IR 05000321/1998009
| ML20207D329 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/24/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20207D318 | List: |
| References | |
| 50-321-98-09, 50-321-98-9, 50-366-98-09, 50-366-98-9, NUDOCS 9903090304 | |
| Download: ML20207D329 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION 11 Docket Nos: 50-321,50-366 License Nos: DPR-57, NPF-5 Report Nos:
50-321/98-09,50-366/98-09 Licensee:
Southern Nuclear Operating Company, Inc. (SNC)
Facility:
E. l. Hatch Plant, Units 1 & 2 Location:
P. O. Box 2010 Baxley, Georgia 31515
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Dates:
December 13,1998 - January 23,1999 Inspectors:
J. Munday, Senior Resident inspector J. Canady, Resident inspector T. Fredette, Resident Inspector R. Chou, Reactor Inspector (Section E8.1)
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D. Forbes, Radiation Specialist (Section R1.2)
Approved by:
P. Skinner, Chief, Reactor Projects Branch 2 Division of Reactor Projects
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9903090304 990224
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Enclosure PDR ADOCK 05000321 O
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EXECUTIVE SUMMARY
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Hatch Nuclear Plant, Units 1 & 2 NRC Inspection Report 50-321/98-09,50-366/98-09
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This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspection and inspections in the area of Health Physics and Independent Spent Fuel Storage Installation by Region based
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inspectors.
Operations
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A review of the Unit 2 Standby Gas Treatment system identified that the system was
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properly aligned and labeled, procedures were adequate and the associated drawings reviewed were accurate. No longstanding system or component problems were identified. (Section O2.1).
A Non-cited violation was identified for failure to follow a procedure which rendered
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Emergency Diesel Generator 2C inoperabla. A plant equipment operator placed a ventilation damper switch to OFF and defeated support ventilation for the diesel.
(Section O4.1).
Plant equipment operators demonstrated satisfactory knowledge of the Standby Liquid
Control system and the surveillance procedure. This was demonstrated through their
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use of the procedure and suggestions for procedure improvements. Additionally, the
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operators selected the appropriate portions of the procedure to return the system to it's correct standby lineup following the termination of a surveillance test. (Section 04.2).
Enaineerina Engineering support to operations and maintenance was not thorough or detailed as
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evidenced by the root cause analysis of the 2A Emergency Diesel Generator low lube oil temperature problem. The analysis failed to take into account multiple equipment or component failures and their resultant effect on the oil temperature. (Section E2.1)
The design calculations for the spent fuel cask concrete pad and related structures were
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performed well and in accordance with industrial standards and code requirements.
(Section E8.1)
Plant Sucoort The licensee's As Low As Reasonably Achievable (ALARA) pre-planning for a leak
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repair on a Unit 2 Reactor Water Cleanup system heat exchanger drain line valve was not thorough. The dose goal was set at approximately twice the actual dose accumulated; consideration was not given to having appropriate spare tools available so that inappropriate tools would not have to be used; and the decision to not hold a post-job critique pre.sented a missed opportunity to identifying lessons learned to ensure the dose received for future jobs would be ALARA. (Section R1.1).
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Based on observations and surveys during tours, the licensee had effectively labeled,
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controlled, and stored radioactive material as required by 10 CFR Part 20.1904 and effectively posted radioactive material storage areas as required by 10 CFR Part
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20.1902. Contamination control practices reviewed were consistent with licensee procedures and were acceptable. (Section R1.2)
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Report Details
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Summary of Plant Status
Unit 1 cperated at essentially 100 percent rated thermal power (RTP) during this report period.
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Unit 2 operated at essentially 100 percent maximum operating power (MOP) or 98 percent RTP ~
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l. Operations
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Conduct of Operations j
.01.1 General Comments (71707)
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Using Inspection Procedure 71707, the inspectors conducted frequent reviews of '
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ongoing plant operations. On January 6, the inspectors performed a cold weather check '
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i of the intake structure, the diesel generator building, and the fire protection pump house j
while the ambient temperature was below freezing. The inspectors did not identify any
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adverse affects of the below freezing conditions. In general, the conduct of operations was professional and safety-conscious; specific events and observations are detailed in
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the sections below.
i 01.2 Pre-iob Briefina for Ground Fault Troubleshootina Activities (71707) (71750)
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The inspectors reviewed procedure AG-MGR-21-386N, " Evolution Pre-job Brief Guidelines," Revision (Rev.) 1, and attended the pre-job brief for activities supporting p
the on-line troubleshooting and isolation of the 2A Station Service Direct Current
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subsystem ground.
The inspectors observed that the pre-job brief detailed the actions that were to be taken
j in an attempt to isolate the ground to the "C" safety relief valve (SRV) circuitry inside the Unit 2 drywell. The inspectors reviewed special work instructions that were to be used, j
and found them to be thorough. The inspectors observed that there were no Health
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Physics (HP) personnel in attendance even though the activity involved personnel
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entering a potentially contaminated area in the Unit 2 Reactor Building. The inspectors j
discussed this observation with Operations management. Management personnel
acknowledged that a procedure expectation was to have all personnel involved in the activity present and that an HP representative should have been at the briefing to
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address radiological and As Low As Reasonably Achievable (ALARA) concerns.
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The inspectors concluded that the pre-job brief was generally thorough and addressed issues related to the activity through the use of the appropriate procedure checklist.
Equipment operability and Technical Specification (TS) requirements were reviewed and j
discussed. No HP personnel were in attendance to address radiological and ALARA
concerns at the pre-job brief.
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O2 Operational Status of Facilities and Equipment t
O2.1 Enoineered Safety Feature Walkdown - Unit 2 Standby Gas Treatment System (SBGT)
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The inspectors used Inspection Procedure 71707 and walked down portions of the SBGT system. The inspectors reviewed the TS, Updated Final Safety Analysis Report (UFSAR), and selected procedures used to operate the SBGT system during both normal and abnormal plant conditions. The system was walked down to verify proper valve position, labeling, and general material condition of system components, The inspectors verified that the locations of jumpers which have to be installed during emergency conditions were adequately described in the procedures and properly labeled in the panels. No deficiencies were identified. Maintenance work history was reviewed to determine if there existed any longstanding system or component problems.
Applicable electrical and logic diagrams were also reviewed and no deficiencies were identified.
The inspectors reviewed procedure 34SV-T46-001-2S, " Standby Gas Treatment -
l System Operability," Rev.10, and specifically a calculation for converting system flow, as read from a portable manometer, from inches of water differential pressure to cubic feet per minute. The inspectors questioned licensee personnel about the derivation of the calculation. Licensee personnel were unable to produce the original calculation and at I
the end of the report period had implemented actions to generate a new calculation. A similar calculation existed for Unit 1. Pending the licensee's resolution of the flow conversion calculation issue, this item will be tracked as Inspector Followup Item (IFI)
50-321,366/98-09-01, Review of SBGT Flow Conversion Calculation.
i The inspectors concluded that the Unit 2 Standby Gas Treatment system was properly aligned and labeled, procedures were adequate and associated drawings reviewed were accurate. No longstanding system or component problems were identified.
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Operator Knowledge and Performance
04.1 Inoperability of 2C Emeraency Diesel Generator (EDG) Due to Operator Action
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Inspection Scope (71707)
The inspector reviewed the events and procedures surrounding an incident in which the
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2C EDG was rendered inoperable when a support ventilation system switch was operated.
b. Observations and Findinas
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On January 17, the Unit 2 Main Control Room received an alarm indicating that the 2C EDG lube oil temperature was low. Maintenance personnel determined that a 208 volt
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electrical switch had failed. This resulted in a loss of the EDG standby lube oil pump j
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and subsequent cool down of the lube oil. The licensee considered the EDG inoperable if the lube oil cooled down to 80 degrees F. Operations began monitoring the temperature every four hours.
On January 18, a plant equipment operator (PEO), assigned the task of monitoring the tube oil temperature, independently determined that the temperature decrease could be slowed if the ventilation dampers for the room were closed. The PEO placed the local control switch for the room dampers to the OFF position which kept the closed dampers from opening. The PEO did not use a procedure or consult with his supervisor or other control room personnel prior to taking this action.
On January 19, a member of licensee management observed that the damper control switch was in the OFF position and that this rendered the EDG inoperable. The switch was immediately returned to its normal alignment which restored the operability of the EDG. The licensee determined that a PEO assigned to monitor the lube oil temperature had placed the switch in the OFF position. The PEO stated that he was not aware that this would render the EDG inoperable. The EDG had been inoperable for approximately 9.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
Administrative Control Procedure 30AC-OPS-003-OS, " Plant Operations," Rev. 20, section 8.11, states, in part, that occasionally situations arise which necessitate the operation of equipment without a procedure. However, it requires that in those cases, the system must be in a configuration that allows the component to be operated without adversely affecting the system. In addition the Shift Supervisor must approve the action and the action must be recorded. In a case involving a safety related system, a second individualis required to review the actions. These requirements were not met.
Technical Specification 3.8.1.8, requires that surveillance requirement 3.3.1.1 be performed within one hour of the EDG becoming inoperable and once per eight hours thereafter. Because control room personnel were unaware that the EDG was inoperable, these required surveillances were not completed. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. This issue is identified as Non-cited Violation (NCV) 50-366/98-09-02, Inoperability of 2C Emergency Diesel Generator Due to Operator Actions that Rendered the Diesel Ventilation Support System inoperable.
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Conclusions A Non-cited Violation was identified for failure to follow procedure which rendered Emergency Diesel Generator 2C Inoperable. A plant equipment operator placed a ventilation system damper switch to OFF and defeated support ventilation for the diese '
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04.2 Observation of Operator Performance Durina Unit 1 Standby Liauid Control Pumo Operability Test (71707)
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The inspectors observed PEOs during the performance of surveillance procedure 34SV-C41-002-1S, " Standby Liquid Control Pump Operability Test," Rev.12. The inspectors observed the operators perform the test using the appropriate revision of the procedure, peer and self checking techniques, and independent verification. The inspectors questioned the PEOs with respect to knowledge, skills, and abilities used to correctly operate the system. The operators were knowledgeable about system operation and test performance. When a condition was identified that required securing the "A" pump prior to completion of the surveillance, the PEOs correctly selected the appropriate steps and completed the remaining portions of the procedure that were required to return the system to the standby lineup. The inspector observed that the PEOs did this without guidance from supervision. The inspectors also observed that the PEOs identified several areas in the procedure that could be enhanced as part of the operations overall procedure enhancement initiative.
The inspectors concluded that the PEOs demonstrated satisfactory knowledgeable of the Standby Liquid Control system and the surveillance procedure. This was demonstrated through their use of the procedure and suggestions for procedure improvements.
Additionally, the operators selected the appropriate portions of the procedure to return the system to it's correct standby lineup following the termination of a surveillance test.
M8 Miscellaneous Maintenance issues (92700) (92902)
M8.1 (Closed) Licensee Event Report (LER) 50-366/98-05: Administrative Error Results in Missed Technical Specifications Surveillance This LER is discussed in Section M3.2 of Integrated inspection Report 50-321,366/98-07. No new information was presented in the LER. This item is closed.
Ill. Enaineerina E2 Engineering Support of Facilities and Equipment i
E2.1 2A Emeraency Diesel Generator (EDG) Jacket Coolant and Lube Oil Temperature Monitorina and Compensatory Actions a.
Inspection Scope (37551)
The inspectors observed licensee actions in response to a recurring problem with low lube oil and jacket water temperatures on the 2A EDG. The inspectors reviewed operability assessments, repair plans, and compensatory activities as a result of this issue.
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Observations and Findinas
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In October 1998, operations personnel discovered that the Plant Service Water system
(PSW) outlet valve,2P41-F339A, from the 2A EDG was leaking by the seat. This problem manifested itself in an at> wally low engine Jacket coolant temperature.
Maintenance work order 2-98-2736,,Js initiated to repair the valve, however -
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i investigation revealed that the valve needed to be replaced. Maintenance personnel
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initiated a new work order to perform a complete replacement of the valve when a spare
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was made available and a suitable window of opportunity to take the EDG out of service was identified.
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Subsequently, in December,1998, a problem with intermittent tripping of the EDG
standby tube oil pump on low lube oil temperature began to manifest itself through
repeated low lube. oil temperature annunciators in the control room. The engineers only
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assessment of this issue was that the leaking PSW valve was causing the low lube oil temperatures. Engineering management informed the inspectors that the initial root
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cause determination and diagnosis of the problem was not thorough or detailed. The inspectors observed that the engineering evaluation determined that tube oil temperature
was the most limiting parameter for EDG availability, but that the EDG was operable as long as lube oil temperature remained greater than 80*F.
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actions were initiated to monitor jacket coolant and lube oil temperatures at regular
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intervals, and to execute procedure 52PM-R43-016-0S " Diesel Generator Lube Oil
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Pumps Major Inspection / Overhaul," Rev. 3, as required to restore the standby lube oil pump, establish normal oil temperature and "bar" the engine per procedure. This i
evolution was required on several occasions, resulting in an inoperable EDG for several i
i minutes on each occasion. The licensee later identified that a contributing problem was that the control room annunciator response procedure contained the wrong setpoint for
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low lube oil temperature (105*F vice 115*F). This incorrect setpoint was used by
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personnel in an attempt to understand the problem, initiate troubleshooting activities, and
Identify possible corrections. The procedure error led to confusion and misdiagnosis of l
the problem. The incorrect value was also used to establish compensatory actions. The I
licensee corrected this problem on December 30. The inspectors reviewed
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Compensatory Action Control Sheet #2-98-48 and observed that there was no guidance
for operators as to what actions to take if EDG lube oil temperature approached or decreased below 80*F.
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The inspectors observed that throughout this time period, maintenance personnel tried to i
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determine if there might be some underlying equipment problem that could be causing
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the low temperature conditions. It was theorized by maintenance personnel that the
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i leaking PSW valve alone should not have resulted in such abnormally low lube oil and
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Jacket coolant temperatures. Maintenance personnelidentified that a faulty heater i
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control temperature switch,2R43-N044A, may be causing the problem. This faulty switch was confirmed when the EDG was finally taken out of service for repairs of the PSW valve on January 7.
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Conclusions
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The inspectors concluded that engineering sepport to operations and maintenance was -
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temperature issue. The analysis failed to take into account multiple equipment or.
l component failures and their resultant effect on the EDG lube oil temperature.
l E2.2 Review of Maintenance Rule Status for 4kV Circuit Breakers (37551I
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l The inspectors reviewed the licensee's corrective actions and resolution for a series of r
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maintenance preventable functional failures (MPFFs) affecting safety-related 4kV circuit
t breakers in November,1997. These MPFFs, associated with Westinghouse Type DHP metal-clad switchgear, required that the 4kV circuit breakers be classified as an."(a)(1)"
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j subsystem in accordance with 10 CFR 50.65. Past performance on these breakers required increased monitoring. Based on a review of the licensee's actions, the
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i inspectors determined that the licensee has completed training items associated with
maintenance and overhaul of these circuit breakers, no subsequent related failures of this equipment item have occurred, and breaker maintenance and overhaul continue to
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be accomplished on schedule.
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The inspectors concluded that the licensee was complying with 10 CFR 50.65
req'iirements and that the reliability program insti'.uted for this equipment class had been l
effective.
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E8 Miscellaneous Engineering issues
E8.1 Concrete Pad Desian Activities for Indcoendent Soent Fuel Storaae Installation (ISFSI)
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Inspection Scope (60851)
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The inspectors reviewed design calculations and drawings and other related activities to
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determine if these activities met industrial standards and regulatory requirements.
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Observations and Findinas
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The inspectors discussed the ISFSI project with the licensee's engineers and reviewed the following calculations and specifications:
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i Calculation No. SCNH 98-029, " Laboratory Soils Test Data for Dry Cask Storage i-Pads", Rev. O, Calculation No. SCNH 98-032, " Stability of Loaded Crawler Against Bearing _
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Capacity Failure", Rev. O,
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Calculation"No. SCNH 98-034, " Analysis and Liquefaction Potential for Soils to i
150-foot Depth for Dry Cask Storage Pads", Rev. O, Calculation No, SCNH 98-057, " Static Analysis and Pad Design for Dry Cask [
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p Storage Pads," Rev. 0,
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O Calculation No. SCNH 98-059, " Static Bearing Capacity of Singular Pad for Dayj j
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Calculation No. SCNH 98-067, " Maximum Seismic Responses for ISFSI Pad -
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Design," Rev. O, holtec Report No. HI-981878, " Specification Document for ISFSI Concrete Pad at E. l. Hatch," Rev. 3, and Specification No. SC-DS-98-08, " Specification for ISFSI Pads for Hatch Nuclear j
Plant," Rev. 3.
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l The input and results of the reviewed calculations were checked against the limitations and guidance contained in Holtec Report No. Hl-981878, American Concrete Institute i
code, American Nuclear Society code, industry standards, and regulatory requirements.
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The calculations were performed well with regard to the details, accuracy, and order.
Minor discrepancies were discussed with the licensee.
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The licensee was required to have a separate Quality Assurance (QA) program or modify their existing QA program to include the ISFSI. The licensee developed " Project
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Quality Plan for Spent Fuel Rack Addition and ISFSI," Revision 0, as an intermediate r
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measure. The licensee planned to integrate this document into their current QA
program.
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Conclusions l
The design calculations for the spent fuel cask concrete pad and related structures were j
j well performed and in accordance with industrial standards and code requirements.
i IV. Plant Support j
R1 Radiolwical Protection and Chemistry Controls
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R1.1 ALARA Oversiaht for the 2A Reactor Water Cleanuo (RWCU) Heat Exchanaer Drain
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Line Leak Repair (71750)
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i The inspectors attended the pre-job briefing, reviewed the ALARA review package, and
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l witnessed the drain line valve repair remotely with a video camera on December 11.
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j The job was to inject liquid leak sealant into the valve.
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The inspectors noted that the ALARA review package indicated a total accumulated dose goal of 172.5 mrem. HP oversight of the job was provided by stationing a _
1 technician in the room where the work was performed. The dose goal for the work was based on the technician accumulating 60 mrem, which accounted for slightly more than
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one-third of the total expected dose. ' The inspectors noted that on two previous
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occasions for similar work the estimated dose for the HP technician was 25 and 32
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mrem. ALARA personnel were not able to explain the increase in the estimated dose for
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. the HP. technician during this job. The inspectors did observe the technician wait in a low
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The inspectors noted that the ALARA review did not consider staging redundant tools in the event tool problems occurred. Workers were required to work at a distance of one foot from the source during part of the work activity after equipment failed to operate properly. Prior to the equipment failure workers were able to stay about 7 feet 4way from the source term. The total dose accumulated for the job was 93 mrem. The inspectors had observed that a larger portion of the dose was received following the failure of the air operated equipment planned for the job use. The inspectors discussed this observation with the licensee management. The ALARA staff was not able to explain why the total estimated job dose was 172.5 mrem and actual dose received was 93 mrem.
The inspectors inquired whether or not a post-job critique would be held. The licensee stated that post-job critiques were not generally held for jobs where dose accrued was low. The inspectors reviewed ALARA procedures which indicated a post-job brief should be considered for work in areas where dose rates were greater than 100 mrem per hour and total dose accumulated was estimated to be equal to or greater than 1 rem. The inspectors questioned whether or not lessons learned from previous jobs could be captured to ensure ALARA for future jobs with the absence of a post-job brief.
The inspector concluded that licensee's ALARA pre-planning for a leak repair on the 2A RWCU heat exchanger drain line was not thorough. The dose goal was set at approximately twice the actual dose accumulated; consideration was not given to having spare tools available so that hand tools would not have to be used; and the decision to not hold a post-job critique presented a missed opportunity to capture lessons learned to ensure the dose received for future jobs would be ALARA. (Section R1.1).
R1.2 Tour of Radioloaical Protected Areas a.
Inspection Scope (83750) (84750)
The inspectors reviewed implementation of selected elements of the licensee's radiation protection program as required by 10 Code of Federal Regulations (CFR) Parts 20.1902, and 1904. The review included observation of radiological protection activities for control of radioactive material, personnel frisking, postings and labeling, and the use of survey instrumentation.
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Observations and Findinas The inspectors reviewed survey data of radioactive material storage areas. During tours of the plant, all radioactive material storage areas observed were appropriately posted as required by 10 CFR Part 20.1902. All material observed was labeled to meet the requirements of 10 CFR Part 20.1904. The inspectors also performed radiation and contamination surveys of material storage locations outside of the Radiological Control Area (RCA) but inside the protected area. Surveys were also performed at the onsite landfill. Wood and other debris being placed into the landfill was surveyed for contamination. Results of surveys of the locations and material were consistent with established background levels.
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The inspectors reviewed personnel contamination control practices for personnel exiting the RCA including personnel entering satellite break areas attached to the RCA during outages. Based on observations, procedural reviews, and interviews, the inspectors determined that the licensee's practices for contamination control when exiting an RCA were consistent with licensee procedures and were acceptable practices.
During plant tours, the inspectors verified that the various types of instrumentation in use for frisking material and personnel were currently calibrated and source checked as required by licensee procedures. The inspectors also interviewed four technicians on the use of the instruments for frisking material and observed the use of the instrumentation. The technicians interviewed were knowledgeable on the use of the instruments. The inspectors reviewed training lesson plans for contractor health physics technicians used to staff exit points from the RCA during outages. Training on the use of the instrumentation was covered during le-tures.
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Conclusions Based on observations and surveys during tours, the licensee had effectively labeled, controlled, and stored radioactive material as required by 10 CFR Part 20.1904 and effectively posted radioactive material storage areas as required by 10 CFR Part 20.1902. Contamination control practices reviewed were consistent with licensee procedures and were acceptable practices.
V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management at the conclusion of the inspection on February 3,1999. Interim exit meetings were held on December 18,1998, and January 15,1999 to discuss the findings of Region based inspections. The licensee acknowledged the findings presented.
The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.
PARTIAL LIST OF PERSONS CONTACTED l
Licensee
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Betsill, J., Assistant General Manager - Operations Davis, D., Plant Administration Manager Fornel, P., Performance Team Manager Fraser, O., Safety Audit and Engineering Review Supervisor Googe, M., Performance Team Manager Hammonds, J., Engineering Support Manager Kirkley, W., Health Physics and Chemistry Manager Lewis, J., Training and Emergency Preparedrass Manager l
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Madison, D., Operations Manager Moore, C., Assistant General Manager - Plant Support Roberts, P., Outage and Planning Manager Thompson, J., Nuclear Security Manager Tipps, S., Nuclear Safety and Compliance Manager Wells, P., General Manager - Nuclear Plant Other licensee employees contacted included office, operations, engineering, maintenance, chemistry / radiation, and corporate personnel.
INSPECTION PROCEDURES USED IP 37551:
Onsite Engineering IP 60851:
Design Control of ISFSI Components IP 61726:
Surveillance Observations IP 62707:
Maintenance Observations IP 71707:
Plant Operations IP 71750:
Plant Support Activities IP 83750:
Occupational Exposure IP 84750:
Radioactive Waste Treatment, Effluent, and Environmental Monitoring IP 92902:
Followup - Maintenance / Surveillance IP 92700:
Onsite Follow-up of Written Reports of Nonroutine Power Reactor Facilities
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ITEMS OPENED. CLOSED. AND DISCUSSED ODer0d 50-321,366/98-09-01 IFl Review of SBGT Flow Conversion Calculation (Section O2.1)
Closed 50-366/98-09-02 NCV Failure to Follow Procedure Rendered Emergency Diesel Generator 2C Inoperable (Section 04.1)
50-366/98-05 LER Administrative Error Results in Missed Technical Specifications Surveillance (Section M8.1)
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