel Plate. The licensee contractor changed the cutting process from a saw cut

to a torch cut for removal of the rebar in the shield building wall. The licensee approved

this change based on a contractor demonstration of the capability of the torch cutting

process to maintain under the maximum allowed 1/4 inch rebar cut gap.

b.6.2 Containment Vessel Access Opening

The inspectors observed the licensees contractor performing demonstration cuts for the

containment vessel on a containment mockup. The contractor used a large flat vertical

steel plate with the same thickness as the containment vessel for this mockup. The

contractor used an oxyacetylene torch head mounted on a motorized track assembly to

produced an accurate and repeatable cut line on the mockup plate. The contractor had

welding personnel demonstrate proficiency by performing several practice cuts on the

mockup plate. The inspectors also observed the welders completing welds that

attached steel I-beams to the mockup plate. These welds were intended to simulate

conditions during the installation of reinforcement I-beams added to the periphery of the

access cut to stiffen the containment vessel.

The inspectors observed the licensee contractor performing torch cutting of the

containment access opening using a track mounted cutting torch. The licensees

contractor followed the work order number 02-003545-013 and WIPR C-CLP-02

controlling this process to cutout the rectangular 13 feet high by 18 feet wide access

plate in the 1.5 inch thick containment vessel. The inspectors did not identify any

deficiencies in the containment cutting process observed.

b.7 Containment Access Opening Restoration

The licensee reinstalled the plate section removed from the containment vessel for the

temporary construction opening using a manual shield metal arc welding (SMAW)

process. The licensee reinstalled this 1.5 inch thick plate with full penetration butt

welds, such that the original ASME Code (Section III, of the 1968 Edition, 1969 Summer

Addenda) vessel construction requirements were met.

For the shield building, the licensee reinstalled original rebar removed during the

construction of the access opening and poured new concrete to close the shield

9

building. The reinstallation of rebar and concrete conformed to original design

requirements except for the requirement to test samples of the production cadweld

splices as discussed below.

b.7.1 Welder Qualifications for Containment Closure Welding

The inspectors observed the licensee contractor performing qualification of welders

used to fabricate the containment access closure weld. The inspectors reviewed the

RT film of welds performed by seven welders during the qualification process. The

contractor applied qualification weld acceptance criteria in accordance with procedure

96-RT-005, General Radiographic Procedure Per ASME Section V Article 2. The

inspectors confirmed that this procedure contained welder qualification requirements

and acceptance criteria which were consistent with the requirements of the ASME Code,

Section IX, 2001 Edition. The licensees contractor had made conservative decisions in

applying weld acceptance criteria for qualification welds. The licensees contractor

followed Code requirements, which included retesting welders with initial unsatisfactory

welds. Two of the seven welders failed to produce satisfactory welds for this manual

SMAW process and were not qualified by the licensees contractor.

b.7.2 Containment Vessel Access Opening Closure Welding and Radiographic Examination

The NRC inspectors observed welding of the root pass on the closure plate to the

access opening in the containment performed by the licensees contract welders. The

inspectors confirmed that the welding electrodes E7018, 1/8 inch diameter (designated

as PCI 3229) used during this activity were of the correct material with appropriate Code

records (e.g. Certified Material Test Reports). The licensees contractor properly stored

weld electrodes in a holding oven at 270 degrees Fahrenheit. The required base-metal

preheat temperature was measured and confirmed by licensee Quality Control (QC)

inspectors to be within acceptable range (294 degrees Fahrenheit). The licensees

QC inspectors also measured the heat input parameters (amperage, voltage and

travel speed) used during the welding activities using calibrated meters. Based on

these measured values (for welder M-991), the weld heat inputs used were well below

the maximum of heat input of 130,909 Joules per inch required by the procedure

(1 MN-GTAW/SMAW-1). The NRC inspectors concluded that the overall weld quality

appeared good and that the activities were being appropriately monitored by licensee

QC inspectors. Additionally, the NRC inspectors confirmed that welders observed were

qualified in accordance with Code requirements.

The licensees contractor performed SMAW of the containment closure plate, in

accordance with weld procedure 1 MN-GTAW/SMAW-1. The inspectors confirmed that

this welding procedure met qualification requirements from the ASME Code Section IX

and impact testing requirements as specified in the original construction Code (ASME

Code Section III, 1968 Edition, Summer 1969 Addenda). The containment access

opening closure weld, was fabricated such that the original Code requirements

(e.g. basemetal thickness and minimum preheat requirements) were met allowing the

licensee to exempt a post weld heat treatment on the containment vessel.

10

The inspectors reviewed RT records of the containment closure welds to confirm that

the these records met ASME Code acceptance criteria. The inspectors identified minor

porosity and slag which had not been recorded on the reader sheets for these RT

records. The inspectors also identified a base metal indication adjacent to weld

CS-01D, view 48-60, which had not been recorded or evaluated. The licensee staff

subsequently visually verified the indication as an acceptable surface indication and

noted this on the reader sheet. An indication on weld CS-01C, view 156-168 was

documented as surface contour, however visual verification was again not recorded on

the reader sheet. The licensee subsequently visually verified this surface indication and

annotated this on the reader sheet. The inspectors considered these examples to

constitute minor documentation discrepancies. The key quality requirements for these

RT records such as selection/placement of the penetrameter and the readily visible 2T

penetrameter hole were in accordance with Code requirements. Overall, the inspectors

concluded that the quality of the RT records was good, weld interpretation was generally

conservative, and that indications identified were well within Code acceptance limits.

b.7.3 Welder Qualification for Shield Building Rebar

The inspectors observed the licensee contractor performing qualification of two

welders for shield building reinforcing bar (rebar) welds using number 8 and number 11

rebar. The contractor used procedure P1-Rebar (0.64 CE) with E9018 filler material

and a SMAW process to fabricate the qualification welds. The qualification testing

included pull tests and acid etch testing as specified by the American National

Standards Institute/American Welding Society D1.4-98, Structural Welding

Code-Reinforcing Steel. The inspectors confirmed that acceptance criteria for this

qualification testing met Code requirements. The inspectors noted that the licensee

was using paragraph 6.1.2.2, of D1.4-98 which allowed using this Code instead of the

previous Code American Welding Society D.12.1 referenced in Section 3.8.2.7 of the

Updated Final Safety Analysis Report (UFSAR). The licensee had appropriately

reviewed this change to the UFSAR referenced Code as documented in the 10 CFR

50.54f screening dated August 24, 2002.

The inspectors observed the licensee contractor performing qualification tests of four

cadwelders using number 8 and number 10 rebar. The contractor used procedure

CP-C-2 Cadweld Rebar Splices with a ferrous filler materials designated PBF 105 for

number 10 rebar and PBF 70 for the number 8 rebar. The inspectors identified that the

licensee did not have the vendor documentation on the job site which confirmed which

material should be used with a given size of rebar. Specifically, the licensee was using

longer sleeves with a different filler number referenced than the standard configurations

identified on the vendor table. The licensees contractor reportedly had discussed this

configuration with the cadweld vendor, but did not have documentation from the cadweld

vendor accepting the specific configuration that was being used. This issued prompted

the licensee to issue a stop work order for the contractor that remained in affect until

improvements in management and QA oversight were implemented. The licensee

subsequently contacted the cadweld vendor and obtained documentation to confirm that

the correct cadweld splice and filler material configuration was being used. The

licensee documented this issue in nonconformance report number 009, condition report

(CR) 02-05486 and CR 02-05548.

11

Tensile testing of a sample of the production cadweld splices was required during the

original construction of the shield building (reference Appendix 3B of the UFSAR). The

licensee did not conduct testing of the production cadweld splices for the rebar

reinstalled during restoration of the temporary containment shield building construction

opening. Instead, the licensee chose to conduct tensile tests of cadwelds performed on

removable "sister" splices, which are made using the same method and at the same

location. The licensee performed an evaluation under requirements of 10 CFR 50.54

which allow changing the plants QA program requirements. In this evaluation, the

licensee concluded the change to adopt the 1995 Section III Code, Paragraph

CC-4333.5.2, requirement to test sister splices, instead of production cadweld splices,

did not constitute a reduction in commitments. This conclusion was based on NRC

review and approval for this alternative in support of the D.C. Cook steam generator

replacement project (reference NRC safety evaluation dated November 7, 2000). The

inspectors discussed this application of 10 CFR 50.54 requirements with NRR staff and

no deficiencies were identified.

b.7.4 Containment Shield Building Concrete Restoration

The inspectors observed the delivery and placement of concrete used to restore the

access opening in the shield building. The licensee contractor performed this evolution

in accordance with WP&IR C-SWR-01, Shield Building Restoration. The inspectors

noted that this activity was observed by licensee and contractor QA personnel. The

licensee QA personnel questioned the assumptions made by the contractor regarding

the reduction in air content as the concrete was pumped from the truck to the point of

placement. This prompted the licensee contractor to take additional samples to confirm

these assumptions. The inspectors considered that this action demonstrated an active

oversight role by the licensee.

The concrete used for shield building restoration was required to meet acceptance

criteria of Specification 12501-C-321, Technical Specification for Purchase of Safety

Related Ready Mix Concrete, for slump and air content at the point of delivery. After

1/3 of the first concrete truckload was placed inside the forms for the access opening,

the licensee contractor made the measurements for slump and air content used to

accept the concrete. The contractor measured the air content at 2.8 percent, which

was below the required range of 3 to 6 percent. The licensee contractor subsequently

resampled the concrete from the same wheelbarrel used for the first sample, and got

an acceptable reading of 3.4 percent. The licensee issued nonconformance report

number 017 to record the initial out of specification reading. The inspector noted that in

accordance with ASTM C 94/C 94M-00, Standard Specification for Ready-Mix

Concrete, paragraph 16.6, if the second sample had been outside specified limits, the

concrete shall be considered failed. Because the second sample passed, the licensee

considered the concrete acceptable. However, the licensee conservatively chose to not

install the remaining 2/3 of the first truckload of concrete. The inspectors discussed this

issue with cognizant NRR staff and no technical concerns were identified. Additionally,

the inspectors reviewed the licensees vendor report 150-20129-34, Report of Tests on

Cylinder Compressive Strengths, which documented the shear strength of the concrete

cylinder samples from concrete used in restoration of the shield building. In this report,

the licensees vendor documented the that these concrete cylinder samples had

12

compressive shear strength in excess of 5000 psi after only 7 days. This value

exceeded the minimum 4000 psi minimum strength required at 28 days as discussed in

Section 3.8.2.7, Materials, of the UFSAR. Therefore, the inspectors concluded that

the strength of the concrete used in the containment shield building restoration

exceeded the minimum design requirements.

After removal of the concrete forms, the licensee identified several voids exposing rebar

at six areas on the inside face of the shield building wall and two areas on the outside

face. These areas were typically near the top of the construction access opening and

the deepest void measured 8 inches in depth (reference CR 02-07472 and 02-07080).

On October 3, 2002, NRR and Region III staff held a tele-conference with licensee

personnel to discuss the cause and corrective actions for this condition. The licensee

stated that the voids observed were caused by air trapped at the top of the construction

opening that prevented a complete fill. In addition, the licensee identified areas where

the concrete surface had a rough honeycomb texture. The licensee stated that the

honeycomb areas were caused by inadequate vibration of concrete in areas between

the forms and rebar mats. The licensees planned corrective actions included chipping

the honeycomb/voids back to sound concrete and filling the cavity with a concrete grout.

The inspectors confirmed that the licensee corrective actions proposed for the voids was

consistent with the governing procedure CP-C-1, Concrete Operations. The licensee

staff concluded that no internal voids could exist because of the adequate consolidation

(e.g., no trapped air) of the concrete. This conclusion was based on the relatively large

area between the inner and outer rebar mats which provided adequate access for the

concrete vibration tools used to consolidate the concrete.

b.8 Oversight of Containment Construction Opening and Restoration

The licensee managers and QA personnel performed independent observation of

contractor activities associated with construction and restoration of the containment

access opening. The QA observations included contractor activities which occurred on

backshifts and weekends. Based on these observations, the licensee identified lapses

in contractor oversight of work activities related to the construction of the temporary

containment opening. For example, the licensee QA personnel identified that the

contractor had not met acceptance criteria during a trial test run for concrete delivery to

the site and that the contractor was not providing quality control personnel to monitor

backshift evolutions (CR 02-05108). Additionally, the NRC inspectors identified a lack of

documentation associated with a nonstandard configuration used in the qualification of

cad-welders. In response to these issues, the licensee initiated a stop work order

(CR 02-05548) until the contractor placed additional quality control oversight on work

activities. These actions indicated that the licensee was actively engaged in oversight

of contractor activities associated with construction and restoration of the containment

access opening.

c. Findings

No findings of significance were identified.

13

.3 Reactor Head Replacement Project Security Measures

a. Inspection Scope

During the baseline inspection conducted from July 29, 2002, through August 2, 2002,

the inspectors reviewed security plans onsite with the licensee Security Manager and

observed the areas where the additional physical protection measures would be

established.

On July 8, 2002, the licensee provided the inspectors a detailed description of the

security measures planned to address the Reactor Head Replacement Project via

secure telephone. The inspectors evaluated the adequacy of these measures.

b. Observations

The inspectors concluded that the additional physical protection measures appeared to

be consistent with the licensees plans and provided appropriate interim security

measures. Further, the licensee's description and plans for the security measures to

address the Reactor Head Replacement Project appeared to be appropriate and well

designed.

c. Findings

No findings of significance were identified.

.4 Conclusions on Reactor Vessel Head Replacement Activities

The licensee records were adequate to confirm that the RPVCH was designed and

fabricated in conformance with ASME Code requirements and that the original ASME

Code Section III N-stamp remained valid. Further, the licensees vendor corrective

actions were adequate to resolve the lack of QA controls identified in the 10 CFR

Part 21 notification for the RPVCH.

The licensee performed a detailed engineering evaluation of the work activities

associated with construction of the temporary containment access opening which

supported the head replacement. This evaluation included a calculation for the

construction opening which considered appropriate loads and demonstrated that

stress in the containment shell materials would not exceed design limits.

The licensee restored the temporary containment vessel construction, such that, the

original ASME Code construction requirements were maintained. The inspectors

confirmed that the licensee staff adhered to Code requirements during welder

qualifications and containment closure welding.

The licensee work activities to construct and restore the temporary containment

opening and closure occurred in a controlled manner and in accordance with procedure

requirements.

14

The licensee managers demonstrated an active oversight role for the control of the

contractors on the containment building temporary construction opening. Specifically,

the QA personnel performed independent observation of contractor activities associated

with the temporary containment access opening and initiated appropriate actions to

improve contractor QA for lapses in the quality of work activities that were identified.

At the conclusion of this inspection, the licensee had not yet completed the final

acceptance pressure tests for the vessel head and containment vessel. Therefore, NRC

Restart Checklist Item No. 2a, associated with the adequacy of the reactor vessel head

replacement and NRC Restart Checklist Item No. 2b associated with the adequacy of

the containment vessel restoration following head replacement will remain open pending

completion of this testing.

4OA5 Other

.1 Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles

(Temporary Instruction(TI)- 2515/145)

a. Inspection Scope

On August 3, 2001, the NRC issued Bulletin 2001-01, Circumferential Cracking of

Reactor Pressure Vessel Head Penetration Nozzles, in response to circumferential

cracking identified in CRDM penetration nozzles at Oconee Nuclear Station Units 2 and

3, along with axial cracking in the J-groove welds of additional CRDM nozzles at these

facilities and at Oconee Nuclear Station Unit 1 and at Arkansas Nuclear One Unit 1.

This phenomenon raised concerns regarding the potential safety implications of the

active degradation mechanism (PWSCC) and compliance with applicable regulatory

requirements. Therefore, the NRC issued TI-145, to implement an NRC review of

licensees activities in response to NRC Bulletin 2001-01. The Davis-Besse Nuclear

Power Station was in the sub-population of plants (Bin 2) that have high susceptibility to

vessel head penetration cracking (e.g., susceptibility ranking of less than 5 effective full

power years from the Oconee Unit 3 condition).

From February 20, 2002, through March 6, 2002, the inspectors performed a review of

the licensees activities in response to commitments made to NRC Bulletin 2001-01. To

assess the licensees efforts in conducting an effective examination of the reactor

vessel head penetration nozzles, the inspectors review included:

C observation of the licensees UT and visual examination of the reactor vessel

head penetrations,

C interviews with the licensees contract NDE personnel,

C review of NDE procedures, and

C review of the head inspection NDE reports.

Additionally, the inspectors observed the repair activities implemented on the cracked

vessel head penetration nozzles.

15

b. Observations

Summary

The licensee identified five penetration nozzle locations (1, 2, 3, 5 and 47) with axial

crack indications. The licensee also determined that penetrations 1, 2 and 3 contained

through-wall cracks based on UT. In addition, for penetration nozzle number 2, the

licensee identified a circumferentially oriented indication just above the J-weld which

extended for about 35 degrees. The licensee initiated repairs on the five cracked

penetrations discussed above, and during the machining process on nozzle number 3,

the licensee identified movement of the nozzle. The licensee subsequently cleaned

boric acid deposits from the head during investigation of this phenomena and

discovered a large cavity in the vessel head. An NRC augmented inspection team

performed an inspection of this issue and the NRC teams conclusions were

documented in NRC IR 50-346/02-03.

Evaluation of Inspection Requirements

In accordance with requirements of TI-145, the inspectors evaluated and answered the

following questions:

a. Was the examination:

1. Performed by qualified and knowledgeable personnel? (Briefly describe the

personnel training/qualification process used by the licensee for this activity.)

Top of Vessel Head Visual Examinations

Yes. The licensee conducted remote visual examination of the head with

knowledgeable personnel certified to Level II or III as VT-1 and VT-2 examiners in

accordance with programs meeting the American Society for Nondestructive Testing

(ASNT) Recommended Practice SNT-TC-1A and CP 189.

UT of Penetration Nozzles

Yes. The licensee conducted UT with personnel certified to Level II and Level III in

accordance with programs meeting ASNT Recommended Practice SNT-TC-1A and CP

189. A portion of the licensees UT personnel also had Electric Power Research

Institute Performance Demonstration Initiative qualifications which met ASME Code

Section XI, Appendix VIII requirements. Further, the lead UT analyst had experience

analyzing CRDM penetration UT data at the Oconee Units.

2. Performed in accordance with approved and adequate procedures?

Top of Vessel Head Visual Examinations

Yes. The licensee conducted visual examinations in accordance with

procedure 54-ISI-367-03, Procedure for Visual Examination for Leakage of

Reactor Vessel Head Penetration. The licensees visual inspection scope included

all vessel head penetrations and the visual examination method met visual quality

16

standards established for remote VT-1 examinations as defined in Section XI of the

ASME Code.

Ultrasonic Penetration Examinations

Yes. The licensees contractor conducted UT in accordance with procedure

54-ISI-100-08, Remote Ultrasonic Examination of Reactor Head Penetrations.

This procedure included instructions for UT equipment setup, calibration and sizing

of indications. The licencees contractor performed an on-site demonstration of the

effectiveness of this procedure at detecting PWSCC using RPVCH penetration nozzles

(removed from an Oconee Unit) that contained PWSCC.

3. Adequately able to identify, disposition, and resolve deficiencies?

Top of Vessel Head Visual Examinations

No. Due to the presence of boric acid and corrosion deposits, the licensee was unable

to inspect 12 CRDM nozzle locations. The remaining penetrations were partially

obscured such that none of the penetrations could be positively excluded as a potential

source of RCS leakage. The licensee subsequently removed the boric acid deposits

and identified a large cavity around penetration nozzle number 3.

Ultrasonic Penetration Examinations

Yes. The licensees contractor performed UT system calibrations at 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> intervals

on calibration standards which contained outside diameter notches. These UT

examinations were conducted from the inside of the penetration and data was recorded

from at least 1 inch above the nozzle J-weld groove weld to the end of the penetration

tube. The licensee examined each vessel head nozzle penetration tube with a blade

type UT probe. This probe head contained UT transducers setup for time-of-flight-

diffraction oriented such that it provided maximum sensitivity for circumferentially

oriented cracks near the outside diameter of the tube. Based on this examination, the

licensees contractor identified six penetrations with flaw/crack type indications. The

licensee subsequently used a rotating head UT probe installed from above the head at

each of the six penetrations with indications. The rotating UT probe contained several

transducers set up for time-of-flight-diffraction which were designed to maximize

response of cracks oriented in both the circumferential and axial direction. Additionally,

this probe contained a 0 degree and 60 degree shear wave transducers. Based on the

UT examination using the rotating probe, the licensee confirmed that cracks existed in

five of the six penetrations identified by the blade UT probe. The licensee subsequently

initiated repairs on these five cracked penetration nozzles.

4. Capable of identifying the primary water stress corrosion cracking phenomenon

described in the bulletin?

Top of Vessel Head Visual Examinations

No. The vessel head nozzle penetrations were obscured by boric acid and corrosion

deposits such that the licensee could not exclude any nozzle from having potential

RCS leakage.

17

Ultrasonic Penetration Examinations

Yes - for the vessel head penetration nozzle tubes. The licensee used an UT technique

demonstrated on nozzle penetration tubes removed from the Oconee Nuclear Power

Station, to be effective for identifying PWSCC in the penetration tube materials.

Further, the licensee used a calibration standard of similar material and dimensions as

the head penetration tubes. This standard contained both axial and circumferential

oriented notches located at the outside surface. Therefore, the inspectors concluded

that the UT method used would be effective at detecting PWSCC in the penetration

nozzle tubes.

No - for the J-welds. The UT technique used by the licensee was not designed to detect

PWSCC within the J-weld region attaching the nozzle to the RPVCH. The primary

inspection technique (blade probe) relied on a pitch-catch type UT method, in which the

crack interferes with the sound path reflecting off the back-wall of the nozzle tube.

However, at the J-weld location, PWSCC could exist beyond the back-wall of the nozzle

which would not be within the sound path demonstrated as effective for detection of

PWSCC. Therefore, the inspectors concluded that PWSCC could not reliably be

detected if it was entirely contained in the J-weld region attaching the penetration nozzle

tube to the vessel head.

b. What was the condition of the reactor vessel head (debris, insulation, dirt, boron from

other sources, physical layout, viewing obstructions)?

Top of Vessel Head Visual Examinations

The reactor head was covered with reflective metal insulation panels installed on a

support structure over the top of the reactor head. The licensee conducted the remote

camera visual inspection under the insulation support structure using a camera mounted

to a pole and other cameras mounted to a remote crawler. The as-found head condition

was not sufficiently clear of boric acid deposits to determine if these deposits may have

been the result of RCS leakage through cracked RPVCH nozzles.

Ultrasonic Examinations

The surface of the inner bore of the penetration nozzle tubes was sufficiently smooth,

such that the UT was not affected and the licensee was able to achieve full coverage of

each penetration nozzle.

c. Could small boron deposits, as described in the bulletin, be identified and

characterized?

Top of Vessel Head Visual Examinations

No. The inspectors observed deposits of boric acid and corrosion products at each

nozzle that precluded a meaningful determination of which nozzles could be sources of

RCS leakage. Therefore, the licensee relied on the effectiveness of the UT to detect

nozzle cracking and associated RCS leakage.

d. What materiel deficiencies (associated with the concerns identified in the bulletin)

were identified that required repair?

18

Of the six penetrations that the licensee identified as having UT indications, five were

selected for repair. For penetration nozzle number 58, the UT blade probe had detected

a small axial indication. However, the licensee used the top-down rotating UT probe to

confirm that cracking was not present in this nozzle and thus, did not require repair. For

the remaining five penetration locations (1, 2, 3, 5 and 47) the licensee identified axial

crack indications. The more significant axial crack indications typically traversed the full

width of the J-weld. In addition, for penetration nozzle number 2, the licensee identified

a circumferentially oriented indication just above the J-weld which extended for about

35 degrees. The licensee concluded that the axial crack indications in penetrations 1, 2,

and 3 were through-wall based on analysis of the UT data and reported this condition to

the NRC on February 27, 2002 (in notification number 38732). The specific number

and orientation of cracks in each nozzle was documented in NRC inspection report

(IR) 50-346/02-03.

The licensee initiated repairs on the five cracked penetrations discussed above. This

repair process included roll expanding the penetration nozzle, grinding out the affected

portion of the penetration nozzle to a location above the J-weld and addition of a temper

bead weld metal buildup beginning at machined nozzle end-prep. The next step

included finish machining on the inside bore, followed by UT and PT examinations of the

weld. However, on March 5, 2002, during the machining process on nozzle number 3,

the licensee identified movement of the nozzle. The licensee subsequently cleaned

boric acid deposits from the head during investigation of this phenomena and

discovered a large cavity in the vessel head. An NRC augmented inspection team

performed an inspection of this issue and the NRC teams conclusions were

documented in NRC IR 50-346/02-03. The NRC findings associated with this issue

were documented in NRC IR 50-346/02-08. The licensee subsequently decided to

replace the vessel head as discussed in Section 4OA3 of this report.

e. What, if any, significant items that could impede effective examinations and/or

As-Low-As-Reasonably-Achievable issues were encountered?

The inspectors did not identify any significant impediments to the UT conducted from

below the head as discussed above. The licensees initial visual examination of the

head penetrations was not effective as discussed above because of boric acid and

corrosion products on the surface of the head. The licensee stated that the actual

dose received for this job was about 3.5 Roentgen Equivalent Man (REM). This dose

was below the licensees projected dose for the ultrasonic examinations on the head of

5.5 REM.

c. Findings

No findings of significance were identified. However, the NRC performed

additional followup inspection in this area and the results are discussed in NRC

IR 50-346/02-03 and NRC IR 50-346/02-08.

19

.2 Reactor Pressure Vessel Head and Vessel Head Penetration Nozzles (TI 2515/150)

The objective of TI-150 was to review licensees activities in response to NRC Bulletin

2002-02, Reactor Pressure Vessel Head and Vessel Head Penetration

Nozzle Inspection Programs. This TI implements the NRC inspections needed to

confirm that the licensee meets vessel head examination commitments associated with

Bulletin 2002-02, including procedures, equipment, and personnel demonstrated to be

effective in the detection and sizing of PWSCC in vessel head penetration nozzles. The

Davis-Besse replacement RPVCH had never been operated, and thus, had not been

exposed to the hot plant operating environmental conditions necessary to initiate

PWSCC. Therefore, PWSCC does not currently exist in the replacement vessel head

penetration nozzles and the NRC will complete TI-150 during the next scheduled

Davis-Besse RFO.

4OA6 Meetings

.1 Exit Meeting

The inspectors presented the inspection results to Mr. L. Meyers and other members of

licensee management at the conclusion of the inspection on October 24, 2002. The

inspectors asked the licensee whether any materials discussed as potential report

material should be considered proprietary. No proprietary information was identified.

20

KEY POINTS OF CONTACT

Licensee

L. Myers, Vice President - Nuclear

L. Pearce, Vice President - Oversight

R. Fast, Plant Manager

D. Baker, Project Manager

J. Reddington, Supervisor Quality Assurance

S. Loehlein, Manager Quality Assurance

A. Alford, Regulatory Affairs

S. Saunders, Senior Engineer

T. Swim, Engineer

J. Cunnings, Supervisor Mechanical Engineering

T. Chambers, Containment Health Manager

R. Mende, Containment Health Engineer

Vendor - Bechtel

S. Fox, Senior Project Manager

Vendor - Framatome ANP

E. Mayhew, Vice President Quality, US Region

V. Montalbano, Manager, Nuclear Services Quality

T. Werner, Lead Quality Specialist

M. Gerlich, QA Engineer

M. Morgan, Manager, Quality Assurance Audits and Programs

S. Dasgupta, QA Consultant

H. Behnke, Technical Consultant, Component Engineering

F. Snow, Project Engineer

H. Harrison III, Engineer

Nuclear Regulatory Commission

C. Thomas, Senior Resident Inspector

D. Simpkins, Resident Inspector

A. Mendiola, Project Manager, NRR

J. Ma, Division of Engineering, Civil Engineering and Mechanics Section, NRR

A. Ashar, Division of Engineering, Component and Containment Reliability Section, NRR

21

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None

Closed

None

Discussed

None

LIST OF ACRONYMS USED

ASME American Society of Mechanical Engineers

CAC Containment Air Cooler

CFR Code of Federal Regulations

CR Condition Report

CRDM Control Rod Drive Mechanism

DBHRP Davis-Besse Head Resolution Plan

DHR Decay Heat Removal

EWR Engineering Work Request

MT Magnetic Particle Examination

NCV Non-Cited Violation

NDE Nondestructive Examination

NRC Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

PT Dye Penetrant Examination

PWSCC Primary Water Stress Corrosion Cracking

RE Radiation Element

QA Quality Assurance

QC Quality Control

REM Roentgen Equivalent Man

RCS Reactor Coolant System

RPVCH Reactor Pressure Vessel Closure Head

RT Radiographic Examination

SDP Significance Determination Process

SMAW Shield Metal Arc Welding

TI Temporary Instruction

UT Ultrasonic Examination

UFSAR Updated Final Safety Analysis Report

WPIR Work Plan and Inspection Record

22

LIST OF DOCUMENTS REVIEWED

ASME Code Data Packages

F-ANP QA Data Midland Replacement Reactor Pressure Vessel Closure Revision 0

Package No. 23- Head ASME B&PV Code Section III Data Package

5018698-00

F-ANP QA Data Midland Replacement Reactor Pressure Vessel Closure Revision 0

Package No. 23- Head ASME B&PV Code Section XI Data Package

5019258-00

Calculation

12501-C-003 Evaluation of Containment Vessel for Construction Revision 1

Opening

Condition Reports

6015661 Supplemental examinations needed to aid in the evaluation of

radiography film.

02-00891 Reactor Vessel Head

02-00932 Reactor Vessel Head

02-01053 Number 3 nozzle machining tool moved approximately 15 degrees

02-05108 Bechtel Quality Assurance Oversight Concerns

02-05548 Issues have been raised regarding the effectiveness of Bechtel Quality

Assurance program that reveal a negative trend

02-05486 The testing and qualification of cadwelders for the restoration of

containment was started without written direction of the vendor.

02-07472 Containment Shield Building-Annulus Side

02-07080 Containment Shield Building

Drawings

12501-C-102 Reactor Pressure Vessel Head Replacement Revision 2

Project Containment Shield Building Construction

Opening Details

12501-C-103 Reactor Pressure Vessel Head Replacement Revision 4

Project Containment Shield Building Barrier and

Lug Details

12501-C-201 Reactor Pressure Vessel Head Replacement Revision 4

Project Temporary Opening in Containment

Vessel Wall

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F-ANP Midland 2 RVH CRDM nozzle modification Revision 1

5018608B-01

F-ANP Revision 1

6015305B-1 CRDM nozzle flange modification go gauge

F-ANP Revision 1

6015470B-1 CRDM nozzle flange modification no go gauge

F-ANP Modified CRDM flange nut ring installation Revision 2

5018780B-2

F-ANP Midland head x-axis keyway block modification Revision 1

5018900B-1

F-ANP Midland -2 service structure support skirt Revision 0

5018532E-0 openings

Engineering Work Request

02-0146 Provide Opening in the Containment Structure to Supplement 2

Remove/Replace the Reactor Vessel Head

Field Change Requests

FCR-C-003 EWR 02-0146 August 8, 2002

FCR-C-008 EWR 02-0146 September 6, 2002

FCR-C-016 EWR 02-0146 September 22, 2002

FCR-C-017 EWR 02-0146 September 23, 2002

FCR-C-019 EWR 02-0146 September 07, 2002

FCR-C-022 EWR 02-0146 September 14, 2002

FCR-C-027 EWR 02-0146 September 23, 2002

FCR-C-028 EWR 02-0146 September 24, 2002

Inspection Plans

Davis-Besse Reactor Head Resolution Plan Revision 0

Davis-Besse Reactor Head Resolution Plan Revision 1

Nonconformance Reports

NCR-009 CP-11 September 4, 2002

NCR-017 Specification 12501-C-321 September 24, 2002

24

ML023370100

Text

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