ML16161A857
| ML16161A857 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/05/1987 |
| From: | Blake J, Economos N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML16161A858 | List: |
| References | |
| 50-269-87-28, 50-270-87-28, 50-287-87-28, IEB-83-05, IEB-83-06, IEB-83-5, IEB-83-6, NUDOCS 8708170310 | |
| Download: ML16161A857 (8) | |
See also: IR 05000269/1987028
Text
C;'pgREGJ
UNITED STATES
o'
NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTA STREET, N.W.
ATLANTA, GEORGIA 30323
Report Nos.:
50-269/87-28, 50-270/87-28, and 50-287/87-28
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 Conduc d:
July
, 1987
Inspector:
N
c omo
/ia / Si ned
Approved b :
7
/
J. .
lIake, Section Chief
D'ate Signed
E gi
ering Branch
ivi ion of Reactor Safety
SUMMARY
Scope:
This routine,
unannounced inspection was conducted in the areas of
licensee action on previous open items,
including follow-up on unusual event
report(s) (LER), and response to IE Bulletin 83-05.
Results:
No violations or deviations were identified.
8708170310 870807
PDR ADOCK 05000269
0
REPORT DETAILS
1. Licensee Employees Contacted
- M. S. Tuckman, Station Manager
- T. B. Owen, Superintendent Maintenance
- R. J. Brackett, Senior Quality Assurance (QA) Engineer
K. G. Rohde, Nuclear Production Engineer
B. Stengel, Station Training Instructor
B. W. Carney, Jr., Mechanical Technical Support Engineer
W. R. Hunt, ISI Coordinator, Oconee
C. R. Henson, Technical Support Welding
G. L. Blubaugh, QA Technician, Welding/Nondestructive Testing
F. E. Owens, Regulatory Compliance Specialist
- Attended exit interview
2. Exit Interview
The inspection scope and findings were summarized on July 16, 1987, with
those persons indicated in the above paragraph. The inspector described
the areas inspected and discussed in detail the inspection findings. No
dissenting comments were received from the licensee.
The following
previously identified open items were closed at this time.
(Closed) Unresolved Item 50-269,
270, 287/87-06-01, Material Traceability
for UT Bolt Calibration Block 40363, paragraph 3.
(Closed)
IFI 50-269,
270,
270,
287/87-06-03,
Evaluation of Changing
Nondestructive Examination Requirements of Class C, B31.7 Pipe Welds,
paragraph 7.
(Closed)
Unresolved Item 287/85-02-01, Leak Rate Acceptance Criteria for
Individual Containment Isolation Valves, paragraph 3.
(Closed)
Licensee Event Report 270/87-03,
Reactor Shutdown
due to
Nonisolated Leak in Reactor Coolant System, paragraph 6.
(Closed) 83-BU-05 ASME Nuclear Code Pumps and Spare Parts Manufactured by
the Hayward Tyler Pump Company, 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 (92701)
(Closed)
Unresolved Item
(UNR)
50-269,
-270,
-287/87-06-01,
Material
Traceability for UT Bolt Calibration Block No. 40363 (73755)
2
This item was identified when the licensee could not provide objective
evidence to verify that UT Calibration Block No.
40363 was made from ASTM
A-540, Grade B-24 material.
Results of a chemical analysis performed on
the block material by Law Engineering Industrial Services showed the block
to be ASTM A-540, Grade B-23 material.
The analysis was consistent with
material specification requirements for the RCP main flange studs for
which the subject block was used as a calibration standard.
The grade
number on the calibration block drawing was changed from GR-24 to GR-23 to
make it consistent with the analysis results.
(Closed) UNR 269, 270, 287/85-02-01, Leak Rate Acceptance Criteria for Individual
Containment Isolation Valves. (92706)
This item was identified as a result of a review of the licensee's
procedure for leak testing Containment Isolation Valves (CIVs).
The
applicable
code
was
identified
(80W80),
Subsections IWV-3426
and -3427.
The procedure under discussion was
identified as PT/1/A/0150/06,
September 27,
1984 Mechanical Penetration
Leak Rate Test.
Although this procedure applies to Unit 1, it
is
essentially the same as the procedures used on the other two units.
Based on discussions and the review conducted during the inspection
documented in RH Report No. 50-287/85-02, the NRC inspector reported that
the licensee used an approach in assigning individual valve leakage limits
that appeared consistent with the intent of the NRC accepted approach.
However,
the inspector found that the individually assigned acceptance
limits were not true acceptance limits in that the procedure permitted
them to be exceeded upon evaluation.
Moreover, the inspector stated that he understood that the evaluation may
permit leakage above the individual limits so long as the total
containment leakage limits were not exceeded.
The procedure did not
require the evaluation to be documented.
At the time, licensee test
personnel informed the inspector that, in practice, the individual valve
leakage limits were only permitted to be exceeded if it was not practical
(e.g.,
due to long lead time in obtaining a valve part) to correct the
condition before returning to power and if it did not appear that further
degradation of leakage limits would occur and result in violation of
containment leakage limits.
The inspector informed the licensee that he
was concerned that:
-
The actual acceptance limits permitted through their evaluation
exceeded limits the inspector understood to be considered generally
acceptable to the NRC.
-
Evaluations of excessive, individual CIV leakage rates were not
documented.
During this inspection,
the inspector discussed the aforementioned
concerns with cognizant licensee personnel and reviewed changes in the
aforementioned
procedure that resolve the concerns described above.
3
Revisions to the procedures were completed on August 8, 1985.
4. Unresolved Items
Unresolved items were not identified during this inspection.
5. IE Bulletins (92703)
(Closed) 83-BU-05: ASME Nuclear Code Pumps and Spare Parts Manufactured by
Hayward Tyler Pump Company.
By memorandum dated September 8, 1983, the licensee responded to the
action items of this bulletin. The licensee's response stated that:
a. The only pump manufactured by the Hayward Tyler Pump Company (HTPC)
currently installed at the Oconee Nuclear Station is located in
Units 1 and 2 Spent Fuel Cooling System.
Twelve (12)
other pumps
manufactured by HTPC were earmarked for installation in the radwaste
facility which was considered to be non-safety-related and therefore,
exempt from the action items required by this bulletin.
b. The HTPC
pump,
installed in the spent fuel pool cooling system of
Oconee Units 1 and 2 is included in the Oconee Pump and Valve
inservice testing program and is tested as required by:
ASME Section XI, IWP, 1980 Edition.
At the time of installation, motor rotation and phase current were
checked and declared satisfactory.
From the time of installation to
the time of the licensee's response on September 8, 1983, the motor
had no malfunctions and no problems had been experienced with the
motor switchgear due to overload. A motor current check conducted in
response to this bulletin showed 38 amperes per phase which was
acceptable for the design load.
c. Acceptance tests run at the time of installation, July 1981, showed
pump performance to be satisfactory and the pump has continued to
operate satisfactorily since that time.
Vibration amplitude is one
of the parameters being monitored and documented as required by Code.
A 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> endurance test required by this bulletin was completed
successfully on August 12, 1983.
d. The subject pump and associated piping of the Spent Fuel Pool Cooling
System were hydrostatically tested per Code requirements prior to
system turnover with no leaks observed. The Spent Fuel Cooler area
is checked during each shift.
Inspection items include
identification of operating pump and evidence of leakage. Leaks, if
observed, are noted and necessary repairs are arranged.
e. At the time of the response, no spare parts manufactured by HTPC had
been installed in any pumps at Oconee.
The recommendation given in
4
Attachment 3 of the Bulletin, Installation of Replacement Parts, was
incorporated into the disassembly/assembly procedure of the HTPC
Spent Fuel Cooling Pump prior to September 16,
1983.
The subject
pump was identified as pump IC, Model 4x6x12,
N3; Serial Number
827401.
Oconee station procedures related to pump maintenance and
testing which were reviewed for content were as follows:
-
MP/O/A/300/32, April 28,
1986 Pump - Hayward Tyler - Spent Fuel
Coolant - Disassembly, Repair and Reassembly.
-
PT/1/A/0251/02, December 19, 1986 Spent Fuel Pool Cooling Pumps
Performance Test
-
MP/O/A/1720/10, Hydrostatic Test
NSM 1212, Part D-004
-
Printout of Work Requests on the subject pump from July 1, 1981
until the present.
(Open)
83-BU-06:
Nonconforming Materials by Tube-Line Corporation
Facilities at Long Island City,
and Carol
Stream, Illinois.
By memoranda dated November 18 and December 9, 1983 the licensee responded
to the action items requested by this Bulletin. The response stated that
the only Tube-Line (T-L) material supplied to Oconee was installed in the
Unit 3 Auxiliary Feedwater System.
The material in question involved
weld-neck flanges, end caps and reducing tees - all of which were made of
carbon steel material. Tests performed on a sample of SA-234 WPB material
produced from the same heat as the caps and tees, in Oconee 3, showed the
material properties equaled or exceeded the ASME
Code requirement.
However,
in the case of the three inch (3") weld-neck raised face
flanges, which are located on the risers that lead to the Auxiliary
Feedwater Steam Generator Nozzles, laboratory tests conducted by Babcock
and Wilcox (B&W), and T-L indicated that the material properties did not
meet B&W's purchase specification requirements.
Specifically, the yield
and ultimate strength properties were
below the
standards.
The Code of Record, the 1967 USAS B 31.1 Code,
does not
require a specific flange analysis.
However,
it does require that the
material meet ASTM and B 16.5 standards. The laboratory test results were
essentially identical to the material properties for materials which are
acceptable per ANSI B 16.5 for Oconee.
The reported test results and
Code/ANSI standard requirements are as follows:
5
Yield
Tensile
KSI at Ambient To
KSI at Ambient To
B&W Specified Material,
ASME,SA-105
36
70
USAS, B 31.1/B 16.5
Requied Material Properties
ASTM A-105Gr.1
30
60
Flange Material
31.1
59.8
Properties
On the basis of these results, the licensee concluded the material meets
B31.1 Code requirements for ASTM A-105 Gr.1 but not ASTM A-105 Gr-11/ASME
SA-105 as specified.
Results of a rigorous analysis used to evaluate the subject flanges,
caused the licensee to conclude that one of the 12 installed flanges did
not meet Code stress allowables for the flange stresses and flange bolt
stresses during operating conditions.
However,
the licensee concluded
that, on the basis of these evaluations, Unit 3 could return to power
until the spring 1984 Refueling Outage.
During that outage all the
flanges were visually inspected and checked for hardness.
All material
except one flange was found acceptable and the one that failed was
replaced.
Because the licensee could not provide stress analysis
calculations for the inspector review and a description of the general
concerns as expressed in paragraph 4 of the bulletin, i.e. short-term and
long-term corrective actions as applicable was not included in the
licensee's reponse, this item will remain open.
Within the areas inspected no violatins or deviations were identified.
6. Nonroutine Events (92700)
(Closed)
Licensee Event Report (LER) 270/87-03 Reactor Shutdown Due to
Non-Isolable Leak in the Reactor Coolant System
This work effort was conducted as a follow-up to that performed earlier by
the Oconee resident inspectors and documented in Report 50-270/87-03.
As
stated in the subject LER,
on April 6, 1987, the unit was taken off-line
and brought to 240 degrees F and 170 psig to determine the efficiency of
the Decay Heat Coolers. While at 240 degrees F and 170 psig, maintenance
personnel entered the Reactor Building to measure for a pipe support.
At
0600 on April 8, 1987,
personnel observed water coming from a welded
connection.
The licensee's investigation concluded that section of pipe
containing the leak could not be isolated. Therefore, at 0607 an unusual
event was declared and at 0708 the unit was taken to cold shutdown.
6
The root cause of this event has been attributed to a weakening of the
pipe wall from stresses induced by material frequency vibration.
The
circumstances that caused the pipe to react in this manner stem from an
error made by a pipe-fitter during fabrication.
Apparently,
the
pipefitter cut the pipe to length without taking into account the radius
of the 12" connecting pipe and thereby cut the pipe section 6" too long.
The pipe being too long caused the vibration problem.
The leak emanated from a crack in the heat affected zone of the pipe to
coupling weld of one of the RVLIS level transmitters.
The pipe was
approximately one inch in diameter with a minimum wall thickness of .219
inch (Schedule 160).
The crack propagated circumferentially about 180
degrees around the pipe.
The licensee's records showed that the repair was performed in accordance
with construction Code B31.7 and ASME Section XI requirements.
The
inspector reviewed the weld process control sheet and other related QA/QC
records for completeness and accuracy.
Other quality records reviewed
included welder qualifications, filler metal certifications and
records.
Other corrective actions planned and subsequently implemented were as
follows:
-
Develop training for craft personnel,
management,
and all other
personnel
who install modifications to interpret the correct
dimensions of piping installation from isometric drawings.
-
Inspect all modifications that were installed with isometric drawings
that have the possibility of similar consequences.
-
Perform safety analyses and initiate changes to modifications where
appropriate.
Records on file show these actions have been implemented.
7. Inspector Follow-up Items
(Closed) IFI 259, 260, 287/87-06-03 Evaluation of Changing Non-destructive
Examination Requirements of Class C, B31.7 Pipe Welds (55050)
This item was identified when the inspector noted that Variation Notice
VN-0587 had been issued to document a switch from B31.7 to ASME
Section III non-destructive examination requirements on Duke class "C"
By Revision 13 to Piping Installation Specification OS-0243, 00-00-0001,
the licensee has adopted the requirements of ASME Section XI repairs and
replacements, which meet the following Section III requirements and are
exempted from certain NDE requirements as described below (Ref.Section XI
7
a)
Duke Class C (B31.7 Class III) welds, for which both the welds and
welders meet the examination and qualification requirements of ASME
Section III 1974 edition, Summer 1975 Addendum, are not required to
meet the random radiograph requirements of B31.7 chapter 3-VI.
b)
Duke Class C welded repairs of defects in materials which meet the
exemption and examination requirements of ASME Section III subarticle
ND-4130 of the 1974 edition, winter 1976 addendum,
are not required
to be radiographed.
This revision became effective as of July 1, 1987.