ML20209A723
| ML20209A723 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 01/23/1987 |
| From: | Bruce Bartlett, Cummins J, Hunter D, Mullikin R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20209A650 | List: |
| References | |
| 50-482-86-34, IEIN-86-001, IEIN-86-014, IEIN-86-037, IEIN-86-057, IEIN-86-061, IEIN-86-077, IEIN-86-092, IEIN-86-093, IEIN-86-094, IEIN-86-096, IEIN-86-1, IEIN-86-14, IEIN-86-37, IEIN-86-57, IEIN-86-61, IEIN-86-77, IEIN-86-92, IEIN-86-93, IEIN-86-94, IEIN-86-96, NUDOCS 8702030423 | |
| Download: ML20209A723 (20) | |
See also: IR 05000482/1986034
Text
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APPENDIX B
U. S. NUCLEAR REGULATORY COMMISSION
REGION IV
NP.C Inspection Report: 50-482/86-34
LP: hPF-42
Docket: 50-482
Licensee:
Kansas Gas and Electric Company (KG&E)
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Post Office Box 208
Wichita, Kansas 67201
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Facility Name:
Wolf Creek Generating Station (WCCS)
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Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas
Inspection Conducted: December 1 to 31, 1986
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Inspectors:
@.ME. Cummins,1Mor Resident Inspector,
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Operations
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B. L. Bartlett',' Resident Reactor Inspector,
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R.' i'. 'Mullikih,' Project Engineer
Date
Approved:
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K[Iunt
, C{3 ef, Projbet Section B
Date
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Reac'tbr Iroject Branch
8702030423 870129
ADOCK 05000482
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Inspection Summary
Inspection Conducted December 1-31, 1986 (Report 50-482/86-34)
Areas Inspected: Routine, unannounced inspection including plant status;
followup of previously identified NRC items; operational safety verification;
ESF system walkdown; monthly surveillance observation; monthly maintenance
observation; cold weather preparation; review of LERs; IE Information Notice
followup; 10 CFR Part 21 Report followup; onsite event followup; plant startup
from refueling; thermal-hydraulic anomaly; and inservice inspection of safety
valves.
Results: Within the fourteen areas inspected, two violations were identified
(failure to have an adequate procedure for draining the reactor coolant system,
paragraph 12 and a violation of TS-fire suppression system surveillance,
paragraph 4). Two open items are identified in paragraphs 14 and 15.
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DETAILS
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1.
Persons Contacted
Principal Licensee Personnel
B. D. Withers, President, Wolf Creek Nuclear Operating Corporation
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J. A. Bailey, Interim Site Director
- F. T. Rhodes, Plant Manager
- G. D. Boyer, Deputy Plant Manager
- R. M. Grant, Director-Quality
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- C, M. Estes, Superintendent of Operations
- M. D. Rich, Superintendent of Maintenance
+*M. G. Williams, Superintendent of Regulatory, Quality, and
Administrative Services
O. L. Maynard, Manager Licensing
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- K. Peterson, Licensing
+*G. Pendergrass, Licensing
+*W. M. Lindsay, Supervisor Quality Systems
+ C. J. Hoch, QA Technologist
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+ W. J. Rudolph, QA Manager-WCGS
- A. A. Freitag, Nuclear Plant Engineering Manager-WCGS
- M. Nichols, Plant Support Superintendent
+ S. Wideman, Technical Staff Specialist
+ A. S. Mah, Superintendent of General Training
+ W. L. Mutz, itanager of Nuclear Operations Support
NRC Personnel
+*J. E. Cummins, Senior Resident Inspector
- B. L. Bartlett, Resident Inspector
-+ R. P. Mullikin, Project Inspector
The NRC inspectors also contacted other members of the licensee's staff
during the inspection period to discuss identified issues.
+ Denotes those personnel in attendance at the exit meeting held on
November 18, 1986.
- Denotes those personnel in attendance at the exit meeting held on
January 6,1987.
2.
Plant Status
During this report period, the first refueling cutage (October 16 to
December 20) was completed and the plant was returned to operating status
in Mode 1.
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3.
Followup On Previously Identified NRC Items
(Closed) Open Item (482/8508-10):
Reactor Engineering Surveillance
Procedures. This open item tracked the establishment of surveillance
procedures for performing surveillances in Section 3.10 of the TS. At the
time the open item was written, the licensee's matrix referenced
Procedures RXE 04-100 and 200 for performing the required surveillances.
However, this numbering system was not used and since TS limiting condition
for operation (LCO) for Surveillances 4.10.1.1, 4.10.1.2, 4.10.2.1,
4.10.2.2.a, 4.10.2.2.b, 4.10.4.1, and 4.10.4.3 apply only during the
infrequent suspension of the LCOs to support testing, the licensee has
elected to write these surveillance procedures on an as-needed basis, when
and if they are ever needed. The table below shows the applicable procedure
for the remaining TS surveillances in Section 3/4 10, "Special Test Exceptions,"
of the TS:
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TS Requirement
Procedure
4.10.3.1
RXE 01-002
4.10.3.2
RXE 01-002
STS IC-235
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STS IC-236
STS IC-241
STS IC-242
STS IC-243
STS IC-244
4.10.3.3
RXE 01-002
STS IC-201
STS IC-202
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STS IC-236
STS IC-241
STS IC-242
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STS IC-243
STS IC-244
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4.10.5
STS RE-007
This item is closed.
(Closed) License Condition 2.C(1): This item was also tracked as a 10 CFR Part 21 Report-Colt Industries Inc. dated March 28, 1984, " Emergency
Diesel Generator Lube Oil Keep Warm Pumps."
The NRC inspector verified by review of documents and discussions with
licensee personnel that the emergency diesel generator lube oil keep warm
pumps had been replaced with pumps which satisfy ASME Section III, Class 3
requirements. The pumps were replaced on Wolf Creek (WC) Work
Requests (WR) 03157-86 and 03158-86. This item is closed.
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(Closed)OpenItem(482/8511-06): Microbiologically Induced Corrosion
(MIC) Inspections. This open item tracked a licensee commitment to
perform MIC inspection of selected systems during refueling outage. The
licensee has established and implemented Administrative Procedure ADM
01-100, Revision 0, "MIC Control Program," which provides instructions for
performing activities related to monitoring, evaluating, and controlling
MIC. The NRC inspector verified by review of the WC WRs listed below that
the licensee has inspected the selected systems:
o
WR 00688-86, Inspected closed cooling water heat exchangers
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WR 60142-86, Circulating water system piping
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WR 60143-86, Component cooling water (CCW) heat exchangers and
essential service water piping at CCW heat exchanger water box
o
WR 60145-86, One each of the diesel generator heat exchangers end'
spool piece between the two diesel generator "B" heat exchangers
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WR 60144-86, Fire protection system piping section in circulating
water screen house
This item is closed.
(Closed) Open Item (482/8515-01): Testing of Ruskin Fire Dampers During
Normal Air Flow Conditions. This item concerned a 10 CFR Part 21 Report
filed by Ruskin Manufacturing Company regarding the potential for certain
fire dampers to not completely close against normal ventilation system
flow.
Even though the licensee had administrative controls to shut off
HVAC flow during a fire, they agreed to test TS fire dampers during normal
flow conditions. The licensee has tested the fire dampers and either
modified or replaced those that have failed. The NRC inspector reviewed
the applicable completed work requests. This item is closed.
(Closed) Open Item (482/8511-16):
Reports Describing the Vendor Interface
Program and Reactor Trip Breaker Testing Frogram. This item required the
licensee to submit reports describing the vendor interface program and the
reactor trip breaker testing program, the results of the testing, and the
impact of the results on the breakers demonstrated life cycle and periodic
replacement interval. This is in regard to NRC Generic Letter 83-28. The
licensee has submitted the reports to the Office of Nuclear Reactor
Regulation (NRR) on May 21 and December 10, 1986.
This item is closed.
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4.
Operational Safety Verification
The NRC inspectors verified that the facility is being operated-safely and
in conformance with regulatory requirements by direct observation of
licensee facilities, tours of the facility, interviews and discussions
with licensee personnel, independent verification of safety system status
and limiting conditions for operations, and reviewing facility records.
The NRC inspectors, by observation of randomly selected activities and
interviews of personnel, verified that physical security, radiation
protection, and fire protection activities were controlled.
By observing accessible components for correct valve position and
electrical breaker position, and by observing control room indication, the
NRC inspectors confirmed the operability of selected portions of safety-
related systems. The NRC inspectors also visually inspected safety
components for leakage, physical damage, and other impaiments that could
prevent them from performing their designed functions.
Selected NRC inspector observations are discussed below:
On November 25, 1986, during a routine review of Defect / Deficiency
Report (DDR) No.86-104, the NRC inspectors observed that KG&E Quality
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Assurance (QA) Audit TE: 50140-K142, " Fire Protection (Annual)," had
written Quality Program Violation (QPV) 11/86-180 stating that maintenance
surveillance procedures did not cover TS surveillance requirement 4.7.10.3.b.
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TS 4.7.10.3.b requires halon systems to be demonstrated operable "at least
once per 18 months by verifying the system, including associated ventilation
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system fire dampers . . . actuates manually and automatically . . . ."
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Procedures STS MT-032, Revision 4, " Single-Zone Halon System Checkout,"
STS MT-036, Revision 3. "Two-Zone Halon System Checkout," and STS MT-037,
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Revision 3 "Six-Zone Halon System Checkout," did not address the verification
of damper actuation, either automatically or manually, during halon system
surveillance testing. This failure to meet TS requirements is a
violation (482/8634-02).
10 CFR Part 2, Appendix C, Section V.a states,
in part, NRC will not generally issue a Notice of Violation (NOV) for a
violation that is self-identified, that is a Level IV or V, that was
reported if required, that was or will be corrected, and that was not a
violation that could reasonably be' expected to have been prevented by the
licensee's corrective action for a previcus violation. NRC
Violation 482/8541-01, " Violation of Technical Specification-Fire
Suppression System Surveillance," issued Hovember 29, 1985, states, in
part, "STS FP-005, Revision 0, monthly sprays / sprinkler valve position
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verification, does not require an inspection of these two ESF transformer
water spray systems. Thus, the required TS monthly surveillances have not
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been performed." KG8E's response to this violation, KMLNRC 85-277, dated
December 20, 1985, states, in part, " Internal Operations Program
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Deficiency (IOPD) Number 85-11 was generated to review the fire protection
program comitments to ensure they are all met by our administrative
program." Violation 482/8634-02 is a violation that could reasonably be
expected to have been prevented by the licensee's corrective action for a
previous violation.
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5.
Engineered Safety Features System Walkdown
The NRC inspector verified the operability of the ESF system by walking
down selected accessible portions of the system. The NRC inspector
verified valves and electrical circuit breakers were in the required
position, power was available, and valves were locked where required. The
NRC inspector also inspected system components for damage or other
conditions that could degrade system performance.
The ESF system walked down during this inspection period and the documents
utilized by the NRC inspector during the walkdown are listed below:
System
Documents
High Pressure Coolant
Drawing M-12BG03(Q), Revision 2,
InjectionSystem(EM)
Piping and Instrumentation Diagram
Chemical and Volume Control System
Drawing M-12EM02(Q), Revision 1,
Piping and Instrumentation Diagram
High Pressure Coolant Injection
System
No violations or deviations were identified.
6.
Monthly Surveillance Observation
The NRC inspectors observed selected portions of the perfonnance of
surveillance testing and/or reviewed completed surveillance test
procedures to verify that surveillance activities were performed in
accordance with TS requirements and administrative procedures. The NRC
inspectors considered the following elements while inspecting surveillance
activities:
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Testing was being accomplished by qualified personnel in accordance
'with an approved procedure.
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The surveillance procedure conformed to TS requirements.
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Required test instrumentation was calibrated.
Technical Specification limit.ing conditions for operation (LCO) were-
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satisfied.
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Test data was accurate and complete. Where appropriate, the NRC
inspectors performed independent calculations of selected test data
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to verify their accuracy.
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The performance of the surveillance procedure conformed to applicable
administrative procedures.
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The surveillance was performed within the required frequency and the
test results met the required limits.
Surveillances witnessed and/or reviewed by the NRC inspectors are listed
below:
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STS KJ-100A, Revision 3, " Integrated D/G and Safeguards Actuation
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Test-Train A," conducted on December 5-10, 1986
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STS RE-004, Revision 5 " Shutdown Margin Determination," conducted on
December 10, 1986
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STS PE-013, Revision 5, " Personnel Air Lock Seal Test," conducted on-
December 10, 1986
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STS BG-002, Revision 5, "ECCS Valve Check and System Vent,"
conducted on December 9 and 10, 1986
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STS AL-005, Revision 6, " Auxiliary Feedwater Automatic Pump Start and
Valve Actuation," conducted on December 11, 1986
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STS KJ-0018 Revision 3 " Integrated D/G and Safeguards Actuation
Test-Train B," conducted on December 11 through 13, 1986
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STS AL-103, Revision 4, " Turbine Driven Aux FW Pump Inservice Pump
Test," conducted on December 16, 1986
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STS AL-210, Revision 5, " Auxiliary Feedwater System Flow Path
Verification and Inservice Check Valve Test," conducted on
December 16, 1986
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STS PE-19E, Revision 0, "RCS Isolation Check Valve Leak Test,"
conducted on December 10, 1986
No violations or deviations were identified.
7.
Monthly Maintenance Observation
The NRC inspector observed maintenance activities perfonned on
safety-related systems and components to verify that these activities were
conducted in accordance with approved procedures, Technical Specifications,
and applicable industry codes and standards. The following eleraents were
considered by the NRC inspector during the observation and/or review of
the maintenance activities:
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LCOs were met and, where applicable, redundant components were
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Activities complied with adequate administrative controls,
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Where required, adequate, approved, and up-to-date procedures were
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used.
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Craftsmen were qualified to accomplish the designated task and
technical expertise (i.e., engineering, health physics, operations)
was made available when appropriate.
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Replacement parts a'nd materials being used were properly certified.
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Required radiological controls were implemented.
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Fire prevention controls were implemented where appropriate.-
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Required alignments and surveillances to verify post maintenance
operability were performed.
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Quality control hold points and/or checklists were used when
appropriate and quality control personnel observed designated work
activities.
Selected portions of the maintenance activities accomplished on the work
requests (WR) listed below were observed and related documentation
reviewed by the NRC inspector:
No.
Activity
WR 01253-86, ALV054 Investigate Valve Internal Components for Degradation
and Valve Backleakage
WR 05132-86, KC System, Fire Damper Operated By ETLs Receiving Signals
from Halon System Must Be Physically Checked
No violations or deviations were identified.
8.
Cold Weather Preparation
The NRC inspector verified by reviewing licensee procedure (performed on
September 19 and 20,1986) STN GP-001, Revision 1 " Plant Winterization,"
that the licensee had taken action to protect vital areas and equipment
that could be subjected to freezing conditions.
The procedure verified
that heat and heat tracing was available and energized where appropriate
to the areas and equipment. The NRC inspector walked down the areas
adjacent to the refueling water storage tank and the condensate storage
tank and verified that lines (heating and instrument) were insulated where
they could be subjected to freezing conditions, and where appropriate heat
was being supplied to the area.
No violations or deviations were identified.
9.
Review of Licensee Event Reports (LER)
During this inspection period, the NRC inspectors performed followup on a
Wolf' Creek LER. The LER was reviewed to ensure:
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Corrective action stated in the report has been properly completed or
work is in progress,
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Responses to the events were adequate,
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Responses to the events met license conditions, commitments, or other
applicable regulatory requirements,
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The information contained in the report satisfied applicable
reporting requirements.
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Generic issues were identified.
The LER discussed below was reviewed:
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482/85-017-00, " Technical Specification Violation-Ruskin Fire
This LER addressed the concerns listed in the Part 21 Report for
Ruskin Fire Dampers which have been reviewed in paragraph 11 of this
inspection report. This LER is closed.
Previous NRC inspection reports contained typographical errors in the LER
review paragraph. These errors and the correct designation are listed
below:
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In paragraph 8 of NRC Inspection Report 50-482/86-28, LER 86-028
should be 86-029.
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In paragraph 9 of NRC Inspection Report 50-482/86-24, LER 85-007
should be 66-007, LER 85-018 should be 86-018, LER 86-073 should be
85-073, and LER 85-001 should be 86-001.
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In paragraph 9 of NRC Inspection Report 50-482/86-24, the DDR for LER
85-064 should be DDR 85-116 rather than DDR 85-117, and the DDR for
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LER 85-065 should be DDR 85-117 rather than DDR 85-121.
No violations or deviations were identified.
10.
IE Information Notice (IN) Followup
The NRC inspector, by review of documents and discussions with licensee
personnel, verified that the ins listed below had been reviewed and
appropriately acted upon by the licensee:
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IN 86-01, " Failure of Main Feedwater Check Valve Caused Loss of
Feedwater System Integrity and Water-Hammer Damage"
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IN 86-14. "PWR Auxiliary Feedwater Pump Turbine Control Problems"
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IN 86-37, " Degradation of Station Batteries"
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In 86-57, " Operating Problems With Solenoid Operated Valves At
Nuclear Power Plants"
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IN 86-61, " Failure of Auxiliary Feedwater Manual Isolation Valve"
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IN 86-77, " Computer Program Error Reporting Handling"
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IN 86-92, " Pressurizer Safety Valve Reliability"
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IN 86-93, "IEB 85-03 Evaluation of Motor-Operators Identified
Improper Torque Switch Settings"
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IN 86-94, "Hilti Concrete Expansion Anchor Bolts"
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IN 86-96, " Heat Exchanger Fouling Can Cause Inadequate Operability of
Service Water Systems"
The licensee has developed a program called the Industry Technical
Information Program to ensure that lessons learned from industry technical
experience are translated into corrective actions, if required, to improve
safety and reliability. This program reviews industry reports as well as
NRC Information Notices for applicability to Wolf Creek. The NRC
inspector's review of this program showed it to be a well organized and
implemented program. The licensee has shown initiative by reviewing
information received even before an IN was issued.
No violations or deviations were identified.
11.
10 CFR Part 21 Report Followup
The NRC inspector, by review of documents and discussions with licensee
personnel, verified that the 10 CFR Part 21 Reports discussed below had
been reviewed and appropriately acted upon by the licensee.
a.
(Closed) Potential Malfunction of Reactor Protection System
-Permissive P-10, dated February 26, 1986.
This 10 CFR Part 21 Report
identified a potential problem concerning the function of the P-10
permissive during operation with one or more of the power range
neutron flux measurement channels inoperative or removed from service.
The P-10 permissive, which is comprised of the four power range
neutron flux channels has two functions. The first function of the
permissive is to enable a manual block of certain functions during
power escalations when at least two of the four power range neutron
flux channels reach a value corresponding to the P-10 setpoint
(typically 10 percent nuclear power). The second function of the
permissive is to provide an automatic reinstatement of these
functions when three of the four power range neutron flux channels
indicate power level is below the setpoint.
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The licensee issued WCGS Standing Order 17 which addressed the
recommendations issued by Westinghouse in their letter notifying the
licensee of the potential malfunction. The NRC inspector reviewed
Standing Order 17, Revision 2, issued December 18, 1986, and verified
the licensee had implemented the Westinghouse recomendations as
appropriate, taking into account Final Safety Analysis Report (FSAR)
commitments and TS requirements.
b.
(Closed)RuskinFireDampers
There were three 10 CFR Part 21 Reports associated with Ruskin Fire
Dampers. These reports were:
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Certain fire dampers may not completely close against normal
system ventilation flow.
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Welds used to join damper sections to damper section and damper
section to mullion plate were 1 inch long on 9 inch centers
rather than 1 inch long on 6 inch centers as specified by design
drawings.
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Some fire dampers were found to have less than the design
clearance between the damper frame and the penetration sleeve.
The subject of closure against system flow is addressed in the
closcout of Open Item 482/8515-01 in paragraph 3.
The NRC inspector reviewed documentation from Underwriters
Laboratories Inc. (April 19,1985) that stated that a weld spacing of
9 inches would not adversely affect the perfonnance of the fire
dampers. This resolves this 10 CFR Part 21.
The licensee inspected all TS fire dampers for clearance problems.
The NRC inspector reviewed all of the WR associated with the dampers
which required either replacement or modification. All work has been
completed. This resolves this 10 CFR Part 21.
c.
(Closed) Emergency Diesel Generator Lube Oil Keep Wann Pumps, dated
March 28, 1986
This item was also tracked as a licensing condition and is discussed
and closed in paragraph 3 of this inspection report.
12. Onsite Event Followup
The NRC inspector performed onsite followup of nonemergency events that
occurred during this report period. The NRC inspector (when available)
observed control room personnel response, observed instrumentation
indicators of reactor plant parameters, reviewed logs and computer
printouts, and discussed the event with cognizant personnel. The NRC
inspector verified the licensee had responded to the event in accordance
with procedures and had notified the NRC and other agencies as required in
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a timely fashion. Engineered safety feature actuations that occurred
during the report period are listed in the table below. Where applicable,
the NRC inspector will review the LER for each of these events and will
report any findings in subsequent NRC inspection reports.
Date
Event
Plant Status
Cause
12/02/86 RHR Pump
Mode 6-
Reduced water level in reactor
Cavitation
vessel
12/05/86 CRVIS*
Mode 6
Spot on detector paper
12/10/86 P-14 Actuation
Mode 5
Low level in level instrument
reference legs on steam
generator (S/G)
12/20/86 Reactor Trip
Mode 1
S/G 10-10 level during
(6%)
power increase
12/22/86 Reactor Trip
Mode 1
Failed S/G feedwater flow
(30%)
instrument
12/22/86 Feedwater
Mode 1
Hi-Hi S/G water level
Isolation
Mode 1
Chlorine monitor component
12/29/86 CRVIS*
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failed
- CRVIS - Control Room Ventilation Isolation Signal
Selected NRC inspector observations are discussed below:
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The NRC inspector performed onsite followup of the RHR pump air
binding event that occurred on November 29, 1986. At the time of the
event, the plant was in Mode 6 with the refueling cavity being pumped
down in order to put the reactor vessel head on the vessel. RCS
recirculation flow was through RHR Pump "A" at approximately 3,000 gpm
and RHR Pump "B" was taking a suction from the RCS and rejecting
flew to the refueling water storage tank (RWST) at approximately
1,500 to 2,000 gpm. An operator had been stationed on the upper edge
of the refueling pool to monitor level. A 1/2 loop level meter was
valved out-of-service and a tygon hose used for level indication
which connected the drain line on RCS loop 1 crossover leg to a
pressurizer relief loop seal drain was not valved in. At
approximately 1:48 p.m. CST, both RHR pumps lost suction and were
secured. The "B" RHR pump was restarted on RCS recirculation.
However, flow could not be increased above 1,000 gpm until its
suction was vented, approximately 3:15 p.m. CST.
It took until
3:55 p.m. CST to get "A" RHR pump filled, vented, and on RCS
recirculation.
Refueling pool level was raised a few feet and then
the pumpdown was restarted, after the tygon level tube and the 1/2
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loop level meter were placed in service. At approximately 4:34 p.m.
CST, "A" RHR pump again cavitated and was secured. The pump was
vented and returned to service at 10:11 p.m. CST. The refueling pool
was again flooded and subsequent letdown was performed using
chemical and volume control system (CVCS) letdown at approximately
100 gpm. Ho further problems occurred. There was no observed
increase in RCS temperature as a result of this event. The licensee
wrote Wolf Creek Event Report (WCER) No. 86-82, dated November 29,
1986, to initiate corrective action.
Events involving loss or degradation of RHR capability have been
reported in five NRC IE Information Notices.(80-20, 81-09, 81-10,
82-17, and 82-45), one NRC IE Bulletin (80-12), four INPO significant
operating experience reports (SOERs) and 14 INP0 significant event
reports (SERs).
In addition, IN 86-101 " Loss of Decay Heat Removal
Due To Loss of Fluid Levels In Reactor Coolant System," was issued on
December 12, 1986. NRC Inspection Report 50-482/86-24, paragraph 8,
described a previous RHR pump cavitation or. air binding event at this
facility. Plant Procedure SYS EC-200, Revision 6 " Changing Level in
the Spent Fuel Pool or Refueling Pool," was used to lower level in
the refueling pool on November 29, 1986. This failure to have an
adequate procedure in accordance with TS 6.8.1, Regulatory Guide 1.33,
and ANSI N18.7-1976/ANS-3.2 is an apparent violation (482/8634-01).
o
Wolf Creek Event Report 86-91
On December 20, 1986, with irradiated fuel assemblies in the spent
fuel storage pool, the storage pool water level was accidentally
lowered approximately 15 inches below TS 3.9.10.2 limit of 23 feet of
water over the top of the irradiated fuel assemblies. The event was
caused when a hose used for adding water to the storage pool was left
with its open end' suspended approximately 2 feet into the storage
pool. After the addition of water to the storage pool, the valve
lineup for the filling operotion was not isolated and water siphoned
from the storage pool to a recycle hold up tank until the end of the
hose was uncovered.
The licensee took immediate action and refilled the pool within the
TS action stateirent time limit. To prevent a reoccurrence, the
licensee is revising Procedure SYS HB202, Revision 3, " Waste
Evaporation Cooldown and Pumpdown," so that, af ter using this method
for filling, the hose is removed from the storage pool and the valve
lineup is returned to a configuration that would prevent siphoning of
the storage pool.
13. Plant Startup From Refuel _ing
The NRC ins]ectors observed selected portions of the plant startup,
following t1e first refueling outage.
Portions of the activities
discussed below were observed:
'.
,
-15-
o
Mode changes performed in accordance with TS requirements and the
procedures listed below:
GEN 00-001, Revision 6, " Mode 5-F111 and Vent of the RCS"
.
GEN 00-002, Revision 9, " Cold Shutdown to Hot Standby"
.
GEN 00-003, Revision 12. " Hot Standby to Minimum Load"
.
GEN 00-004, Revision 8 " Power Operation"
.
o
Control rod testing performed in accordance with Surveillance
Procedure STS RE-007, Revision 2. " Red Drop Time Measurements"
RXE 01-002, Revision 9 g performed in accordance with Procedure
Low power physics testin
o
Reload Low Power Physics Testing"
The NRC inspector performed post maintenance walkdowns of the coolant
charging system and the emergency diesel generators.
The NRC inspector's review of core power physics test data will be
documented in a future inspection report.
No violations or deviations were identified.
14.
Thermal-Hydraulic Anomaly
On December 30, 1986, while operating at 100 percent reactor power,
licensee personnel confirmed the existence of a thermal-hydraulic anomaly.
This anomaly was first discovered at the Callaway Plant on November 17,
1986, and is discussed in Callaway, NRC Inspection Report 50-483/86-20.
The anomaly is characterized by random increases / decreases of up to
0.5 percent on individual power range detectors, up to 5.8 percent on
reactor vessel level indication (RVLIS), random increases of up to
1.2 percent on individual loop flows, and random decreases up to 1.2
percent on incore thermocouples.
In response to the ancmaly, the licensee contacted Westinghouse who
provided an evaluation. Westinghouse stated that the observed deviations
were within the limits of the WCGS TS and within the bounds of the WCGS
FSAR. There is no evidence of any loose part flow blockage, or internal
abnormality. The NRC inspectors observed the strip chart recordings of
the selected plant parameters, discussed the data with the Callaway NRC
inspectors, and followed the licensee's actions.
The cause of the thermal-hydraulic anomaly has not been determined. NRC
and the licensee are continuing their investigation of this anomaiy. This
itera is open pending further NRC review (0 pen Item 482/8634-03).
No violations or deviations were identified.
_
_
_
'
-
.
-16-
15.
Inservice Inspection (ISI) of Safety Valves
Duringthefirstrefuelingoutage(October 16toDecember 20,1986),the
licensee performed inservice inspection of selected safety / relief valves
per subsection IWV-3500 of ASME Code Section XI, 1980 Edition. The
licensee implemented the ISI program September 3,1985, and uses this date
as the startup date for scheduling ISI inspections. Based on this startup
date and the valve testing schedule delineated in subsection IWV-3500 of
the ASME Code Section XI, the licensee elected to not test any of the
three pressurizer safety valves, BB 8010A, BB 80108, and BB 8010C, during
this refueling outage. The setpoints for the pressurizer safety valves
were last tested by the vendor on the dates shown below:
Valve
Date Tested
BB 8010A
April 25, 1978
BB 8010B
April 26, 1978
BB 8010C
April 20, 1978
Pending further NRC review of this situation where the pressurizer safety
setpoints have not been retested for an interval of time in excess of 8.5
years, this is an open item (482/8634-04).
'
No violations or deviation; were identified.
16. Open Items
Open items are matters which have been discussed with the licensee, which
will be reviewed further by the inspector, and which involve some action
on the part of the NRC or licensee or both. Open items disclosed during
the inspection are discussed in paragraphs 14 and 15,
17. Exit Meeting
The NRC inspectors met with licensee personnel to discuss the scope and
findings of this inspection on January 6,1987. The NRC inspectors also
attended entrance / exit meetings of the NRC region based inspectors
identified below:
Inspection
Area
Inspection
Period
Inspector
Inspected
Report No.
12/9-12/86
R. Stewart
Surveillance
86-29
12/01-05/86
G. Pick
Surveillance
86-32
12/01-05/86
J. Kelly
Security
86-33
12/15-19/86
R. Mullikin
Followup
86-34
.
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