IR 05000321/1990001
| ML20012C476 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/02/1990 |
| From: | Brockman K, Menning J, Randy Musser NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20012C473 | List: |
| References | |
| 50-321-90-01, 50-321-90-1, 50-366-90-01, 50-366-90-1, NUDOCS 9003220008 | |
| Download: ML20012C476 (17) | |
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>*h88uq UNITE 3 STATES
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g NUCLEAR REGULATORY COMMIS$lON
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REGION il r
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101 MARIETTA STREET,N.W.
- t AT LANT A, GEORGI A 30323
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Report Numbers: 50-321/90-01 and 50-366/90-01 Licensee: Georgia Power Company P.O. Box 1295 Birmingham, AL 35201 e
Docket Numbers:
50-321 and 50-366 License Numbers:
DPR-57 and NPF-5 Facility Name:
Hatch Units 1 and 2
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Inspection Dates: January 6 - February 16, 1990 Inspection at Hatch site near Baxley, Georgia Inspectors:
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J E.Menning}.SeniorResidentInspector Date Signed
$ m' A A 'AS-f0
.s M6ndall A. Musser. Residsnt Inspector Date Signed Approved by:
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J - >2 -[8 Mef1Tieth E. Brockman. Chiif Date Signed Reactor Projects Section 3B Division of Reactor Projects SUMMARY
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Scope:
This routine inspection was conducted at the site in the areas of Operational Safety Verification. Maintenance Observation. Surveil-
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lance Testing Observation. ESF System Walkdown. Reportable
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Occurrences. Operating Reactor Events. Installation and Testing of Modifications. Preparation for Refueling. Inspection for Verification of Quality Assurance Request Regarding Diesel Generator Fuel Oil (Multi-Plant Action Item A-15. TI 2515/93), and Visit to Local Public Document Room.
Results: Two non-cited violations were identified during this reporting period.
The first NCV (paragraph 2) was for failure to have a HPCI system valve positioned as specified on an equipment clearance sheet.
The second NCV (paragraph 6) was for an inadequate operations shift turnover briefing.
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No specific strengths or weaknesses of licensee programs were identified based on the inspectors' findings and observations in the 900322000sSlo'oo$21 f'
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i REPORT DETAILS
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Persons Contacted Licensee Employees
C. Coggin. Training and Emergency Preparedness Manager
- D. Davis. Manager General Support
- D. Edge Nuclear Security Manager P. Fornel. Maintenance Manager i
- 0. Fraser. Site Quality Assurance Manager G. Goode. Engineering Support Manager
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- M. Googe. Outages and Planning Manager J. Lewis. Acting Operations Manager
- C. Moore. Assistant General Manager - Plant Support
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H. Nix. General Manager - Nuclear Plant
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- H. Sumner. Assistant General Manager - Plant Operations S. Tipps. Nuclear Safety and Compliance Manager
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R. Zavadoski. Health Physics and Chemistry Manager
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Other licensee employees contacted included technicians, operators.
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mechanics, security force members, and office personnel.
I NRC Resident Inspectors
- J. Menning-
- R. Musser NRC management on site during inspection period:
K. Brockman. Chief. Reactor Projects Section 3B. Region 11 L.-Crocker Project Manager. Hatch. Project Directorate 11-3. NRR
A. Herdt Chief. Reactor Projects Branch 3. Region II D. Matthews. Director. Project Directorate 11-3. NRR J. Milhoan. Deputy Regional Administrator. Region II
- Attended exit interview Acronyms and initialisms used throughout this report are listed in the last paragraph.
2.
Operational Safety Verification (71707) Units 1 and 2
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Unit I continued operating at power during this reporting period.
At 0700 on February 16, 1990 shutdown of the unit commenced in preparation for a scheduled condenser retubing/ refueling outage.
The unit's main generator
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time. Unit I had operated continuously on-line for over 423 days.
Unit 2 l-l
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started the reporting period operating at power.
At 1710 on January 12 1990, the unit automatically scrammed because the MSIVs were less than 90 percent open.
The MSIVs had closed on a Group I PCIS signal which resulted from a false low condenser vacuum signal.
(The Unit 2 scram is discussed in paragraph 7 of this report.) Restart of Unit 2 commenced at 223b on January 14. and criticality was achieved at 0140 on January 15.
1990.
The unit's main generator was synchronized with the grid at 0342 on January 16. 1990.
Unit 2 achieved rated power at 0337 on January 17 1990 and ended the reporting period operating at power.
The inspectors kept themselves informed on a daily basis of the overall plant status and any significant safety matters related to plant operations. Daily discussions were held with plant management and various members of the plant operating staff. The inspectors made frequent visits to the control room.
Observations included control room manning, access
control. operator professionalism and attentiveness, adherence to procedures. adherence to limiting conditions for operation, instrument readings, recorder traces, annunciator alarms, operability of nuclear instrumentation and reactor protection system channels, availability of power sources, and operability of the Safety Parameter Display system.
These observations also included log book entries. tags and clearances on equipment, temporary alterations in effect. ECCS system lineups, containment integrity, reactor mode switch position, conformance with technical specification safety limits, daily surveillances, plant chemistry, scram discharge volume valve positions, and rod movement controls. This inspection activity involved numerous informal discussions with operators and their supervisors.
The operability of selected safety-related systems was confirmed on essentially a weekly basis.
These confirmations involved verification of proper valve and control switch positioning, proper circuit breaker and fuse alignment and the operability of related instrumentation and support systems.
Major components were also inspected for leakage, proper lubrication, cooling water supply, and general condition.
On January 9.
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1990, the inspector confirmM the operability of the Unit 2 Post LOCA Hydrogen Recombiner system.
Proper switch, electrical, and valve alignments were confirmed using Attachments 1. 2. and. 3 to procedure 3450-T49-001-25.
On January 10 and 11.1990, the inspector confirmed the operability of the Unit 1.HPCI system.
Proper switch, breaker, and valve lineups were confirmed using Attachments 1. 2. and 3 to procedure 3450-E41-001-1S.
On January 23. 1990, the operability of the Unit 1 RCIC
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system was confirmed.
Proper switch, breaker, and valve positions were l
verified using Attachments 1. 2 and 3 to procedure 34S0-E51-001-15.
On February 1. 1990, the inspector confirmed the operability of the."A" loop of the Unit 2 Core Spray system.
Switch. breaker, and valve lineups were verified using Attachments 1, 2 and 3 to procedure 3450-E21-001-2S, On February 15. 1990. the operability of the Unit 1
"A" Core Spray system loop was confirmed.
Switch, breaker, and valve lineups were confirmed using Attachments 1. 2. and 3 to procedure 3450-E21-001-15.
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i General plant tours were conducted on at least a weekly basis.
Portions of the control building, diesel generator building, intake structure.
turbine building. reactor building, and outside areas were toured.
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Observations included general plant / equipment conditions, fire hazards, fire alarms fire extinguishing equipment. emergency lighting. fire barriers, emergency equipment. control of ignition sources and flammable r-materials, and control of maintenance / surveillance activities in progress.
Radiation protection controls. implementation of the physical security program, housekeeping conditions / cleanliness. control of missile hazards, and instrumentation and alarms in the main control room were also observed.
The inspectors observed selected operations shift turnover briefings to confirm that all necessary information concerning the status of plant systems was being addressed.
Each briefing was conducted by the oncoming
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0S05.
The inspectors noted that each OSOS discussed existing plant problems, activities that were anticipated for the shif t, and any new standing orders or management directives.
Radiological and industrial safety were generally stressed.
The STAS discussed any recent procedure revisions that impacted on the attendees.
The inspectors attended shift turnover briefings on the following dates and shifts: January 14, 1990 -
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Day. January 17. 1990 - Day January 18, 1990 - Day, January 19. 1990 -
Day, January 21, 1990 - Day. January 30, 1990 - Day. February 1. 1990 -
Day, and February 16, 1990 - Day.
Several safety-related equipment clearances that were active were reviewed to confirm that they were properly prepared and placed.
Involved circuit breakers, switches, and valves were walked down to verify that clearance tags were in place and legible and that equipment was properly positioned.
Equipment clearance. program requirements are specified. in licensee procedure 30AC-0PS-001-0S. " Control of Equipment Clearances and Tags." On January 8, 1990. Unit 1 equipment clearance 1-90-33 was walked down. This clearance was placed to support maintenance on the
"B" Train of the Standby Gas Treatment System.
On January 11. 1990. Unit 2 equipment clearance 2-89-2035 was walked down.
This clearance was placed to electrically disconnect the directional control valves for HCU 26-31.
On January 30, 1990 Unit 1 equipment clearance 1-90-411 was walked down.
This clearance was placed to support maintenance on the Unit 1 HPCI system.
On January 30. 1990 while walking down equipment clearance 1-90-411. the inspector observed the control switch for valve 1E4',-F051 (torus suction valve) to be in the closed position, while the equipment clearance sheet required it to be in the open position.
Additionally the danger tag placed on the control switch incorrectly specified the valve position as closed. This discrepancy was brou9ht to the attention of the Unit 1 Shift Supervisor.
Upon confirming the discrepant condition, corrective action was initiated by placing the control switch in the open position and by
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replacing the incorrect danger tag.
This corrective action was observed by the inspector.
The incorrect positioning of the control switch for valve IE41-F051 is a violation of Technical Specification 6.8.1.a.
Technical Specification 6.8.1.a requires that written procedures be implemented covering the activities recommended in Appendix A of Regulatory Guide 1.33. Revision 2. February 1978.
Section 1 of Appendix A ef Regulatory Guide 1.33 recommends procedures for equipment control (e.g., locking and tagging).
However, this violation meets the criteria specified in Section V of the NRC Enforcement Policy for not issuing a Notice of Violation and, therefore, is not being cited.
This matter, identified as NCV 321/90-01-01, is considered to be closed.
Implementation of the licensee's sampling program was reviewed by the inspector.
This review involved observation of sampling activities (reactor coolant and tank sampling) and chemistry surveillance.
Related records were also reviewed.
During this inspection period, the inspector monitored the following activities.
On January 12, 1990, the inspector observed the performance of the monthly main stack source check in accordance with procedure 62CI-CAL-007-05.
On January 19. 1990, the inspector observed the acquisition and analysis of a Unit 2 Pre-Treat Offgas sample in accordance with procedure 64CH-SAM-001-0S.
On February 14, 1990, the inspector observed the sampling and subsequent analysis of the Unit 2 Chemical Waste Sample Tank "A" in accordance with procedures 64CH-SAM-004-05 and 64CH-RCL-001-05.
Additionally, since the tank contents were to be discharged into the river, the inspector observed the preparation of the discharge permit.
The licensee's deficiency control system was reviewed to verify that the system is functioning as intended.
Licensee procedure 10AC-MGR-004-0S.
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" Deficiency Control System." establishes requirements and responsibilities for the preparation, processing, review, and disposition of deficiency reporting documents. This procedure applies to all deficiencies affecting equipment, procedures, or personnel.
Deficiencies are reported on Deficiency Cards.
On January 16, 1990, the inspector reviewed recently prepared DCs.
The inspector verified that DCs had been prepared as required by the controlling procedure and that several deficiencies that were noted in the Shift Supervisors' logs had been documented on DCs.
More specifically, the inspector observed that DC 1-90-0255 had been prepared to document that main control room annunciators for Reactor Building Stack Radiation monitors 1D11-K619A and B were erroneously
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indicating that high-high radiation conditions existed.
It was also noted that DC 2-90-0142 had been generated in response to the downscale failure of SRM 2C51-K601A.
On January 22, 1990, the inspector also reviewed recently prepared DCs and verified that problems observed in the plant had been properly documented.
The inspector observed that DC 1-90-0373 had been prepared to document the unanticipated tripping of LPCI inverter 1R44-S002.
The inspector also noted that DC 2-90-0282 had been generated to document the inability of operations personnel to select control rod 42-43 on the rod select matrix on control room panel 2H11-P603.
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February 5,1990. DCs were again reviewed to verify that problems observed in the plant had been properly documented.
The inspector noted that DC 1-90-0574 had been prepared to document unusual sounds coming from the turbocharger area of the "1A" D/G during surveillance testing.
The inspector also noted that DC 2-90-0389 had been generated to document an unexpected increase in pressure on the #2 seal of the Unit 2
"A" recirculation pump.
Selected portions of the containment isolation lineup were reviewed to confirm that the lineup was correct. The review involved verification of proper valve positioning verification that motor and air-operated valves were not mechanically blocked and that power was available (unless blocking or power removal was required), and inspection of piping upstream of the valves for leakage or leakage paths.
On January 9.1990, the inspector reviewed the following Unit 2 containment isolation valves:
2T48-F333A. 2T48-F334A and B 2T48-F340. 2T48-F341. 2T48-F361A and B.
2T48-F364A and B. 2T49-F002A and B. 2T49-F0048, 2B31-F020, 2D11-F050, and 2D11-F052.
On January 22, 1990, the inspector reviewed the following Unit I containment isolation valves:
1E41-F111. 1E41-F121. IE41-F122. IP33-F002.
IP33-F004 IP33-F005. IP33-F006. IP33-F010 IP33-F012, 1P33-F013.
IP33-F014. IP41-F050, 1P42-F051, IP42-F052. IP70-F066, and IP70-F067.
On February 12. 1990, the inspector reviewed the following Unit 2 containment isolation valves:
2E11-F023. 2E11-F0418, 2E11-F0410. 2P33-F002, 2P33-F004. 2P33-F005. 2P33-F010, 2P33-F012, 2P51-F513, 2P64-F045.
2P70-F066.-2P70-F067. 2T23-F004, and 2T23-F005.
During this reporting period the inspector reviewed tne licensee's controls on overtime of personnel who perform safety-related functions.
Section 6.2.2 9 of the technical specifications establishes requirements for the control of such overtime, and Section 8.4 of licensee procedure 30AC-0PS-003-05. " Plant Operations." provides implementing instructions to support the technical specification requirements.
On January 17, 1990 the inspector reviewed an Operations Department Overtime Report for the
- month of November and determined that technical specification and procedural requirements had been met.
On January 17, 1990, the inspector verified that all required notices to workers were appropriately and conspicuously posted pursuant to 10 CFR 19.11.
The licensee has established posting locations at the Waste Separation and Temporary Storage Facility. S'mulator Building near-the breakroom. Service Building near the cafeteria, and Security Building.
The inspector reviewed the postings at these locations and observed no discrepancies.
One NCV.was identifie.
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3.
Maintenance Observation (62703) Units 1 and 2 During the report period, the inspectors observed selected maintenance i
activities.
The observations included a review of the work documents for i
adequacy. adherence to procedure, proper tagouts, adherence to technical specifications. radiological controls, observation of all or part of the actual work and/or retesting in progress, specified retest requirements, and adherence to the appropriate quality controls.
The primary
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maintenance observations during this month are sunmarized below:
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Maintenance Activity Date
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a.
Calibration of Meter Relay GE 01/12/90 Type 195. in accordance with MWO
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2-90-0080 and procedure 57CP-CAL-015-2S
(Unit 2)
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Preventive Maintenance on Reactor 01/17/90 Building Vent Monitor 2011-P002A, in accordance with MWO 2-89-5725
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and procedure 52PM-011-001-0$
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(Unit 2)
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36-Month Preventive Maintenance 01/21/90 on valve IP41-F4028. in accordance i
with MWO 1-89-3970 and procedure 52PM-MNT-005-05 (Unit 1)
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Electrical Compartment Inspection 01/30/90 on Limitorque Operator of valve 1E41-F042. in accordance with
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MW01-89-5652(Unit 1)
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Replacement of the HPC1 System 01/30/90 Rupture Disks, in accordance with MWO 1-89-6238 and procedure 52PM-E41-001-0S (Unit 1)
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36-Month Preventive Maintenat.ce 02/13/90 on valve IE11-F026A. in accordance with procedures 52GM-MEL-022-05 and 52PM-MNT-005-05 and MWO 1-89-5574 (Unit 1)
No violations or deviations were identified.
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4.
Surveillance Testing Observations (61726) Units 1 and 2 The inspectors observed the performance of selected surveillances.
The
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observation included a review of the procedure for technical adequacy.
p conformance to technical specifications. verification of test instrument calibration, observation of all or part of the actual surveillances, removal from service and return to service of the system or components affected. and review of the data for acceptability based upon the acceptance criteria..The primary surveillance testing observations'during
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this month are sunnarized below:
Surveillance Testing Activity Date a..
SBGT Ventilation and Operability 01/09/90 and SBGT Damper Operability tests.
in accordance with procedures
34SV-T46-003-2S and 34SV-T46-002-25 (Unit 2)
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Msiv Leakage Control System Blower 01/12/90 Operability Testing. in accordance procedurt 34SV-E32-002-25 (Unit 2)
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Main Turbine Weekly Surveillance 01/16/90 Testing, in accordance with procedure 345V-N30-001-2S(Unit 2)
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HPCI Pump Operability Testing, in 01/18/90
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accordance with procedure 345V-E41-002-25 (Unit 2)
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GE Numac Main Steam Line Logarithmic 01/21/90 Radiation Monitor Functional Test, in accordance with procedure S75V-D11-016-IS (Unit 1)
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Standby Liquid Control Recirculation 01/29/90 Test, in accordance with procedure 34SV-C41-001-15 (Unit 1)
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RCIC Pump Operability Testing.
01/30/90 in accordance with procedure 34SV-E51-002-15 (Unit 1)
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RHR Valve Operability, in 01/30/90 accordance with procedure
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34SV-E11-002-IS (Unit 1)
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L Surveillance Testing Activity Date (Continued)
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Diesel Generator "1A" Monthly 02/01/90 Operability Test in accordance with procedure 345V-R43-001-IS (Unit 1)
No violations or deviations were identified.
5.
ESFSystemWalkdown(71710) Unit 1 The inspectors routinely conducted partial walkdowns of ESF systems. Valve and breaker / switch lineups and equipment conditions were randomly verified both locally and in the contro; room to ensure that lineups were in accordance with operability requirements and that equipment material conditions were satisfactory.
During this reporting period, accessible portions of the Unit 1 HPCI system were walked down in detail.
This effort involved confirmation that system lineup requirements in procedure 34S0-E41-001-IS. "High Pressure Coolant Injection System." were consistent with the as-built configuration and the applicable plant drawings (H-16332 and H-16333).
The detailed walkdown also involved confirmation that valves were properly positioned and that material condition was satisfactory.
No violations or deviations were identified.
6.
Reportable Occurrences (90712 and 92700) Units 1 and 2 A number of LERs were reviewed for potential generic impact to detect
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trends and to determine whether corrective actions appeared appropriate.
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Events which were reported immediately were also reviewed as they occurred to determine that technical specifications were being met and the public health and safety were of utmost consideration.
Unit 1:
89-18 Miscommunication During Shift Turnover Results in an Engineered Safety Feature Actuation This LER related to two unanticipated isolations of the RWCU system that occurred as operations personnel were attempting to return the system to service.
Both isolations resulted from system high differential flow signals.
The root cause of the events was determined to be personnel error.
On-coming day shift personnel erroneously believed that the RWCU system had been filled and vented during the previous shift.
In fact, the system had not been filled and vented.
This resulted in a high differential flow condition when attempts were made to return the system to servic ye
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I The events of this LER were identified by the licensee and properly reported to the NRC.
Corrective action
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emphasizing the need for effective. accurate. and
complete communication of plant status during shift
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turnover briefings. On January 9.1990. the inspector
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reviewed memorandum LR-0PS-020-1189 from the Acting
Manager of Operations and confirmed that the intended
corrective action had been completed.
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Technical Specification 6.8.1.a requires that L
procedures be implemented for the activities
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I recommended in Appendix A of Regulatory 1.33.
Revision 2. February 1978.
Section 1 in Appendix A of
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Regulatory-Guide 1.33 recommends procedures for shift l
turnovers. Licensee procedure 31G0-0PS-007-05. " Shift
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Logs and Relief of Personnel." provides instructions for the shift relief and turnover of operations l
personnel.
This instance of inadequate transfer of.
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plant status information during a shift turnover is a violation of Technical Specification 6.8.1.a.
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However, this licensee-identified violation meets the
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criteria in Section V of the NRC Enforcement Policy for not issuing a Notice of Violation and, therefore.
is not being cited.
This matter, identified as NCV 321/90-01-02. is considered to be closed.
Review of this LER is also closed.
Unit 2:
89-09 Equipment Failure Results in a Reactor Protection System Actuation This LER concerned an RPS actuation that occurred as a result of a SDV high level condition.
No control rod movement resulted from the actuation since all control
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rods were full-in at the time.
The Unit 2 SSACs were in the process of being replaced to improve i
reliability.
At the time of the event. the "B" and
"C" SSACs had been replaced and the "C" SSAC was in service.
A temporary air compressor had been installed and maintained in standby as a backup to the
"C" SSAC.
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"C" SSAC tripped on loss of system
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cooling.
Delays were encountered-in placing the
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temporary compressor in service and low system air pressure resulted.
The low system air pressure allowed the scram valves to drift open resulting in a discharge of water to-the SDV.
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The root cause of the event was determined to be equipment failure.
At ~ the time of the event.
troubleshooting activities were in progress on a temperature switch in the closed cooling water system dedicated to the SSACs. The "C" SSAC tripped due to a
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multimeter lead which disengaged from the multimeter
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and grounded the cooling water control circuit.
Corrective action involved placing the temporary air compressor in service and returning the
"C" SSAC to service following the replacement of a blown fuse in the SSAC cooling water control circuitry.
Additionally. the temporary air compressor was placed in an automatic mode of operation.
In this mode. the compressor automatically provides air to the station service air system on low system pressure.
The
operating procedure for the temporary air compressor was also to be revised to address the automatic mode of operation.
On January 10, 1990, the inspector reviewed Revision 1 of procedure 34SP-090889-QX-1-2N.
" Operation of Temporary Station Service Air Compressor." and confirmed that the procedure had been nodified.
Since corrective actions have been completed, review of the LER is closed.
90-01 Component Failure and Inadequate Design Cause Group I Isolation and Scram
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This LER concerned the Unit 2 automatic scram that occurred on January 12. 1990. The events of the scram are discussed in paragraph 7.
As stated in para-
. graph 7. this LER will remain open pending review of an update to the LER and of the implementation of DCR 1H90-009.
One NCV was identified.
7.
OperatingReactorEvents(93702) Unit 2 The inspectors reviewed activities essociated with the ' reactor scram discussed below.
The review included determination of cause, safety significance, performance of personnel and systems, and corrective action.
l The ' inspectors examined instrument recordings. computer printouts, operations journal entries, scram reports and had discussions with operations. maintenance, and engineering support personnel, as l
appropriate.
At 1710 on January 12.1990. Unit 2 automatically scrammed from rated power.
The reactor scram occurred because the MSIVs were less than 90 percent open. The MSIVs had closed on a Group I PCIS signal that resulted from a false low condenser vacuum signal.
Vacuum switches 2B21-N0560 i
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and D, which provide input to the "A" and "B" trip systems of the Group 1 PCIS logic. respectively, sensed low condenser vacuum after root isolation valve 2N61-F5880 failed and isolated the switches' common sensing line.
Subsequent investigation revealed that the disc in 2N61-F588D separated i
from its stem as a result of vibration-induced wear.
Following the scram. reactor vessel water level was restored via the automatic initiation of HPCI and the manual initiation of RCIC.
(RCIC was initiated manually prior to its automatic initiation setpoint.)
Reactor water level decreased to approximately minus 40 inches indicated during i
the transient.
The SRVs operated in their relief mode and. later. LLS
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mode to control reactor pressure.
Reactor pressure peaked at approxi-
mately 1117 psig and LLS maintained pressure betweer, 850 and 990 psig.
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a thereafter. as-per design.
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Plant systems and equipment responded properly during the transient with the exception of HPCI injection valve 2E41-F006.
Both HPCI and RCIC tripped per design at the high reactor water level setpoint (approximately-52 inches indicated) following the initial restoration of water level.
When operations personnel subsequently attempted to manually initiate HPCI for reactor water level control. HPCI injection valve 2E41-F006 could not be opened using the remote manual switch in the main control room.
Operations personnel then used RCIC and both CRD pumps to recover and maintain reactor water level.
The cause of 2E41-F006 failing to open was determined to be component failure.
The heater strip of a thermal overload relay in the valve motor's local starter failed.
The heater strip fused as attempts were made to open the valve resulting in an open
circuit to the motor.
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The licensee implemented several corrective actions in response to this
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Root isolation valves 2N61-F588B and D (in the sensing lines for Group I. isolation logic vacuum switches) and 2N61-F061 and F064 (in the sensing lines for turbine trip logic vacuum switches) were removed and replaced.
These valves were replaced with gate valves that are not
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susceptible to the failure mode of 2N61-F5880.
The valves were also installed upside down so that a disc-stem separation would not result in isolation of a sensing line. The condenser vacuum sensing lines were also i
reconfigured such that each of the four sensing lines off the condenser now has one Group I isolation logic vacuum switch and one turbine trip logic vacuum switch.
The new arrangement is single failure proof and i
intended to prevent spurious trips due to single failures.
The licensee also reviewed the Unit I design to determine if it was vulnerable to failures in any one condenser vacuum sensing line.
This review revealed
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that the root isolation valve, sensing line, and logic arrangements were the same as previously existed in Unit 2.
Consequently, the licensee generated DCR 1H90-009 to implement changes to Unit 1 similar to those implemented in Unit 2.
This DCR is expected to be implemented during the upcoming Unit 1 condenser retubing/ refueling outage.
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As discussed in the licensee's report on the scram (LER 50-366/1990-001),
investigation into the 2E41-F006 heater strip failure is continuing. The licensee committed in the LER to provide an update detailing the cause of the heater strip failure and any additional corrective actions that will be taken as a result of the investigation results.
Review by the resident inspectors of the heater strip failure and implementation of DCR 1H90-009 will be tracked with the LER.
Within the areas inspected, no violations or deviations were identified.
8.
Installation and Testing of Modifications (37828) Unit 1 This inspection effort involved the examination of a minor modification to verify that installation work was performed in accordance with approved instructions. procedures, and drawings and that specified post-installa-tion testing was appropriate and properly performed.
Although this inspection effort involved the review of documentation, primary emphasis was placed on the direct observation of activities in progress. The minor modification selected for examination was DCR 87-213.
This DCR was generated to connect a fill pipe to the loop seal in each main control room AHU drain line.
The fill pipes are intended to allow surveillance and maintenance of these seals to prevent the undetected occurrence of MCR air in-leakage through this pathway.
(Concerns about the undetected occurrence of MCR air in-leakage were discussed in NRC Information Notice 86-76.)
The inspector initially reviewed the package for DCR 87-213 which included the safety evaluation performed pursuant to 10 CFR Part 50.59.
Maintenance Work Order 1-90-134 and WPS 87-213-P001 were prepared for.the modification of the drain piping for AHU 1Z41-B003A.
Maintenance Work Orders 1-90-133 and WPS 87-213-P002 were prepared for the modification of the drain piping for AHU 1Z41-B003B, Modification of the drain piping for AHU 1Z41-B003C was to be controlled by MWO 1-90-132 and WPS 87-213-P003.
The DCR package specified that the post-installation functional testing of each of the three AHUs was to be performed in accordance with procedure 421T-TET-004-0S. " Operating Pressure Testing of Piping and Components."
During the period January 13 to February 15, 1990, the inspector observed selected installation activities associated with DCR 87-213.
The inspector verified that craft personnel at the job site had and were using approved drawings, procedures, and instructions.
Installed piping was visually examined and measured to verify that the installation conformed with applicable drawings (S-02573 and FCR 87-213-2).
Installed materials were also compared with the applicable Stock Material Issue Sheets.
The inspector verified that appropriate QC hold points had been established and were being observed.
The inspector also verified that the acceptance criteria for post-installation testing were satisfie i
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The inspector concluded that the installation activities associated with
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DCR 87-213 were properly controlled and performed in accordance with
i approved procedures, instructions, and drawings.
It was also concluded p
that the acceptance criteria for post-installation functional testing were
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satisfied and the testing was appropriate for the DCR.
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L No violations or deviations were identified.
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9.
Preparation for Refueling (60705) Unit 1 i
This inspection effort was perfonned to ascertain the adequacy of the
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licensee's procedures and administrative requirements for the control of
refueling operations.
The first aspect of this effort focused on procedure review for the purpose of verifying the technical adequacy of existing approved refueling-related procedures.
The first review wa; conducted in the area of receipt. inspection, and storage of new fue'.
The inspector reviewed the-following procedures and found them to be satisf actory:
62RP-RAD-010-05. " Receipt of Radioactive Materib1" 42FH-ENG-002-1. " Receiving New Fuel" 42FH-ENG-004-IS. "New Fuel Inspection (8 X 8)"
42FH-ENG-005-1. " Storage of New Fuel"
The next review was conducted in the area of fuel handling, transfer, and core verification.
The inspector _ reviewed the following procedures and found them to be satisfactory:
DI-0PS-37-0889N. " Fuel Movement Rules"
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42FH-ENG-030-05. "Special Nuclear Material Inventory
& Transfer Control" 42FH-ENG-001-1S. " Criticality Rules" 42FH-ERP-014-05. " Fuel Movement Operation" The final review was conducted in the area of inspection of_ irradiated fuel.
The_ inspector reviewed the following procedure and found it to be satisfactory:
246-GP-03 " Fuel Bundle and Individual Rod Visual Examination" During this inspection effort, the inspector determined that the licensee was in the process of combining several refueling-related procedures.
The procedures covering criticality rules, receipt of new fuel, the inspection of new fuel, and the storage of new fuel were being combined into procedure 42FH-ERP-012-05. "New Fuel and New Channel Handling."
The inspector reviewed the final unapproved (not yet approved by the PRB)
t version of this procedure and found that it satisfactorily incorporated the requirements of the individual procedures.
The licensee plans to issue this new procedure during the upcoming Unit 1 condenser retubing/
refueling outag iT
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The second aspect of this inspection effort focused on a review of the administrative requirements for the control of refueling operations.
Several aspects of the licensee's administrative requirements for the control of refueling operations are covered in Attachment 1 (Unit 1 Fuel Movement Prerequisites) of procedure 42FH-ERP-014-05. " Fuel Movement Operation."- This attachment requires the licensee to verify numerous administrative controls prior to fuel movement.
The following administrative controls are included in the fuel movement prerequisites:
(1) the establishment of two-way communications between the control room and the refueling bridge. (2) the proper briefing of the refueling crews, and (3) the checkout of critical equipment such as the refueling bridge.
The prerequisite sheet additionally _ requires the licensee to verify that the requirements specified in the Refueling section of the Unit I technical specifications are satisfied prior to the commencement of refueling operations.
Additionally, the inspector
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verified the existence of casualty procedures.
Specifically, the procedures for Irradiated Fuel. Damage During Handling (34AB-0PS-031-15)
and Decreasing Reactor Well/ Fuel Pool Water Level (34AB-0PS-047-IS) were
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reviewed and found to be satisfactory.
Finally, the inspector verified the adequacy of the licensee's control of foreign material over the open reactor vessel.
Procedure 51GM-MNT-002-05. " Maintenance Housekeeping and Tool Control." provides the administrative controls necessary to maintain the required levels of cleanliness during refueling operations.
No violations or deviations were identified.
Based on the reviews conducted, the inspector concluded that the licensee has adequate procedures and administrative requirements for the control of refueling operations.
10. - Inspection for Verification of Quality Assurance Request Regarding Diesel Generator Fuel Oil - Multi-Plant Action item A-15 (Tl 2515/93) (71707)
Units 1 and 2 On January 7. 1980. NRR requested all-licensees to check their QA programs with respect to D/G fuel oil and to include D/G fuel oil in their QA programs or provide justification for not doing so.
In a response to NRR dated March 7, 1980, the licensee stated that D/G fuel oil was not on the
"Q" list at Plant Hatch.
However, the licensee further stated that D/G
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fuel oil was considered to be in the Plant Hatch QA program because of
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controls placed by the technical specifications and the inclusion of D/G fuel oil in the QA audit program.
(Testing requirements for D/G fuel oil are included in Sections 4.9. A.2.d and 4.8.1.1.2.c of the Unit 1 and 2 Technical Specification:. respectively.)
Since the licensee is testing and auditing D/G fuel oil consistent with its response to NRR. review of this matter by the resident inspectors is close,
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11. Visit to Local Public Document Room On February 7.1990 the NRR Project Manager visited to Appling County Public Library to check the status of the LPDR for Plant Hatch.
The library has one section (approximately 63 linear feet of shelf space)
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devoted to matters pertaining to Plant Hatch. The shelves appeared to be about 95 percent full.
Included in the shelf displays are an updated FSAR for each unit and folders of correspondence and inspection reports pertaining to the plant. A microfilm file cabinet containing more generic material, such as NUREG reports, regulatory guides. Commission issuances.
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NRC Bulletins, etc.. together with a microfilm reader / printer are also available.
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The librarian indicated that several hours per week are required to
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maintain the LPDR material up-to-date.
She also related that requests for LPD9 material are received about once every three months.
The librarian
indicated that the library could provide an adequate service with much
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less effort and considerably less library space if all of the material
.were on microfiche or available via a terminal with access to the NRC i
document system.
The PM will discuss this matter with the NRC Office of Administration. LPDR Branch.
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The PM concluded that_ the overall condition of the Hatch LPDR compares very favorably with other LPDRs that he has visited.
It was also concluded that the facility is adequate to meet. local needs for
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information regarding Plant Hatch.
12.
ExitInterview(30703)
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The inspection scope and findings were sumarized on February 16. 1990, with those persons indicated in paragraph 1 above.
Particular emphasis was placed on the NCVs discussed in paragraphs 2 and 6.
The licensee was also advised that review of LER 366/90-01 discussed in paragraph 6 would remain open.
The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during this. inspection.
Dissenting comments were not received from the licensee.
Item Number Statu_s.
-Description / Reference Paragraph s
321/90-01-01 Opened and NCV - Incorrect Valve Position Closed (paragraph 2)
321/90-01-02 Opened and NCV - Inadequate Operations Shift Closed Turnover Briefing (paragraph 6)
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L 13. Acronyms and Abbreviations Air Handling Unit AHU
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Code of Federal Regulations CFR
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Deficiency Card DC
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. Design Change Request
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p ECCS - Emergency Core Cooling System ESF Lngineered Safety Feature
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General Electric Company GE.
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Diesel Generator D/G
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FCR Field Change Request
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FSAR -' Final Safety Analysis Report
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HCU. - Hydraulic Control Unit HPCI - High Pressure Coolant Injection Licensee h ent Report LER
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LLS.- Low' Low set LOCA - Loss of Coolant Accident i
LPCI - Low Pressure Coolant injection LPDR - Local Public Document Room
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NCR Main Control Room
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MSIV - Main Steam Isolation Valve Maintenance Work Order MWO
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NCV Non-Cited Violation
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Nuclear Regulatory Commission NRC
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Office of Nuclear Reactor Regulation NRR
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OSOS - On-Shift Operations SupervisorL PCIS - Primary Containment Isolation System PM
- Project Manager
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PRB' - Plant Review Board QA Quality Assurance
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Quality Control QC
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RCIC - Reactor Cort-Isolation Cooling
- - Residual Heat Removal Systea RHR RPS. - Reactor Protection System RWCU - Reactor Water Cleanup System SBGT - Standby Gas Treatment System SDV - Scram Discharge Volume SRM Scurce Range Monitor
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SRV Safety /kelief Valve
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SSAC - Station Service Air Colapressor STA Shift Technical. Advisor
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TS Technical Specifications
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WPS Work Process Sheet
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