ML023370100
| 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) | |
See also: IR 05000346/2002007
Text
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
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JCreed (4OA3.3)
DATE
11/13/02
11/14/02
11/25/02
11/13/02
OFFICE
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
SPS1
RidsNrrDipmIipb
GEG
CST1
C. Ariano (hard copy)
DRPIII
DRSIII
PLB1
JRK1
DB0350
U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Docket No:
50-346
License No:
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
2
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.
3
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.
4
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:
procurement and certification;
RPVCH modifications;
temporary fuel removal;
access through the Davis-Besse containment opening;
installation of new RPVCH;
restoration, inspection and testing of the RPVCH and containment;
storage and disposition of original RPVCH; and
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.
5
.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
6
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
7
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:
visual examination of the entire RPVCH to identify signs of degradation or
evidence of welding while the head was in storage at Midland;
dye penetrant examination (PT) records of all 69 CRDM J-groove welds;
ultrasonic examination (UT) Examination of all 69 CRDM nozzles;
PT of the head cladding at 6 sample areas;
PT & RT of all 69 CRDM flange-to-nozzle welds;
RT of closure head-to-flange weld;
magnetic particle examination (MT) of the head lifting lug attachments;
UT and MT of the closure head-to-flange weld; and
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
8
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.
9
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
10
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.
11
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.
12
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
13
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.
14
.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.
15
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:
observation of the licensees UT and visual examination of the reactor vessel
head penetrations,
interviews with the licensees contract NDE personnel,
review of NDE procedures, and
review of the head inspection NDE reports.
Additionally, the inspectors observed the repair activities implemented on the cracked
vessel head penetration nozzles.
16
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
17
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.
18
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?
19
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.
20
.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.
21
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
22
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None
Closed
None
Discussed
None
LIST OF ACRONYMS USED
American Society of Mechanical Engineers
Containment Air Cooler
CFR
Code of Federal Regulations
CR
Condition Report
Control Rod Drive Mechanism
DBHRP
Davis-Besse Head Resolution Plan
Engineering Work Request
Magnetic Particle Examination
Non-Cited Violation
NRC
Nuclear Regulatory Commission
Office of Nuclear Reactor Regulation
Dye Penetrant Examination
Primary Water Stress Corrosion Cracking
RE
Radiation Element
Quality Assurance
Quality Control
Roentgen Equivalent Man
RPVCH
Reactor Pressure Vessel Closure Head
Radiographic Examination
Significance Determination Process
Shield Metal Arc Welding
TI
Temporary Instruction
Ultrasonic Examination
Updated Final Safety Analysis Report
WPIR
Work Plan and Inspection Record
23
LIST OF DOCUMENTS REVIEWED
ASME Code Data Packages
F-ANP QA Data
Package No. 23-
5018698-00
Midland Replacement Reactor Pressure Vessel Closure
Head ASME B&PV Code Section III Data Package
Revision 0
F-ANP QA Data
Package No. 23-
5019258-00
Midland Replacement Reactor Pressure Vessel Closure
Head ASME B&PV Code Section XI Data Package
Revision 0
Calculation
12501-C-003
Evaluation of Containment Vessel for Construction
Opening
Revision 1
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
Project Containment Shield Building Construction
Opening Details
Revision 2
12501-C-103
Reactor Pressure Vessel Head Replacement
Project Containment Shield Building Barrier and
Lug Details
Revision 4
12501-C-201
Reactor Pressure Vessel Head Replacement
Project Temporary Opening in Containment
Vessel Wall
Revision 4
24
F-ANP
5018608B-01
Midland 2 RVH CRDM nozzle modification
Revision 1
F-ANP
6015305B-1
CRDM nozzle flange modification go gauge
Revision 1
F-ANP
6015470B-1
CRDM nozzle flange modification no go gauge
Revision 1
F-ANP
5018780B-2
Modified CRDM flange nut ring installation
Revision 2
F-ANP
5018900B-1
Midland head x-axis keyway block modification
Revision 1
F-ANP
5018532E-0
Midland -2 service structure support skirt
openings
Revision 0
Engineering Work Request
02-0146
Provide Opening in the Containment Structure to
Remove/Replace the Reactor Vessel Head
Supplement 2
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
September 4, 2002
NCR-017
Specification 12501-C-321
September 24, 2002
25
Other Documents
21896-002
Mockup of Containment Cutting for Welding
Revision 0
Work Order
02-003545-010
A Temporary Access Opening in the
Containment Shield Building is Required to
Support Replacement of the RPV Head
August 10, 2002
WIPR C-CRA-02
Removal of Concrete in the Temporary
Construction Opening Through the
Containment Shield Wall
Revision 0
Work Order
02-003545-013
A Temporary Access Opening in the
Containment Shield Building is Required to
Support Replacement of the RPV Head
August 21, 2002
WIPR C-CLP-02
Cut Vessel Plate
Revision 0
50-5018614-00
Process Traveler Modification and
Preparation of Midland-2 RVCH
June 4, 2002
54-PT-6-07
Visible Solvent Removable Liquid Penetrant
Examination Procedure
August 3, 2000
54-1027734-05
Radiographic Testing
June 3, 2002
Midland Reactor Pressure Vessel Control
Rod Drive Mechanism Dissimilar Weld
Radiography Supplemental Examination
Plan
June 12, 2002
02-0146-00
Provide access opening in the containment
structure to remove/replace the reactor
vessel head
August 2, 2002
Letter from J.
Mallay (Framatome
ANP INC.) to NRC
Closure of Interim Report Concerning a
Potential Safety Concern on the Condition
of the Midland Reactor Vessel Head
Proposed for Use at an Operating Plant
September 9, 2002
Nonconformance
report 009
Cadweld Qualification Testing
September 4, 2002
WP&IR C-SWR-01
Shield Building Restoration
Revision 0
Bechtel
Specification
12501-C-321
Technical Specification for Purchase of
Safety Related Ready-Mix Concrete
Revision 2
Other Documents
26
Davis Besse Nuclear Power Station
Replacement Project Concrete Placement
and Test Plan
Revision 0
Certified Material
Test Report
Atom Arc 7018, 1/8," 9950 lbs, lot number
4G113A07
August 21, 2001
PSI report 150-
20129-34
Report of Tests on Cylinder Compressive
Strengths
October 1, 2002
David Besse 13 RFO CRDM Nozzle
Examination Report
March 11, 2002
Letter Serial
Number 1-128
FirstEnergy letter to Mr. James E. Dyer,
Administrator - Replacement of the Reactor
Pressure Vessel Head at the Davis Besse
Nuclear Power Station
August 9, 2002
F-ANP Document
03-5018636-00
CRDM housing flange modification drill
fixture operating instruction, Midland 2
Revision 0
F-ANP Technical
Document 33-
5019877-00
ASME Stress Report, Davis Besse Original
Closure Head Replacement Design Report,
Davis Besse Unit 1
August 15, 2002
F-ANP Document
51-5018522-10
Midland Closure Head Dedication Plan
July 17, 2002,
F-ANP Document
51-5019457-00
Davis Besse RV Closure Head
Replacement Reconciliation
August 16, 2002
F-ANP Document
51-5019457-01
Davis Besse RV Closure Head
Replacement Reconciliation
August 22, 2002
02-046u
UFSAR Change Notice for EWR 02-0146
August 24, 2002
Bechtel
Specification No.
12501-C-310
Technical Specification for Purchase of
Safety Related Ready-Mix Concrete
Revision 2
Bechtel
Specification No.
12501-C-321
Technical Specification for Installation of
Cadweld Rebar Splices
Revision 2
Bechtel
Specification No.
12501-C-322
Technical Specification for Purchase of
Safety Related Ready-Mix Concrete
Revision 2
Other Documents
27
Bechtel
Specification No.
12501-C-101
Technical Specification for Materials
Testing Services
Revision 3
Quality Assurance Program Manual
Revision 3
Williams Concrete
Inc.
Letter
Mix Design Submittal Information
July 29, 2002
Bechtel
Document
Concrete Placement and Test Plan
September 24, 2002
Bechtel WP&IR
Shield Building Restoration
September 14, 2002
12501-SC-025-
PQR-721A-02
Procedure Qualification Record
Revision 1
12501-SC-025-
PQR-627A-02
Procedure Qualification Record
Revision 3
Specification 7749-
C-37
Containment Vessel Technical Specification
Revision 19
Erico Concrete
Reinforcement
Products
Memorandum
CADWELD rebar splice configuration
September 5, 2002
Procedures
96-RT-005
General Radiographic Procedure
Per ASME Section V Article 2
Revision 5
Concrete Operations
Revision 0
Cadweld Rebar Splices
Revision 0
Testing of Cadweld Rebar
Splices
Revision 0
54-ISI-367-03
Procedure for the Visual
Examination for Leakage of the
Reactor Vessel Head
Revision 3
54-ISI-100-06
Remote Ultrasonic Examination
of Control Rod Drive Mechanism
(CRDM) Nozzles
Revision 6
1 MN-GT-GTAW/SMAW-1
Welding Procedure Specification
Revision 14
96-RT-005
General Radiographic Procedure
Per ASME Section V Article 2
Revision 5
Concrete Operations
Revision 0
Cadweld Rebar Splices
Revision 0
Testing of Cadweld Rebar
Splices
Revision 0
54-ISI-367-03
Procedure for the Visual
Examination for Leakage of the
Reactor Vessel Head
Revision 3
54-ISI-100-06
Remote Ultrasonic Examination
of Control Rod Drive Mechanism
(CRDM) Nozzles
Revision 6
28
Radiographic Records
CS-01A
Containment Vessel Weld
September 30, 2002
CS-01B
Containment Vessel Weld
September 30, 2002
CS-01C
Containment Vessel Weld
September 30, 2002
CS-01D
Containment Vessel Weld
September 30, 2002
W-7
Reactor Vessel Closure Head to
Head)
W-9
Control Rod Drive Mechanism
NiCrFe Body-to -Stainless Steel
Weld (Replacement Head)
Surface Examination Records on Replacement Head Welds
W-13
Dye Penetrant Record of
J-groove Buttering Weld
W-15
Magnetic Particle Record of
Service Structure Segments to
Closure Head Weld
WH-17
Magnetic Particle Record of Lift
Lug to Closure Head Weld
WH-25
Dye Penetrant Record of CRDM
Nozzle J-grove weld
WH-27
Dye Penetrant Record of Arrow
to Closure Head
96-RT-005
General Radiographic Procedure
Per ASME Section V Article 2
Revision 5
Concrete Operations
Revision 0
Cadweld Rebar Splices
Revision 0
Testing of Cadweld Rebar
Splices
Revision 0
54-ISI-367-03
Procedure for the Visual
Examination for Leakage of the
Reactor Vessel Head
Revision 3
54-ISI-100-06
Remote Ultrasonic Examination
of Control Rod Drive Mechanism
(CRDM) Nozzles
Revision 6
29
Welder Qualifications
PCI Welder M990
Certified on August 19,
2002
PCI Welder M991
Certified on August 19,
2002
PCI Welder M989
Certified on August 16,
2002
PCI Welder M987
Certified on August 13,
2002
PCI Welder M985
Certified on August 9,
2002
PCI Welder M983
Certified on August 9,
2002
Bechtel Welder IW-1
Certified on August 29,
2002
Bechtel Welder IW-3
Certified on August 29,
2002
Bechtel Welder IW-7
Certified on August 27,
2002
Bechtel Welder IW-8
Certified on August 27,
2002
Bechtel Splicer 7423
Certified on
September 15, 2002
96-RT-005
General Radiographic Procedure
Per ASME Section V Article 2
Revision 5
Concrete Operations
Revision 0
Cadweld Rebar Splices
Revision 0
Testing of Cadweld Rebar
Splices
Revision 0
54-ISI-367-03
Procedure for the Visual
Examination for Leakage of the
Reactor Vessel Head
Revision 3
54-ISI-100-06
Remote Ultrasonic Examination
of Control Rod Drive Mechanism
(CRDM) Nozzles
Revision 6
30
Bechtel Splicer 8578
Certified on
September 15, 2002
Bechtel Splicer 6792
Certified on
September 15, 2002
Bechtel Splicer 8715
Certified on
September 15, 2002
Bechtel Splicer 5251
Certified on
September 15, 2002
Bechtel Splicer 5243
Certified on
September 15, 2002
Bechtel Splicer 6906
Certified on
September 15, 2002
31
DOCUMENTS REQUESTED
Information to provide to M. Holmberg for on-site inspection beginning on August 13,
2002.
A.
For the containment vessel access cut in support of head replacement provide a copy
of:
1)
Detailed schedule for containment access cut and restoration including
description of related activities such as welding, nondestructive testing, welder
qualification and/or mockup training.
2)
ASME Code repair/replacement plan identifying Construction Code and Code
Cases used for the containment vessel access cut. Specifically, identify the
applicable Code Section(s) and Edition applicable to the containment closure
weldment and the acceptance criteria for the applicable nondestructive testing.
3)
Fabrication and weld construction drawings for the containment vessel.
Drawings associated with the containment access cut and restoration.
4)
List identifying the weld process, procedure and applicable revision for each new
weld on the containment vessel.
5)
List of welders or weld operators that are to be used to perform welding on the
containment vessel.
6)
List of design change packages and safety evaluations associated with the
containment vessel access cut.
7)
List of condition reports (beginning in January of 2002) and non-conformance
reports associated with containment vessel, with a brief description of the
condition.
8)
The containment vessel design specification and containment coating design
specification.
9)
Containment modification package (EWR 02-0146) including 50.59 evaluation
and supporting containment vessel analysis for the temporary containment
opening.
10)
The welding procedures and supporting qualification documents (PQRs and test
reports) used to close the temporary vessel access cut.
11)
List of procedures/work orders (including description) that control the work
activities and non-destructive testing.
12)
Procedure(s) that identify the quality control hold point and witness checks for
containment access work (installation and testing) as specified by the on-site
quality control organization.
Enclosure (1)
32
INFORMATION REQUESTED ON 12/18/2001 BY E-MAIL (To R. Cook )
A.
Please provide the following information to Melvin Holmberg at the Region III NRC office
located at 801 Warrenville Rd, Lisle IL 60532, no later than January 7, 2002, to support the
NRC Inservice Inspection (IP 71111.08 and TI-145) scheduled at the Davis Besse plant for
February 20, 2002 - March 8, 2002.
1)
A detailed schedule of nondestructive examinations planned for Class 1 & 2 systems and
containment, performed as part of your ASME Code ISI Program during the scheduled
inspection weeks. Provide a detailed schedule of vessel head examinations which fulfill
NRC commitments made in response to NRC Bulletin 2001-01. Provide a detailed schedule
of steam generator (SG) tube inspection and repair activities for the upcoming outage,
2)
A copy of the procedures used to perform the examinations identified in A.1. For ultrasonic
examination procedures qualified in accordance with Appendix VIII, of Section XI of the
ASME Code, provide documentation supporting the procedure qualification (e.g. the EPRI
performance demonstration qualification summary sheets). Also, include documentation of
the specific equipment to be used (e.g. ultrasonic unit, cables, and transducers including
serial numbers).
3)
A copy of any ASME Section XI, Code Relief Requests applicable to the examinations
identified in A(1).
4)
A list identifying nondestructive examination reports (ultrasonic, radiography, magnetic
particle, dye penetrant, visual (VT-1, VT-2, VT-3)) which have identified relevant indications
on Code Class 1 & 2 systems in the past two refueling outages (both Units).
5)
List of welds in Code Class 1, and 2 systems which have been completed since the
beginning of the last refueling outage (both Units and identify system, weld number and
reference applicable documentation).
6)
For reactor vessel weld examinations required by the ASME Code, that are scheduled during
the inspection, provide a detailed description of the welds to be examined, extent of the
planned examination and a copy of your responses to the NRC, associated with Generic Letter 83-15.
7)
Provide a list with description of ISI and steam generator related issues entered into your
corrective action system beginning with the date of the last refueling outage (both Units).
8)
Copy of any part 21 reports submitted beginning with the date of the last refueling outage.
9)
Copy of SG history documentation given to vendors performing eddy current (ET) testing of
the SGs during the upcoming outage.
10)
Copy of procedure containing screening criteria used for selecting tubes for in-situ pressure
testing and the procedure to be used for in-situ pressure testing.
11)
Copy of previous outage SG tube operational assessment completed following ET of the
SGs.
12)
Copy of the document defining the planned ET scope for the SGs and the scope expansion
criteria which will be used.
13)
Copy of the document describing the ET probe types, and ET acquisition equipment to be
used, including which areas of the SG (e.g. top of tube sheet, U-bends) each probe will be
used in. Also, provide your response letter(s) to generic letters 95-03, 95-05, 97-05, and 97-
06.
14)
Copy of document describing actions to be taken if a new SG tube degradation mechanism
is identified.
Enclosure (1)
33
15)
Identify the types of SG tube repair processes which will be implemented for defective SG
tubes. Provide the flaw depth sizing criteria to be applied for ET indications identified in the
SG tubes.
16)
If tube leakage was identified during the previous operating cycle, provide documentation
identifying which SG was leaking and planned corrective actions.
17)
Provide a copy of the EPRI Technique Specification Sheets which support qualification of
the ET probes to be used during the upcoming SG tube inspections.
18)
Provide a copy of the guidance to be followed if a loose part or foreign material is identified
in the SGs.
19)
Detailed scope of the planned nondestructive examinations (NDE) of the vessel head which
identifies the types of NDE methods to be used on each specific part of the vessel head to
fulfill NRC commitments made in response to NRC Bulletin 2001-01. Also include
examination scope expansion criteria and planned expansion sample sizes if relevant
indications are identified.
20)
Copy of NDE procedures to be used for performing vessel head inspections that fulfill NRC
commitments in response to NRC Bulletin 2001-01.
21)
Identify what standards or requirements will be used to evaluate indications identified during
NDE examinations of the vessel head.
B.
Information to be provided on-site to the inspector at the entrance meeting:
1)
For welds selected by the inspector from A.5 above, provide copies of the following
documents:
a)
Document of the weld number and location (e.g. system, train, branch).
b)
Document with a detail of the weld construction.
c)
Applicable Code Edition and Addenda for weldment.
d)
Applicable Code Edition and Addenda for welding procedures.
e)
Applicable weld procedures (WPS) used to fabricate the welds.
f)
Copies of procedure qualification records (PQRs) supporting the WPS on selected
g)
Copies of mechanical test reports identified in the PQRs above.
h)
Copies of the nonconformance reports for the selected welds.
i)
Radiographs of the selected welds and access to equipment to allow viewing
radiographs.
j)
Copies of the preservice examination records for the selected welds.
2)
For the replacement activities selected by the inspector provide a copy of the records of the
repair or replacement required by the ASME Code Section XI Articles IWA -4000 or IWA
7000.
3)
Provide a list of NDE personnel performing inspections of the vessel head and the
qualification records for these personnel.
4)
Copies of commitments made to the NRC for performing vessel head examinations.
5)
Copy of the most recent quality assurance department audit, which included the ISI program
and activities. Copies of documents resolving findings in this audit.
6)
Updated schedules for item A.1.