IR 05000482/1984027
| ML20140D616 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/10/1984 |
| From: | Bruce Bartlett, Bundy H, Carpenter D, Guldemond W, Martin L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20140D583 | List: |
| References | |
| 50-482-84-27, NUDOCS 8412180582 | |
| Download: ML20140D616 (21) | |
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APPENDIX B
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U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
NRC Inspection Report:
50-482/84-27 CP:
CPPR-147 Docket: 50-482
. Category: A2 Licensee:
Kansas Gas and Electric Company (_KG&E).
Post Office Box 208 Wichita, Kansas 67201 Facility Name: Wolf Creek Generating Station (WCGS)
Inspection At: Wolf Creek Site, Coffey County, Burlington, Kansas Inspection Conducted: August 10 to October 11, 1984
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[B. L. Bartlett, Rdtident Reactor Inspector, Yh Inspectors:
- -AV- $F Date Operations, Wolf Creek Task Force (pars.
2, 8, 9A, 98, 9C, 9D, 9E, 11, and 12)
//- Af -WV H. F. Bundy, Resident Reactor Inspector, Date Operations, Wolf Creek Task Force (pars. 2, 3, 4, 7, 8, 9A, 98, 9C, 9D, 9E, 10,11,and.13).
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D. R. Carpenter, Resident Reactor Inspector, Date Operations, Reactor Project Section IA (~ pars. 7 and 9A).
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'W. G. Guldemand, Chi 6f, Resident Inspection Date
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Program, Wolf Creek Task Force (pars. 2, 5, 6, and 8)
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8412180582 B41212
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PDR ADOCK 05000482 l
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Approved:
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-W." E. Guldemond ef,' Resident Inspection Dare [ '
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Program, Wolf reek Task Force l,
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LYE'. Martin,'
tlon Chief, Wolf Creek patpf
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Task Force
Inspection Summary Inspection Conducted August 10 to October 11, 1984 (Report 50-482/84-27)
Areas Inspected: Routine unannounced inspection including site tours;~ reactor '
coolant pump seal leakage; plant procedures; IE Circular followup; Safety Evaluation Report review; conduit separations; quality concern inspection activities; preoperational testing; emergency preparedness; event followup; comparison of as-built plant to FSAR description; and operations followup of training on modification. The inspection involved 491 inspector-hours onsite by four NRC inspectors.
Results: Within the twelve' areas inspected, two violations were identified (failure to operate plant by procedure, paragraphs 9A and 98) and four open.
items were. identified (paragraphs 3, 4, 9E and 138).
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DETAILS
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1.
Persons Contacted Principal Licensee Personnel
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- F.T. Rhodes,PlantMdnager
- R. M. Grant, Director-Quality C. A. Snyder, Quality First Manager 0. Maynard, Licensing Supervisor
- M. G. Williams, Supt. of Regulatory, Quality, and Administrative Services
- F. D. McLaurin, Asst. Startup Manager
- K. Ellison, Startup Technical Support Supervisor G. Sausman, Hot Functional Test Director C. Landstrom, System Startup Engineer D. Kinoshita, System Startup Engineer F. Faist, Asst. Lead Test Supervisor
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T. Mitchell, Hot Functional Test Director C. Alderson, Hot Functional Test Director J. Molnar, Hot Functional Test Director R. Gass, System Startup Engineer B. Weber, System Startup Engineer
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J. Behen, System Startup Engineer J. Pickett, System Startup Engineer
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l J. Zell, Supt. of Operations
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H. Campbell, System Startup Engineer i
R. Hotstream System Startup Engineer l
- C. J. Hoch, QA Technologist
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- W. M. Lindsay, Quality Systems Supervisor
- R. L. Stright, Licensing
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Other Personnel
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M. P. Crimboli, Westinghouse Reactor Coolant Pump Field Engineer l
- R. P. Denise, NRC, Wolf Creek Task Force Director
- R. Smith, NRC, Reactor Inspector
- B. Breslau, NRC, Reactor Inspector Other licensee and contractor personnel were also contacted during the l
course of this inspection activity.
- The above identified personnel attended the exit meeting held on l
October 11, 1984.
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' Site Tours The hRC' inspectors toured the site at various times during the inspection period. ' Ongoing construction and test activities were observed to ensure conformance to applicable requirements or procedures. Areas were inspected included:
Housekeeping
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Fire Protection
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Logbook Entries
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Maintenance Activities
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Tag Outs
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Temporary Modification Control
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It was observed that housekeeping was marginally satisfactory for a plant under construction; however, it would have been considered unsatisfactory for an operating plant. An example observed during one tour on September 20, 1984, was a mop head, coil of wire, and various other items of debris lying at the base of the stairs by the elevator in containment. A substantial improvement was observed following the identification of concerns to plant management.
Balancing of D Reactor Coolant Pump in accordance with Wolf Creek Work Request 11656-84 dated September 1, 1984, was observed. The.instruc-
-tions in the work request appeared adequate with consideration to the training of the technicians performing the work. The shaft vibration was observed to decrease from 14 to 6 mils after balancing.
Shift' supervisor and supervising operator turnover checklists were reviewed on numerous occasions a'.: no problems were identified. Standing
Order No. 31, Rev. 4 dated August 29, 1984, was found in the control room i
book with no review initials by the supervising operators.
..ie operations i
superintendent was informed. On September 9, 1984, Standing Orders 36 I
and 37, both dated June 29, 1984, were listed in the index dated
August 29, 1984, but were not in the control room book. The' operations
coordinator was informed. No similar problems were observed in sub -
j sequent inspections.
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On September 17, 1984, a significant leak was observed by an NRC inspector
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on pressure transmitter CHB PI-1 and reported to the control room for i
isolation and repair. Damaged flexible conduits to instruments TLF 18 and SJ HV015 were observed on September 20 and 21, 1984, respectively, and j
reported them to maintenance for repair.
l No violations or deviations were identified.
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Reactor Coolant Pump Seal Leakage-The ' licensee observed excessiseal leakage on' the Number 1 seal for the C and D reactor coolant pumps at the beginning of hot functional testing.
The resident reactor inspector (RRI)iinspected the defective seals after-removal and observed evidence of-rubbing on the inner face of both.the seal. runner and ring for each seal. The: analysis of why this occurred has not been completed. The seal leakage returned to expected values
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after-replacement. This will be tracked as an open item.
(50-482/8427 02)
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No violations or deviations were identified.
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Plant Procedures All nuclear department policies and a substantial number of plant administrative procedures were reviewed to establish compliance with
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American National Standard ANS 3.2-1976.
It was noted that Policy I.3.0 contains no requirement for periodic review and update of the manual.
This was brought to the attention of the Quality Director and a procedure change notice was issued to require a review of policies
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every two years.
Policy III-36, dated April 1984, Section 36.4.3.1 requires identification of measuring and test equipment by labels on markings indicating the
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calibration status.
It was observed that this requirement is not being i
implemented for installed plant instruments in the control room and was brought to the attention 'of the Quality Director.
He requested an internal surveillance. which resulted in issuance of a KG&E violation.
The plant response to this violation was that the exception had been taken to the calibration status labeling requirements of ANS 3.2-1976 for installed plant instrumentation.
Prior to the exit meeting conducted on October 11, 1984.. the NRC considered this failure to label installed plant instrumentation as to calibration status as an item of noncompliar.ce. This had been conveyed
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to the attention of the resident staff until the exit meeting.
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As a result of the new informatior presented at the exit meeting, a
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meeting was held between the Chief of the Wolf Creek resident program, the Wolf Creek plant manager, and members of the plant manager's staff.
The purpose of the meeting was to clarify the status of installed plant inatrumentation relative to the requirements for measuring and test equipment contained in ANS 3.2-1976.
As a result of this meeting the
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following points were-agreed upon.
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The accuracy of installed plant instrumentation ~1s' not necessarily:
the same as portable test instrumentation; however, that-accuracy-imust be such as to allow verification of conformance to Technical:
Specification surveillance operability requirements if relied upon
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to do so.'
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The range of. installed plant instrumentation need not conform to however,quirements associated with portable test instrumentation; range re it must be sufficient to demonstrate conformance to Technical' Specification surveillance operability mquirements' if
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-called upon to do so.
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Accuracy and range. adequacy of installed plant instrumentation will be verified by.the licensee during initial operational surveillance
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For those cases where installed plant instrument capabilities are inadequate,. alternate instrumentation will be relied upon until appropriate resolution is obtained.
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Calibration status for installed plant instrumentation relied upon to demonstrate conformance with Technical Specification surveillance-operabilit requirements will ue available through computer call-up to the operations staff.
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A list of instruments and surveillance procedures for which they are used will be prepared to allow timely performance of the-evaluation specified by ANS-3.2-1976 for instrumentation found inoperable or out of calibration.
Based on these understandings, the Chief of the Wolf Creek resident program agreed that no citation would be issued for the observed lack of calibration-status markings.
Items 3 and 5 above will be tracked as a single open item.
(482/8427-03)
Policy III-29, dated February 1984 Section 29.5.1 implied that an employee must first contact his management for resolution of a problem prior to contacting the NRC for assistance.
This was brought to the attention of the Quality Director and a procedure change notice was issued to clarify.
an employec's right to take quality concerns directly to the NRC without fear of. recrimination. This policy was covered by the KG&E legal-staff in meetings with all' site managers.
It was stressed that management does not have to be informed when an employee desires to refer a concern to the NRC.
Policy III-38, dated June 1984, required biannual audits on all quality program elements in Section 38.6.4.1 and biennial audits in Sec-tion 38.6.4.2.
This possible anomaly was referred to the Quality Director to establish whether_ the requirements are twice a year or every two years.
A procedure change notice was issued to clarify the requirements as every two years.
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. The following administrative. procedures were reviewedi o
.ADM 01-001, Rev. 10-ADM 01-002, Rev. 11 ADM 01-003L Rev. 1 i
ADM 01-004. Rev. 2-ADM 01-005, Rev.
1-ADM 01-006, Rev. 1
.ADM 01-007. Rev.- 2 ADM 01-019, Rev.
~ADM 01-020,~Rev. 1 L-4ADM 01-021,-Rev.
-ADM 01-022, Rey.
ADM 01-023 Rev. 2-ADM 01-025 Rev.
ADM 01-029 Ren
ADM 01-030.-Rev. l'
ADM 01-031. -Rev. 4 ADM 01-033 Rev.
ADM 01-034, Rev. 4-
'ADM 01-037. Rev.
ADM 01-039. Rev.. 1 ADM 01-041,1Rev. 2
'ADM 01-042. Rev.
ADM 01-047 Rev.
O ADM 01-048, Rev. 2 ADM 01-049.- Rev.
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- ADM 01-050, Rev. O ADM 01-051,'Rev. O
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ADM 01-053 'Rev.
ADM 01-057, Rev.
<ADM 01-058.-Rev. 0
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ADM 01-059. Rev. 2
'ADM 01-061,'Rev.
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'ADM 01-063. Rev. O ADM 01-106, Rev.
ADM 02-001, Rev.
ADM 02-002, Rev. 2 ADM 02-003. Rev.
-ADM 02-004, Rev.
ADM 02-005, Rev. 1 ADM 02-008. Rev.
ADM 02-010. Rev.
ADM 02-012, Rev. 0
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ADM 02-020. Rev.
'ADM 02-021. Rev.
ADM 02-023. Rev. 0
.ADM 02-030. Rey. O ADM 02-100. Rev.
6-ADM 02-101, Rev. 7
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ADM 02-102, Rev.
ADM 02-103.-Rev.
ADM 02-104. Rev. O l
ADM 02-110, Rev.
3-ADM 02-200, Rev.
ADM 02-400, Rev. 4 Of the administrative procedures reviewed. no significant problems were
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l discovered. ~ However, this review is ongoing.
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No violation or deviations were identified.
5.
IE Circular Followup A.
The following IE Circulars were reviewed, were determined to be not applicable to the Wolf Creek facility, and are considered closed:
78-01 - Loss of Well Logging Source
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78-10 - Control of Sealed Sources in Radiation Therapy
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78-11 - Recirculation M-G Set Overspeed
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79-01 - Administration of Unauthorized Byproduct Material
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to Humans 79-06 - failure to Use Syringe and Bottle Shields in
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Nuclear Medicine
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.79-07 - Unexpected Speed Increase of Reactor Recirculation
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~ M-G Set Resulted in Reactor Power Increase
.79-14 -' Unauthorized Procurement and Distribution of Xenon-133
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80-06 - Control and Accountability Systems for Implant
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Therapy Sources 80-08 - BWR Technical Specification Inconsistency
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80-19 - Noncompliance With License Requirements for Medical
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Licensees 80-24 - AECL Teletherapy Unit Malfunction
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For the following IE Circulars '(IEC) the inspector determined that the -licensee had taken those actions necessary to resolve the identified issues.: These IECs are considered closed:
78-12 - HPCI Turbine Control Valve Lift Rod Bending: This
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IEC described a Terry Turbine lift rod bending problem unique to certain turbines and control systems. This equipment is not employed at Wolf Creek.
79-05 - Motsture Leakage in Stranded Wire Conductors: This
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IEC described a potential problem with leakage of-moisture between the wires and insulation on stranded wire conductors exposed to a differential pressure under accident conditions. The licensee has reviewed equipment inside primary containment and the main steam tunnel for susceptibility to this problem. Such equipment has' been -included in the environmental qualification program and has either been qualified or is still in the qualification process.
79-10 - Pipefittings Manufactured from Unacceptable Material:
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This IEC described certain lot number fittings manu-factured by Tube Turn Company which were made from a steel possessing an incorrect carbon content. The licensee, through its constructor and architect /
engineer has determined that the discrepant fittings were not employed at Wolf Creek.
79-11 - Design / Construction Interface Problems: This IEC
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described interface problems between licensees and,
architect / engineers at multi-unit facilities which resulted in misorientation of reactor vessel internals and other su) ports. The contents of this IEC were reviewed wit.1 the Wolf Creek architect / engineer and NSSS vendor and the conclusion was reached that sufficient controls exist to prevent this type of problem from occurring at Wolf Creek.
.79-12 - Potential Diesel Generator Turbocharger Problems:
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This IEC described a post operation turbocharger lubrication problem with diesel generators supplied by EMD.
EMD diesels are not emp1pyed at Wolf Creek.
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~ 79-13
~ Replacement of Diesel Fire Pump Starting Contactors:
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This IEC described a problem with certain ~ starting contactors on diesel fire pumps supplied by Cummins.
Cunmiins diesel fire pumps are not employed at Wolf Creek.s
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79-16 -l Excessive Radiation Exposure to Members of the
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-General Public and a Radiographer: This IEC described a problem with rt.diographic source control and as such is not applicable to Wolf Cnek.
79-17 - Control Problem in S8-12 Switches 'on General Electric
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Metal Clad Circuit 1 Breakers: This-IEC described a manufacturing problem involving excessive tolerances on the subject switches. The Itcensee determined that none of.the subject switches is employed in a safety-related application at Wolf Creek.
6.
Safety Evaluation Report Review The following list of open items was identified by review of the Wolf Creek Safety Evaluation Report including Supplements 1 through 4 and the Callaway Safety Evaluation Report.. These items have been; categorized as follows for timeliness of closure:
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Category 1 - Must be closed prior to fuel load.
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Category 2 - Long term issue or dated completion
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Category 3 - Must be closed prior to startup following
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the first refueling outage.
Category 4 - Must be closed prior to initial criticality.
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Category 5 - Must be closed prior to 5% of fuel power.
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NUMBER SUBJECT CAT 84-00-01 SER 2.3.3 Upgrade meteorological measurement prog' ram
84-00-02 SER 2.4.2.5 Establish admin controls on ESWS pumphouse pressure doors
84-00-03 SER 2.5.6.8 Evaluation of main dam seepage
84-00-04 SER 3.6.1 Augmented ISI of HE piping in containment pene-tration break exclusion region
84-00-05 SER 3.9.3.2 Audit of pump and valve operability assurance program
84-00-06 SER 3.10 Onsite audit to assess adequacy of SQRT program
84-00-07 SER 3.11 Submittal of EQ status
84-00-08 - SER 3.11 Staff EQ audit -
1-84-00-09 SER 4.2.3.2(2) Licensee verification of reload cladding collapse analysis bounds
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84-00-10 ' SER 4.2.5(2) Perfom online fuel failure monitoring;and
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Post irradiation surveillance l'
.84-00-11 SER 8.1 Staff audit of installation and arrangement of
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electrical equiunent.
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84-00-12 SER 9.5.1.1 F.>. control and sectionalization valves not electrically supervised under admin control
84-00-13 SER 9.5.1;2 Installation of 3-hour fi m penetration' seals
84-00-14 SER 9.5.1.2 Installation of fire protection on cable tray supports
84-00-15 SER 9.5.1.4 Installation of RCP oil collection system -
seismic cualification'
L 84-00-16-Assess acequacy of diesel oil. day tank collection dike.
capacity with tank' auto refill
84-00-17 SER 9.5.4.1 Prior to startup the licensee will implement a training and retraining program for personnel mspon-sible for maintenance and availability of D/G
84-00-18 SER 12.5.1 Qualify health physics supervisor to
ANS 3.1-1979
84-00-19 SER 12.5.1 Admin controls to ensure either health -
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physicists or health physics supervisor availability
F 84-00-20 SER 12.5.4 Instruction to female workers per Reg.
Guide 8.13
1 84-00-21 SER 13.1.2 Implementation of program management group
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for operations / engineering interface
84-00-22 SER 13.1.2 Nuclear service branch staffing
84-00-23 SER 13.1.3.1 Maintenace department staffing
84-00-24 SER 13.l.3.1 Mechanical supervisor training in nuclear codes and standards
84-00-25 SER 13.1.3.2 Support department staffing
.1 84-00-26 SER 13.1.3.2 Training department personnel licensing
.1 84-00-27 SER 13.1.3.3 Technical support department staffing.
I 84-00-28 SER 13.1.3.4 0)erations department staffing / strike support 1 84-00-29 SER 13.1.3.4 S10 licenses for operations and surveillance coordinators
84-00-30 SER 13.1.3.4 Shift consultants on shift for 1 year
84-00-31 SER 13.1.3.4 Issue a management directive emphasizing duties and authority of shif t supervisor. Vice President to review annually
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84-00-32 SER 13.2.2 Annual nonlicensed employee training
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84-00-33 SER 13.4 per 07371.B.I.2 Implement 7 member ISEG
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spending time on backshifts weekly
84-00-34 SER 13.5.2.1 Review of applicant's plant for emergency operating procedures per TAP 1.C.B or 0737 I.C.1
84-00-35 SER 14 Repeat natural circulation test for training per TAP 1.G.1
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I 84-00-36 SER 22.1/0737 I.A.2.3 Training supervisor and engineer SR0 licensed
84-00-37 SER 22.1/0737 I. A.2.3 Operator license program instructors participation in requalification program
84-00-38 SER 22.1/0737 I.A.2.3 Operator training instructors participation in periodic onshift assignments, review of operating and emergency procedures as they are developed, and instructor certification program
84-00-39 SER 22.1/0737 I.C.7 Vendor review of emergency operating procedures
84-00-40 SER 22.1/0737 I.D.1 Correct human factors deficiencies for panels RL013 and RLO14 per 6/29/82 KG&E submittal to NRC
84-00-41 SER 22.1/0737 I.D.1 Implement human factors deficiencies corrections per APP. E of SNUPPS TER
84-00-42 SER 22.1/0737 I.D.1 Report on Human factors deficiencies in APP. D to SNUPPS TER
84-00-43 SER 22.1/0737 I.D.1 Submit resolution of human factors deficiencies in APP. C of SNUPPS FSAR 120 days prior to licensing
84-00-44 SER 22.1, SSER4 22.1/0737 II.B.3 Prior to exceeding 5%
power, have installed and operational post accident accident system
84-00-45 SER 22.1, SSER4 22.1/0737 II.B.3 Prior to exceeding 5% power, submit core damage assessment procedure
84-00-46 SER 22.1, SSER4 22.1/0737 II.B.3 Prior to exceeding 5%
power, demonstrate applicability of procedures and instrumentation in the post accident water chemistr.v and radiation environment
84-00-47 SER 22.1, SSER4 22.1/0737 II.B.3 Semi-annual operator retraining on post accident sampling systems
84-00-48 SER 22.1/0737 II.B.4 Complete training for mitigating core damage prior to fuel load
84-00-49 SER 22.1, SSER4 22.1/0737 II.E.1.1 Establish the security of the CST outlet manual isolation valve (forAFW)
84-00-50 SER 22.1, SSER4 22.1/0737 II.K.1 Review all valve positions, positioning requirements, positive controls, and related test and maintenance procedures to assure proper ESF functioning
a 84-00-51 SER 22.1, SSER4 22.1/0737 II.K.1 Review and modify as required procedures for removing and restoring safety-related systems from service to assure operability status is known
84-00-52 SER 22.1, SSER4 11.1/0737 III.A.I.2 Post implementation appraisal of ERFs
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84-00-53 SER 22.1, SSER 22.1/0737 III. A.2 Successful completion-of a full scale emergency exercise
84-00-54 SER 22.1/0737. III.D.3.3 Implementation of a training program for airborned radioiodine monitoring
84-00-55' SER APP C Item A-1 Perform a steam generator water hamer test using normal plant procedures
84-00-56 SER APP C Item A-43 Demonstrate that the containment sump will perform as expected following a LOCA including housekeeping and emergency procedures
84-00-57 SER APP C Item A-44 Procedures and training for station blackout
84-00-58 SSER1 3.7.4 Installation of discrete response spectrum recorder at the containment foundation with the capabil-ity to provide immediate control room indication
84-00-59 SSER 1, 4 7.3.2.2 Installation of modified P-4 interlock testjacks
84-00-60 SSER 1, 4 7.3.2.7 Installation of the automatic indi-cation of the block signal initiating AFW on loss of both main feedwater pumps
84-00-61 SSER 1, 4 7.3.2.8 Installation of 2 of 4 logic on HI steam generator level to isolate main feedwater flow I
84-00-62 SSER 1, 4 7.5.2.2 Installation of the modification to the bypass indication occurring when a valve leaves its required position
84-00-63 SSER1 Sec 18 Item 2 Obtain computer codes and training employees in plant nuclear-thermal-hydraulic behavior
84-00-64 SSER2 3.9.6 Incorporation into T S of allowable pressure isolation valve leakage
84-00-65 SSER2 4.2.3.1(10) Surveillance of control rods
84-00-66 SSER2 7.5.2.3 Establish a Reg. Guide 1.97 implemen-tation schedule
84-00-67 SSER2 13.3.2.5 Implementation of prompt notification ar;d alert system or compensatory measures prior to fuel load
84-00-68 SSER2 13.3.2.6 Monthly, quarterly, annual test of communication equipment
84-00-69 SSER2 13.3.2.6 Quarterly communications drills
84-00-70 SSER2 13.3.2.15 Annual emergency response training for offsite agencies
84-00-71 SSER2 13.3.2.15 First aid training for first aid team members
84-00-72 SSER2 13.3.2.15 Annual retraining for emergency response personnel
84-00-73 SSER2 13.3.2.16 Annual independent review of emergency preparedness program and report to NRC
84-00-74 SSERI 13.3.2.16 Quarterly update of telephone numbers in emergency procedures
84-00-75 SSER3' 4.5.2 Replace control rod drive guide tube support pins prior to fuel load
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84-00-76 : SSER 3, 4.7.3.2.9 Installation of additional alarms and
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indicators'of the plant computer.
'84-00-77 SSER3 --7.5.2.1 Install single failure proof RCS hot and cold leg temperature indication on the auxiliary shutdown panel.
84-00-78 SSER3 7.6.7.1 Installation.of interlocks for RCS
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' pressure control during low temperature operation
84-00-79 SSER3 ~ 7.6.7.3 Implement connitted-to design for termination of boron dilution-seismically and environ-
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.' mentally qualify SR and IR-
84-00-80 SSER3 ' 8.2.2.1 - Adequacy of fire protection of trans-fonners XNB01 and XN802
84-00-81 SSER3 8.2.2.3 Resolve re?' ability of load sequence for on and off site power 1-84-00-82 SSER3 8.3.1.2 Submit analysis and test results.of voltage drops / degraded grid voltage
84-00-83 SSER3 8.3.1.6 Establish independence of load sequences
84-00-84 SSER3 22.1/0737 II.D.1 Staff review of applicant submittal on PORV and block valves
84-00-85 SSER4 ' 3.9.3.4. Applicant investigation into stiff pipe clamp use pursuant to IEN 83-80
84-00-86 'SSER4 3.11.3.1 Address compliance with Reg. Guide 1.97 for Cat. I and 2 instruments supplied from non-1E power supplies
84-00-87 SSER4 3.11.3.1 Staff review of justification for exclusion of certain 0737 items from Reg. Guide 1.97 requ'irements
84-00-88 SSER4 3.11.3.1 Staff review of mechanical equipment
. qualification files
84-00-89 SSER4 3.11.3.2.2 Subinittal of mechanical equipment qualification information-1 84-00-90 SSER4 3.11.3.3.5 Implement a program to detect degradation of safety-related motor cables
84-00-91 SSER4 '3.11.3.3.5 Implement a surveillance program for safety-related non-motor cables inside containment
84-00-92 SSER4 3.11.3.4 Submittal of< analyses for safe operation with items not fully environmentally qualified
84-00-93 SSER4 3.11.3.4 Complete replacement or retest of unqualified equipment or justify interim operation
84-00-94 SSER4 3.11.3.4 Resolution of outstanding concerns on Rockbestos cable qualification
84-00-95 SSER4 3.11.3.4 Provide justification for interim operation with mechanical equi > ment for which qualifi-cation has not been fully esta)11shed
84-00- % SSER4 3.11.4.3 Applicant review / upgrade of qualification
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documentation
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'84-00-97 SSER4-3.11.4.3 Submit a description of the approach used to assure all equipment items are installed in a manner which does not invalidate their qualification
84-00-98 SSER4 3.11.4.3 Demonstrate that excluding pre-aging from the qualification program does not invalidate qualification status
84-00-99 SSER4 3.11.4.3 Seal the wire entrance cavity on limitorque SMB-000 operators or demonstrate satis-factory unsealed perfonnance
84-00-100 SSER4 3.11.4.3 Submission of qualification information-on Marathon 1600 series terminal blocks
84-00-101 SSER4 5.4.2.2 Inclusion of steam generator inservice inspection requirements in T S
84-00-102 SSER4 7.2.2 Staff review of applicant's response to Generic Letter 83-28
84-00-103 SSER4 7.2.2.2 Implementation of 18 month calibration of FIS for RTO 8/P loops
84-00-104 SSER4 9.5.7 Review D/G test results to ensure temp.
i limits on rockor arm lube oil system are not exceeded
84-00-105 SSER4 13.3.2.1 Provide letters of agreement with offsite support agencies 180 days prior to fuel load
84-00-106 SSER4 13.3.2.7 Distribution of public information brochures prior to fuel load
84-00-107 SSER4 13.3.2.8 Establish an assessment with a first order station of the national weather service to supply weather information
84-00-108 SSER4 17.3 Staff review of applicant supplied infor-mation on Ray Miller material Bulletin 83-07
84-00-109 SER 3.8.1 Perform a containment SIT at 1.15 time:;
design pressure
84-00-110 SER 3.9.2.1 Preservice and Preoperational testing of snubbers per June 9, 10, 1981, meeting minutes summary
84-00-111 SER 3.9.2.2 Perfonn a post lift vessel internals inspection for evidence of vibration, wear, and loose parts
84-00-112 SER 3.9.3.4 Staff review of applicant response to IED 79-02 on pipe support baseplace flexibility
84-00-113 SER 4.4.4.1 Staff verification of acceptable accommodation of Rod Bow penalties
84-00-114 SER 6.2.3 GDC-57 Compliance-demonstrate the need for continuous operation of the minipurge system by plant operating experience
84-00-115 SER 6.3.3 Control power lockout for the following valves-accumdschg(open);SIcoldlegdschg(open);SImini-flow IS0(closed); SI hot leg dschg(closed); RilR hot leg dschg(closed); RiiR cold leg dschg(open)
84-00-115 SER 6.3.3 Locking of two manual isolation valves-RWST suction to ECCS(open); RilR test return to RWST(closed)
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84-00-116 SER 6.3.3 Locking of two manual isolation valves RWST suction to ECCS(open); RHR test return to RWST(closed)
84-00-117 SER 7.2.2.5 Inclusion into T S, a test requirement for both the UV and SHUNT trip breaker trip functions
84-00-118 SER 7.2.2.6 Incorporation of lead / lag time constants into T S
84-00-119 SER 7.3.2.1 Verification of ESF reset during preop-erational testing IEB 80-06
84-00-120 SER 7.3.2.3 Resolution of concerns on level measurement errors due to environmental temperature effects on level instrument reference legs IEB 79-21
84-00-121 SER 7.3.2.4 Staff review of safety system setpoint methodology
84-00-122 SER 7.4.1 Installation of safety guide AFW flow indi-cation on the auxiliary shutdown control panel
84-00-123 SER 7.4.3.1 Licensee submittal of information on the capability for safe shutdown following loss of a bus supplying power to instrument and controls IEB 79-27, IEC 79-02
84-00-124 SER 7.7.11.3 Submittal of a plant specific analysis on the affect of a HELB on rod control systems outside containment IEN 79-22
84-00-125 SER 8.3.1.1 D/G load rejection test criteria clarifi-cation for preop
84-00-126 SER 8.3.2.1 Installation of a battery discharge alann on the plant computer in the control room
84-00-127 SER 8.3.3.1.2 Inclusion of related valve motor thermal overload setpoints in T S
84-00-128 SER 8.3.3.3 Installation of color coded tags on non-1E cables in cabinets or panels associated with a single separation group
84-00-129 SER 8.3.3.6 Licensee reevaluation of electrical pene-tration assembly protective device settings and capabilities
84.nn-130 SER 9.1.2 Incorporation of Zone 2 spent fuel storage requirements in T S
84-00-131 SER 9.1.4 Implement interim actions identified in Enclosure 2 to the 12/22/80 genericletter(NUREG-0612)
84-00-132 SER 9.1.4 Implement admin controls on transport of loads lighter than a fuel element over the spent fuel pool
84-00-133 SER 15.3.8 Develop and train on ATWS procedure
84-00-134 SER 22.1/0737 II.B.1 Submittal, review, and approval for operation of RCS vents
84-00-135 SER 22.1/0737 II.D.1 Qualification of relief and safety valve block valves by 6/1/82
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84-00-136 SER 22.1/0737 II.E.1 Implement valve lineup procedures including second checks for AFW following extended outage, testing. or maintenance
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84-00-137 SER 22.1/0737 II.E.1 Submit AFW pump 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> test run data for, evaluation '
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84-00-138 Ser 22.1/0737 II.E.1 Staff confirmation that AFW initiation signals and circuits are safety grade
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~84-00-139 SER-22.1/0737 II.E.4.2 Qualification of containment
' purge isolation valves prior'to their use above Mode 5
.1 84-00-140 SER 22.1/0737 II.F.2. Installation of R.V. level instruments and incore thermocouple / core cooling monitoring systems prior to fuel load
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1 84-00-141 SER 22.1/0737 II.F.2 Staff review and approval of
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inadequate core cooling emergency procedures and control display, prior to license issuance
84-00-142 SER 22.1/0737 III.D.I.1 Submittal of detailed.
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information on integrity of systems outside containment 4 months prior to fuel load
84-00-143 SER 22.1/0737 III.D.1.1 Post implementation of leak reduction program
.3 7.
Conduit Separations t
The NRC inspectors noted that various electrical conduits in the plant
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have hold tags installed indicating violation of electrical separation l
criteria. They expressed a concern to startup management that the
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validity of certain preoperational tests could be compromised as a result of actions to be taken to correct the separation problems. The assistant lead system test supervisor and assistant startup manager
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stated that it had been determined that none of the mentioned hold tags-would affect preoperational test acceptance criteria. The deficiencies will-be corrected under the authority of construction work pemits
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(CWPs).
Each CWP will be reviewed by the appropriate system startup engineer (SSE) to detemine testing status.
If the associated components were previously tested, functional testing will be required after completion of the work.
In these instances, the Joint Test Group will review the retest requirements.
No violations or deviations were identified.
8.
Quality Concern Inspection Activities The'RRIs reviewed and closed several concerns regarding the preoperational
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testing program.
These inspections were performed informally at the request of the concerned parties who requested anonymity. The RRI also discussed the' status of the Quality First investigative activities periodically with the KG&E Quality First Manager.
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No violations or deviations were identified.
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Preoperational Testing
. The following tests were reviewed and selected parts were witnessed except as noted. Significant observations and findings are discussed
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when appropriate:
A.
SU3-8805, Rev. 2. Reactor Coolant S_ystem Hot Preoperational Test (HFT)
The NRC inspector observed that there were some anomalies between 5U3-B805 and supporting procedures and in some instances between supporting procedures. These were identified to the licensee who initiated appropriate resolutions. The licensee also' established a task force to review relevant procedures for. anomalies in advance of testing.
Some of the supporting procedures reviewed by the NRC inspector were:
GEN 00-001, Rev. O, Mode 5 Fill and Vent of RCS
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GEN 00-002, Rev.1. Cold Shutdown to Hot Standby
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SYS B8-201, Rev. 1. Reactor Coolant Pump Operation
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STS BB-011, Rev. O. Reactor Coolant System and Pressurizer
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Heatup/cooldown Surveillance SYS BB-110, Rev. O, Reactor. Coolant System Fill and Vent
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During performance of Section 71 of SU3-BB05, the plant was being
heated up from ambient to a 150 plateau under the control of GEN 00-031. Rev. O, " Mode 5-Fill and Vent of RCS." Step 4.4 of GEN 00-001 had originally required RCS pressure to be maintained between 110 and 120 psig. This step had been changed by Procedure Change Form to ".. 110 to 200 psig." This pressure would not allow main coolant pumps to be run for heatup under Procedure SYS BB-201, Rev.1. " Reactor Coolant Pump Operation " so Step 4.4 of GEN 00-001 was changed again to ".. 110 to 340 psig." The NRC inspector was observing the heatup in the control room and observed the RCS pressure at 364 psig.
Data Sheet 1 of Procedure SYS BB-011 was reviewed and several readings above the 340 psig level were-noted. The NRC inspector infomed the Shift Supervisor and the HFT Test Director of the limit violation. The licensee stopped the test, reduced the pressure, and held an evaluation meeting. The Procedure GEN 00-001, Step 4.4 was again changed to ".. 110 to 425 psig." The test was restarted after appropriate review.- No action was determined to have an impact on the safety of the plant systems.
The limits were not too restrictive.
The various procedures were not properly intergrated before starting SU3-BB05. This failure to follow written approved procedures is an apparent violation.
(482/8427-01) This violation was transmitted to the licensee by previous separate correspondence. (Letter NRC Denise to KG&E Koester
. dated August 28.1984)
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SU3-ALO2, Rev. O Auxiliary Feedwater Turbine-Driven Pump and Valve
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The RRI observed ~during perfomance of this test that Step -7.4.19 called for closing ~ valves HV6, HV8, HV10,'and HV12; however, the
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SSE stopped the auxiliary feedwater pump turbine KFC02 by directing.
4-the operator to press close push button FC-HIS-312A on RL005.' The SSE subsequently stated that he did not know at the time that Step 7.4.19 would teminate auxiliary feedwater-flow to the_ steam
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generators and he was concerned that he was exceeding the original 80; percent narrow range level indication ' limit imposed upon him by the shift supervisor. The SSE did not satisfactorily explain why he teminated the. test in this manner when he had been exceeding -
the shift supervisor's limit on "D". steam generator for 10 minutes -
and on "A" steam generator for 5 minutes. He also stated he was unaware the shift supervisor had raised that limit thus giving him more time in which to complete the test evolution. This was a failure in communications. Additionally, the SSE ' appeared to fail to realize that a test deficiency was required to be written until it was pointed out to him by the RRI. These actions constitute a.
failure to perfom testing in accordance with written test procedures.
This is a violation.
(482/8427-02)
C.
SU3-ALO4, Rev. 2. Auxiliary Feedwater System Water Hammer Test The RRI witnessed the test briefing and subsequent to the briefing asked questions to determine the collective understanding of test precautions and operating limits by test participants. Nobody except the SSE knew the steam generator water level test limits.
Although the operator hesitated when asked the limit on suction pressure for the motor driven' auxiliary feed pump, the shift supervisor correctly stated that it was 8 psig. The RRI expressed a concern to the HFT Director that precautions and operating limits may not be adequately covered in test briefings.
During observations of the test, two unsuccessful attempts were made-to start B motor driven euxiliary feed pump. The SSE indicated an intention to make a third attempt after a 10-minute wait. The RRI infomed him that a longer waiting period was mquired which he confimed by reference to the test procedum. The RRI then asked the operator what the waiting period was and he responded that the SSE told him it was 10 minutes. These observations were discussed'
with the shift supervisor and the HFT test engineer.
Subsequent briefings observed by the RRI appeared adequate. These inspections were followup to the problems described in Item 9.B above.
No violations or deviations were observed.
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There were no significant adverse observations for the following tests:
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SU3-ALOS, Rey. O Auxiliary feedwater Turbine Driven Pump
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Endurance Test SU4-AC01, Rev. O, Initial Turbine Roll
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SU8-0007.1 Rey.1, Hot Functional Plant Performance Test
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(Reviewed test results only)
SU3-BG06, Rev. 2. Chemical and Volume Control System
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SU3-EM03, Rev. O, Accumulator and Safety Injection System
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Check Valve Preoperational Test SU4-SQ02, Rev. O Loose Parts Monitoring System
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SU3-BB13, Rev. O, Pressurizer Relief Valve and PRT Hot
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Preoperational Test S03-8807, Rev. O, Pressurizer Level Control Test
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SU3-BB15A, Rev. O Reactor Coolant System Leak Rate
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SU9-0025, Rev. O, Reactor Coolant System Heat Loss
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SU3-0004. Rev.1. Power Conversion and ECCS System Thermal
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Expansion Test SU3-BB12, Rev. O, Pressurizer Continuous Spray
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SU3-BB08, Rev. O, Pressurizer Heater and Spray
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SU3-BB04, Rev. O, Pressurizer Pressure Control
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SU3-EM02, Rev. 0. Safety Injection Flow Verifications
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(Procedure review only)
SU3-EN02, Rev. 1 Containment Spray System
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SU3-EJ02, Rev. 0. RHR System (Procedure review only)
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No violations or deviations were observed.
E.
SU3-0008, Rev. O, Cooldown From Hot Standby External To The Control Room The primary objective of this test was to verify the adequacy of plant off-normal operating procedures to cool the reactor coolant system to a point where the residual heat removal system can be placed in service.
It was observed that adequate drawings and procedures to support plant cooldown were not available in the auxiliary shutdown room.
A file cabinet was available and the shift supervisor stated the drawings had been ordered.
The shift supervisor was making a list of additional supplies, pump curves, procedures, etc. which would be required.
The RRI will follow up on this item at a later date. This is an open item.
(482/8427-04)
Valve BG-FK-121 was inadverently closed when placed in manual control which caused a momentary low seal injection flow. The operations superintendent stated an operating aide would be placed at the local controller to preclude this from happening in the future.
No violations or deviations were observed.
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. 10.. Emergency Preparedness The entrance interview for the emergency preparedness team inspection was attended by the RRI. Also, an interview hy an inspector with shift operating personnel to determine their response to an accident scenario was witnessed by the RRI.
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No violations or deviations were identified.
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11. Event Followup
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The RRI was placed on the distribution for Wolf Creek Event (WCE) Reports and reviewed selected reports to determine:
Adequacy of operator response
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Adequacy of analysis to determine cause
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Adequacy of corrective action to prevent recurrence
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Among those reviewed was WCE No. 84-73, LOCA Outside Containment.
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The RRI also witnessed certain operating events and response of operating personnel in recovery operations. Among the events were complete loss of electric' Bus NB01 and loss of the startup transformer.
Involved personnel reacted calmly and knowledgeably; however, the RRI observed that operators were slow in referring to the appropriate off-normal procedures to deter-k mine if they had completed all required actions. No reference-to the procedures actually used was made in the operating logs. This concern was expressed to plant management.
No violations or deviations were identified.
12.
Comparison Of As-Built Plant To FSAR Description Inspection was performed to determine, on a sampling basis, that the
"As-Built" plant conforms to commitments contained in the FSAR hy inspecting mechanical and fluid systems to determine that the physical installation is in agreement with P& ids contained in the FSAR and reviewing control and logic instrumentation to determine if it conforms to the description of instrumentation and controls contained in the FSAR. This inspection is incomplete.
No violations or deviations were identified.
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13. Operations Followup Training On Modifications On August 30, 1984, an NRC consultant observed that operating personnel did not understand the significance of an illuminated amber light for the position indication of pressurizer safety valves.
Further inves-tigation revealed that the SSE also did not understand the significance.
After researching the drawings, it was discovered that the illuminated amber light indicated that the safety valve was in a mid-position.
Inasmuch as this was a modification to original design, the incident raised a question regarding the effectiveness of training operations personnel on the effects of design modifications.
Tne RRI discussed this potential problem with the operations super-intende1t who made the following points:
A.
Plant Procedures ADM 01-041, " Plant Modification Request" and ADM 01-031, " Operating Experience Review Program" both have require-ments which would cause followup training and procedure changes to be implemented if appropriate.
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Operations is in' the process of reviewing all procedures for current applicability. Training will be conducted as appropriate.
These measures appear to be adequate to assure the required training is
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accomplished. Cc 11etinn of the ongoing review will be tracked as an open item.
(482/8427-05)
No violations or deviations were identified.
14. 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 3, 4, 6, 9E, and 13.
15. Exit Meeting NRC inspectors met with the licensee personnel on October 11, 1984, to discuss the scope and findings of this inspection.
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