IR 05000387/1992025
| ML17157C086 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/25/1992 |
| From: | Jason White NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17157C085 | List: |
| References | |
| 50-387-92-25, 50-388-92-25, NUDOCS 9212020065 | |
| Download: ML17157C086 (33) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION I
Inspection Report Nos.
50-387/92-25; 50-388/92-25 License Nos.
Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Facility Name:
Inspection At:
Susquehanna Steam Electric Station Salem Township, Pennsylvania Inspection Conducted:
September 29, 1992 - November 9, 1992 Inspectors:
G. S. Barber, Senior Resident Inspector, SSES D. J. Mannai, Resident Inspector, SSES Approved By:
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J. White, Chief,'eactor Projects 'Sec n
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station activities during normal and backshift hours, including: plant operations, radiation protection, surveillance and maintenance, and safety assessment/quality verification.
One non-cited violation was identified pertaining to inoperable fire doors, as identified by the licensee.
9212020065 92ii25 PDR ADQCK 05000387
EXECUTIVESUMMARY Susquehanna Inspection Reports 50-387/92-25; 50-388/92-25 September 29, 1992 - November 9, 1992 Operations Both Susquehanna units were operated in a safe manner.
Operators effectively controlled plant evolutions and identified plant problems.
Unit 2 outage activities were conducted in a safe manner.
(Reference Sections 1.0 and 2.0)
Radiological Controls Observations during the period indicated that workers generally had good radiological controls practices.
One worker was exposed to a 32 millirad per hour hot particle while installing temporary shielding in the drywell. Upon exit from the drywell, the personnel contamination monitor alarmed.
A Cobalt-60 particle was found on the back of the individual's left knee.
The worker was assessed a skin dose of approximately four rem, which was below 10 CFR 20 and licensee administrative limits.
(Reference Section 3.2)
Maintenance/Surveillance
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The licensee exercised good control of maintenance and surveillance activities.
No scrams were attributable to maintenance or surveillance activities.
Only one ESF actuation occurred during a maintenance activity.
(Reference Sections 1.3 and 4.0)
Emergency Preparedness The inspector completed followup of concerns relative to the implementation of the licensee's personnel accountability procedures during the July 28 Emergency Drill. It was determined that personnel accountability was properly completed in accordance with the licensee's pi'ocedu res.
Engineering/Technical Support The licensee experienced two failures of Potter & Brumfield (P&B) Model MDR relays during surveillance testing while Unit 2 was shutting down for the refueling outage.
Previously, the NRC issued Information Notice 92-04 documenting failures of P&B relays at several nuclear plants.
In response, the licensee's initial evaluation determined 567 P&B relays existed within both units and the common plant; and eight P&B relays had failed previously.
In response to the continuing failures, the licensee initiated an extensive reassessment.
System engineers performed an evaluation of each relay failure and the impact on operability.
Thirteen relays were identified as having potential operability impact on
Unit 1, which was operating at the time.
These relays have been scheduled for replacement within the 'next few months.
The licensee replaced the equivalent relays during the Unit 2 Outage.
Allremaining normally energized safety related and non-safety related P&B MDR relays have been prioritized for replacement.
(Reference Section 7.0)
Safety Assessment/Assurance of Quality The inspector reviewed three Licensee Event Reports during the period.
The inspector identified one non-cited violation related to inoperable fire doors.
(Reference Section 8.0)
SUMMARYOF OPERATIONS 1.1 Inspection Activities The purpose of this inspection was to assess licensee activities at Susquehanna Steam Electric Station as they related to reactor safety and worker radiation protection.
Within each inspection area, the inspectors documented the specific purpose of the area under review, the scope of inspection activities and findings, and conclusions, as appropriate.
This assessment is based on actual observation of licensee activities, interviews with licensee personnel, measurement of radiation levels, independent calculations, and selective review of applicable documents.
Abbreviations are used throughout the text.
Attachment 1 provides a listing of these abbreviations.
1.2 Susquehanna Unit 1 Summary Unit 1 operated at or near full power for the duration of the inspection period.
Operators conducted several routine power reductions during the period to facilitate control rod pattern adjustments, surveillance testing, and maintenance.
No ESF actuation occurred during the inspection period.
1.3 Susquehanna Unit 2 Summary At the start of the inspection period, the Unit 2 reactor was defueled (Condition 5) and in day 18 of its fifth refueling outage.
Routine preventive and corrective maintenance generally progressed as planned.
The Unit finished the period in Condition 4 following a reactor scram for scram time testing of control rods.
The significant events during the outage included:
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On October 19,,the service water (SW) effluent radiation monitor was declared inoperable when elevated background radiation levels sealed in an upscale alarm.
The elevated background levels were caused by a crud burst in the reactor cavity when operators swapped residual heat removal shutdown cooling loops.
The SW effluent monitor recorded a 45 mr/hr dose rate, normal background is 8 mr/hr.
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On October 27, shutdown cooling was lost when the RHR inboard isolation and injection valves closed.
An Engineered Safety Features actuation was initiated when a technician accidentally grounded a loose lead, which deenergized a PCIS relay.
Operators restored shutdown cooling after 51 minutes.
The reactor coolant temperature increased 5'
to a maximum temperature of 112'.
2.0 OPERATIONS The inspectors verified that the facility was operated safely and in conformance with regulatory requirements.
Pennsylvania Power and Light (PP&L) Company management control was evaluated by direct observation of activities, tours of the facility, interviews and discussions with personnel, independent verification of safety system status and Limiting Conditions for Operation, and review of facility records.
These inspection activities were conducted in accordance with NRC inspection procedure 71707.
Unless otherwise noted, the inspector determined that Unit 1 operational and Unit 2 outage activities were conducted safely.
The inspectors performed 12.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> of deep backshift inspections during the period.
These deep backshift inspections covered licensee activities between 10:00 p.m. and 6:00 a.m. on weekdays, and weekends and holidays.
3.0 RADIOLOGICALCONTROLS 3.1 Inspection Activities PP&L's compliance with the radiological protection program was verified on a periodic basis.
These inspection activities were conducted in accordance with NRC inspection procedure 71707.
3.2 Inspection Findings Observations of radiological controls during maintenance activities and plant tours indicated that workers generally obeyed postings and Radiation Work Permit requirements.
No inadequacies were noted.
Hot Particle Exposure On October 14, the licensee identified an unexpected exposure to a worker as a result of a hot particle.
The affected individual had been applying temporary shielding in the dryweH.
Components that were shielding included:
the N2B and N2C nozzles (738'lev.), the equipment drain tank (704'lev.), pipes around the drywell floor drain sump (704'lev.),
and the CRDM laydown area.
After performing this work, the affected individual exited the drywell and alarmed the personnel contamination monitor (PCM).
Health Physics surveyed the worker and found a 32 millirad per hour (mrad/hr) particle on the back side of the left knee.
The particle could have been there for as long as 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
The licensee decontaminated the individual and restricted his access.
The hot particle was recovered and found to be primarily Cobalt-60.
VARSKIN(a computer-based skin dose program) indicated that the individual received a total exposure of 4.085 rem to the skin; of this, 0.335 rem was whole body exposure.
This exposure did not exceed any administrative or regulatory limi t a
The VARSKINprogram was developed by Pacific Northwest Laboratories for NRC (NUREG/CR-4418) to calculate radiation dose to the skin from a given amount of contamination.
This code accounts for the most significant factors affecting skin dose and willcompute the beta radiation dose at a given skin depth from up to five different radionuclides, Personnel Contamination Report (PCR)92-355 documented the event.
The inspector reviewed the circumstance surrounding this exposure and noted that the licensee took prompt actions to locate and quantify the hot particle.
The licensee restricted the affected individual's access and conducted surveys inside the drywell to determine the source of contamination.
Very high levels of contamination were found on and around the equipment drain tank.
The licensee identified that location as the most probable source and decontaminated it to acceptable levels.
The affected individual was given a verbal and written briefing on the relative risks and hazards of hot particle exposures.
The inspector reviewed the licensee's actions and had no further questions, 4.0 MAIN'IXNANCE/SURVEILLANCE 4.1 Maintenance and Surveillance Inspection Activity
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On a sampling basis, the inspector observed and/or reviewed selected surveillance and maintenance activities to ensure that specific programmatic elements were being met.
Details of this review are documented in the following sections.
4.2 Maintenance Observations The inspector monitored maintenance activities to determine ifthe work was conducted in accordance with approved procedures, regulatory guides, Technical Specifications, and industry codes or standards.
The following items were considered, as applicable, during this review: Limiting Conditions for Operation were met while components or systems were removed from service, required administrative approvals were obtained prior to initiating the work, activities were accomplished using approved procedures, quality control hold points were established where required, functional testing was performed prior to declaring the involved component(s) operable, activities were accomplished by qualified personnel, radiological controls were implemented, fire protection controls were implemented, and the equipment was verified to be properly returned to service.
These observations and/or reviews included:
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WA 13337, Inspected Reactor Core Isolation Cooling System Governor Valve Internals, dated October 15.
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WA 26086, Installation of New Division II Suppression Pool (SPOTMOS) Resistance Temperature Devices, dated October 2 I
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WA 20221, Steam Separator Installation, dated October 23.
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WA 20491, Reactor Water Level Shutdown Range Calibration with Reactor Pressure Vessel Head Installed, dated October 25.
4.3 Surveillance Observations The inspector monitored surveillance tests to determine ifthe following criteria, when applicable, were met:
the test conformed to Technical Specification requirements, administrative approvals and tagouts were obtained before initiating the surveillance, testing was accomplished by qualified personnel in accordance with an approved procedure, test instrumentation was calibrated, Limiting Conditions for Operations were met, test data was accurate and complete, removal and restoration of the affected component was properly accomplished, deficiencies noted were reviewed and appropriately resolved, and the surveillance was completed at the required frequency.
These observations and/or reviews included:
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SI-283-325, Eighteen Month Calibration of Main Steam Isolation Valve RPS Limit Switches, dated October 13.
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SO-281-001, Weekly Refueling Platform Operability, dated October 16.
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SM-259-001, Eighteen Month Vacuum Relief Breaker Valve Position Switch Channel Calibration, dated October 21.
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TP-250-004, Reactor Core Isolation Cooling (RCIC) Turbine Overspeed Trip Testing, dated 26 October.
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SE-200-002, Reactor Pressure Vessel Leak Check, dated October 28.
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TP-252-006, High Pressure Coolant Injection Turbine Overspeed Trip Testing, dated October 29.
4.4 Inspection Findings The inspector reviewed the listed maintenance and surveillance activities.
The review noted that work was properly released before its commencement, that systems and components were appropriately tested before being returned to service, and that surveillance and maintenance activities were conducted by qualified personnel.
Where questionable issues arose, the inspector verified that the licensee took the appropriate action before system/component operability was declared.
The inspectors had no further questions on the listed activitie.0 FMERGENCY PREPAREDNESS 5.1 Inspection Activity The inspector reviewed licensee event notifications and reporting requirements for events that could have required entry into the emergency plan.
5.2 Inspection Findings No events were identified that required emergency plan entry.
No inadequacies were identified.
Emergency DrillAccountability The inspector reviewed the licensee's actions during a July 28 Emergency Drillto determine ifthe emergency procedures to verify personnel accountability were properly implemented.
The inspector questioned security personnel, including supervisors, and independently determined that an accountability roster was properly completed in accordance with the licensees procedures.
The inspector determined that the licensee's procedures are designed to affect accountability of personnel within the protected area in the event of an emergency.
Personnel working outside the protected area, such as in a warehouse, are not normally subject to accountability in accordance with the procedure.
In order to assure that personnel are knowledgeable of the requirements of the procedure, a supervisor briefed warehouse personnel of the bases and requirements affecting personnel accountability in an emergency situation.
Based on the above, the inspector had no further questions.
6.0 SECURITY 6.1 Inspection Activity PP&L's implementation of the physical security program was verified on a periodic basis, including the adequacy of staffing, entry control, alarm stations, and physical boundaries.
These inspection activities were conducted in accordance with NRC inspection procedure 71707.
6.2 Inspection Findings The inspector reviewed access and egress controls throughout the period.
No unacceptable conditions were note.0 ENGINEERING/TECHNICALSUPPORT 7.1 Inspection Activity The inspector periodically reviewed engineering and technical support activities during this period.
The on-site Nuclear Systems Engineering (NSE) organization, along with the Nuclear Technology department in Allentown, provided engineering resolution for problems during the inspection period.
NSE generally addressed the short term resolution of engineering problems.
NSE also interfaced with the Nuclear Modifications organization to schedule modifications and design changes, as appropriate, to provide long term corrective action.
The inspector verified that problem resolutions were thorough and directed at preventing recurrences.
7.2 Inspection Findings Potter & Brumfield MDR Rotary Relay Failures The licensee experienced two failures of Potter &Brumfield (P&B) Model MDR relays during the shutdown of Unit 2 for its refueling outage.
The relays failed when the reactor recirculation pump Motor-Generator set "2A" and "2B" drive motor breakers did not automatically open, as expected, during surveillance testing.
The licensee documented this occurrence in Significant Operating Occurrence Report (SOOR) 2-92-092.
~B<~kr ~n The NRC issued Information Notice (IN) 92-04, "Potter &Brumfield (P&B) Model MDR Rotary Relay Failures," on January 6, 1992.
This IN alerted licensees to failures of the subject relays in safety-related systems at certain nuclear power plants. It identified two failure mechanisms:
(1) Mechanical binding of the rotor shaft caused by coil varnish outgassing and the subsequent deposition on the relay rotor shaft.
The contaminants are deposited in the end bell bearings and sleeves, and cause the rotor shaft to bond or stick to the bearing.
This prevents the rotor shaft from fully rotating when the relay coils are energized or deenergized.
The principal contaminant is outgassed material emitted from the brown enamel varnish used to coat the relay coils; and (2) Intermittent continuity of the electrical contacts.
These failure mechanisms apply to P&B MDR relays with a product
"date code" of 90-24 and earlier.
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PP&L's Industry Event Review Program (IERP) reviewed Information Notice 92-04 with a recommended review and assessment in February 1992.
The Nuclear Technology department generated and issued SEA-No. EE-434, "Evaluation of Potter &Brumfield MDR Rotary Relays For Failure Mechanism as Identified in NRC IN 92-04," in June 199 II I
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In their SEA, PP&L identified that 567 total non-latching P&B MDR relays existed in both units.
The licensee determined that the failure modes described in the reference IN did exist at Susquehanna, but that these failure modes were more predominant in environments having elevated temperatures.
The licensee determined that the maximum ambient panel temperatures are 122'F, per Environmental Qualification (EQ) calculations.
Further, the licensee found that normally deenergized relays are not generally subject to outgassing at Susquehanna, based on plant failure history and equipment cabinet temperatures.
Therefore, PP&L concluded that the principal failure mode generally involved normally energized relays.
PP&L identified approximately 213 P&B MDR relays that are normally energized.
Of those relays, 139 are used in control circuits for safety related systems.
Attachment 2 lists the particular relays, systems affected, and potential system impact.
The licensee identified eight P&B MDR relays that have been replaced since 1982, five were normally energized and three were normally deenergized.
The licensee determined the total MDR relay failure rate as of June 12, 1992, to be 2.306 x 10 ~ failures per component hour.
The energized MDR relay failure rate was 3.82 x 10~ failures per component hour.
To further minimize these low failure rates, the licensee identified recommended corrective actions in their SEA.
These included:
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Cycling of all continuously energized MDR relays; and
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For normally energized MDR relays, replacement of all old varnish coil type relays with epoxy coil type MDR relays having a product date code 90-25 or later.
In addition to the two relay failures identified during the Unit 2 shutdown, the licensee found two additional failed relays during their planned end-of-life environmental qualification program replacement.
These findings contributed to a higher than expected failure rate for this type relay.
As a result, the licensee performed the following:
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Field walkdown of relays to identify "date codes;"
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System engineering review and documentation of the effect on system operability for each relay failure, including the frequency of relay cycling;
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A comprehensive document search of past Work Authorizations, Non-Conformance Reports, Licensee Event Reports, Significant Operating Occurrence Reports, etc. to form a plant database containing historical information;
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Date code check of relays staged for use in the current outage; and
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Purge of spare relays in storage with date codes of 90-24 or earlie The NSE performed a field walkdown and identified an additional 21 relays beyond the 213 relays identified in SEA-No. EE-434.
Of the total 234 normally energized relays identified, 33 safety related ("Q") relays were in Unit 1, 47 "Q" relays were in Unit 2, and 97 "Q" relays were common.
The balance of the relays were non-safety related (non "Q"). Work Authorizations were written for all normally energized "Q" and non "Q" P&B MDR relays and priorities set for replacement.
Since the recent Unit 2 shutdown, the licensee has evaluated 130 "Q" relays in Unit 1 and common for operability, and identified 13 Unit 1 relays as having a potential for affecting system operability.
The failure of the relays would result in the following: (1) With a loss of offsite power, the RHR pumps would not load onto their respective buses in the proper time sequence, which could result in an emergency diesel generator lockout; or (2) Primary containment isolation components may not automatically change position upon receipt of an initiation signal, however, manual initiation capability would be unaffected.
The licensee found that the remaining relays may have minor system impact but should not affect system operability.
All 13 of the identified relays were DC powered.
PP&L searched plant history and failed to locate any failures of DC relays that were in service for less than seven years.
Twelve of the 13 relays in question were replaced in 1989, under the EQ program.
All 13 relays are scheduled to be replaced with new P&B MDR relays within the next few months; and within within the expected service life experienced at Susquehanna.
The licensee evaluation concluded that a failure of any of the 13 relays is not likely to occur within the next few months based on plant specific performance history for the affected relays.
To date, PP&L has experienced 12 total failures of P&B MDR relays, nine of which were normally energized.
Failure history shows that the less frequently the relay is cycled, the greater the potential for failure per the mechanisms described in IN 92-04.
The frequency of cycling for the failed, normally energized relays was as follows:
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3 relays
- 18 month cycle
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2 relays
- 5 year cycle
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1 relay
- no cycling
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2 relays
- cycled frequently
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1 relay
- cycled quarterly PP&L determined that the four most recent failures were due to binding of the relay.
Approximately four of eight failures identified in the original SEA were also due to binding.
During the Unit 2 refueling outage, the licensee replaced the 13 relays having the potential to impact operability.
Twelve of the 13 relays having potential operability impact were part of a group of 20 relays that had been previously identified for replacement under the EQ program.
Consequently, a total of 21 P&B MDR relays were replaced during the outag I V
e
The licensee concluded that operability of the systems containing the 13 referenced relays remained unaffected.
This was based on a review of the plant historical data, walkdown of energized MDR relays, system engineering review of relay failure effects, and the Nuclear Technology department's independent review.
PP&L identified no other engineering basis for failure.
Based on these activities, the licensee also concluded that this condition was not reportable.
The licensee plans to replace all of the remaining normally energized safety related and non-safety related P&B relays in both units by the end of the next outage for each plant.
The next Unit 1 outage is scheduled to commence in September 1993 and the next Unit 2 outage is scheduled to commence in March 1994.
PP&L intends to send failed relays out for testing by an independent laboratory.
The specific scope of testing is under review.
The Nuclear Technology department is currently updating SEA-No. EE-434, and performing an independent review of Nuclear System Engineering relay walkdown results.
Beyond normal surveillance testing and the EQ program replacement, the licensee does not plan any special preventive or corrective maintenance practices for this type relay.
Surveillance frequencies vary from monthly to 18 months for safety related "Q" relays.
7.3 Inspector Conclusion The licensee response was prompt and thorough, and the inspector concurred with the licensee's operability and reportability determinations.
The inspector concluded that the extensive reassessment performed by Nuclear System Engineering following the relay failures experienced during the Unit 2 shutdown was comprehensive and commensurate with the safety significance of the problem.
Although the licensee replaced the 13 relays of particular concern during the Unit 2 refueling outage, the inspector observed that the SOOR resolution failed to document what corrective action was taken or planned for Unit 2 relays.
The licensee committed to update the SOOR resolution to include the corrective action for Unit 2.
The schedule of relay replacement was aggressive considering the large number of relays.
Although not addressed in the SEA, the licensee is evaluating normally deenergized P&B MDR relays in an Engineering Discrepancy Report.
The inspector willcontinue to monitor the P&B MDR relay failure issue resolution as new information becomes available.
8.0 SAFETY ASSESSMENT/QUALITY VERIFlCATION Licensee Event Reports The inspector reviewed Licensee Event Reports (LERs) submitted to the NRC office to verify that details of the event were clearly reported, including the accuracy of the description of the cause and the adequacy of corrective action.
The inspector determined
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whether further information was required from the licensee, whether generic implications were involved, and whether the event warranted onsite followup. The following LERs were reviewed:
2-11-
"Fir D r In in Pr hi i T
hni
ifi in" On June 30, 1992, with both units at 100% power, contractor plant fire protection auditors discovered three fire doors having mechanical problems with the closing and latching mechanisms.
The auditors contacted the PP&L fire protection engineer and it was decided that the doors were operable, since they were closed and latched.
However, the surveillance requirements include the operability of the door hold-open, release, and closing mechanisms; thus, the licensee should have entered the action statement for TS LCO 3.7.7.
On July 1, 1992, the licensee identified that this criterion was not reviewed during the initial operability assessment, and subsequently entered the applicable action statement.
As corrective action, the licensee willconduct training for work groups on the importance of notifying shift supervision upon discovery of problems with TS related equipment.
The licensee is also developing a Fire Barrier Required Action cross reference, for fire doors and dampers, to assist in future operability determinations.
The safety significance of this event is low since the licensee verified the doors were closed and latched daily, and the appropriate action was taken after they discovered the missed operability requirement.
This violation willnot be subject to enforcement action because the licensee's efforts in identifying and correcting the violation meet the criteria specified in Section VII.Bof 10 CFR Part 2, Appendix C.
2-14-
"Mn f rr'n ll in n
L ri n fr ln idVlv i n M n 1R ired In rr-rin On September 9, 1992, licensee engineers became aware that Circle Seal solenoid valves used in both units could be susceptible to deterioration of their 0-rings.
The manufacturer's manual incorrectly recommended the use of Vaseline Petroleum Jelly (containing hydrocarbons) could be used with ethylene propylene elastomers.
This event was reviewed in NRC Inspection Report 50-387/92-22.
2-
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n 1 nn FA i n-RHR h>
wn lin I
i n" On September 26, 1992, with Unit 2 in Condition 5, the RHR Shutdown Cooling outboard containment isolation valve closed.
There was no valid isolation signal present and there were no safety consequences since alternate decay heat removal methods were available.
No unacceptable conditions were identifie.0 MANAGFMENTAND EXITMEETINGS 9.1 Resident Exit and Periodic Meetings The inspector discussed the findings of this inspection with station management throughout, and at the conclusion of, the inspection period.
Based on NRC Region I review of this report, and discussions held with licensee representatives, it was determined that this report does not contain information subject to 10 CFR 2.790 restrictions.
9.2 Inspections Conducted By Region Based Inspectors 10/5-10/9/92 Security 10/26-10/30/92 Electrical Followup Inspection ggggr~g 92-26 92-28 Reporting
~ng~gr G. Smith C. Woodard
i
ATl'ACHMENTI r viati n Li
,
AD ADS ANSI ASME CAC CFR CIG CRDM CREOASS DG DX ECCS EDR EP EPA EQ ERT ESF ESW EWR FO FSAR HVAC IERP ILRT I&C JIO LCO LER LLRT LOCA LOOP MSIV NCR NDI NPE NPO NQA NRC NSE
- Administrative Procedure
- Automatic Depressurization System
- American Nuclear Standards Institute
- American Society of Mechanical Engineers
- Containment Atmosphere Control
- Code of Federal Regulations
- Containment Instrument Gas
- Control Rod Drive Mechanism
- Control Room Emergency Outside Air Supply System
- Diesel Generator
- Direct Expansion
- Emergency Core Cooling System
- Engineering Discrepancy Report
- Electrical Protection Assembly
- Environmental Qualification
- Event Review Team
- Engineered Safety Features
- Emergency Service Water
- Engineering Work Request
- Fuel Oil
- Final Safety Analysis Report
- Heating, Ventilation, and Air Conditioning
- Industry Event Review Program
- Instrumentation and Control
- Justifications for Interim Operation
- Limiting Condition for Operation
- Licensee Event Report
- Local Leak Rate Test
- Loss of Coolant Accident
- Non Conformance Report
- Nuclear Department Instruction
- Nuclear Plant Engineering
- Nuclear Plant Operator
- Nuclear Quality Assurance
- Nuclear Regulatory Commission
- Nuclear Systems Engineering
revi i n Li n
ATI'ACHMENT1 OI OOS PC PCIS PMR PORC PSID QA RB RBCCW RCIC RG RHR RHRSW RPS RWCU SGTS SI SO SOOR SPDS SPING TS TSC WA
- Open Item
- Out-of-Service
- Protective Clothing
- Primary Containment Isolation System
- Plant Modification Request
- Plant Operations Review Committee
- Pounds Per Square Inch Differential
- Quality Assurance
- Reactor Building
- Reactor Building Closed Cooling Water
- Reactor Core Isolation Cooling
- Regulatory Guide
- Residual Heat Removal Service Water
- Standby Gas Treatment System
- Surveillance Procedure, Instrumentation and Control
- Surveillance Procedure, Operations
- Significant Operating Occurrence Report
- Safety Parameter Display System
- Sample Particulate, Iodine, and Noble Gas
- Technical Specifications
- Work Authorization
Attachment
CATEGORIES OF FAILURE IMPACT OF CONTINUOUSLY ENERGIZED MDR RELAYS 1.
Diesel Generator Supply Fans - 4 Relays ACTION:
Prevents shutdown of Diesel Generator Vent Supply Fans.
2.
instrument Air Compressor
- 4 Relays ACTION:
Failure of relay would prevent shutdown of Instrument Air Compressor.
3.
Instrument Air Dryer - 1 Relay ACTION:
Failure of relay would prevent shutdown, of Instrument Air Dryers 1F116C 5 D when required.
4.
RBCCW Pump Failure to Shutdown -
1 Relay
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ACTION:
Relay failure to change state will maintain RBCCW 2P210A sealed in the Run position, 5 ~
RB Zone 3 Isolation Damper Failure to Close When Required - 12 Relays ACTION:
Relay failure to change state along with a RB Zone 3 Isolation Signal - Div 1 would keep RB Zone 3 dampers from closing on demand.
6.
Control Room AC, Fan Prevented from Stopping - 2 Relays ACTION:
Relay failure may prevent fan from being stopped.
7.
Core Spray Bus Undervoltage - 5 Relays ACTION:
Failure of relay could affect timing of when Core Spray pumps start.
8.
RHR Pump Start Circuit - 4 Relays ACTION:
'Relay failure to deenergize wiii keep RHR pumps stoppe Attachment p
9.
4KV Dead Bus, Lack of Indication - 6 Relays ACTION:
Relay failure may cause RHR pumps to be aligned to dead buses.
10, Reactor and Turbine Building Chillesr - 25 Relays ACTION:
May prevent normal remote operation of Chilled Water System Chiller Compressor motors, 11.
Drywell Area Unit Coolers - 32 Relays ACTION:
Relay failure will prevent Drywell Area Unit Coolers from going into Fast mode of operation.
Failure will not prevent tripping of Fans, however..
12.
Suppression Pool - 2 Relays ACTION:
Failure of relay prevents shutdown of Suppression Pool Water Filter Pumps 13.
Diesel Generator - 2 Relays ACTION:
Failure of relay impacts availability of Biesel Generator A upon loss of normal 4KV power.
14.
Containment Isolation - 6 Relays ACTION:
Failure of relay impacts completion of Containment Isolation 15.
Recirculation Pump Trip - 8 Relays ACTION:
Tripping of Reactor Recirculation Pumps is prevented.
Also provides false permissive for RX Recirc Pump MG Set 1B Drive Motor. Also, indication to Transient Monitoring System is incorrect.
16.
Reactor Building HVAC Fans, Safety Function Prevented
- 5 Relays ACTION:
Failure of relay may allow continued operation of RB Ventilation Supply Fans during LOCA condition Attachment
17.
ESW Pump and RHR Pump Operation Start Permissive Present when no Power Available - 6 Relays ACTION:
Failure of relay gives potential for ESW pumps OP504C Bc D and RHR SW Pumps 1P506A and 1P506B to be lined up to a dead Bus.
18.
Erroneous Start Permissive for RHR Pumps - 4 Relays ACTION:
Relay closes switch to indicate normal power available which is a permissive to start RHR pumps. Relay failure would allow pump to be aligned to a deenergized bus.
19.
Suppression Pool Nitrogen Make Up Valves - 2 Relays ACTION; Failure of the relay to open contacts when required would maintain Nitrogen Make Up valves open.
20.
ESSW and RHR SW pumps supply and house fans - 8 Relays ACTION; Failure may prevent fans from stoppin flECFIVEi>-BFGION !
'92 NOY 17 P2:05