ML20153B893
| ML20153B893 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 02/05/1986 |
| From: | Ebneter S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | Fiedler P GENERAL PUBLIC UTILITIES CORP. |
| References | |
| NUDOCS 8602180166 | |
| Download: ML20153B893 (2) | |
See also: IR 05000219/1985014
Text
.
'
-
FEB 05192s'
Docket No. 50-219
License No. DPR-16
GPU Nuclear Corporation
ATTN: Mr. P. B. Fiedler
Vice President and Director
Oyster Creek Nuclear Generating Station
P
O. Box 388
Forked River,- NJ 08731
Gentlemen:
Subject:
Inspection No. 50-219/85-14
This refers to your letter dated November 7,1985, in response to our letter
dated September 4, 1985.
Thank you for informing us of the corrective and preventive actions documented
in your letter.
These actions were further discussed in the special meeting conducted by
Mr. J. P. Durr of this office and Mr. R. F. Wilson of your staff on November 13,
1985. Commitments made by your staff for the resolution of certain inspection
findings were addressed in our letter of December 16, 1985. Although our staff
has performed a preliminary review of some nf your specific criteria and pro-
cedures relating to'the 1985 reverification effort for piping and pipe supports,
a formal evaluation will be conducted during a future inspection of your
licensed facility.
This inspection will also include examination of the correc-
tive actions outlined 'in your letter of November 7,1985 and commitments made
in the November 13, 1985 meeting.
Your cooperation with us is appreciated.
Sincerely,
.
.
Stewart D. Ebneter, Di
ctor
Division of Reactor Safety
cc:
M. Laggart, BWR Licensing Manager
Licensing Manager, Oyster Creek
Public Document Room (PDR)
local Public Document Room (LPDR)
Nuclear Safety Informatio1 Center (NSIC)
NRC Resident Inspector
State of New Jersey
" ICIAL RECORD COPY
RL OYS C 85-14 - 0001.0.0
g[Q
01/24/86
,
.1E0/
L
-
- -
n
,
..
- '
-GPU Nuclear Corporation.
2
bcc:
Region I Docket Room (with concurrences)
Management Assistant, DRMA (w/o encl)
Section Chief, DRP
L
!-
l
!
I
.
i
(j
RI:DRS
RI:DRS
RI:DRS
Wiggins
Ourr_
dEP,44P7ef ?
/
l
Manoly/ms
,)
\\J
"slf,(66
'
W 1/Zif86
1/q/86
1/q/86
i
0FFICIAL RECORD COPY
RL OYS C 85-14 - 0002.0.0
01/24/86
i-
L
c
-
.
.
.
.
.
GPU Nuclear Corporation
U
MF
Post Ofbce Box 388
Route 9 South
Forked River.New Jersey 087310388
609 971-4000
Writer's Direct Dial Number:
November 7, 1985
Mr. Stewart D. Ebneter, Director
'
'
Division of Reactor Safety
U.S. Nuclear Regulatory Commission
Washington, DC 20655
Dear Mr. Ebneter:
Subject: Oyster Creek Nuclear Generating Station
Docket No.60-219
IE Inspection Report 85-14
Attachment I to this letter provides our responses to the Notice of
Violation and Notices of Deviation contained in Appendices A and B of your
letter dated September 4,1985. Due to the scope and complexity of the
response, and the preparation required to respond to the SALP, this
submittal has been delayed. The NRC Senior. Resident Inspector has been kept
informed of our progress in this area. By discussion and telecon an
extension to November 8,1985 was granted.
GPUN's preliminary responses to the concerns addressed by the identified
Notice of Violation and Notices of Deviation were presented in a management
meeting between the NRC staff and GPUN on May 20, 1985.
This meeting was
documented in IE Inspection Report 50-219/85-15. A second meeting was held
on June 13, 1985 and documented in IE Inspection Report 50-219/85-19. As a
result of this second meeting, a letter was sent from Mr. P. R. Clark,
President - GPUN to Dr. Thomas E. Murley, Region I Administrator on June 14,
1985. A follow-up meeting between GPUN and the NRC staff on GPUN's
completed and proposed corrective actions was held on June 25, 1985.
This
third meeting was documented in IE Inspection Report 50-219/85-22.
/ ^i
,A
9
D b uli M
i
GPU Nuclear Corporahon is a subsidiary of the General Pubhc Uhhties Corporation
'
- '
Mr. Stewart D. Ebneter, Direct r
-
Division of Rzctor Safety
-
Page 2
.-
In the event that any additional consents or questions should arise,
please contact Mr. John Rogers of sty staff at (609)971-4893.
Very truly yours,
k
R'. F. Wil. hm
sn
Vice P. resident
Technical Functions
--
RFW/JR/ dam
(0lllA)
Attachments
cc: Mr. Harry B. Kister, Chief
Division of Project and Resident Programs
U.S. Nuclear Regulatory Commission
Region I
631 Park Avenue
"
King of Prussia, PA .19406
Dr. Thomas E. Murley, Administrator
Region I
U.S. Nuclear Regulatory Commiss' ion
631 Park Avenue
.
King of Prussia, PA 19406
Mr. Jack N. Donohew, Jr.
U.S. Nuclear Regulatory Commission
7920 Norfolk Avenue, Phillips Bldg.
Bethesda, MD 20014
Mail Stop No. 314
.
NRC Resident Inspector
_
Oyster Creek Nuclear Generating Station
Forked River, NJ 08731
y
_ _ _
_ . .
.
-
-
-.
. _ ,
.
ATTACHMENT I
In response to IE' Inspection 50-219/85-14, a project was estabilsned to
verify both the seismic acceptability of the Oyster Creek Nuclear Generating
Station piping and pipe supports, and the validity of 'the results obtained
from the original 1979/1980 inspection effort. A field walkdown of piping
and pipe supports produced two (2) significant and related results:
1.
The Oyster Creek piping within the scope of IE Bulletin 79-02/14
has no identified inadequacies, except as noted herein for the
Emergency Service Water (ESW) System.
2.
Conclusions drawn from the original 1979/1980 effort were correct,
but the effort was weak in scope and especially in the
documentation of the specifics of tne work whicn was accomplished.
-
Over 700 supports' have been inspected to criteria which encompass and
"
exceed those defined in IE Bulletin 79-14. Less than 2% of the inspected
hangers have had discrepancies which required immediate repairs. Only one
hanger was found which had significant deficiencies resulting in the
,
implementing of Technical Specification time limits for repairs. A recent
inspection of the ESW system identified the need for installing additional
shims to minimize seismic gaps.
.
The 1985 reinspection effort is being conducted with fully approved
procedures by personnel qualified to ANSI N45.2.6 (1978) including
supplemental training in ASME section XI VT-3 or ASNT SNT-TC-1 A, (June 1980)
VT-3. _ Discrepancies between the as-found plant and respective as-designed
drawings are being identified and evaluated on three (3) levels. Firstly, a
conditional evaluation of the individual support is performed for
acceptability. Secondly, an anchor to anchor analysis is performed for
portions of piping. systems to assure that no combination of adjacent or
proximate hanger discrepancies affect tne operaoility of the system.
Thirdly, a computer analysis of the piping system and detailed pipe support
calculations will integrate all of the identified discrepancies to determine
-
the aggregate effect on an entire system.
The reverification program (and any responses'to the identified IE
Bulletins) is being conducted in accordance with a quality assurance program
as outlined in the GPUN Quality Assurance Plan, Section 2.3.1.1.
Reinspections for IE Bulletin 79-14 are now completed inside the drywell
with the exception of certain Recirculation Piping and pipe supports and
outside the drywell with the exception of 5 supports and associated piping
on the ESW system below the intake structure.
The remaining Recirculation
pipe and pipe supports are scheduled to be reinspected during the 11R
refueling outage, currently planned to commence in April 1986. Refer to
GPUN letter R. F. Wilson to S. D. Ebneter dated October 28, 1985. The
remaining ESW pipe and pipe supports will be completed prior to restart from
the 10M outage. The reverification for IE Bulletin 79-02 will commence
following the 10M outage and is scheduled to be completed prior to restart
from the 11R outage.
y
%-
.:
..,~.,
__
_
- -
.
.
.
Vinlation:
Criterion III of Appendix B to 10 CFR 50 requires the establishment of
measures to assure that applicable regulatory requirements and design
bases for important to safety structures and components are correctly
translated into sper.ifications, drawings, procedures and instructions.
Section IV, Part 1, Subsection A, of Revision 2 of the Jersey Central
' Power and Light / General Public Utilities (JCP&L/GPU) Operational Quality
Assurance Plan, the plan in effect at the time of the work, states, in
part, "the Manager-Generation Engineering is responsible for controlling
design work and administering the design control activities relating to
-
modification or addition of structures, components, and systems."
Section X Part 1, Subsection F, of Revision 2 of the JCP&L/GPU
Operational Quality Assurance Plan stated, in part, "the
Manager-Generation Engineering is responsible for the issuance and
approval of specification, drawings and installation requirements
associat.ed with modifications and additions..."
t
Contrary to the above, activities relating to IE Bulletins 79-02 and
79-14 were performed either: (1) with no documented engineering
procedures and specifications, or (2) with procedures and instructions
~
which were not controlled in accordance with the Operational QA plan.
Examples of tne first category include: Calculation No. 8.13.309 -
tnermal expansion analysis of line No. 6-ND-10; Calculation No. 8.31.208
- containment spray system pipe support calculations; calculation for
thermal analysis of emergency condenser line 10-NE-S from the reactor to
NE01-A, B.
Examples of the second category include: Field walkdown
check list for implementation of IE Bulletin 79-14; 80-44 procedure for
~
the modification and repair of pipe supports; 80-4S procedure for the
relocation of pipe supports; PP-001 procedure for the implementation of
IE Bulletin 79-14; Special Procedure No. 79-31 for inspection, test and
installation of concrete expansion anchor bolts in seismic piping
systems at OCNGS.
This is a Severity Level IV violation (Supplement 1) applicable to
License No. DPR-16.
Response:
GPUN concurs in the violation.
,,
1.
Corrective actions which have been taken and the results dchieved:
a.
GPUN has contacted Burns & Roe, Oyster Creek Nuclear
Generating Station's Architect / Engineer, and acquired copies
of documents and records relating to the 1979/1980 IE Bulletin 79-02, 79-14 field walkdown and seismic acceptability
verification. The procedures and specifications which were
utilized during the original 1979/1980 effort are now
available,
b.
Specific criteria and procedures relating to the 1985
reverification effort for piping and pipe supports have been
drafted by GPU Nuclear and reviewed by the NRC.
These
procedures are currently either approved and in use, or are in
the final review and approval cycle.
_
<
__
.
,
c.
All checklists and mark:;d up drawings relattd to cr resulting
.
from the 1985 field walkdowns have bean indexed and retained
to provide the requisite documentation of the reverification
'
effort.
2.
Corrective actions which will be taken to avoid further violations:
a.
Pipe support specifications and acceptance criteria are being
formalized and will be approved and released through the
Engineering Design and Configuration Control (ED&CC) document
control system.
- Date when full compliance will be achieved: Scheduled to be
released in November 1985
b.
Anchor bolt testing specifications and acceptance criteria are
being fonnalize' and will be approved and released through the
d
ED&CC document control system.
- Date when full compliance will be achieved: Scheduled to be
released by November 15, 1985.
c.
Configuration control piping isometric drawings will be
generated and GPUN procedures will be revised to assure future
system modifications have documented and readily retrievable
design-basis calculations and drawings supporting the
as-installed plant configuration.
.
- Date when full compliance will be achieved: Presently
projected to have: 1) a realistic estimate of the scope of 'the
redrawing effort; and 2) a draft schedule for completion
<
presented to GPU Nuclear upper management for funding, by
August 31, 1986. This date is contingent upon the actual
completion dates of the 1985 reinspection sub tasks. Further,
an interim program to control piping and pipe support design
will be in place to support the llR outage.
~' '
Deviation No.1:
Action Item No. 4 of IE Bulletin No. 79-02 specifies a design
requirement verification for each concrete anchor bolt with respect to
anchor bolt preload, design size and type, and proper installation.
If
these requirements cannot be met from existing quality control
documentation, a testing program should be initiated to assure that each
seismic category I system will perform its intended function. This
action item suggests two sampling techniques for testing of concrete
expansion anchors. The licensee s IE Bulletin 79-02 response dated
July 6,1979, states, "An in-plant inspection and test program has been
initiated to verify proper anchor bolt installation and correct size and
type." It also states, "The test metnod involves a pull test of the
anchor bolt to a minimum test load of 125% of the calculated bolt design
load. The sampling technique being used is the first method listed in
Appendix A of the supplement to the bulletin", i.e. testing one bolt per
base plate.
,
, . . ,
- _ ,.. .
. , . - -
_
.
_ __
-
---
-
.
'
Contrary t3 the abov1, infomaticn c:ntained in Attachment 8 of Burns
--
- -
and Roe document "R2 view cf Cenfomanca to USNRC IE Bullctins 79-02 and
79-14' indicates that bolts were tested on only seven of the twelve
'
. seismic Category I piping systems. The five systems which were excluded
from the anchor bolt testing were the Cleanup Demineralizer, Control Rod
Drive, Feedwater, Main Steam and the Shutdown Cooling Systems.
.
Therefore, the functionality of these five untested systems was not
assured. Furthermore, the anchor bolt testing on the seven tested
systems was neither conducted according to the first sampling ~ method
-
proposed by the bulletin and committed by the licensee in the July 6,
1979 response, nor according to the second sampling method which
requires a statistical sampling of anchor bolts ca each safety related-
-
system such that a 95 confidence level is provided to establish an
evidence of operability. of these systems.
' Response:
Notice of Deviation No. I contains an error. Although JCP&L origina11y '
prop'osed inspecting. twelve safety related systems. . subsequent
,
evaluations in 197.9 determined that one system (Reactor Building Closed
Cooling Water).was not a. safety related system. This yields a total
scope of eleven systems. These systems are: Liqui'd Poison', Shutdown
Cooling, Core Spray, Isolation Condenser, Emergency Service Water,
-
Control Rod Drive / Scram Discharge, Volume, Containment Spray, Feedwater,
,
Main Steam, Cleanup Demineralizer, Recirculation. JCP&L notified the
USNRC of our intent not to test the five systems identified in Notice of
Deviation No. I due to the high success rate of the seven tested systems
by letter dated December 7,1979. GPUN was not notified by the USNRC of
'the unacceptability of this decision'until May 1985.
GPUN' concurs in tile deviation.
1.
Corrective actions which have been taken and the results achieved.
The IE Bulletin 79-02 records resulting from the original 1979/1980
inspection effort have been acquired from the architect / engineer
and reviewed for completeness and adequacy of record traceability.
Although the majority of work which was performed is accurate, an
expanded scope of reinspection has been planned and is in the final
stages of development prior to implementation.
2.
Corrective actions which will be taken to avoid further deviations.
.
a.
. Completion of the requisite anchor bolt and baseplate
inspections will be conducted to verify their acceptability.
- Date wnen full comformance will be acnteved: The
reinspection is presently projected to be completed prior to
restart from the llR refueling outage,
b.
Individual anchor bolt pull tests will apply a load equal to
or greater than 125% of the calculated tensile loads and
determine the acceptability of slip behavior in this range
(less than 1/8" slip for shell type anchors and less than 3/8"
for wedge type anchors).
These specifications will be
controlled and proceduralized.
,
n-
.
.
.
t
'.
.
.
.
- Dat] when full comfcnnanca will be achiev;d: The
-
specifications.are presently scheduled to be approved and
-
issued by November 15, 1985
Deviation No. 2:
- IE Bulletin 79-14 states that the licensee dould assure that seismic
analyses of safety related piping systems accurately reflect the'
as-built configuration of the plant. The design specifications,
drawings, and other data used for these analyses should be considered in
this investigation. Among the bulletin statements are the following:
Identify the piping system parameters which were input into the
.
seismic analyses.
Identify the specific design document from which values of these
.
parameters were.obtained for the seismic ana' lyses.
Establish acceptance criteria which as built values of the
.
parameters must meet.
In addition, action items No. 2 and No. 3 require that, for each safety
related system, the ' licensee should verify by physical inspection that
the inspection elements meet the acceptance criteria.
In the August 31,
1979, and the February 22, 1980, letters to the USNRC, the licensee
provided reports of inspections conducted at the plant in accordance
with the IE Bulletin 79-14.
These reports state, "The inspection
.
consisted of comparing .the as-found condition of. piping, valves,
s'upports, and restraints with the design documents used as input to the-
seismic analysis." The' licensee-also provided the scope of wnat would
be considered in the inspection with a general summary.
Contrary to the above:
,
No specific design documents applicable to the seismic evaluation
of as-built safety related piping could be located.
Although an inspection checklist was established, it provided only
one acceptance criterion regarding hanger location which was not
substantiated by analysis. The cnecklist presented was not
,
~
sufficiently complete to compare the as-found conditions With a
specific acceptance criteria.
In addition, no evidence was
provided, during the inspection, to show that the checklist was
applied in a controlled and consistent manner. Furthermore, the
licensee did not provide, during the inspection, such documentation
as marked-up walkdown drawings, field inspection data sheets or
other first generation documents resulting from a physical
inspection.
Response:
GpVN concurs in the deviation.
1.
Corrective Actions which have been taken and the results achieved:
r
. , . ,
-
- ,
. , . ,
. . ,
-
-
,
. , . , , , . _ . .
~.
-
-
-
,
,
-
e.
, . .
'
.,
-
.
,
'
-
'
The identifiid c,issing document'aticn w;s acquired from the
. . . . .
.
a.-
architect / engineer and-ev&luated for. completeness and .
-
.
All first 'ge'neration documents'resulting!from
.
. acceptability.
the original field walkdown are available for inspection and
.
,
-
audi t. .
'
'
b.
Two GPUN specificatio'ns, SP-1302-12-208 "1985 IE Bulletin
79-02/14' Inspection Program Design Input for Piping Stress
Analysis" and SP-1302-12-212 "1985 IE Bulletin 79-02/14
Inspe'ction Program Design Input for Piping Support", have been
,
originated identifying the design document.applic'able to the
.
seismic analysis and are presently'in the review and approval
-
cycle..
.
' All inspectors 'and mai.ntenance ' personnel involved in the 1985
'
c.
reinspection effort have received trainin'g'in the use of the
inspection procedures and detailed explanations of the
,acceptanc'e criteria.
Corrective actions wh'ich will, be tak'en to avoid further deviations:
.
.
'2.
The 1985 in'spection criteria. exceed those req' ired by IE
u
'
.a.
Bulletin 79-14 and.have been'well documented.and approved.
.
They will be ' indexed and. filed witn the respective
reinspection packages.
- Date when full conformance will be achieved - The identifed
documents will be placed in duplicate storate as they are
.-
completed, presently projected, except for Recirculation
Piping supports, for after the-llR refueling outage.
.
Marked up first generation field walkdown copies of records
b.
will be retained in duplicate storage as they complete the~
review and approval cycle.
-
- Date when full comformance will be achieved - All records
will be indexed and stored af ter tney have received final
approval, presently projected for after the llR refueling
outage.
.
.