IR 05000483/1989015
| ML19325C824 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 09/15/1989 |
| From: | Defayette R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML19325C822 | List: |
| References | |
| 50-483-89-15, IEB-87-002, IEB-87-2, NUDOCS 8910170312 | |
| Download: ML19325C824 (15) | |
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.U..S.: NUCLEAR REGULATORY-C'MMISSIONI O
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REGION III'
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1ReportNo[50-483/89015(DRP)
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Docket No. 50-483 License No. NPF-30
- Licensee:. Union Electric Company
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. Post Office Box:149 - Mail Code 400
^5t. Louis, MO 63166
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Facility-Name: -Callaway Plant,-Unit 1
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Inspection'at: ;Callaway Site, Steedman, Missouri
Inspection Conducted: : July 16 through August 31, 1989 Inspectors:
B. H, Little L
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o-C. H. Brown
'Approv_ed By:
Rob
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Reactor Projects Section 3A Date - '
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-Inspection Summary
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-Inspection from July 16 through August 31,1989 (Report No. 50-483/89015(DRP))
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. Areas Inspected: A routine unannounced safety inspection of non-routine tevents, plant operations, maintenance and surveillance, and regional requests was-performed.
Results: Two violations-were identified. -The first identified three examples
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of the. licensee's -failure to report diesel generator failures (Paragraph 2.a).
i The other involves an environmental qualification deficiency associated with Target Rock solenoid operated valves (Paragraph 2.b).
Other results included:
tidentification of avoidable plant transients and/or safety system actuations associated with performance / procedural weaknesses (Paragraph 2.c); and observations of good' radiological controls, conditions, and practices by
- health physics staff (Paragraph 3.d).
A conservative safety attitude was demonstrated during work planning, work performance and in response to problems associated with fuel reconstitution activities (Paragraph 3.f).
The W
- licensee was responsive to " fastener testing" - TI 2500/27 (Paragraph 5).
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DETAILS i
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Persons-Contacted
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D. F. Schnell, Senior Vice President, Nuclear
- G. L. Randolph, General _ Manager, Nuclear Operations J. D. Blosser, Manager, Callaway Plant C. D. Naslund, Manager, Operations Support
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- J V. Laux, Manager, Quality. Assurance
- J. R. - Peevy, Assistant Mcnager, Technical Services
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- W. R. Campbell, Manager, Nuclear Engineering M. E. Taylor, Assistant Manager, Work Control D. E. Young, Superintendent, Operations
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- W. R._ Robinson, Assistant Manager, Operations and Maintenance R. R. Roselius,- Superintendent, Health Physics-
- T. P. _ Sharkey, Supervising Engineer, Site Licensing G. J. Czeschin, Superintendent, Planning and Scheduling W. H. Sheppard, Superintendent, Maintenance
- G. R. Pendegraff, Superintendent, _ Security L. H. Kanuckel, Supervisor, Quality Assurance Program
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G. A. Hughes, Supervisor, Independent Safety Engineer Group i
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- J. C. Gearhart, Superintendent, Operations Support, Quality Assurance
- C, S. Petzel, Quality Assurance Engineer
- J. A. McGraw, Superintendent, Design Control
- J. M. Brown, Licensing Fuels-Engineer
- Denotes those present at one or more exit interviews.
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In addition, a number of equipment operators, reactor operators, senior
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- reactor operators, and'other member:, of the quality control, operations, l
maintenance, health physics, and engineering staffs were contacted.
2.
Reports of Non-Routine Events (9270M
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Diesel Generator (D/G) Failures
An inspection in this area was performed to assess the licensee's f
documentation, evaluation, and reporting of D/G failures. The i
inspection was-initiated following discussions with NRR staff,
- 3 regarding the licensee's classification of the D/G failure l! {
documented in the Special Report Number 89-03 and questions as to p
the reportability of the D/G failure described in NRC Inspection l
Report Number 483/89009(DRP).
On July 13, 1989 the resident inspectors met with utility management and discussed the apparent reporting discrepancies relating to the
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D/G failures.
The licensee expressed the desire for further
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communication with NRC staff regarding the D/G failure classificati.on. On July 17,1989, during a telephone conversation, the licensee was advised by the NRC staff that it disagreed with the classification of the D/G failure as reported in Special Report
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Number 89-03. IThe NRC-' staff position is that the discovery of
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W conditions that are automatically alarmed or found during
inspections, when the D/G unit ~1s on standby, that would have
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b resulted in the' failure of the -diesel generator unit during test.or during response to a bona fide signal, should be considered a valid test and failure. - The NRC staff position,-regarding the D/G failure
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described in NRC Inspection Report Number 83/89009(DRP),.. is that :
the failure is'reportabi-n an " invalid" failure. The above D/G
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failures are discussed Aly in the paragraphs that follow.
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'Special Report (SR) 89-03 (Invalid Diesel Generator:' A'
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Failure) documented that on March 30, 1989 the' diesel generator s
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'A' was declared inoperable due to cooling water loaking from the exhaust valve cooling water jumper 0-rings on cylinder #8 -
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ir.to the rocker arm lube oil-reservoir. This condition was
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E discovered when the rocker arm lube oil reservoir high-high l
level' alarm sounded.
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The _ report also stated that this condition would eventually
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fail, and that the engine could not be relied upon to perform-its safety function. However, the report classified this' event
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as an invalid failure since this failure did not. occur during a test.
On July-24, 1989 the licensee issued Revision 1 to SP 89-03 which reclassified the above failure as a valid D/G 'A'
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failure.
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NRC Inspection Report Number 483/89009(DRP) documents the
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inspector's observations during the licensee's performance of surveillance procedure ISP-SA-2413A (Train 'A' Diesel Generator and Sequencer Test). The report indicated that the initial
' test failed due to a card failure.
The licensee documented the-failure in an Incident Report (IR) Number 89-079 on April 1, 1989. 'Through' trouble shooting, the licensee determined that g
the D/G ' A' failure to start during-the test was due to a
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i failed driver relay card in the load' shedding and emergency
l load sequencing (LSELS) circuit. The'11censee dispositioned the IR as "not reportable", because the failed card was not part of the defined D/G unit design and the special report was not submitted within the allotted time.
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On July 19, 1989, the licensee issued SP 89-07 which
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appropriately reported the failure as an " invalid" failure of
D/G 'A'.
Technical Specification (T/S) 4.8.1.1.3 requires that all diesel
generator failures,. valid or non-valid, be reported in a special report,tp the Commis.sion pursuant,to. Specification 5.9.2 within
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30 days.
Reports of diesel generator' failures shall includp the information recommended in Regulatory Position C.3.b. of Regulatory Guide 1.108, Revision 1, August 1977.
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.The A'. D/G-f ailure to start.on' April 1,1989 was not reported until
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July 19,1989. The licensee's failure to report all D/G failures
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'within~30 days is a violation of T/S 4.8.1.1.3(483/89015-01(A)(DRP)).
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Upon becoming aware of the NRC position relating to D/G failures,-
.the. licensee initiated a document review to assure that prior D/G
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included previous SPs and the IR database for D/C system deficiencies occurring since the issuance of the operating licensee f
(June 1984).
The review identified two additional D/G failures that
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had not been reported as required by T/S 4.8.1.1.3.
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(1) On February 15, 1989 D/G train 'B' was declared inoperable when the fuel oil transfer pump failed to autostart during a surv:'11ance test.
Troubleshooting determined the cau:;e to be-a
.ity Main Control Board (MCB) switch. The MCB switch had 6,aoled the fuel oil day tank level switches and the local handswitch. Consequently, the transfer pump was not capable of
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-automatically supplying fuel to the day tank.
The licensee's evaluation determined that the deficiency had
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existed since February 7, 1989. On that date an equipmo t
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operator (EO)-discovered that the transfer pump failed to start f
using the local handswitch.
The E0 notified the control room licensed reactor operator (RO) of the problem.
The R0 operated
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the pump from the MCB switch. Due to the short period taken to fill the standpipe, there was no indication of a problem with the MCB switch.
The E0 verified that the standpipe level was satisfactory and wrote a work request for the-local handswitch.
The licensee determined that the D/G had been inoperable from about 4:00 a.m. on February 7, 1989 until 10:25 p.m. on February 15, 1989.
Technic 61 specification 3.8.1.1 actions "b" and "d" were not initiated until discovery of the cor.dition at 4:00- a.m. on February 15, 1989 and, therefore, exceeded the allowed time limits.
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Licensee Event Report (LER) Number 89001 issued'on March 9, 1989 appropriately reported the violation of T/S 3.8.1.1 but did not provide the D/G failure details, e.g. ; classification and f ailure history, as required by T/S 4.8.1.1.3.
LER 89001-01 issued on July 21, 1989 classified the failure as
" valid" D/G 'B' failure and provided the specified "special report" details.
The D/G " failure" discovered on February 15,
1989 was not reported until July 21, 1989. The licensee's failure to report all D/G failures within 30 days is another example of a violation of T/S 4.8.1~.1.3(483/89015-01(B)(DRP)).
(2) On April 3~,1989 during the performance of surveillance test procedure ISP-SA-2413B,.D/G
'B'. failed to start; This failure
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was documented in IR Number 89-084. The licensee's evaluation showed that the operator failed to depress the undervoltage
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(UV) pushbuttons long enough for the logic circuit time: delays
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to " time out" and as a result the D/G did not receive a start-i l
signal. The operator error was attributed to'a procedural
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~ inadequacy in that the procedure step did not specify a minimum h
time that the pushbuttons must be held.
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I On July 19, 1989 the licensee' issued SP 89-08 which classified l
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the failure as an " invalid" failure'of-D/G 'B'.
The 0/G-
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R failure discovered on April 3, 1989 was not reported until'
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July 19, 1989.
The licensee's failure to report all D/G failures within.30 days is another example _of a violation of T/S 4.8.1.123 (483/89015-01(C)(DRP)).
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1The inspector reviewed the irs associated ~with the above events and interviewed licensee personnel involved in the event evaluation and P
resolution process.
The inspector determined that'although i
reporting errors were'made, the deficiencies were documented in irs.
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The -irs-underwent licensee event cause evaluation -and appropriate corrective action was taken, including action in response to an i
engineering reliability study-(0-rings) and in response to a failure -
analysis (relay cards).
The licensee's corrective action included the following:
Replaced 0-rings on all cylinders for both D/G units during the
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refueling outage.
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Instrumentation and control preventive maintenance (PM) tasks
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(Numbers P460225 and P459502) were issued. These tasks specify
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monthly verification of proper relay driver card operation.
Operations surveillance procedures (ISP-SA-2413A and B) were
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revised to provide caution of the 1.2 second time delay
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A temporary change to operations surveillance procedures
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(OSP-NE-00002 and ODP-ZZ-00016) was issued which changes the D,
minimum level of the D/G day tanks to 3.0 feet.
This allows the automatic control operation of D/G fuel oil transfer pumps.
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In' addition, D/G reportability issues were discussed with site licensing personnel. The NRC staff position relating to D/G failures has been entered in the licensee's event " reference file" to preclude future reportability determination errors.
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The inspection showed that the licensee had taken appropriate corrective action for the specific events and action to prevent future reporting violations.
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The violations meet the tests of.10 CFR 2, Appendix C, Sect. ions V.A and V.G;. consequently,.no Notice of Violation will be issued.and
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th'is matter is considered closed.
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y-LThis inspection closes the associated items listed below:
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Special' Report 89-03-
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Special Report 89-07
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Special-~ Report' 89-08:.
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LER 89001 and~89001-01
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Unresolved Item 483/89011-02(DRP)
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(Closed) Unresolved item Number 483/89011-01(DRP): Target Rock Solenoid Operated Valves (SOVs) Used in 10 CFR 50 49 Applications p
' Carolina Puwer and Light (CP&L) Licensee Event Report (LER)88-026, dated October. 6,1988, documented equipment qualification (EQ)
deficiencies associated with Target Rock (TR) SOVs at the Shearon
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Harris nuclear plant.
In November 1988, th'e Nuclear Utility Group
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on EQ (NUGEQ) notified its members, by a memorandum, dated November 29, 1988, of the specific deficiencies identified by CP&L.
The;EQ discrepancies noted were; 1) the cracking of Ristance or Markel' reed switch lead wires; 2) the lack of traceability between the EQ-tested' components and the components actually installed in
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the S0Vs.(qualification of the SOVs had been established through the testing of Bolden wire and Kulka terminal blocks,- and the valves installed in.the-plant were found to contain either Ristance or Markel wires and Beau terminal blocks) and; 3) the cracking of r
terminal blocks.due to the tightening of jumper wire lugs on the
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blocks. - CP&L replaced the reed switch lead wires with qualified
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Kapton wires and the Beau terminal blocks were replaced with qualified'Kulka terminal blocks.
Due to.the EQ deficiencies identified above, Union Electric (UE)
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The-inspection identified unqualified Beau
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' terminal blocks and Ristance or Markel lead wires installed in the
SOVs.
The licensee initiated a~ Justification for Continued
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Operations (JCO) which addressed the qualifiability and operability i
of 14 suspect SOVs. installed in the plant. The valves were reported to be installed in the BB system (head vent), BG system (excess letdown), and the EP system (accumulator vent).
The licensee stated, in its JCO, that the valves were operable because the TR housing was a sealed enclosure and would prevent moisture intrusion.
In addition,'the licensee reported that alternate equipment and instrumentation indications were available to mitigate possible S0V failures.
During the April 1989 refuel outage the licensee performed an
inspection of the suspect TR SOVs installed in the plant.
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licensee identified the unqualified Ristance or Markel wires and i
Beau terminal blocks installed in the SOVs. The licensee reported
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that the components were found in good condition and that no cracking or breakage was discovered..The licensee replaced the unqualified wires :nd terminal blocks with the qualified Kapton wires and Kulka terminal blocks.
The NRC inspectors concluded that
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this re;) resented a violation of 10 CFR 50.49 requirements
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(483/E9015-02(DRS)).
However, this violation meets the tests of
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10 CFR.Part 2 Appendix C, Section V.G.1; consequently, no Notice of
. Violation will oe issued, and this matter is considered closed..
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Inspectionjf Licensee Event Reports
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Through direct observations, discussions with licensee personnel,
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and a review of records, the following licensee event reports were reviewed to determine that reportability requirements were fulfilled, that immediate corrective action was accomplished, and
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that corrective action to prevent recurrence was accomplished in accordance with Technical Specifications (T/Ss). The LERs listed
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below are considered closed.
(1) (Closed) LER 89002:
Engineered Safety Features (ESF)
Actuations W(.
Two unplenned ESF actuations occurred.
In the first event, a control room ventilation isolation signal (CRVIS) occurred due to an electrical spike on GK-RE-04. The spike occurred due to
.a shield breakage of the GK-RE-04 coaxial' cable.
In the second event, a CRVIS and a fuel building isolation signal (FBIS) occurred due to a technician's inadvertent actions w5tle troubleshooting the ESF automatic test insertion (ATI) system.
In this event, t% technician accideatally pushed in an ESF actuation pushbstton located adjacent to the-ATI decoder module card he was reinserting.
(2) (Closed) LER 89003:
Engineered Safety Features Actuation on High Steam Generetor Level During a plant shutdown in preparation for the Refuel III outage, a feedwater isolation signal (FWIS) and an auxiliary feedwater actuation signal (AFAS) occurred due to a high water -
level in stsam generator 'A'.
Reactor power was at three percent and decreasing. The main turbine had been manually tripped eight minutes prior to the event.
The root cause of the event was attributed to personnel error, in which the reactor operator did not take sufficient actions in reducing feedwater flow to stop the 'A' steam generator level increase. A contributing factor was the fact that automatic level control for steam generas
'A' was not availab e, due to a problem with the contioller card.
(3) (Closed) LER 89004-01:
Engineered Safety Features Actuation on l
,High Steam Generator Level
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A fe'edwater tsolat. ion occurred due to a' h'igh water' 1evel in
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steam generator (S/G) ' A'.
The plant was being heated up following Refuel III. The
'A' main steam isolation valve
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(MSIV) was opened, and the level in S/G 'A' began ti swell-
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rapidly to the high level setpoint.
The 'A' MSIV bypass valve had been opened 14 minutes earliea to equalize pressure across the MSIV.
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The root cause of this event was the failure to adequately equalize pressure across the' MSIVs prior to opening the valves.
(4) (Closed) LER 89005:
Two Safety Injection (SI) Actuations e
A SI actuation on train 'A' was received due.to a low steamline pressure signal during a reactor trip breakt.r trip actuating device operational test.
Five minutes later durin7 restoration, a second SI signal was received. The plant was in Mode 4, Hot Shutdown at the time of the event, r
The root cause of the first SI event is cognitive personnel error. A licensed operator turned a logic switch in the wrong direction, unblocking the low steamline pressure SI. When the mode switch was turned from " test" to " operate", the i
established SI signal was processed, causing the train 'A'
SI plant equipment to actuate.
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The second $1 was due to a licensed operator inadvertently pressing the " pressurizer pressure SI reset" button, instead of the " manual SI reset". A second SI signal was processed, but all equipment was already in its SI position.
(5) (Closed) LER 89006:
Reactor Trip / Turbine Trip Due to High Power Start-Up FTux Rate Trip On May 29, 1989, a reactor trip occurred on a power range
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neutron flux rate trip signal. An incore/excore calibration surveillance was in progress. At the request of a reactor engineer, an I&C technician removed the leads that were connected to power range channels N41 anc N42. When a
reconnecting, the N42 leads were dropped causing a ground. A ground in one of the N41 leads also occurred, which gave the two-out-of-four coincidence for the trip signal.
The cause of this event was tne lack of detail provided in
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plant precedures, which resulted in personnel not being aware of the pctential consequences, Neither the technician nor the reactor engineer s aalized the risks associated with the configuration due to the multiple channel arrargement and the non-isolated test points.
(6 (Closed) LER 89007: Auxiliary Feedwater Actuation (AFA)
During the performance of surveillance procedure OSP-SA-0015A,
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an AFA of the motor driven auxiliary feedwater pumpA'
occurred. Near the end of the procedure, the operator is
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required to reset a steam generator blowdown isolation signal.
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The button is covered with a "Stop and Think" cover. The
operator lifted the cover, then he inadvertently pressed
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" Actuate" instead of " Reset", resulting in an AFA.
The above events occurred between February and June 1989.
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inspectors determined that the events were appropriately docusi $ntcd y
and reported.
Plant systems and equipment functioned as designed, f
The events were attributed to a combination of performance errors,
procedural weaknesses, and test hardware deficiencies; and involved different personnel and work activities.
The licensee performed a post trip and event review for each occurrence.as part of Callaway's Plant Event Reduction Program.
This review included a human performance evaluation..The evaluation
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resulted in a critical assessment of root cause and recommendations for improvements. Action taken included managtment/ crew dissJssions and " lessons learned" training to increase personnel awareness,
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procedure revisions (steps and cautions) and hardwaee improvements (non-conductive material for test instrument tables).
The above events resulted in avoidable plant transients and/or
safety system actuations.
The inspectors determined that the licensee is actively involved in overall event reduction activities including corrective and improvement initiatives.
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Other than avoidable actuation of safety systems, all activities were
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cor. ducted in an adequate and safe manner.
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Plant Operations (71707)
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Oyerational Safety Verification
P Inspections were restinely performed to ensure that the licensee conducts activities at the facility safely and in conformance with
regulatory requirements.
The inspections focused on the
implementation and overell effectiveness of the licensee's control of operating activities, and on the performance of licensed and non-licensed operatot 4 and shift technical advisors.
The inspections
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included direct observation of activities, tours of the facility, interviews and discussions with licensee personnel, independent verification of safety system status and limiting conditions of operation (LCO), and reviews of facility procedures, records, and reports.
The following items were considered during these inspections:
Adequacy of plant staffing and supervision.
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Control room professionalism, including procedure adherence,
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-operator attentivenesst and response to alarms, events, and-
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off-n'ormal conditions.
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Operability of selecte'd safety-related sy t' ems, including
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attendant alarms, instrumentation, and controls, Maintenance of quality records and reports.
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The inspectors observed that control room supervisors, shift
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. technical advisors, and operators were attentive to plant
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conditions, performed frequent panel walkdowns and were responsive r
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Off-Shift Inspection of Control Room
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. The inspectors performed routine inspections of the control room during off-shift and weekend periods; these included inspections i
between the hours of 20:00 p.m. and 5:00 a.m.
The inspections were
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conducted to assess overall crev performance and, specifically,
control room operator attentiveness' during night shifts.
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F operators were; attentive to their duties, and that the
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administrative controls relating to the conduct of operation were.
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Plant' Material = Conditions / Housekeeping-
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The inspe: tors performed routine plant tours to assess material
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plantwide' housekeeping.
The inspectors also accompanied the
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licensee's management on monthly plant tours, j
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Radiological Control,s The licensee's radiological controls ar.d practices were routinely observed by the. inspectors during plant tours and during the inspection of. selected work activities.
The inspection included
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direct observations of health physics (HP) activities relating to radiological surveys and monitoring, maintenance of radiological control signs and barriers, contamination, and radioactive waste controls.
The inspection also included a routine review of the licensee's radiological and wcter chemistry control records and reports.
- Good radiological controls, conditions and practices were observed.
Overall radiological conditions in the plant have improved through the licensee's " contaminated area reduction" program.
The total-contaminated area had been reduced to approximately 5,000 square feet (an area of approximately one-half the area existing in 1988).
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The HP staff provided close supervision and effective control of fuel reconstitution activities in the fuel building.
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Security
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The licensee's security activities were observed by the inspectors during routine facility tours and during the inspectors' site
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arrivals and departures.
Observations included the security
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personnel's performance associated with access con' trol, security
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checks, and surveillance activities., and focused on the adequacy of security staffing, the security response (compensatory measures),
and the security staff's attentiveness and thoroughness.
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Speat Fuel Pool Work Activities
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During August 1989, licensee and Westinghouse personnel oerformed a
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variety of work activities in the spent fuel pool. The work was.
associated with the inspection and testing of Hafnium Rod Cluster-
. Control Assemblies (RCCA), the inspection of fuel pins in standard
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low parasitic (LOPAR) fuel assemblies, and fuel reconstitution of Vantage 5 fuel assemblies, The following deficiencie', were identified during the performance of these activities.
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On August 18, 1989 an attempt to_ remove RCC4 R-41 from the fuel
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assembly failed, apparently due to binding. The licensee i
directed that no further attempts be made to move Rod R-41 pending further evaluation.
(Incident Report IR 89-310).
On August 21 and 23, 1989 two broken fuel pins were found
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during visual inspection of fuel pins in LOPAR fuel assemblies.
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The licensee placed a hold on fuel pin removal pending further evaluation (IR 89-316).
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On August 26, 1989 while performing work associated with fuel
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reconstitutior of Vantage 5 fuel assembly E07, visual examination found that eight of the 24 guide tube flexures had sustained excessive bending.
The bending was sucn that
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locktubes, which secure the nozzle to the assembly, could not be inserted.
Reconstitution work was stopped and is currently undergoing event cause and corrective action evaluation.
(IR 89-321)
The inspectors performed frequent observations of ongoing work in the spent fuel. pool area.
The inspectors determined that the licensee provided close supervision and control of the work activities, and demonstrated a conservative safety attitude during the work planning, work performance and in response to problems. The deficiencies were appropriately documented and are undergoirg event cause and corrective action evaluation, g.
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An inspection of emergency preparedness activities was performed to assess the licensee's implementa; ion of the emergency plan and implementing procedures.
The inspection included monthey
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observation of emergency facilities and equipment, interviews with licensee staff, and a review of selected emergency implementing procedures.
The Callaway Plant Radiological Emergency Response Drill was
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tonducted Augu'st-1.6, 1989.
Thy On-Sh11);.On-Site; Emergency.
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v Operations Facility (EOF), and Public Inf-orm;ation Emergency
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Risponse Organizations were' oper'ational during the drill.
Participation by off-site orgarizations was limited to an abbreviated response by state personnel at the Joint Public L
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f t
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i Information Center and the State Forward Command Post in the EOF.
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f The primary objective for this drill was to provide drill practice'
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and experience for the identified emergency response personnel. The L
majority of on-site drill participants selected had not been g-involved in prior emergency preparedness exercises.- Efforts were-
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power (including the loss of the public address system and plant-l made to reduce the amount of simulation, however, the loss of all.AC
- 3 tW emergency alarms) wa, simulated.
This loss of power identified
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various communication problems and delayed the personnel
accountability process.
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The. inspector observed drill participant's performance in the
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control room and the Technical Support Center, and attended the n
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initial post-drill critique.
Dri?) participants demonstrated good F
teamwork and a positive attitude, The critique included a c-itical p",
self-assessment of drill performance weaknesses. The drill ppeared to be challenging, satisfying the drill objective.
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A11' activities were. conducted in an adequate and safe manner.
L 4.: !bintenance/ Surveillance (62703) (626)
t b
Selected portions of the plant surveillance, test and maintenance
activities on safety-related systems and cc.nponents were observed or r.
reviewed-to ascertain that the activities were performed in accordance with approved procedures, regulatory guides, industry codes and p
standards, and the Technical Specifications. The following items were
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' considered during these inspections:
the limiting conditions for
. operation.'were met while components or systems-were' removed from service; approvals were obtained prior to initiating the work; activities were accomplished ut.ing approve'i procedures and were inspected as applicable; functlanal testing and/or calibration was performed prior to returning the components or systt:ms to service; parts and macerials that were used were properly certified; and appropriate fire prevention, radiological, and housekeeping conditions were maintained.
The observed ongoing maintenance and surveillance activities were found to be properly authorized and were being'. performed using approved procedures.
The activities were noted to be scheduled and requiref isolations and tagging were found to be correctly carried out.
The limiting conditions for operation were adhered to during the performance
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of these activities.
In general, the workmanship was fcund to be satisfactory.
.a.
Maintenance
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The reviewed maintenance activities included:
Work Request No.
Activity WR P449961 Clean / inspect load center breaker ner procedure E-017-00397.
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VR A405639A-Install and remove freeze seal'for recycle.
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evaporator condenser component cooling water
ap outlet relief, per procedure MDP-ZZ-FS001.
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WR P405639 Test pressure relirf valve for recycle
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~ evaporator condent r component cooling water.
i outlet relief.
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WR W1115990 Replace temperature switches, essential
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L service water ultimate heat sink cooling tower
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D ce- ' 'A';
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Surveilla; ice P
The reviewed surveillances included:
pe Procedure No.
Activity p
ISL-GS-00A2B Containment hydrogen concentration analysis j
transmitter, train 'B'.
L ISF-BB-0P458
. Reactor coolant system pressurizer
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g" protection 'B' pressurizer transmitter, functional.
b ISF-EF-00P43 Esseatial service water train ' A' to compressed air compressor differential pt essure-transmitter.
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. Control rod partial movement test.
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i ISF-AL-00P37 Condensate storage tank to auxiliary
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feedwater pump suction header pressure t
transmitter.
ESP-ZZ-00025 Reactor vessel delta temperature a
measurement.
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'A' diesel generator one hour load test.
OSP-SA-0017A Slave relay test, load shedding emergency load sequencing train 'A', centrifugal charg'ng pump 'A'
and 'A' diesel auto start.
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OSP-SB-00001 Reactor trip breakers, trip actuating devicu
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operational test.
f OSI-ZZ-00001 Control room shift and daily log readings and l.
channel, checks.
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Loop tim'erature,' loop 2 delta temperature
- ISL-BB-AE421-p input to steam generator low-low level.
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ISL-NF-NB010'
Loop, miscellaneous NB01C degraded and under-
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voltage to load. shedding emergency load-sequencer, All' activities were conducted in an adequate and safe manner.
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p-5.
Regional Requests-O 2701)
N Temporary Instructicn (TI) 2500/27 - Inspection. Requirements for NRC Compliance to Bullet n 87-02, " Fastener Testing to Determine Conformance
. ith Applicable Materici Specifications".
s, w
TI 2500/27 provided inspection requirements and guidance specific to
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piants identified in the TI.
Inspection requirements specified for Callaway were Paragraphs 04.02, 04.05 and 04.06.
Paragraph 04.05-identified the lice.nsees that had not satisf'ed the sample testing requested by Bulletin 87-02. Callaway was 1.iadvertently listed in that group.
Callaway's sample testing. included 37 safety
related and 12 non-safety related fasteners and nuts, which exceeded the minimum requested by the bulletin.
L The Tl paragraphs applicable to Callaway'are 04.02 and 04.06. These
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relate to the licensee's root cause analysis and corrective actions pertaining to the. three non-safety related nuts which faileri the hardness testing requirements. All three of these nuts were. from a single stock number and were sent to Callaway as a spare parts replacement by Dresser Industries for use-on its non-safety related I"-600 pound carbon steel-pste vahes.
The intended end use was as a replacement gland packing nut F
which is considered non-pressure retaining.
The' licensee attributed the root cause to the lack of a requirement for vendors to supply information on material specification or grades.on non-safety related spare parts. As such, non-safety related spare parts
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do not receive a receipt inspection for this attribute.
The licensee's corrective action includes the following:
-The'three non-safety related nuts which failed the hardness test
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were scrapptd along with the entire batch supplied by the vendor.
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Callaway has developed and implemented a training course " Good
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Bolting Practices" (T67.0120.6) that the mechanical craft personnel were required to complete as part of their annual requalification.
This was implemented in the second half of 1988. Part of this course covered the recognition of fastener specifications and grade c
markings.
Inventory at Callaway is split into two categories comprising of
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General Stock and Spare Parts.
Cc.1away does not maintain any non-safety related SA-194, grade 2H nuts in general stock. Only safety related nuts, which require material certifications and receive a hardness check as part of receipt inspection are available from the-storeroom as general stock.
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Piping class sheets, 10466-MS-2, were revised to require safety
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related ASME SA-194, grade 7 nuts in lieu of ASTM A-194, grade 2H nuts
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on non-safety related Main Steam and Feedwater piping systems a
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as well as other systems. This was done as a combination of
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<F.
. ensuring superior quality nuts on piping systems and :onsolidating
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inventory.. SA-194, grade 7 nut material has the same physical
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strength ~ requirements as SA-194, grade 2H but is more corrosion
' resistant.
No backfit program is anticipated for this item.
The inspector determined that the licensee was responsive to NRC requests t. -
in this matter.
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T1 2500/27 is. considered closed.
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.6.
Violations for Which a " Notice of Violation" Will Not be Issued
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The NRC'uses the Notice of Violation as a standard method for formalizing
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the existence of-a violation of a legally binding requirement.
However, because the'NRC wants to encourage and support licensee initiatives for self-identification and correction of problems, the NRC will not
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generally issue a Notice of Violation for a violation that meets the
tests of 10 CFR 2, Appendix C, Section V.G.I.
These tests are:
(1) the
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violation was ident.ified by the licensee; (2) tha violation would be categorized as Severity Level IV or V; (3) the violation was reported to the NRC, if-required; (4) the violation will be corrected, including
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measures to prevent recurrence, within a reasonable time period; and
(5) it was not.a violation that could reasonably be expected to have been t
. prevented by the licensee's corrective action for a previous violation.
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In addition, for isolated Severity Level V violations, a Notice of Violation normally will not be issued regardless of who identifies the violation provided that the licensee has initiated appropriate corrective action before the inspection ends.
(10 CFR 2, Appendix C, Section V.A).
Violations for which a Notice of Violation will not be issued are identified in Paragraph 2.a and 2.b of this report.
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ExitMeeting(30703}
The inspectors met d i licensee representatives (denoted under Persons
. Contacted) at intervais during the inspection period. The inspectors summarized the scope and findings of the inspection. 1he licensee representatives acknW1 edged the findings as reported herein.
The inspectors also discussed the likely informational content of the inspection report wit", regard to documents er processes reviewed by the inspectors during the inspection. The licensee did not identify any such documents / processes as proprietary.
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