ML16161A867
| ML16161A867 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/07/1987 |
| From: | Jape F, Matt Thomas NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML16161A866 | List: |
| References | |
| 50-269-87-30, 50-270-87-30, 50-287-87-30, NUDOCS 8708270297 | |
| Download: ML16161A867 (6) | |
See also: IR 05000269/1987030
Text
SREGo
UNITED STATES
0
oNUCLEAR
REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Report Nos.:
50-269/87-30, 50-270/87-30, and 50-287/87-30
Licensee:
Duke Power Company
422 South Church Street
Charlotte, NC 28242
Docket Nos.: 50-269, 50-270,
License Nos.: DPR-38, DPR-47, and
and 50-287
Facility Name:
Oconee 1, 2, and 3
Inspection Con ucted:
July 21-24, 1987
Inspector: ___
-
jIrazkz8187
M. Thodas
Date Signed
Approved by::__
F. Jape, Chief
Odte Signed
Test Programs Section
Division of Reactor Safety
SUMMARY
Scope: This routine, unannounced inspection was in the areas of followup on
questions concerning operation with low pressure service water (LPSW)
system
inlet temperature greater than the design basis value given in the FSARs and
followup on previous open items.
Results:
One violation was identified.
Failure to perform a 10 CFR 50.59
evaluation for operation with LPSW inlet temperature above 75F, paragraph 5.
8708270297 870814
ADOCK 05000269
0
REPORT DETAILS
1. Persons Contacted
Licensee Employees
- P. M. Abraham, Design Engineer, Safety Analysis Section
- R. C. Bucy, Design Engineer, Safety Analysis Section
- P. F. Guill, Licensing Engineer
- D. M. Hubbard, Performance Engineer
- M. D. McIntosh, Acting General Manager for Nuclear Stations
- J. F. Norris, Design Engineer, M&N Division
- F. E. Owens, Regulatory Compliance
- R. L. Sweigart, Operations Superintendent
- M. S. Tuckman, Station Manager
- E. M. Weaver, Design Engineer, Station Support
Other licensee employees contacted included test coordinators, engineers,
operators,
and office personnel.
NRC Resident Inspectors
- J. C. Bryant, Senior Resident Inspector
- P. H. Skinner, Senior Resident Inspector
- L. D. Wert, Resident Inspector
- Attended exit interview.
- Attended NRR/Region II meeting on July 21, 1987.
- Attended NRR/Region II meeting by phone on July 21, 1987.
2. Exit Interview
The inspection scope and findings were summarized on July 24, 1987, with
those persons indicated in paragraph 1 above. The inspector described the
areas inspected and discussed in detail the inspection findings.
No
dissenting comments were received from the licensee.
The following new
item was identified during this inspection.
Violation 50-269, 270, 287/87-30-01, Failure to Perform a 10 CFR 50.59
Evaluation for Operation with LPSW Inlet Temperature Above 750 F,
paragraph 5.
The licensee did not identify as proprietary any of the materials provided
to or reviewed by the inspector during this inspection.
3. Licensee Action on Previous Enforcement Matters
This subject was not addressed in the inspection.
2
4.
Unresolved Items
Unresolved items were not identified during this inspection.
5. LPSW Inlet Temperature in Excess of FSAR Value
The inspector followed .up on the Unresolved Item
(UNR)
269,
270,
287/87-23-01,
concerning plant operation with LPSW inlet temperature
greater than the design basis value as specified in the FSAR.
Because the lake water temperature is expected to exceed the FSAR value
of 750 F by August 1, 1987,
a meeting was held at the Oconee site on
July 21, 1987, to discuss this issue.
The meeting was attended by NRC
representatives from NRR and Region II,
and Duke Power Company (DPC)
representatives from the site and the General Office.
At the meeting, the
licensee stated that the following actions are being considered due to the
combination of high lake water temperatures and the degraded capabilities
of the Units 1 and 2 low pressure injection (LPI)
and reactor building
cooling units (RBCUs)
coolers (which have not-been fully cleaned and
tested):
(a) A modification to the Confirmatory Order (issued by NRC on April 10,
1987 for Unit 1) will be issued to DPC.
(b) The licensee will submit a Technical Specification (TS)
change for
Unit 2 for NRC approval prior to the lake water temperature exceeding
800F.
(c) Unit 3 can operate at 100% power,
based on the results of the
licensee's 10 CFR 50.59 evaluation.
(d) The licensee will submit the 10 CFR 50.59 evaluation to the NRC
prior to the lake water temperature exceeding 750 F.
During the meeting, the licensee stated that the lake water temperature
had exceeded 750F for various lengths of time during nine of the past 11
years.
It was further stated that, prior to the question being raised by
the inspector (NRC
Inspection Report 269,
270,
287/87-17),
lake water
temperature above 750 F had not been evaluated for impact on plant
operation nor on the design bases accident analyses.
The inspector
questioned licensee personnel
as to why the elevated lake water
temperature had not been previously evaluated, and whether there was a
mechanism for ascertaining when FSAR values that are used in procedures
or accident analyses have been exceeded.
The licensee stated that FSAR
values used in design calculations, procedures,
etc., are constantly
being reviewed as a result of the emergency condenser circulating water
(ECCW)
problem; NRC Appendix R inspections; and safety system functional
inspections performed by both NRC and DPC.
However, the licensee stated
that the higher lake water temperature had not been previously evaluated
prior to the question being raised by the inspector, because of an
apparent oversight.
The inspector informed the licensee that failure to
3
perform a safety evaluation to determine if operation with higher lake
water temperature constituted an unreviewed safety question or not, is a
violation of NRC requirement 10 CFR 50.59.
Therefore, unresolved item
50-269,
270,
287/87-23-01 will be closed and a violation will be
identified.
The violation will be tracked as
item 50-269,
270,
287/87-30-01, Failure to Perform a 10 CFR 50.59 Evaluation for Operation
with LPSW Inlet Temperature Above 750F.
No other violations or deviations were identified.
6.
Followup on Open Items (92701)
a.
(Closed)
IFI 269/85-40-01,
Concerning Followup on Licensee Actions
to Resolve the Main Steam Relief Valve (MSRV)
Blowdown Problems.
The inspector reviewed the licensee's post trip review reports for
Unit 1 reactor trips which occurred on January 21, 1986 and May 10,
1986.
Per the trip report of January 21,
1986, work requests (WR
20653D/53688D) were written for MSRV IMS-8 because the valve did not
reseat properly (approximately 975 psig).
This did not appear to be
a problem in that the blowdown pressure setpoint for valve IMS-8 to
reseat is 977 psig.
There were not reported problems with any of
the other MSRVs.
This indicates that the licensee's program for
rebuilding the MSRVs appears to have solved the problem.
This item is considered closed.
b. (Closed) UNR 269/85-40-02, Concerning a Determination by the Licensee
as to the Reportability of the Pressurizer Code Safety Valves Lifting
Outside of their Setpoint Tolerances During Testing.
The licensee
evaluated the valves with high lift
setpoints and concluded that the
valves did not represent a safety concern nor a technical
inoperability.
Per intrastation letters, dated June 9, 1986,
and
May 6, 1987, and with the implementation of the Problem Investigation
Report Station Directive, the licensee has more clearly defined and
documented guidelines and criteria to evaluate for reportability
pressurizer code safety valves that have been tested with high lift
setpoints.
This item is considered closed.
c.
(Closed)
IFI 287/85-31-01,
Concerning Repairs to Unit 3 Reactor
Building Grating.
The licensee replaced all the missing tie-down
clips and secured the grating per work request WR 53478D. This was
done prior to Unit 3 restarting from the refueling which it was in
at the time the concern was identified.
The licensee stated during
this inspection that a program for securing reactor building grating
during refueling has been implemented for all three units.
This item is considered closed.
~r,.
4
d.
(Closed)
IFI 269,
270,
287/86-06-01,
Concerning Review of Final
Resolution to Discrepancy Identified During Performance of Periodic
Test PT/0/A/0610/02.
The inspector reviewed the periodic test in
question, which was approved March 3, 1986.
All test discrepancies
were resolved and test acceptance criteria met.
This item is considered closed.
e.
(Closed)
TI 2500/16, Concerning Licensee Handling of IE Information
Notice 85-45,
Determine If a Seismic Interaction Exists Between
Movable Incore Flux Mapping Systems and Seal Table at Westinghouse
Designed Facilities or Facilities with a Similar Design.
The
licensee evaluated Information Notice 85-45 and concluded in a
memorandum, dated August 19, 1985, that the notice was not applicable
to Oconee since Oconee is a Babcock & Wilcox (B&W)
designed plant.
It
was also stated that no seismic interactions or installation
deficiencies were found that would hinder proper functioning of the
Unit 1 incore detector system (which is similar to the flux mapping
system used in Westinghouse designed plants).
This item is considered closed.
AUG 0 7 1987
Docket Nos. 50-269, 50-270, 50-287
License Nos. DPR-38, DPR-47, DPR-55
Duke Power Company
LATTN:
Mr. H. B. Tucker, Vice President
Nuclear Production Department
422 South Church Street
Charlotte, NC
28242
Gentlemen:
SUBJECT:
INSPECTION REPORT NOS. 50-269/87-28, 50-270/87-28 AND 50-287/87-28
This refers to the Nuclear Regulatory Commission (NRC)
inspection conducted by
N. Economos on July 13-16,
1987.
The inspection included a review of
activities authorized for your Oconee facility.
At the conclusion of the
inspection, the findings were discussed with those members of your staff
identified in the enclosed inspection report.
Areas examined during the inspection are identified in the report.
Within
these areas, the inspection consisted of selective examinations of procedures
and, representative records,
interviews with personnel,
and observation of
activities in progress.
Within the scope of the inspection,
no violations or deviations were
identified.
In accordance with Section 2.790 of the NRC's "Rules of Practice," Part 2,
Title 10,
Code of Federal Regulations, a copy of this letter and its enclosure
will be placed in the NRC Public Document Room.
Should you have any questions concerning this letter, please contact us.
Sincerely,
8708170297 670807
Alan R. Herdt, Chief
ADOCK 05000269
Engineering Branch
0
Division of Reactor Safety
Enclosure:
NRC Inspection Report
cc w/encl:
LM
S. Tuckman, Station Manager
8
bcc w/encl:
Resident Inspector
W. Pastis,
State of South Ca olina
N conomos:gb
iake
Peebles
08/3 /87
8/4 /87
08/ /87