ML20244C800
| ML20244C800 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 06/02/1989 |
| From: | Singh A, Stetka T, Wagner P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20244C796 | List: |
| References | |
| 50-298-89-19, IEB-85-003, IEB-85-3, IEIN-87-004, IEIN-87-4, NUDOCS 8906150143 | |
| Download: ML20244C800 (23) | |
See also: IR 05000298/1989019
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APPENDIX 8
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-298/89-19
Operating License: DPR-46'
Docket: 50-298
Licensee:
Nebraska Public Power District (NPPD)
P.O.' Box 499
-Columbus. NE . 68602-0499
Facility Name: Cooper Nuclear Station (CNS)
cInspection At: CNS. Brownsville, Nebraska
Inspection Conducted: May 1-5 and 15-19, 1989
Inspectors:
W Me
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P. C. Wagner, Reactor Inspector, Plant Systems
Date
Section.-Division of Reactor Safety
\\i b L d
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A. S'ingh, Rfactor Inspect (r. Plant Systems
Datb
Section, Division of Reactor Safety-
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Approved:
T. F. Stetka, Chief, Plant Systems Section
Date
Division of Reactor Safety
Inspection Summary
Inspection Conducted May 1-5 and 15-19,'1989 (Report 50-298/89-19)
Areas Inspected: Routine, unannounced inspection of the licensee's actions in
response to NRC requirements for motor operated valve (MOV) testing and the -
licensee's programs for instrument calibration and piping supports. The
inspection included gathering survey information related to drywell
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temperatures and diesel fuel oil storage and handling, and discussions of the
implementation of Regulatory Guide 1.97 instrumentation. The NRC inspectors
also reviewed the corrective actions related to the Type SJO electrical cable
problems and the actions completed in response to previous NRC inspection-
findings.
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Results: Within the scope of the inspection, one violation of_NRC requirements
was identified (paragraph 3.b). The violation involved two examples where
component testing was performed utilizing instrumentation which had not been
calibrated. The NRC inspectors found the MOV testing program to'be good but
limited in scope in that only' those valves covered by IE Bulletin' 85-03 were -
being tested. .The NRC inspectors also considered the basis for the repair of
the SJO cable to be less substantial than would have been desirable.
The NRC inspectors determined that the calibration program'for instruments
listed in the Technical Specifications to be a good program.
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DETAILS
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Persons Contacted
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++L. Bray, Regulatory Compliance Specialist
+ R. Brungardt, Operations Manager
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+ B. Fehrman, Project Manager
- J. Flaherty, Plant Engineering Supervisor
+*S. Freborg, Assistant Plant Engineering Supervisor-
+*R. Gardner, Maintenance Manager
B._ Jansky, Outage and Modification Manager
- J. Hall', Mechanical Supervisor
- +*G. Horn, Plant Manager, Nuclear Operations
+ H. Jantzen. Instrument and Controls (I&C) Supervisor
+ L. Kohles, Nuclear Project and Construction Manager
+*E. Mace, Engineering Manager
+*J. Meacham, Senior Manager of Operations
- D. Robinson, Quality Assurance (QA) Supervisor ^
NPPD - Columbus Office (via telephone)
4 J. Branch, Supervisor, Engineering
+ A. Heymer, Manager, Configuration Management
+ S. McClure Manager, Nuclear Engineering
+ H. Parris, Vice President Production
+ G. Smith, Licensing Supervisor
+ G. Trevors Division Manager, Nuclear Support
+ K. Walden, Licensing Manager
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+ R. Wilber, Division Manager, Nuclear Engineering and Construction
+ V. Wolstenholm, Division Manager. QA
NRC, Region IV
+ J. Jaudon, Deputy Division Director, Division of Reactor. Safety.
- L. Ellershew, Reactor Inspector
- W. McNeill, Reactor Inspector
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+ L. Gilbert, Reactor Inspector
- G. Pick, Resident Inspector
+ W. Bennett, Senior Resident Inspector
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- Denotes personnel present at the May 5, 1989, exit meeting.
+ Denotes personnel present at the May 19, 1989, exit meeting.
The NRC inspectors also contacted and interviewed other NPPD personnel
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during the performance of this inspection.
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2.
Motor Operated Valves (25573)
Because of problems identified with the torque switch and limit switch
settings of important motor operated valve (MOV) actuators, the NRC issued
IE Bulletin (IEB) 85-03. The.IEB was provided to ensure that switches
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were set and maintained so as to accomodate the most severe loadings
expected during a design basis event. NRC inspection guidance was
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provided in Temporary Instruction (TI) 2515/73.
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NRC inspection of the licensee's response to IEB 85-03 was initiated in
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~ November 1987 and documented in NRC Inspection Report 50-298/87-30. The
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IEB remained open pending completion of the MOV testing. The MOV testing
at CNS was performed utilizing the "M0 VATS" system and equipment.
Subsequent to the initial NRC inspection, NPPD provided additional'
information in response to IEB 85-03 in letters dated June 2 and
September 20, 1988. The NRC requested some clarification to the NPPD
submittals by letter dated November 9,1988, which NPPD provided in their
January 9, 1989, respcnse.
The NRC technical review of the licensee's response is contained in-
subparagraph a. below. The NRC inspection of the remaining requirements of
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TI 2515/73 are contained in subparagraphs b. through d. below. 'A
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discussion of NPPD review results related to NRC Information Notices is
contained in subparagraph e. below,
a.
Technical Review of IEB 85-03 Submittals,
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Review of the licensee's January 9,1989, response and their letter
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of June 2,1988, in response to Supplement 1 of IEB 85-03, indicates
that their selection of the applicable safety-related valves to be
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addressed and the valves' maximum differential pressures meet the
requirements of the IEB and that the program to assure valve
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operability requested by action item e. of the 'IEB and its supplement
is now acceptable.
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b.
Procedure Review
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The NRC inspector reviewed the Maintenance Procedures (MP) related; to
MOVs listed in Attachment 1 to this report. The NRC inspector found
all of the procedures to contain easily understood, detailed,
step-by-step instructions. The procedures also included frequent
requirements for independent verification of actions. There were a.
total of 96 MOVs listed in MP-7.3.36, but only 19 of these MOVs were
included in the IEB 85-03 testing program. The 19 MOVs included in
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the IEB 85-03 program consisted of.10 Reactor Core Isolation
Cooling (RCIC) system and 9 High Pressure Coolant Injection (HPCI)
system valves; all of the valves except 1 RCIC and 1 HPCI were DC
motor operated. During discussions with licensee personnel about the
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M0 VATS program, the NRC inspector was informed that the remaining DC
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MOVs will be MOVATS tested in the fall of 1989 (see paragraph e.
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below).
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,T_rai ni ng
Since paragraph 6.5 of MP-7.3.~a5-1 required qualified personnel to
perform or supervise the MOVATS testing, the NRC inspector reviewed
the training program. lesson plans listed in Attachment 1.
The three
M0V lessons, along with their related 1sberatory exercises, were
prerequisites for the M0 VATS training course. The NPC inspector
found all of the lesson plans to contain sufficient detail for the
functions to Le performed.
The NRC inspector also noted that EQP-011-02-03 included maintenance
update information from Limitorque Corporation (the MOV vendor) and
information on DC motor cable sizing (see paragraph e. below).
d.
MOV Testing
NRC Inspection Report 50-298/87-30 stated that 6 of the 19 IEB 85-03
listed MOVs would be required to be tested under pressure. This
statement was based on an August 3.1987, letter to NPFD f rom M0 VATS
Incorporated; subsequently, NPPD provided MOVATS additional
information, and by letter dated March 22, 1988 MOVATS rescinded
their earlier recommendation on the need to test the valves at
pressure.
The NRC inspector reviewed the records for the MOV testing completed
in April and May 1968, on 10 of the 19 MOVs listed in NP?D's
January 9, 1989, submittal to ensure that prcoer testing had been
performed. While the HRC inspector found the testing records to be
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properly completed with acceptable results, he noted that two valves
had new actuators installed prior to the tests and that five other
valves had the actuator spring packs replaced.
In eddition, all of
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the MOV actuators were cleaned and lubricated prior to testing.
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Therefore, the NRC inspector questioned how the operability prior to
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testing designated in the NPPD letter had been determined.
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personnel stated that the initial operability determination was based
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on the testing performed in 1986; the questions arising from any
abnormal indications during those tests and the results of
evaluations performed subsequent to those tests led to the
replacements noted in the 1988 test records.
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activities were performed to enhance the MOV operations not to
provide operability.
The NRC inspector also questioned how the MOVs had been relubricated,
specifically concerning the possibility of mixing the old and new
lubricants. The licensee personnel stated that the lubrication had
properly consisted of cleaning all of the old grease from the
mechanism and then applying fresh, approved grease.
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e.
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The NRC issued Information Notices (ins) 88-72, " Inadequacies.in the
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Design of DC Motor Operated Valves;" and 89-11. " Failure of DC Motor-
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Operated Valves to Develop Rated Torque Because'of Improper Cable
Sizing," to alert licensees of potential problems. The NRC inspector
discussed the. status of NPPD's efforts to evaluate the conditions
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presented in the ins with the responsible corporate engineer. The
NRC inspector was informed that NPPD had evaluated the active valves
included in the IEB 85-03 listing (i.e., those valves required to
reposition during an accident) and had found them to be acceptable.
The 125VDC valves were evaluated in calculations NED 89-131C and D;
the 250VDC valves in NED 89-131A and B.. The NRC inspector briefly
reviewed those preliminary calculations and found 'them to contain
conservative assumptions. The calculations concluded that adequate
current would be provideo'to the actuator motors to produce the
required torque for valve operation. The torque values were stated .
to have been calculated in accordance with vendor methodology and the
current values were calculated using the presently installed motor
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feeder cables. NPPD engineering personnel stated that some DC motor
starting _ resistors had been previously removed based on a Limitorque
maintenance recommendation. The NPPD engineering. personnel also
stated that the remaining DC motor MOVs would be evaluated after the
in progress refueling outage, and that those evaluations were expected
to be completed in the fall of 1989.
The NRC inspector found the MOV testing program that had been implemented
at the CNS in response to IEB 85-03 to be a good program. The NRC
inspector, however, recommended that NPPD consider expanding the scope of
that program to include additional safety-related valves.
No violations or deviations were identified.
3.
Cal.brations
(56700)
In order to ascertain if the licensee had implemented a program for the
calibration of installed plant instrumentation that was in accordance with
regulatory requirements and accepted industry guidance, the NRC inspector
reviewed selected plant documents. The documents reviewed included
calibration, surveillance and testing procedures, and the. records of the
completion of those procedures. A partial listing of the documents
reviewed is contained in Attachment I to this report.
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a.
Technical Specification Instruments
The NRC inspector reviewed a number of Calibration and Function Test
procedures for Technical Specification (TS) listed instruments and
found them to be acceptable. The majority of the procedures had been
revised within the preceding year into a new, more easily followed,
format. The NRC inspector noted attributes such as: readability;
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easily understood, step-by-step instructions; clearly stated caution
statements, limits and tolerances; and requirements for independent
verification for return to service and for documenting any noted out
of tolerance conditions.
The NRC inspector also noted that the procedures required a
management review of the post-test / calibration data. The NRC
inspector questioned how trending of data was accomplished and was
informed that the systems engineers informally tracked and trended
test results.
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The NRC inspector also noted that most of the TS listed instruments
were required to be functionally tested once per month and calibrated
once per 3 months.
Further, the functional test procedures usually
performed a calibration check.(e.g., a pressure source was connected
to the pressure transmitter and varied to ensure proper response).
The NRC inspector found both the calibration frequency and the
functional test complexity.to be more than usually experienced.
The NRC inspector reviewed the 1988 and 1989 test and calibration
records for those instrument procedures that had been reviewed. The
records were all complete, containing such attributes as proper
sign-offs for completed evolutions and postimplementation
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verifications and' reviews, documentation of the test equipment
utilized, and the initiation of nonconformance reports for out of -
tolerance conditions when discovered. The NRC inspector questioned
how the three times normal setpoint for the Main Steam Radiation
Monitor (Surveillance Procedure (SP) 6.1.4) was determined and was
directed to SP 9.4.3.
Review of SP 9.4.3 indicated that the detector
was calibrated by exposure to a traceable radiation source and that
the normal,100 percent power, background radiation level, in the
area of the main steam line monitor, was determined prior to return
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to service from each refueling outage by calculating an average from
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the previous cycle data.
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The NRC inspector also questioned how the HPCI pump low discharge
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flow setpoint of 2.4 inches water gauge (wg) contained in
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SP 6.2.2.3.13 was determined. A review of the scaling factors showed
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that the TS setpoint of 400 gpm was calculated to represent a
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1.92-inch wg differential pressure at the flow transmitter based on
the flow element curves. The procedure setpoint of 2.4-inch wg
provided an acceptable margin to ensure flows greater than the IS
required minimum.
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Based on the sample reviewed, the NRC inspector found the calibration
program for those instruments listed in the TS to be a good program.
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The NRC inspector had no further questions on the program and no
violations or deviations were identified,
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b.
Nontechnical Specification Instruments
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In order to ensure that the licensee was implementing an acceptable
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calibration program for those instruments which were utilized in
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safety-related functicas but were not specifically identified in the
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TS..the NRC inspector reviewed operating.and surveillance procedures
and selected a sample.of instruments utilized therein to detemine
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proper equipment operability.
(1) Emergency Diesel Generator (EDG) Operability
The NRC inspector reviewed the EDG Operability. Test Procedure
(SP 6.3.12.1) and selected five parameters listed on the data
sheets for review of the calibration status of the instruments
utilized during the testing of No. 2 EDG. The review of the
calibration records disclosed the following:
Jacket Water Temperature - Temperature Indicator DG-JW-3145
was calibrated on a 4-year interval in accordance with
Preventative Maintenance Procedure (PM) 01470.
Lube Oil Filter Inlet and Outlet - Pressure Gauges PI-3143
and -3145 were not on a scheduled calibration program,
but had been calibrated in November 1986 as part of another
work item.
Voltage, Frequency, and Current - Local electrical meters
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were not on a scheduled calibration program and had not had
their calibration checked since initial startup in 1973.
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The NRC inspector reviewed applicable procedures and records and
verified that the protective functions (e.g., low lube oil
pressure, overcurrent, differential overcurrent, etc.) were
calibrated on an acceptable frequency. However, the NRC
inspector also reviewed the procedure for shutdown outside the
control room (Emergency Procedure 5.2.1) and noted that the
local meters and gauges discussed above were utilized to operate
and protect the EDGs during the performance of that procedure.
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Violation (298/8919-01): The failure to have a procedure which
ensured the proper, routine calibration of the local instrumentation
utilized to verify the proper operation of the EDGs is an
apparent violation of TS 6.3.3.D.
(2) Core Spray (CS) System Operability
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The NRC inspector selected another system to evaluate if the.
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instruments utilized during the surveillance test were being
calibrated. The CS Operability Test (6.3.4.1) contained data
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sheets for the inservice test of CS pumps A and B which
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specified the instruments to be utilized during the test. The
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NRC inspector selected pump discharge flow, inlet pressure,
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outlet pressure, and motor current as the parameters-important
to operation.
Review of calibration records disclosed the
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Discharge Flow - FI-50B was calibrated on a quarterly
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Pump Pressures - PI-36B and -48B were calibrated on'an
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annual frequency.
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Motor Current - CSP 1B Ammeter was not included in a
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scheduled calibration program.
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The NRC inspector reviewed the calibration records for the pump
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discharge flow and pressures and found them to be acceptable;;
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however, the motor ammeters had not had calibration verified
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since the initial startup testing in 1973.
The failure to have a procedure to calibrate routinely
instrumentation used during the performance of surveillance
testing of the CS system is an additional example of the
apparent violation described in paragraph 3.b(1) above.
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4.
Electrical Cable Repairs
(92701)
The NRC insper, tor questioried'if the preliminary notifications
(PNO-ADSP-89-01) issued on March 23, 1989, for a problem identified at--the
Browns Ferry Nuclear Facility was applicable to the CNS. The problem
involved the deterioration of the Buna-S rubber insulatior. on the Type SJO
electrical cable provided by General Electric Company (GE). . The conductor
insulation had become hard and brittle and had developed cracks. .NPPD
personnel stated that they had been informed of the problem by both GE and
the Institute of Nuclear Power Operations (INPO) and had identified
approximately 89 installations of the SJO ceble at.the CNS..
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The SJO cables at the CNS were either 2 or 3 conductor, 18 AWG wire size,
and usually provided as interconnecting wire. The cable had the
appearance of normal appliance cord, similar to what would be used on an
electric drill. The problems with the cable at the'CNS included instances .
in which the insulation was completely missing from both conductors. The
exposed portions of the conductor insulation had been brittle and in most
cases had developed radial cracks; however, the portions of the insulation
that had remained covered by the cable's PVC outer jacket material had not
become brittle and remained flexible. NPPD had samples of the removed
cable analyzed by GE and were verbally informed that the unexposed
portions of the Buna-S insulation remained acceptable for continued used.
NPPD evaluated the cable problem and determined that continued use of the
SJO cable pending final corrective action recommendations by GE was ,
acceptable provided the cables were reterminated. The determination
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process included the removal'ofLpreviously exposed portions of Buna-S'
insulation -stripping the cable' jacket back to allow relugging of the
freshly exposed conductors, then reterminating. ,The NRC inspector was
informed that=these interim actions were in accordance with GE
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recommendations. . NPPD also referenced the 3-year expected shelf-storage.. -
life for'SJO cable ~ listed in Military Standardization Handbook for Rubber
Products (MIL-HDBK-695C, 27 March 1985) 'as additional bases for~the
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continued'use of reterminated cables;
Inspector Followup Item-(298/8919-02): While the NRC 'nspector dId not
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determine that continued interim use of SJO ' cable would. be detrimental to
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safe plant operations, followup inspection of the long-term corrective-
actions will.be performed to ensure that the SJO cables are replaced prior
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to exceeding their expected service lifetime.
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Postaccident Monitoring Instrumentation -(25587)
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By letter dated December 17,1982,(Generic' Letter 82-33)theNRCprovided
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all reactor licensees and applicants with the " Requirements for Emergency
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Response Capability."
Included in these requirements was.the application
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of. RG 1.97 " Instrumentation for Light Water Cooled Nuclear Power Plants -
to Assess Plant and Environs Conditions During and Following an Accident."
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The provisions for the instrumentation described in RG I.97 were endorsed
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by the NRC to ensure that nuclear power plant operators w9uld havel
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sufficient and reliable information available for preventing and/or
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mitigating the consequences of a reactor accident. The NRC inspector
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initiated the inspection of this instrumentation in accordance with
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The NRC inspector reviewed the NRC Safety Evaluation' dated October 27.
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1986, and noted that some of the instrumentation being utilized at the CNS
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differed from that described in NPPD's December.4, 1985, submittal. NPPD
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engineering personnel discussed the changes and indicated that the
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installed instrumentation was an enhancement to that originally proposed. .
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NPPD personnel had consnitted' to have the RG 1.97 instrumentation operable
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prior to restart from the in-progress refueling outage and stated that the
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commitment would be met.
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The NRC inspector made'a brief tour of the control room to verify that all
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of the RG 1.97, Category I, instruments were installed, including required
redundancy and recorders. The NRC inspector found that the licensee had'
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installed appropriate instrumentation; however, additional inspection
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activity will be required to ensure that the implementation of RG 1.97 at.
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the CNS meets the NRC acceptance criteria. -NPPD personnel agreed to
update their December 4,1985, response to indicate clearly which
instruments were being utilized to fulfill the CNS commitments. The
updated. response will be submitted by August 21, 1989.
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No violations or deviations were identified.
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Information on Drywell Temperatures '(25598)'
Because of occurrences of higher than anticipated temperatures inside
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containments (PWRs) and drywells'.(BWRs), the NRC issued TI 2515/98 to
obtain plant specific data. The average temperature profiles requested in
Exhibit I to TI 2515/98 were to be'used to determine if the high-
temperatures previously identified were plant specific problems or.
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indicative of a'more generic problem.
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The NRC inspector compiled the data requested by the' TI and attached a
copy of the completed Exhibit I as Attachment 2 to this report.
No violations or deviations were identified.
7.
Information on Diesel Fuel Oil
(255100)
In order to verify that the licensee had implemented a program to maintain
adequate quality of the fuel oil for the EDGs, the NRC inspector
implemented TI 2515/100. This TI also requested information in the form
of appended questions. The survey was completed and is included as
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Attachment 3.
The NRC inspector found the CNS programs to be acceptable but' questioned
the normal alignment of running both portions of the duplex fuel filter in
parallel instead of one side or the other. Licensee personnel stated that
no fuel oil problems had ever been encountered at the CNS.
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No violations or deviations were identified.
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8.
Testing of Piping Support and Restraint' Systems (70370)
The purpose of this inspection was to ascertain whether the licensee had
established an adequate program and procedures pertaining to the
examination and testing,of piping and restraint systems;"
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e.
Procedure Review
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The NRC inspector reviewed the CNS second 10-year Inservice
Inspection (ISI) Plan submitted to, and approved by, the NRC on
January- 27, 1986. The NRC inspector also reviewed the implementing
procedures for the ISI plan listed below:
Procedure No.
Title
Date
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6.3.10.9.1
Surveillance Procedure, Snubber
02/28/89
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Revision 15
Operability
7.2.34.1
Snubber Inspection
'02/09/89
Revision 1
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Pipe Snubbers Removal and
01/17/89
Revision 1
Installation
IV3-W812
Visual Examination VT-3
02/19/88
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Revision 3
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IV4-W812
Visual Examination VT-4
,.02/18/87
Revision 2
ED-88-C4
InserviceInspection(ISI) Activities
03/28/89
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Revision 1
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'7.2.57
Maintenance Procedure for ASME
08/30/88
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Revision 1
Category F-A, F-B, and F-C Components
Supports, Inspection, and Adjustment
7.2.34.7
Grinnell Figure 200/201 Hydraulic
03/15/88
Revision 0
Snubber Functional ~ Test
7.2.34,8
Pacific Scientific Snubber Functional
03/15/88
Revision 0
Test
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60-1
Receipt, Review, and Recording
01/20/89
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Revision 0
Calculations
60-2
Preconstruction Review of Proposed
01/20/89
Revision 0'
Pipe Support Modifications
60-3
Large Bore Piping and Pipe Support
01/20/89
Revision 0
Design Criteria
60-4
Piping Analysis Procedure
01/20/89
Revision 0
Procedures reviewed by the NRC inspector appeared to be adequate.
b.
Field Observations of Pipe Supports
The NRC inspector examined approximately 50 snubbers and pipe
supports of various types and on various systems. Attributes
selected for visual examination were:
deterioration, corrosion, physical damage, or deformations were
not evident;
all required bolts, locking devices, nuts, and washers were
installed;
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extension rods, support plates, and connecting joints were not
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deformed, or loose;
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snubber, settings;
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pipes supports, or other associated equipment or components-
were not restricted or in contact with other surfaces 1 as a-
. result.of thermal expansion;
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springs in hangers were not obstructed by foreign' material; .
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indicators or spring hangers show either " cold"'or " hot"
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position, consistent with plant condition;-and-
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threaded connections'were secured by locknuts, fasteners, and
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Snubbers and supports examined are listed below:.
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MS-SIB
MS-S16A
RH-59
RH-SIO
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SW-H23A through H23H
RH-552
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RH-536
RH-537
MS-513A
MS-S13B
MS-S17-
Supports
RH-H47A
RF-HSS.
MS-H101A
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RCC-S86
SW-S29A
RCC-S67-
SW-H149
RH-H56B
RH-H41A'
BS-H92
RCC-S61
RH-H74
RF-H42A-
RH-H17A
RH-H75
HP-H34
RH-H61A
RCC-S69
SW-H137
MS-H104-
RCC-H65
CRD-BB5
RH-H96A
- RCC-H69~
MS-H113
CS-H19A
RH-H50
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RCC-H137-
SW-H184
,
RCC-H138
SW-H181
RCC-H139
RH-H71
Overall', the pipe supports examined a'ppeared acceptable. The NRC
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inspector also noted, during the walkdown, that several modifications
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c.
Inservice Tests for Snubbers
The NRC inspector witnessed functional testing on the snubbers.
Observations by the NRC inspector indicated the following:
Personnel performing the testing were qualified.
Proper instructions and procedures were followed.
The functional ~ test machine and accessories were calibrated as
required.
As-found drag force, activation / acceleration and as-left drag
force were within acceptable limits.
The licensee's testing program was being conducted by NPPD personnel
and was found to be adequate.
The licensee has established an adequate program and implementing
procedures for the examination and testing of pipe supports,
d.
Operability Evaluation of Essential Piping Systems for CNS
During this inspection, the-NRC inspector also reviewed the
licensee's long-term plan for achieving full code qualification of
all essential large bore pipe supports. The licensee committed to
this plan in a letter to the NRC dated August 12, 1988. The licensee
submitted a contractor prepared report (CYGNA Report 88037A) which
contained a detailed evaluation of 122 supports. This report showed
that those Class IN pipe supports met the design criteria for full
code qualification. However, on May 12, 1989, the licensee reported
that they had been informed by their contractor ~ (CYGNA) that they had
identified two pipe supports included in the above report which
required modifications to satisfy the design criteria. Based upon
this report, the NRC inspector stated his concern about the code
qualification of the remaining supports. As the result of this
concern, the licensee committed to rereview all the remaining
calculations to ensure that they meet the design criteria and, if
required, to implement modifications prior to plant startup.
No violations or deviations were identified.
9.
Action on Previous Inspection Findinos
(Closed) Open Item (298/8627-01):
Completion of design change
package (DCP) reviews for final closecut. An NRC inspector reviewed six
DCPs and noted that they had not been closed even though the physical
modification effort had been completed. The NRC' inspector reviewed the
completion records of the previously reviewed DCPs and found them to be
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acceptable. However, the completion of these DCPs appeared to be
protracted (e.g., the work on DCP 83-023.was finished in Apr11'1986, but-
the DCP was not closed until April 1989). Since the DCPs' referenced in.
the earlier NRC inspection report had been properly. closed, this open-item.
is closed.
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(Closed) Violation (298/8817-01): The use:of " green tags".during the
performance of an integrated. leak rate test (ILRT) was: not" properly -
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implemented (green tags.were used to designate proper alignments for
testingconditions).
In response'to this NRC' finding, NPPD committed'to
. provide, training on the use of the green equipment tags. .The NRC
inspector reviewed the General Employee Industry Safety.Trainir.g Lesson
Plan (Revision 2) and noted the inclusion of a discussion on the reasons.
for, and proper'use of, the_. green tags. The NRC inspector also verified,
through a review of training plans, that the required training had been-
conducted..
10. Exit Meetings (30703)
The NRC inspectors summarized the scope and-findings of the inspection
during exit meetings conducted on May 5 and 19,1989, with the personnel
identified in paragraph 1 above. The licensee acknowledged the NRC
inspectors' findings and agreed to provide the submittals discussed in
paragraphs 5 and 8, of this report. The licensee did not identify as
proprietary any of the material provided to, or reviewed-by,' the NRC;
inspectors during this inspection.
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ATTACHMENT 1
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LIST OF DOCUMENTS REVIEWED
M_otor Operated Valves Procedures-
7.3.35.1, " Testing of Motor Operated Valves Using Motor Operated Valve
Analysis and Testing System (M0 VATS)," Revision 1
7.3.35.2, " Periodic Monitoring of Motor Operated Valves Using M0 VATS Motor
Load Unit," Revision 0
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7.3.35.3, " Periodic Monitoring of Motor Operated Valves Using M0 VATS Motor
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Torque Unit," Revision 0
7.3.36, " Limit and Torque Switch Checkout and Adjustment for Rising Stem
Limitorque MOVs," Revision 3
M0 VATS Training Lesson Plans
Troubleshooting and Repair of MOVs:
EQP 011-02-01 and related laboratory - 051
"
EQP 011-02-02 and related laboratory - 052
EQP 011-02-03 and related laboratory - 053
MOVATS Data Acquisition:
EQP 011-05-01
Calibration and Function Test Procedures
6.1.4, " Main Steam Line Radiation Monitor," Revision 40
6.1.5, "High Reactor Pressure Transmitter," Revision 16
i 9, "Lw Reactor Vessel Transmitter," Revision 23
6.1.4.1, " Main Steam Line Process. Radiation Monitor," Revision 2
6.2.2.3.4, " Emergency CST Level Transmitter," Revision 19
6.2.2.3.13. "SPCI Pump Low Flow Transmitter," Revision 17
7.3.6.2, " Diesel Generator Annual Electrical Inspection," Revision 1
7.3.1.2 " Timed and Instantaneous Overcurrent Relay Testing and
Calibration," Revision 3
7.3.1.3, " Differential Current Reley Testing and Calibration," Revision 2
9.4.3 " Main Steam Line Process Radiation Monitor," Revision 6
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System / Component Surveillance Procedures
6.3.4.1, " Core: Spray System.0perability Test,"iRevision 25
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6.3.4.3, " Sequential Loading of Emergency Diesel Generators," Revision 26:
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6.3.12.1, " Diesel Generator Operability: Test," Revision 27
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6.3.12.3, " Diesel Fuel _ Oil Quality Test," Revision 10
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7.2.53.1, " Diesel Generator Engine Mechanical Inspection," . Revision ~ 0
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14.17.1, "DG-1 Annual. Inspection," Revision 0
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Operating / Emergency Procedures
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5.2.1,"ShutdownfromOutsidetheControlRoom," Revision 718(
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ATTACHMENT 2
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Tl 2516/98 - Exhibit 1
1.
Plant Name: Cooper Nuclear Station (CNS)
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2.
Unit and Docket Number:
50-298
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3.
What are the average containment /drywell (C/D) temperatures during power
operation as recorded by the licensee? Note: We are interested in the
peak operating temperatures during the hottest summer months.
Average weighted temperatures from July 18 through September 27, 1988,
ranged from approximated 130'F to 148'F. These weighted averages were
calculated in accordance with the Daily Surveillance Log
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(Procedure 6.2.4.1).
A discussion of deywell temperatures is included in
3
paragraph 3.1.1 of NRC Inspection Report 50-298/88-200.
4.
Containment temperature at which the plant is licensed to operate (i.e..
operating temperature specified in.the FSAR).
The CNS USAR Section V.2.3.2 states that the drywell is designed for an
internal pressure of 56 psig coincident with a temperature of 281*F.
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5.
Review the temperature readings and provide your assessment n to whether
or not you believe the average temperature readings accurately reflect
y
containment /drywell conditions, or if there is a significant difference,
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due to temperature sensor location or stratification of containment
atmosphere which could produce hot spots.
The temperature readings are a weighted average of the temp'eratures at
five points within the drywell; in addition, seven other drywell
temperature points are recorded.
Temperature sensing locations are
denoted on Figure 1 attached.
6.
What temperature (s) is used by the licensee in its equipment environmental
qualification program when calculating the remaining qualified lifetime
for all equipment inside C/D, and are these temperatures consistent with
temperatures being experienced?
A temperature of 150*F was used in the equipment qualification program. A
review of temperature data showed that some local areas exceed the 150 F
by up to 25 F during the period discussed in 3 above.
7.
Administrative temperature limit for the containment /drywell, if no-
technical specification limit exists.
Although the temperature instruments are listed in the Technical
Specifications, no temperature restrictions are included. An
administrative limit on average temperature is 150*F.
8.
Recent history of temperatures inside containment.
Provide C/D average
air temperature in addition to the containment air temperatures used to
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compute the averaga C/D temperatures for the months of April, May, June,
July, August, and September 1987, if the plant has not operated during
those months, uro an operating period close to these months.
Temperature Element
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1987 : 505A
505B
505C
5050
505E
Apr :
132
129
137
130
143
May :
142
140
147
141
153
Jun :
141'
139
148
142
161
Jul
148
146
155
148
156
Aug :
154
154
163
155
164
Sep :
137
136
146
140
145
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AIR RING HEADER
DRYWELL VENTILATION RECIRCULATION
FIGURE 1
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ATTACHMENT 3
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i
SURVEY OF LICENSEE'S RESULTS TO
SELECTED EDG F0 ISSUES
PLANT NAME AND UNIT:
Cooper Nuclear Station
1.
Has the licensee adequately reviewed and evaluated IE Information Notice 87-04 issued on January 16, 1987, as a result of the AND Unit 2 EDG
F0 starvation event which occurred on June 27, 19867
Review of the NPPD engineering response dated March 6, 1987, and the
Operating Experience Review Transmittal which was closed on March 31,
1987, indicated on adequate review and evaluation.
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2.
Does the licensee have a permanent F0 storage tank recirculation system
which allows for complete F0 inventory cleaning by filtering each
refueling outage to remove accumulated particulate?
No.
3.
Are all F0 storage tanks being cleaned and inspected at a minimum of
10-year intervals in accordance with of Regulatory Guide 1.137?
Yes, in accordance with Section VIII.C.2 of Reference 2.
4.
Does the licensee's F0 program include a regular analysis of F0 samples
and bottom testing for accumulated water, at the lowest point in the F0
day tanks and F0 storage tanks?
Yes, monthly and after use in accordance with Reference 2.
5.
Is a fuel additive being used as a fuel stabilizer which will function to
prevent oxidation and bacterial growth?
Yes, a stabilizer (Power Service Diesel Fuel Supplement) is added in
accordance with Reference 1.
6.
Does the licersee effectively ensure that periodic F0 bottom sampling and
analysis are heing performed to detect high particulate concentrations in
the F0 supply which occurs over long-term storage due to the effects of
oxidaticn, and biological contamination in accordance with ASTM D270-19757
Yes, every 6 months in accordance with Section VIII.C.1 of Reference 2.
7.
Are day tanks and integral tanks being checked for water monthly, as a
minimum, and after each operation of the diesel where the period of
operation was I hour or longer?
Yes, in accordance with Section V.B.5 of Reference 2.
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8.
Is accumulated water removed immediately if it is determined that water is
present in the storage, integral or day tanks?
Yes, in accordance with Section VIII.A.3 of Reference 2.
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9.
Is the licensee replacing F0 in a short period of time (about a week) if
it is determined that the F0 does not meet the applicable specifications?
Yes, in accordance with Section V.B.2 of Reference 2.
10. Are F0 components which may be prone to fouling being routinely monitored
for indications of~ fouling?
Fuel injector nozzles are tested and the fuel filters are replaced during
the performance of Reference 3.
11. Are F0 filters and strainers being cleaned and inspected on a periodic
basis per the vendor recommendations?
Fuel filters are replaced annually (see No.10 above) and fuel strainers
are inspected each cycle in accordance with PM 04779.
12. Does the F0 system utilize dual element filters and strainers which
permits on line cleaning of the elements, in the event of fouling, to
allow continuous operation of the EDG7
There is a single strainer in the suction to each pump)with a duplex
filter in the common discharge line (mounted on engine .
13.
Is there a differential pressure indicator for each duplex filter / strainer
for indication of fouling in accordance with ANSI N195-19767
No.
14. Are F0 alams annunciated in the main control room or incorporated into a
general control room trouble alarm with local individual alams, in
accordance with ANSI N195-1976?
Fuel oil alams are annunciated in the control room and locally.
15. Are any of the instruments that perform a control function and provide an
alarm seismically qualified in accordance with the IEEE Recommended
Practices for Seismic Qualification of Class IE Equipment for Nuclear
Power Generating Station, IEEE-344-1975?
The criginal design did not include seismic qualification; however,
replacement parts are being qualified.
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References:
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OP 2.1.16 Revision 11; " Heating Boiler andiDiesel Fuel Oil Unloading" ,
SP 6.3.12.3. Revision 10, " Diesel Fuel Oil-Quality Test" .. l. Inspection"-
2.
SP 7.2.53.1, Revision 0, " Diesel. Generator Engine Mechanica
3.
4.
Drawing 2011, Sheet 1,;" Diesel Oil System"' .
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5.
Drawing 2077, " Flow Diagram - Diesel Generator Auxiliary Systems"
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