ML20205A748
| ML20205A748 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/15/1986 |
| From: | Blake J, Hallstrom G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20205A715 | List: |
| References | |
| 50-424-86-39, 50-425-86-19, NUDOCS 8608110438 | |
| Download: ML20205A748 (32) | |
See also: IR 05000424/1986039
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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REGION 11
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101 MARIETTA STREET, N.W.
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ATLANTA, GEORGI A 30323
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Report Nos.: 50-424/86-39 and 50-425/86-19
Licensee: Georgia Power Company
P. O. Box 4545
Atlanta, GA 30302
Docket Nos.: 50-424 and 50-425
License Nos.: CPPR-108 and CPPR-109
Facility Name: Vogtle 1 and 2
Inspection Condu ted: May 5-9 and June 23, 1986
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Inspector:
woe,
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G /A.
trorii,
/Datd Signed
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Approved
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,
J./J. plakr, Section Chief
'Date' Signed
Eggineering Branch
ivision of Reactor Safety
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SUMMARY
Scope:
This special announced inspection was conducted on site in the areas of
licensee action on previous enforcement matters (Units 1 and 2), housekeeping
Units 1 and 2), material control (Units 1 and 2), and safety-related piping
Unit 2)
Results:
Two violations were identified - Failure to provide accurate informa-
tion on elimination of intermediate pipe breaks and associated whip restraints;
and Failure to follow procedures for control of welding consumables.
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REPORT DETAILS
1.
Persons Contacted
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Licensee Employees
- R. H. Pinson, Vice President, Construction
- D. O. Foster, Vice President and Project Support Manager
- P. D. Rice, Vice President, Engineering
- C. W. Hayes, Project Quality Assurance (QA) Manager
- E. D. Groover, QA Site Manager, Construction
- C. E. Belflower, QA Site Manager, Operations
- G. A. McCarley, Project Compliance Coordinator
- B. C. Harbin, Manager of Quality Control (QC), Construction
- N. Lankford, QC Support Supervisor
J. E. Seagraves, RR Discipline Manager (Construction)
- W. C. Gabbard, Regulatory Specialist
- H. P. Walker, Manager, Unit Operations
- G. Bockhold, General Manager, Vogtle Nuclear Operations
Other licensee employees contacted included construction craftsmen,
engineers, technicians, operators, mechanics, and office personnel.
Other Organizations
- W. C. Ramsey, Southern Company Services (SCS), RR Project Manager
- 0. Batum, SCS, Deputy to Vice President of Project Engineering
- D. W. Strohman, Bechtel Power Corporation (BPC), Project QA Engineer
D. Niehoff, BPC, Deputy Engineering Group Supervisor
- M. R. Thaker, BPC, RR Civil-Structural Design Team Leader
- R. C. Somerfield, BPC, RR Construction Team Leader
R. Malin, BPC, Civil / Structural Engineering Group Supervisor
D. Sewell, BPC, Plant Design / Pipe Support and Stress
Engineering Group Supervisor
A. Attor, BPC, Senior Pipe Support and Stress Engineer
NRC Resident Inspectors
- H. Livermore, Senior Resident Inspector (Construction)
- J. Rogge, Senior Resident Inspector (Operations)
- Attended exit interview
2.
Exit Interview
T;ie inspection scope and findings were summarized on May 9,1986, with
those persons indicated in paragraph 1 above.
The inspector described the
areas inspected. No dissenting comments were received from the licensee.
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(0 pen) Violation 424/86-39-01, 425/86-19-01, Failure to provide accurate
information for elimination of intermediate pipe breaks and associated whip
restraints - paragraph 3.a.(4).
(0 pen) Inspector Followup Item 424/86-39-02, 425/86-19-02, Field confirma-
tion of distance between welded attachments and the location of intermediate
pipe breaks - paragraph 3.a.(5).
(0 pen) Violation 424/86-39-03, 425/86-19-03, Failure to follow procedures
for control of welding consumables - paragraph 6.b.(4)
The licensee did identify as proprietary some of the materials provided to
and reviewed by the inspector during this inspection; however, details from
those materials are not included in this report.
3.
Licensee Action on Previous Enforcement Matters
References:
(a) Letter, dated May 9,1986, from R. E. Conway (Georgia
Power Company (GPC)) to B. J. Youngblood (NRC Office of
Nuclear Reactor Regulation (NRR)) regarding Vogtle 1 and
2 Intermediate Pipe Breaks.
(b) Memorandum, dated March 18, 1986, from C. E. Rossi
(NRC-NRR) to A. Gibson (NRC Region II) providing an
interpretation of acceptance criteria for elimination of
intermediate pipe breaks.
(c) Letter, dated February 7,1986, from D. O. Foster (GPC)
to J. N. Grace (NRC Region II) responding to unresolved
and inspector followup items described in NRC Inspection
Report 424/85-35.
(d) Letter, dated November 11, 1983, from D. O. Foster (GPC)
to H. R. Denton (NRC) requesting alternate intermediate
pipe break criteria which would allow climination of 182
previously postulated intermediate pipe breaks and 110
associated pipe whip restraints per unit.
(e) Letter, dated April 26, 1984, from D. O. Foster (GPC) to
H. R. Denton (NRC) providing technical information to
justify a request for approval of alternate pipe break
criteria.
(f) Letter, dated June 26, 1984, from T. M. Novak (NRC
Division of Licensing) to D. O. Foster (GPC) providing
an evaluation and acceptance of alternate pipe break
criteria.
(g) Letter dated April 30, 1986, from F. B. Marsh (BPC)
to 0. Batum (SCS) providing details on BPC review of
conformance to alternate intermediate pipe break
criteria.
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a.
(Closed) Unresolved item (424/85-35-03):
Design Control of Interme-
diate Pipe Breaks
(1) Background Information
This item was originated to identify NRC Region II inspectors'
concern that the licensee had not implemented a particular design
commitment in the postulation and mitigation of intermediate pipe
breaks on high-energy piping.
The commitment was part of the
licensee's design bases (Reference (e)) in obtaining NRC accept-
ance (Reference (f)) of deviation of the requirements to postulate
arbitrary intermediate pipe breaks from those included within NRC
Branch Technical Position MEB 3-1 " Postulated Breakage and Leakage
Locations in Fluid System Piping Outside Containment."
The
licensee's proposal of alternate intermediate break criteria was
made to avoid the installation of approximately 110 pipe whip
restraints per unit (Reference (d)) which MEB 3-1 would have
required for the mitigation of design postulated pipe breaks. The
alternate criteria allows elimination of arbitrary intermediate
pipe breaks together with associated whip restraints provided:
Possibility of stress corrosion cracking has been minimized;
Thermal and vibration induced piping fatigue has been
minimized;
Steam / water effects have been minimized;
Local bending stresses from welded attachments have been
minimized.
(2) 5D Criteria
The design commitment of concern was intended to minimize stress
from welded attachments.
As included in Reference (e), this
commitment was to assure that no welded attachments / supports lay
within five pipe diameters (50 criteria) from any postulated pipe
locations that would be eliminated in accordance with the alter-
nate break criteria.
The SD commitment was included as a partial
alternative to the installation of pipe whip restraints. The NRC
had accepted SD criteria at other plants as a reasonable require-
ment to assure that welded supports would not be located close
enough to postulated break points to contribute to the severity of
damage that could occur during unanticipated transients when
stresses in the involved piping can greatly exceed ASME design
,
Code allowables.
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(3) History
Region II inspectors first identified their concern that the
licensee had not implemented the SD criteria during an inspection
of the licensce's Readiness Review work in August,1985.
The
licensee was questioned regarding the absence of any implementing
procedures during the site inspection and again in a telephone
call on September 18, 1985.
The licensee's initial informal
response was that the commitment was only required to be met at
the time of the April 26, 1984, submittal (Reference (e)), and
that it was not intended to be met for later design iterations;
i.e., new break locations due to later changes in the piping
installation.
The inspectors questioned this interpretation as
the written licensee proposal and the NRC acceptance of the
proposal did not support this view.
The inspectors formally
documented their concern to the licensee in NRC Report No.
50-424/85-35 dated December 18, 1985.
The licensee formally
responded by letter to Region II dated February 7,1986 (Refer-
ence (c)).
The licensee contended that the SD commitment had not
been a commitment but that NRC acceptance of minimized stress from
welded attachments for the alternate intermediate break criteria
was based instead on their compliance with the ASME Section III,
Subsections NC/ND-3645, generalized requirements that the design
appropriately consider the effects of welded attachments.
The
inspectors noted that NC/ND-3645 was not adequate to satisfy a
major objective of the SD criteria; i.e., minimizing the contribu-
tion of supports to the severity of damage that could occur from
unanticipated transients with stresses beyond Code allowables.
As part of a Region II inspection on February 24-28, 1986
(Inspection Report No. 424/86-11), the licensee was again
requested to provide technical support for their failure to
implement the 50 criteria as a continuing commitment.
The
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licensee again responded that the SD criteria was not considered a
commitment based on their previous discussions which had been held
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with reviewers from the NRC Office of Nuclear Reactor Regulation
(NRR) while obtaining NRC approval to deviate from MEB 3-1.
As a
followup to inspection 86-11, a formal response from NRR was
requested regarding the licensee's statements that the SD criteria
was not a commitment.
Formal response from NRR (Reference (b))
clarified that, contrary to the licensee's repeated explanations,
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the SD criteria had been considered a commitment and that it had
been material to NRC acceptance of the licensee's proposal for
alternate intermediate break criteria which deviated from ME8 3-1.
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(4) May 4-9, 1986, Region II Inspection
In order to evaluate the extent of the licensee's noncompliance
during the current inspection, the inspector informed cognizant
licensee personnel on April 29, 1986, that this was an announced
inspection and requested that data be gathered for use on inspect-
ing postulated break locations to determine whether the licensee
had complied with the SD criteria when it was originally stated in
April, 1984, and to examine their current compliance.
In response
to this request the licensee performed additional review and
informed the inspector at the beginning of this inspection that
the SD criteria was not currently met and had not been met at the
time it was originally stated.
The licensee provided summary
details (Reference (a)) of the six instances identified to date of
welded attachments located within five pipe diameters from postu-
lated pipe break locations which had been eliminated in accordance
with alternate intermediate break criteria. Four of the intermedi-
ate breaks eliminated had been postulated since April 26, 1984.
Two of the breaks eliminated had been postulated before April 26,
1984.
These were main steam line breaks P-1054-C and P-1055-C.
Consequently, associated pipe whip restraints PBR-217 and PBR-218
were climinated without meeting the SD criteria.
Cognizant
licensee personnel indicated they could not determine the reason
for the inaccurate statements in their April 26, 1984, submittal
regarding pipe breaks P-1054-C and P-1055-C.
The inspector examined internal licensee correspondence dated
April 30, 1986 (Reference (g)) which established that the licensee
was aware as of July, 1984, of their noncompliance with the April,
1984 commitment. Cognizant licensee personnel offered no explana-
tion as to why the NRC had not been informed of this noncompliance
as of the date of this inspection.
The inspector informed the licensee 'that their inaccurate state-
ments regarding pipe breaks P-1054-C and P-1055-C in their
April 26,1984, submittal appeared to be a material false state-
ment as defined in 10 CFR 2 and as such was a violation of NRC
requirements.
Therefore, unresolved item 424/85-35-03 would
be closed and this matter will be identified as violation
424/86-39-01, 425/86-19-01, Failure to provide accurate informa-
tion for elimination of intermediate pipe breaks and associated
whip restraints.
Also, that this violation would be reviewed for
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escalated enforcement action.
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(5) Details of Region II Inspection of SD Noncompliance
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In examining the licensee's findings relative to their noncom-
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pliance with the SD criteria, the inspector completed field
examinations and examined 'oackground documentation to enable
independent verification of the summary data included in
Reference (a), attachment 1.
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The inspector verified that break locations postulated after
April 26, 1984, with welded attachments within five pipe diameters
were as follows:
Stress
Break
Support
Item
Isometric
Calc
Data
Within
X4CP
Point
5 pipe diameter
1
IK3-1206-066-01
7001A
26A
V1-1208-066-H006
2
7063A
800
V1-1314-084-H013
3
7092
88
V1-1208-005-H006
4
IK4-1208-005-02
7092
156
V1-1208-005-H001
During the above examination the inspector noted that support
location tolerances for the small diameter pipe involved amounted
to as much as three pipe diameters. Therefore, the application of
a five pipe diameter minimum distance criteria could not be
assured without field investigation. Cognizant licensee personnel
informed the inspector that the licensee's reviews involved had
been based on design stress isometrics and field verification had
been conducted only for the six instances identified on Attach-
ment 1.
The inspector stressed the need for measurements in the
field to enable proper application of the SD criteria and was
informed that modifications to the confirmation program for IEB 79-14 would be considered so as to assure the BPC review of
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pertinent as-built dimensions from welded attachments to inter-
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mediate break locations.
The inspector informed cognizant licensee personnel that the need
for further NRC examination of proper application of the criteria
including field measurements would be identified as Inspector
Followup Item 424/86-39-02, 425/86-19-02, Field confirmation of
distance between welded attachments and the location of inter-
mediate pipe breaks.
The inspector also verified that break locations postulated prior
to April 26, 1984, with welded attachments within five pipe
diameters were as follows:
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Stress
Break Support
Referenced
Calc
Data
Within
Item
Break No.
Isometric
X4CP
Point 5 pipe diameter
5
P-1055-C
1K5-1301-001-01 7073/74
68 V1-1301-008-H052
6
P-1054-C
1K5-1301-001-01 7073/74
75 V1-1301-008-H055
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During field inspection of main steam line break locations listed
above, the inspector noted that associated pipe whip restraints
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(PBR-217 and PBR-218) had been partially completed since they were
also to function as pipe supports. However, they had been elimi-
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nated as whip restraints without meeting the required SD criteria
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and were not presently able to meet this required design
function.
Cognizant licensee personnel and supporting data
verified that the licensee did not intend to complete fabrication
of these whip restraints as of the date of this inspection.
(6) NRC/GPC Enforcement Conference
Subsequent to this inspection, Region II reviewed the need for
escalated enforcement on this matter.
A decision was reached to
schedule an enforcement conference with the licensee on this
matter.
This conference was held in the NRC Region II office,
Atlanta, Georgia, on June 23, 1986. Attendees were as follows:
Licensee Attendees: R. E. Conway, Vogtle Project Director, Georgia
Power Company (GPC)
P. D. Rice, Vice President, Vogtle Engineer-
ing, GPC
C. W. Hays, Vogtle Project Quality Assurance
Menager, GPC
J. Bailey, Vogtle Project Licensing Manager,
Southern Company Services (SCS)
0. Batum, Deputy to Vogtle, Engineering Vice
President, SCS
F. B. Marsh, Vogtle Project Engineering Manager
Bechtel Power Corporation (BPC)
S. J. Cereghio, Nuclear Group Supervisor, BPC
W. E. Burns, Nuclear Licensing-Nuclear
Operations, GPC
C. W. Whitney, GPC-Vogtle Legal Counsel
NRC Attendees:
R. D. Walker, Acting Deputy Regional
Administrator
L. A. Reyes, Acting Director, Division of
Reactor Projects (DRP)
B. W. Jones, Regional Counsel, RII
V. Panciera, Deputy Director, Division of
Reactor Safety (DRS)
V. L. Brownlee, Chief, Reactor Projects
Branch 3, DRP
A. R. Herdt, Chief, Engineering Branch, DRS
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L. Trocine, Enforcement Specialist, RII
M. V. Sinkule, Chief, Reactor Projects
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Section 3C, DRP
H. H. Livermore, Senior Resident Inspector,
Construction, Vogtle
E. F. Christnot, Vogtle Project Engineer, DRP
E. H. Girard, Reactor Inspector, DRS
G. A. Hallstrom, Reactor Inspector, DRS
S. J. Vias, Reactor Inspector, DRS
M. Miller, Vogtle Project Manager, Office of
Nuclear Reactor Regulation (NRR)
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R. L. Balland, Chief. Engineering Branch,
PWR-A, NRR
G. Bagchi, Section Leader, Engineering Branch,
PWR-A, NRR
E. Holler, Enforcement Specialist, Office of
Inspection and Enforcement (IE)
The meeting was convened at 10:00 a.m., on June 23, 1986.
The
licensee was informed of the Region II decision that this matter
was considered a material false statement, but that the questions
of severity level and potential civil penalty were not yet decided.
GPC Vice President Conway responded that GPC's position has been
and continues to be one of complete, cpen, and honest communica-
tion with the NRC but that there had been a QA breakdown on this
matter.
Vice President Rice then submitted a proposed agenda and
summary of points for discussion.
This licensee submittal is
included as an attachment to this report.
Significant clarifications / amplifications obtained during the
licensee's presentation were as follows:
Arbitraryintermediatepipebreak(AIPB) concept
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Mis-communication had occurred between the licensee
and NRR regarding the need of SD criteria to mitigate
the severity of potential damage from unanticipated
The licensee had understood that conformance to
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NC/ND-3645 would satisfy the need to minimize stress
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from welded attachments.
This view is supported by
other existing design features to accomplish defense on
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depth objectives.
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BPC design engineers had considered SD criteria to be
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non-relevant since conformance to NC/ND-3645 was main-
tained.
Chronology
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The BPC review which commenced in June, 1984, was for
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completeness regarding AIPB requirements and was the
first review with supporting documentation.
Supporting
documentation is unavailable for reviews occurring
before the April 26, 1984, submittal.
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The BPC review completed in July,1984, did establish
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the lack of conformance to SD criteria for main steam
breaks P-1054-C and P-1055-C.
However, this lack of
conformance was not reported to the NRC since the 50
criteria was considered non-relevant as reported above.
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Error in Project Proposal
The initial review of baseline data was not done in
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accordance with formally established GPC policies and
procedures.
The initial review was completed as an
engineering study and no formal policies and procedures
had existed to control this type of activity similar to
those controlling other formal licensee submittals such
as 10.55(e) and Part 21 reports.
For the 11 cases where welded attachments within SD of
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the break point which have been identified through the
June, 1986, licensee engineering review, there were 11
cases in the main steam system.
Of these, only break
P-1054-C is considered by the licensee to require
further deliberations regarding its lack of conformance
to the SD criteria.
Root Cause
The licensee agreed that a QA program weakness had
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existed due to lack of formality in controlling the
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engineering studies associated with the SD criteria.
The licensee agreed that the materiality of the lack of
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conformance with the 50 criteria should have been
identified in July 1984 when the nonconformances were
first identified.
The licensee's failure to identify
materiality was attributed to the failure to properly
interpret the SD criteria versus NC/ND-3645 conformance.
The interpretation failure was attributed to mis-
communication with NRR which would have been rectified
under formal controlling procedures.
The lack of notification to NRC of the inaccuracies in
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the April 26, 1984, submittal which were discovered on
July, 1984, was attributed to the licensee's failure to
recognize the materiality of the 50 requirements.
Corrective Actions
The relevant design criteria (DC-1018) has been revised
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to include the 50 criteria.
However, the revised
DC-1018 will not be implemented until resolution of the
technical issue with NRC.
The licensee's objective is
to avoid imposing the SD criteria due to potential
rework required.
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The QA audit of similar engineering studies / proposals
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included a sample of five which were similar to the AIPB
studies.
These five had not been examined during
readiness review activities.
No discrepancies were
identified.
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The training seminar for project managers who prepare or
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approve regulatory comitments will be conducted by Vice
President Rice.
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The letter to NRC on a proposed AIPB concept to resolve
the technical issue is expected to be transmitted during
the first week of July, 1986.
The meeting was adjourned at 11:50 a.m., on June 23, 1986.
NRC escalated enforcement activities on this matter is not complete.
Therefore no Notice of Violation on this matter is enclosed with this
report.
Separate correspondence will be issued on the results of NRC
deliberations on this matter,
b.
(Closed) Unresolved Item (424/86-03-02, 425/86-02-02):
Polar Crane
Design
This item concerns potentially inadequate design calculations for
seismic qualification of the VEGP Polar Crane. Apparent discrepancies
were identified during the NRC review of the Seismic design calcula-
tion package (BPC leg AXAL01-46-2) provided by the VEGP Polar Crane
Supplier.
BPC specification X4AL01, Revision 1, dated December 14,
1978, requires dynamic analyses for eight different loading conditions
for both Safe Shutdown Earthquake (SSE) and Operational Basis Earth-
quake (0BE).-
Analyses for two of the required loading conditions
(trolley in center of span with full load in mid position and trolley
at end of span with full load in mid position) were not included. NRC
concern was also expressed due to the possibility that the worst case
for hook
height may not have been included in the load conditions
originally specified for analysis.
The technical adequacy of the initial BPC response was questioned on
two points as follows:
The implication that crane girder seismic stresses are essentially
unaffected by the lifted load.
Stated reasoning is that all
Z-component (vertical direction) earthquake stresses are absorbed
in the rope;
i.e., that these stresses are r.ever transmitted to
the Polar Crane Bridge Girder.
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The contentien thatathe up-position of lifted load is the worst
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case for hook height.
The stated basis f6r this contention is
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that the natural . structural frequency (about 2 hertz) for the
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load /ropeicombination in the up position corfesponds to the peak
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acceleration of the vertical seismic response , spectra.
Further
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support of an accurate coincidence of the1 frequencies is necessary
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since lowering the load could increase the seismic acceleration if
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the frequencies do not coincide.
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Further GPC response on this issue was transmitted [GN-892)~.
y letters, dated
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April 17, 1986, (Log: GN-864) and May 1, 1986 (Log:
Engineer-
ing justification in response to the above questioqs was as follows:
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Calculated stresses from model response specfrum { worst-case load)
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analyses were provided to show that the .Y-component (tangential
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direction) stresses are the major contributor to the Crane girder
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seismic stresses.
The vertical (Z-direction) load ~c'ontributes
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primarily to rope stresses and is transferred from the rope to the
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girder.
However, the resultant bending stress is ~ considerably
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less than the Y-component stresses and, therefore, contributes
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little to the Crane girder seismic. stresses.
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Further support of the up-position of- lifted load as the worst
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case for hook height was provided through a comparison of the
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natural frequencies of the polar Crane system as obta'ined through
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finite element model analysis by the polar crane vendor with
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additional calculations completed by BPC on Apr.11 14,1986.
Additional clarification established that. the initial _ response
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spectra reviewed represented the response of a cingle-degree-of-
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freedom oscillator mounted on the lifted load and not the response
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of the polar crane structure.
The basis for seismic analysis of
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the polar crane girders is a set of in-siiructure response spectra
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-developed from a time-history analysis of the b'ilding.
The set
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of vertical response spectra show ' peak responses to lie in the
range from 2 to 10 cps. Below 2 cps, the response is a descending
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ramp. Therefore, as the load is lowered and frequency reduced the
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crane will experience progressively lower levels of , amplification
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in response.
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During review of the above information prior to this inspection further
question was . raised regarding conformance with allowable stress
criteria specified in FSAR paragraph 9.1.5.2.3.1.B versus stress values
stated from the model response spectrum analyses. Additional clarifi-
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cation transmitted with the letter of May 1,1986, indicated acceptable
polar crane component stresses.
During this inspection the inspector' examined background data support-
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ing the engineering justification provided within GPC's April 17, 1986,
and May 1, 1986 letters.
No discrepancies were identified and this
item is considered closed.
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c.
(0 pen) Unresolved Item (424/86-03-03, 425/86-02-03)
High Strength
Bolted Connections
This item concerns ASTM A325 and A490 high strength steel bolts
installed at plant Vogtle.
Discussions with QC inspectors and other
cognizant licensee personnel had identified potential overtensioning of
bolts on mainplate girder #5 at elevation 240' in the Control Building.
The girder was installed in August 1982, and potential overtorquing
was observed due to lack of conformance to " turn-of-nut" installation
requirements within construction procedure CD-T-16. Bolts in girder #5
were replaced.
However, followup discussions with cognizant licensee personnel
had established additional potential for overtensioning due to the
following:
A common philosophy that bolt overtensioning would not present a
problem as long as the bolts did not break during installation. A
technically supported justification of this philosophy was
requested.
An apparent inability of the QC inspection program (past or
present) to identify overtensioning (overtorquing) to near failure
limits.
QC inspection personnel uniformly stated that CD-T-16
specified only that the minimum required torque be checked; i.e.,
no check for potential overtorque is required or conducted and QA
surveillance of " snug tightening" or application of reference
match marks is not required.
Installation of high strength bolts by craftsmen who had not
received training on the " turn-of-nut" method.
The turn-of-nut
method was initiated on Revision 3, dated July 23, 1982, of
CF-T-16.
Initial training of construction craftsman was completed
on September 8,1982.
Statements by cognizant licensee personnel that the minimum time
frame during which installation by turn-of-nut method could be
suspect and overtorque a potential problem is from July 23 to
September 8, 1982.
Further GPC response on this issue was transmitted by letter to
Region II dated April 17,1986, (Log:
GN-864) and reviewed by the
inspector prior to this inspection.
The adequacy of the engineering
justification was questioned on several points and discussed with
cognizant licensee personnel during this inspection. Further justifi-
cation was requested on the following apparent inconsistencies, as a
minimum.
.
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13
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The contention that bolt overtension would have been identified by
i
the QC inspection program; i.e., on overtensioned bolt would turn
(or break) prior to reaching the minimum torque setting on the
calibrated torque wrench.
(This contention apparently ignores
i
test data showing significant remaining bolt tension capacity
above the ASTM proof load (elastic limit) versus CD-T-16 require-
ments that test torque wrenches be calibrated below proof load.)
The contention that the credibility of the GPC bolting program to
identify overtensioned bolts is established by GPC audits and
audits by others.
(This contention apparently ignores the fact
that installation methods and QC procedures which were audited are
4
i
structured to locate and correct undertorqued bolts - overtensioned
bolts are not addressed.)
I
The contention that design in accordance with the 1969 AISC
specification will yield factors of safety of a least 3 and 5,
respectively against ultimate tension and shear failures.
(This
,
contention apparently ignores the 'more germain safety factors
j
associated with the shear capacity of a friction connection using
tensioned bolts and the tensile capacity of these bolts,
i.e.,
j
those safety factors actually prevailing in the design.)
'
The contention that a direct tension load applied to a previously
torque-tightened bolt ultimately approaches the direct tension
load characteristic curve for the bolt; i.e., severely overtorqued
bolts have a tensile load reserve that will provide additional
safety against a failure should additional stress be applied by
,
i
additional strain.
(The test data referenced generally support
l
this conclusion.
However, test data limitations apparently
ignored included thread length of 9/16" within the grip rather
than the more severe 1/8" which would be applicable to the
j
majority of bolts used at plant Vogtle. The 1/8" length provided
'
the significantly more drooping curve.
Therefore, the data
i
referenced is not considered to verify the contention of strength
recovery to the direct tension characteristic for the majority of
bolts used at plant Vogtle.)
'
The contention that initial bolt preload has little effect on the
ultimate shear strength of the bolt in a connection.
(The test
data referenced generally support this conclusion. However, test
data referenced were from bolts in double sheer with both sheer
i
planes passing through the bolt shank. The test jig was designed
to keep the bolt shank in as near to pure shear as possible. No
information was offered relative to the more severe situation of
single shear through the thread root under grip conditions that
,
t
could offer prying tension.
Therefore, additional assurance is
l
needed that the more severe situation is not applicable to con-
l
nections at plant Vogtle.)
, _ . _
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14
During this inspection, additional potential for overtensioning was
identified due to deviation report CD-1417 issued October 8,1981, due
to high tension values in high strength bolted connections within the
turbine building.
A total of 484 connections were inspected with 403
connections exhibiting approximately 850 ft. Ibs. of torque.
Testing
of sample bolts on a hydraulic tensioner established the tensile load
associated with this torque to be equivalent to the ASTM specified
ultimate tensile strength for the 7/8" diameter A325 bolts involved.
The bolts were installed with the calibrated impact wrench method (not
turn-of-nut method) and those methods together with identical impact
wrenches was also used for installation of high strength bolts in the
auxiliary building (levels D and C) during the same time frame.
Therefore, additional engineering justification was requested regarding
any potential long-term adverse affects (stress corrosion cracking, or
other failure mechanisms) for high strength bolts preloaded to near
ultimate tensile strength values.
Additional GPC response was transmitted by letter, dated May 12, 1986
(Log: GN-908) and NRC evaluation is not complete.
This item remains
open.
d.
(Closed) Unresolved Item (425/85-40-02, 425/85-31-02):
Assurance of
adequate backpurge for welding stainless steel piping
This item concerned the use of adequate backpurge when welding stain-
less steel piping.
During previous examinations of welding activities
on stainless steel piping, the inspector noted that oxygen analyzers
were not used to assure the 1% minimum oxygen requirement and that
welders involved were uncertain as to the required argon flow rates and
minimum oxygen required.
Further that clarifying information was not
included on some of the welding technique sheets provided to the
welders involved.
During this inspection, the inspector examined
revised technique sheets and records of additional welder training and
QC surveillance activities which were conducted to assure meeting
backpurge requirements.
The inspector observed uniform use of oxygen
analyzers and adequate backpurge for the welding activities reported in
paragraph 6.b.(1).
This item is considered closed.
4.
Unresolved Items
No unresolved items were identified during this inspection.
'
5.
Independent Inspection Effort
Housekeeping (54834B), Material Identification and Control (429028), and
Material Control (429408)
The inspector conducted a general inspection of Units 1 and 2 containments,
,
,
the control building and the reactor auxiliary building to observe activi-
l
ties such as housekeeping, material identification and control; material
control, and storage,
i
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15
Within the areas examined, no violations or deviations were identified.
6.
Safety-Related Piping (Unit 2)
The inspector examined welding and nonwelding activities for safety-related
piping to determine whether applicable code and procedure requirements were
being met. The applicable code for safety-related piping is the ASME Boiler
and Pressure Vessel Code,Section III,1977 Edition with Addenda through
W77. .
a.
Review of Nonwelding Quality Records (49065)
The inspector selected various safety-related piping components (e.g.,
pipe, fittings and welded-in components) for review of pertinent
records to determine conformance with procurement, storage and
installation specifications and QA/QC site procedures.
Records of the following items were selected for review to ascertain
whether they (records) were in conformance with applicable requirements
relative to the following areas: material test reports / certifications;
vendor supplied NDE reports; Nuclear Steam Service Supply quality
release; site receipt inspections; storage; installation; vendor
nonconformance reports.
Item
Heat / Control No.
System
3/4" dia. SS
S/N B7459
Safety Injection
angle globe valve
3/4" dia. SS
S/N H178 ABE
Safety Injection
SS Gate Valve
1" dia sched 160
8 8607
Chemical and Volume
SS 90 ell
Control
Within the areas inspected, no violations or deviations were identified.
b.
Welding Activities
(1) Production Welding (55050)
The inspector observed in-process welding activities of safety
injection and chemical volume and control system piping field
welds inside of containment as described below to determine
whether applicable code and procedure requirements were being met.
. .
_
_
t
.
16
The below listed welds are examined in-process to verify work
conducted in accordance with traveler, welder identification and
location, welding procedures, WPS assignment, welding technique
and sequence, materials identity, weld geometry, fit-up; temporary
attachments, gas purging, preheat, electrical characteristics,
shielding gas, welding equipment condition, interpass temperature,
interpass cleaning, process control systems, qualifications of
inspection personnel, and weld history records.
Size
Status
2K4-1208-015-02 R/1
112-W-115
1"
Final Pass
112-W-109
1"
First Pass
2K4-1204-030-02 R/3
034-W-109
2"
Root Pass
030-W-136
3/4"
Final Pass
030-W-137
3/4"
Final Pass
030-W-140
3/4"
Second Pass
The following inspector qualification status records and QA/QC
Inspector Qualification / Certification records were reviewed
relative to inspection of the weld joints listed above.
Inspector
Type of Certification
PEM-PPP
VT-II
MSG-PPP
VT-II
(2) Welding Procedures
Welding procedure specifications (WPS) applicable to the weld
joints listed in paragraph 6.b.(1) were selected for review and
comparison with the ASME code as follows:
Procedure Qualification
WPS
Process
PQR Reports
1/
29-111/1-8-08-1
125,132,133
(9/7/83)
l
38-111/1-KI-1
120,121
(12/6/85)
l
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17
General welding standard BWS-111/1 (4/15/85)
1/ GTAW - Gas Tungsten ARC-Welding
The above WPSs and their supporting Procedure Qualification
Records (PQRs) were reviewed to ascertain whether essential,
supplementary and/or nonessential variables, including thermal
treatment, were consistent with Code requirements; whether the
WPSs were properly qualified and their supporting PQRs were
accurate and retrievable; whether all mechanical tests had been
performed and the results met the minimum requirements; whether
the PQRs had been reviewed and certified by appropriate contractor /
licensee personnel; and whether essential were noted.
WPSs are
qualified in accordance with ASME Section IX, the latest edition
and addenda at the time of qualification.
(3) Welder Performance Qualification
The inspector reviewed the PPP program for qualification of
welders and welding operators for compliance with QA procedures
and ASME Code requirements.
The following welder qualification status records and " Records of
Performance Qualification Test" were reviewed relative to the weld
joints-listed in paragraph 6.b.(1).
Welder Symbol
WPS
29-III/I-8-0B-1
CZ1
38-III/I-8-KI-1
(4) Welding Filler Material Control
The inspector reviewed the PPP program for control of welding
materials to determine whether materials were being purchased,
accepted, stored and handled in accordance with QA procedures and
applicable code requirements.
The following specific areas were
examined.
-
Purchasing, receiving, storing, and distribution and handling
procedures; material identification; and inspection of
welding material issuing stations.
-
Welding material purchasing and receiving records for the
following material applicable to current production welding
were reviewed for conformance with applicable procedures and
code requirements.
.
. . . .
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.
18
Tyge
Size
Heat, Lot, Batch /No.
1/16" X 36"
26245
3/32" X 36"
05394
1/8" X 36"
P0443
5/32" X 14"
X45602
During the above inspection, the inspector observed apparent
documentation deficiencies for E309L electrodes stored in the
"doublewide" welding materials distribution center (WMDC).
The
covering of these electrodes were not marked with heat, lot, or
control number and the inspector requested documentation which
independently verified that the electrodes involved were from heat
X45602 as was indicated on the stationary holding oven door. WMDC
personnel informed the inspector that the electrodes involved had
been received via GPC bulk materials requisition for on-site
vendor use to complete repair of ASME code valves (maintenance
work order 12607258) (50.55(e) item 424/425 CDR 85-90).
The
inspector reviewed requisition No. 221686 (dated April 26,1986)
for conformance to PPP procedure VIII-3 " Control of Welding
Consumables" dated February 27, 1985, and GPC procedure MD-T-12"
Receipt Inspection and Storage / Issue of Pipe, Pipe Components, and
Weld Filler Material" dated June 21, 1985. The inspector observed
that requisition No. 221686 did not conform to requirements in
that it did not:
Include "N/A" in the Unit No., Drawing No., Rev., System No.,
Project Class and Material Class spaces
Include the authorized signature for the owners welding
section supervisor, or his designee in the " approved by"
space
Include an "N/A" in the GPC "QC Inspection" space
Include the signature of a contractor's Q.C. representative
in the "Q.A. Appr. Doc." space
Include the Purchase Order No.; Item No.; or Heat, Lot, or
Control No.
The inspector informed cognizant licensee personnel that the above
deficiencies were considered a lack of conformance to 10 CFR 50
Appendix B, Criterion V, and would be identified as Violation
424/86-39-03, 425/86-19-03, Failure to Follow Procedures for
Control of Welding Consumables.
.
_
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_
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19
7.
Previously Identified Inspector Followup Items
a.
(Closed) Inspector Followup Item (424,425/85-14-01):
Clarification of
Liquid Penetrant Inspection Procedure
This item concerned required minimum light intensity at the inspection
site under PPP procedure IX-PT-1-W77.
Cognizant licensee personnel
informed the inspector that all PPP NDE Technicians were aware of the
illumination / lighting requirements to perform penetrant examination to
ASME Section V requirements.
Further, that all technicians were
supplied flashlights to aid in inspections.
Followup discussion with
NDE technicians confirmed that flashlights are universally used for
penetrant inspections in the field. This item is considered closed,
b.
(Closed) Inspector Followup Item (424/85-40-01, 425/85-39-01):
Assurance of Necessary Minimum Clearances for Installed Piping
This item concerned need for assurance of minimum clearances between
installed piping and Unit I containment pipe racks.
There appeared to be potential for contact between the 12" X 12" X 6"
Reducing Tee in Reactor Coolant Line 1K4-1201-036-01 and the top of
column 8 in Rack R0001.
The inspector reviewed revisions to PPP
Procedure 1 X 3 which require a general minimum separation of 1" in the
direction of the obstruction from installed piping. The inspector also
observed completed modifications to the top of column 8 in Rack R0001
to obtain adequate separation from Line 1K4-1201-036-01. The inspector
also completed random inspection on Unit 1 containment and no potential
clearances less than 1" were identified.
This item is considered
closed.
1
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ATTACHMENT
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NRC ENFORCEMENT CONFERENCE
JUNE 23, 1986
^
AGENDA
.
INTRODUCTION
R. E. CONWAY
.
BACKGROUND
P. D. RICE
.
-
AIPB CONCEPT
CHRONOLOGY
-
DISCUSSION OF PROBLEM
P. D. RICE
.
-
EVALUATION
-
ERROR IN PROJECT PROPOSAL
-
ROOT ~CAUSE
CORRECTIVE ACTIONS
P. D. RICE
'
.
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CONCLUSIONS
P. 9. RICE
.
l
CLOSING REMARKS
R. E. CONWAY
.
R. E. CONWAY - SENIOR VICE PRESIDENT AND PROJECT DIRECTOR
P. D. RICE - VICE PRESIDENT PROJECT ENGINEERING
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- - _ - - - - - - . _ _
- --
.
.
ARBITRARY INTERMEDIATE PIPE BREAK (AIPB) CONCEPT
DEFENSE IN-DEPTH CONCEPT FOR EVENTS UNANTICIPATED IN DESIGN
,
POSTULAT!0N OF BREAK POINTS IN HIGH-ENERGY PIPING SYSTEMS
.
WHEN ACTUAL STRESSES ARE BELOW ALLOWABLE STRESSES
,
NRC AND INDUSTRY AGREEMENT IN PRINCIPLE THAT EXISTING
.
DESIGN FLATURES AND CONSERVATISMS ACCOMPLISH THE DEFENSE
IN-DEPTH OBJECTIVES
COORDINATED INDUSTRY AND NRC EFFORTS HAVE JUSTIFIED THE
.
ELIMINATION OF THIS REQUIREMENT FOR A NUMBER OF PLANTS
i
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CHRONOLOGY
Nov '83
GPC LETTER TO NRC PROPOSING ELIMINATION OF ARBITRARY
INTERMEDIATE PIPE BREAKS (AIPB)
.
MAR '84
GPC MEETING WITH NRC ON NOVEMBER, 1983 LETTER
APR '84
GPC LETTER PROVIDING ADDITIONAL JUSTIFICATION FOR
NOVEMBER, 1983 LETTER AND RESPONDING TO MARCH, 1984
MEETING
JUN '84
BPC COMMENCED REVIEW OF AIPB FOR COMPLETENESS IN
ANTICIPATION OF NRC APPROVAL OF GPC PROPOSALS AND IN
PREPARATION FOR FSAR CHANGE
JUN '84
NRC LETTER APPROVED DEVIATION FROM STANDARD REVIEW PLAN
TO USE ALTERNATIVE AIPB CRITERIA
JUL '84
BPC COMPLETED REVIEW STARTED IN JUNE, 1984
l
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AuG '84
FSAR CHANGED TO INCORPORATE JUNE, 1984 NRC APPROVED
CHANGES
I
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,
.
ARBITRARY INTERMEDIATE PIPE BREAKS
EXCERPTS FROM APRIL 24, 1984, GPC LETTER TO NRC
ATTACHMENT
A,
TECHNICAL JUSTIFICATION FOR ELIMINATION OF
ARBITRARY INTERMEDIATE BREAKS, STATES:
"2.
WELDED ATTACHMENTS ARE NOT LOCATED IN CLOSE PROXIMITY TO
THE BREAKS TO BE ELIMINATED.
CONSEQUENTLY, LOCAL
BENDING STRESSES RESULTING FROM THESE ATTACHMENTS WILL
NOT SIGNIFICANTLY AFFECT THE STRESS LEVELS AT THE BREAK
LOCATIONS (REFER TO ATTACHMENT E)."
ATTACHMENT
E,
PROVISIONS FOR MINIMIZING LOCAL STRESSES FROM
WELDED ATTACHMENTS, STATES:
"WE HAVE REVIEWED ALL ARBITRARY INTERMEDIATE BREAK
LOCATIONS TO BE ELIMINATED AND HAVE DETERMINED THAT IN
NO CASES ARE WELDED ATTACHMENTS CLOSER THAN FIVE PIPING
DIAMETERS FROM POSTULATED BREAK LOCATIONS.
AT THIS
DISTANCE, LOCAL BENDING STRESSES INDUCED BY THE ATTACHMENT
WILL NOT AFFECT THE STRESSES AT THE POSTULATED BREAK POINT.
TO ENSURE THAT THIS IS THE CASE, THE LOCAL STRESSES HAVE
BEEN DETERMINED AND ADEED TO THE PRIMARY STRESS REPORT."
,
.
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,
CHRONOLOGY, CONT'D
APR '85
GPC LETTER REQUESTING ELIMINATION OF AIPB IN MAIN
FEEDWATER SYSTEM
JuN '85
NRC LETTER APPROVED APRIL, 1985 GPC MAIN FEEDWATER
'
SYSTEM PROPOSAL
MAR '85
READINESS REVIEW IDENTIFIED LACK OF UPDATE OF DESIGN
CRITERIA IN REGARD TO ELIMINATION OF AIPB CRITERIA
DEC '85/
NRC INSPECTIONS IDENTIFIED PROBLEMS IN IMPLEMENTATION OF
APR '86
ALTERNATIVE AIPB CRITERIA
APR '86
GPC REVIEW OF CURRENT HIGH STRESS POINTS TO DETERMINE
THOSE POINTS WITHIN SD OF A WELDED ATTACHMENT
MAY '86
GPC LETTER TO NRC DOCUMENTING APRIL, 1986 REVIEW AND
DEFINING FURTHER ACTIONS
MAY '86
GPC REVIEW OF HIGH STRESS POINTS AS PROMISED IN MAY,
1986 LETTER
JUN '86
GPC AND NRC MEETING TO DISCUSS RESULTS OF MAY, 1986
REVIEW AND TO DEFINE FURTHER ACTIONS
ACTIONS CONTINUE TO ADDRESS TECHNICAL ISSUES.
GPC
PREPARING RESPONSE TO NRC ON QUESTIONS AND ACTIONS
DEFINED IN JUNE 1986 MEETING
5
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_ . _ , _
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DISCUSSION OF PROBLEM
,
,
EVALUATION
.
ERROR IN PRO.!ECT PROPOSAL
.
ROOT CAUSE
.
6
,
_
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,
.
EVALUATION OF PROBLEM
CONDUCTED REVIEW FOR APPLfCABLE DOCUMENTATION
.
INTERVIEWED PERSONNEL INVOLVED
.
,
PERFORMED REREVIEW OF ARBITRARY INTERMEDIATE PIPE BREAK
.
(AIPB) LOCATIONS BASED ON MARCH, 1984 DESIGN DOCUMENTS AND
CRITERIA
-
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.
i
.
ERROR IN PROJECT PROPOSAL
BASELINE DATA (NOVEMBER 1983 GPC LETTER)
,
576 TOTAL PIPE BREAK LOCATIONS
-
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182 ARBITRARY INTERMEDIATE PIPE BREAKS (AIPB)
-
233 TOTAL PIPE WHIP RESTRAINTS
-
110 PIPE WHIP RESTRAINTS FOR AIPB
ERROR (JUNE 1986 ENGINEERING REVIEW)
,
-
18 CASES WHERE WELDED ATTACHMENTS WERE WITHIN SD OF
BREAK POINT
-
11 CASES IN MAIN STEAM SYSTEM
-
3 CASES IN MAIN FEEDWATER SYSTEM
-
2 CASES IN CHEMICAL AND VOLUME CONTROL SYSTEM
-
1 CASE IN STEAM GENERATOR WET LAY-UP SYSTEM
-
1 CASE IN AuxrLIARY FEEDWATER SYSTEM
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ROOT CAUSE
,
QA PROGRAM WEAKNESS
LACK OF FORMALITY IN DOCUMENTING THE SCOPE, CRITERIA,
.
DETAILED RESULTS, AND SUPERVISORY REVIEWS ASSOCIATED WITH
THE SD REVIEW
TIMELINESS OF IDENTIFICATION
,
FAILURE TO INCORPORATE PROPOSED SD PROVISION INTO
.
ENGINEERING DESIGN CRITERIA
0A AUDITS BASED PRIMARILY ON COMMITMENTS INCORPORATED
.
INTO PROJECT DESIGN DOCUMENTS
i
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9
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.
CORRECTIVE ACTIONS
REVIEW PROJECT AND NRC CORRESPONDENCE RELATED TO
.
ELIMINATION OF ARBITRARY INTERMEDIATE PIPE BREAK (AIPB) FOR
,
ANY SIMILAR ISSUES
0A AUDITED SELECTED PAST ENGINEERING / LICENSING
.
CORRESPONDENCE TO NRC FOR SIMILAR PROBLEMS
0A AUDIT PROCEDURES STRENGTHENED TO EXAMINE FOR PROPER
.
INCORPORATION OF COMMITMENTS MADE IN NRC CORRESPONDENCE
DESIGN CRITERIA REVISED TO REFLECT CURRENT APPROVED STATUS
.
(IMPLEMENTATION ON HOLD PENDING RESOLUTION OF TECHNICAL
ISSUE WITH NRC)
ACTION INITIATED TO STRENGTHEN PROJECT PROCEDURES FOR
.
OFF-NORMAL ENGINEERING REVIEWS
PROJECT POLICY FROCEDURES WHICH CONTROL CORRESPONDENCE TO
,
NRC REVISED TO STRENGTHEN PERSONNEL ACCOUNTABILITY FOR
ACCURACY
l
TRAINING PRESENTATION DEVELOPED AND SCHEDULED FOR
.
ENGINEERING PERSONNEL TO INCLUDE ENGINEERING PROCEDURE
CHANGES AND SENSITIVITY
l
TRAINING SEMINAR DEFINED FOR PROJECT MANAGERS WHO PREPARE
.
.
OR APPROVE REGULATORY CORRESPONDENCE
LETTER TO NRC IN PREPARATION ON PROPOSED AIPB CONCEPT
.
10
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V0GTLE MANAGEMENT PHILOSOPHY
READINESS REVIEW
'
.
.
CONTINUED EMPHASIS ON PROJECT POLICY IRAINING
.
QUALITY CONCERN PROGRAM
.
ANTI-DRUG PROGRAM
.
_
SENIOR CORPORATE INVOLVEMENT (PROJECT MANAGEMENT BOARD,
.
QUALITY ASSURANCE COMMITTEE AND READINESS REVIEW BOARD)
'
NUMEROUS TECHNICAL ASSESSMENTS (3 INP0 CONSTRUCTION
.
ASSESSMENTS, SELF-INITIATED EVALUATION AND DESIGN CONTROL
REVIEW)
CONTINUING ENHANCEMENTS TO PROJECT MANAGEMENT 0RGANIZATION
.
OPENNESS IN DEALINGS WITH NRC
.
11
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CONCLUSIONS
,
WEAKNESS IN QA PROGRAM FOR OFF-NORMAL REVIEWS -
.
LACK OF FORMALITY IN DOCUMENTING THE SCOPE, CRITERIA,
DETAILED RESULTS, AND SUPERVISORY REVIEWS ASSOCIATED WITH
THE SD REVIEW
CONTINUING DIALOGUE BETWEEN GPC AND NRC IS EXPECTED TO
.
SATISFACTORILY RESOLVE TECHNICAL ISSUE
GPC HAS AND WILL CONTINUE TO DEAL WITH ALL ISSUES IN AN
,
OPEN AND FACTUAL MANNER.
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