ML20115J989
| ML20115J989 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/18/1996 |
| From: | Mueller J NEBRASKA PUBLIC POWER DISTRICT |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NLS960138, NUDOCS 9607250179 | |
| Download: ML20115J989 (87) | |
Text
COOPER NUCLEAR STATION P.O. DOX 98, BROWNVLLE. NEBRASKA 68321 Nebraska Public Power District
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u NLS960138 July 18,1996 U. S. Nuclear Regulatory Commission Attention: Document Centrol Desk Washington, D.C. 20555
Subject:
10 CFR 50.59(b) Summary Report Cooper Nuclear Station NRC Docket No. 50-298, DPR-46 Gentlemen:
In accordance with the provisions of 10 CFR 50.59(b)(2) and Paragraph 6.5.1.C.2 of the Cooper Nuclear Station Technical Specifications, the Nebraska Public Power District hereby submits a summary of facility changes, tests, and experiments completed in accordance with the requirements of 10 CFR 50.59(b). This report covers the 1995 time period, as well as those changes incorporated in the fourteenth revision of the Updated Safety Analysis Report.
In accordance with 10 CFR 50.4, the original report is enclosed for your use, and copies are being transmitted to the NRC Regional Office and the NRC Resident Inspector for Cooper Nuclear Station.
Should you have any questions or comments regarding this report, please contact me.
Sincerely, f
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Jt H. Mueller Vice-President, Nuclear Cooper Nuclear Station JHM:lb
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Op Attaclunent
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Regional Administrator USNRC - Region IV Arlington, Texas NRC Resident inspector - CNS
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LIST OF NRC COMMITMENTS l ATTACHM2NT 3 -l
-l Correspondence No: NLS960138 l
The following table identifies those actions committed to by the District in this document. Any.other actions discussed in the submittal represent intended or planned actions by_the District. They are described to the NRC for the NRC's information and are not regulatory commitments.
Please notify the Licensing Manager t
at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITTED DATE COMMITMENT OR OUTAGE
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PROCEDURE NUMBER 0.42 l
REVISION NUMBER 1.1 l
PAGE 9 OF 11 l
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o REPORTABLE SPECI AL PROCEDURES (SPs) /SPECI AL TEST PROCEDURES (STPs)
SP 92-022 TITLE:
Operation / Maintenance of Model 2000 Cask DESCRIPTION: This SP provided instructions for the Model 2000 Transport Package which was used to ship Reactor Pressure Vessel (RPV) surveillance specimens to General Electric for testing. The Model 2000 is a steel encased lead shielded shipping cask. This SP successfully removed and sent offsite the RPV specimens trom the Spent Fuel Pool Storage Pool.
SAFETY ANALYSIS:
ne Model 2000 Transport Package was designed in accordance with the criteria of Federal Regulations 10CFR71 and 49CFR173. This includes meeting both the normal transport and accident condition requirements of the NRC and DOT. Cooper Nuclear Station is a registered user of this cask. The major safety significance was associated with movement of the Model 2000 cask in and out of the spent fuel pool area and on the refueling floor. The Restricted Mode of operation for the Reactor Building crane was used to keep the cask from traveling over the spent fuel storage racks and to allow the cask to travel only over structural members that can syport a cask drop accident. By using *be Restricted Mode, a cask drop accident involving the Model 2000 cask was encompassed by the safety analysis of the IF 300 Spent Fuel Cask and an adequate safety margin was assured. This SP was the coordinating document for use of components in accordance with established procedures, thus ensuring all established safety margins and Technical Specifications were maintained.
SP 95-109 TITLE:
Troubleshooting of Reactor Water Cleanup (RWCU) Channel A Space Temperature Switches DESCRIPTION: This SP provided guidance to troubleshoot the RWCU space temperature switches in response to a spurious % Group 3 isolation that occurred in July 1995. His SP tested the Steam Leak Detection temperature switch logic in the RWCU Pump Room for a possible shorted condition which could have contributed to the spurious isolation of the RWCU system. The test was performed during the 1995 refueling outage and revealed no shorted conditions. The four Division 11 RWCU Steam Leak Detection Temperature Switches were subsequently replaced after the completion of this SP.
SAFETY ANALYSIS:
This SP was performed with the safety function initiated (RWCU containment isolation valves closed) and did not subject any other components or systems to abnormal conditions. The consequences of an accident were unchanged since the RWCU system was already isolated from the primary containment and the RWCU system itself performs no safety function. This SP cycled the RWCU Channel A space temperature switches utilizing permanently installed heat wires. The procedure method was the same as used for the monthly surveillance testing of these switches. This SP did not modify or change the operation of any safety related equipment or modify the plant in such a way that a new initiator for an accident was created. No setpoint changes or equipment chaages were introduced by this SP. The margin of safety as defined in the basis for any Technical Specification was not reduced.
SP 95-114 TITLE:
General Electric Electrical Penetration Repair DESCRIPTION: This SP provided guidance for the repair of three Essential /EQ electrical containment penetrations by General Electric. Penetrations X100A, X101C, and X104D were exhibiting leakage and required resealing for Local Leak Rate Testing concerns. The penetrations were pressurized and leak tested to determine the location of the leakage. Initially, leakage was found on the inboard sides of all three 1
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penetrations. Repairs were performed by General Electric by applying qualified sealant to seal the leak paths. It was determined that penetration X101C was also leaking on the outboard side and it was subsequently repaired by CNS personnel using the General Electric methods. Although leak-free performance was not attained on all three penetrations, substantial improvements were achieved and the repair effort was considered successful.
SAFETY ANALYSIS:
This activity did not increase the probability of occurrence or consequences of an accident previously evaluated in the USAR because the only result of this SP was the restoration of the penetrations to their original analyzed design condition. No equipment operational parameters were changed, no new equipment functions were added, and the per.etrations were restored to their analyzed configurations.
No new failure modes were added and no existing failure modes were changed. This SP did not adversely affect the Technical Specification margin of safety. The margin between the Integrated Leak Rate Test allowable limit and the actual leakage was increased.
SP 95-128 i
TITLE:
ASME Class 1 Hydrostatic Pressure Test DESCRIPTION: This STP provided instructions and acceptance criteria for performing the Class 1 System Hydrostatic Pressure Test of piping and components in accordance with the CNS Second Interval ISI Pressure Test Program,Section XI of the ASME Boiler and Pressure Vessel Code, and Plant Technical Specifications. It also provided instructions for testing portions of Class 2 Systems that are required to be tested in conjunction with the Class I pressure test.
SAFETY ANALYSIS:
This test was performed with the reactor shutdown and all control rods in. Emergency Core Cooling Systems and Secondary Containment were operable as required by Plant Technical Specifications.
Compliance with the Pressure Test Curve in Figure 3.6.2 of the Technical Specificatiom precluded brittle failure of the vessel during the test. The test pressure did not affect the structuralnagrity of the reactor coolant pressure boundary. Although the safety relief valves were gagged during this test, the other safety valves remained operable and provided over pressure protection. Precautions were provided in the event a recirculation pump tripped to prevent temperature stratification. Any through-l wall leakage or high pressure jumper hose failure detected during this test would be bounded by the I
accident analysis for a line break. Therefore, this STP did not increase the probability or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR, nor create the possibility of a different type of accident or equipment malfunction. This test was consistent with the requirements of Technical Specifications 3.6.A.3 and 3.6.G.
STP 90-352C j
TITLE:
SW-MO-36 and SW-MO-37 Dynamic Flow Differential Pressure Test DESCRIPTION: The purpose of this STP was to perform a design basis dynamic flow differential pressure test of SW-MO-36 and SW-MO-37 Motor Operated Valves in accordance with District's MOV program. The differential pressure testing data was transmitted to the MOV group for analysis to provide assurance that SW-MO-36 and SW-MO-37 will operate when subjected to design basis conditions.
SAFETY ANALYSIS:
Implementation of this STP was performed during plant shutdown conditions when the tested portion of the affected system was not required to perform its safety functions as governed by the Limiting Conditions for Operation of the Technical Specifications. The testing and verification in this STP ensured that the affected components will adequately perform their safety and design functions. This STP did not affect the design basis of any systems as described in the USAR; therefore, the probability of occurrence er consequences of an accident or malfunction of equipment were not increased.
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STP 91-066 l
l TITLE:
Electro-llydraulic (EH) Constant Pressure Pump Assembly DESCRIPTION: This STP provided for the installation and performance evaluation of Westinghouse Electric's constant pressure pump assembly on the turbine high pressure fluid system, in place of the existing vane type pumps. The constant pressure discharge pump assembly was installed to reduce pipe vibration, reduce equipment fatigue, and increase system reliability. The EH system was monitored far a period of six months after the installation of the new pump assemblies. Vibration readings taken J
at the end of the six-month period provided sufficient evidence that the constant pressure discharge pumps provided less vibration than the original vane pumps and, therefore, reduced the stress of the Eli System piping. Consequently, an Engineering Work Request was initiated to document the installation of the STP as a permanent change. Reference Minor Modification 95-089.
SAFETY ANALYSIS:
This STP evaluated the performance of an installation that improved the reliability of the Turbine Generator High Pressure Fluid System. The Turbine Generator Fluid System is a nonessential system and does not provide a safety function. Mo equipment important to safety was installed or modified j
by this STP. The Turbine Control Valve Fast Closure (63/OPC) pressure switch is the only essential component that interfaces with the system affected by this STP. By reducing EH piping vibration, system fatigue was also reduced, thus extending the service life of all EH components and connections including the 63/OPC pressure switches. This STP did not increase the probability of occurrence or consequences of an accident or malfunction of equipment previously evaluated in the USAR. The roargin of safety as defined in the basis for the Technical Specifications was not reduced.
STP 94-334 l
TITLE:
RHR-DPIS-125A/B Functional Flow Test l
DESCRIPTION: The purpose of this STP was to: 1) gather data necessary to verify the accuracy of the RHR flow elements at low flow rates,2) determine the flow rate / differential pressure when RHR-DPIS-125A/B actuates to open/close the RHR minimum flow bypass valves, and 3) determine the system's capability in a degraded state. This information was required to verify compliance with Technical Specification requirements and to provide a basis for Design Change 94-332, RHR Minimum Flow i
Bypass Valve Modification.
SAFETY ANALYSIS:
This STP was performed while the plant was shutdown and RHR was not required to be operable.
The division of RHR required for shutdown cooling was not involved in this activity. No operating margins or parameters were exceeded during this STP. The probability of occurrence or consequences of an accident or malfunction of equipment important to safety were not increased by this test. This STP did not reduce the margin of safety as defined in the basis for any Technical Specification because it involved only a portion of the RHR system which is not required for accident mitigation during shutdown conditions.
STP 95-005 l
TITLE:
Fire Detection Supervisory Circuits Verification DESCRIPTION: This STP was performed to ensure that each fire detector associated with Technical Specification 4.14.B is electrically configured to its appropriate detector zone and to ensure that the supervisory circuits associated with each detector are operable. This was accomplished by removing each of the detectors individually and ensuring that the correct trouble annunciation was received. The test pushbuttons for each detector zone were verified operable to allow current testing required by Technical Specification 4.14.B to be validated. Actions were taken to correct any deficiencies identified during performance of this STP.
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SAFETY ANALYSIS:
His STP temporarily disabled affected detectors; however, a fire watch patrol was established prior to the performance of the test in order to ensure protection of plant equipment from fire. This STP i
is specific to the Fire Detection system, and as such does not penain to any accident previously i
evaluated in the USAR. The Fire Detection system is not the source of a postulated accident, nor can j
operation of the system result in a postulated accident. No permanent changes resulted from implementing this STP. The detectors and associated circuitry were restored to their original configuration and tested to ensure that the system was fully functional upon compleiivoitesting.
This STP provides assurance that the Fire Detection system will operate as designed. It does not i
affect any Technical Specification basis.
STP 95-066 TITLE:
As-Build Penetration X-29A Instrumentation DESCRIPTION: The purpose of this STP was to resolve discrepancies between historical and walkdown information in order to determine the arrangement of instruments provided via penetration X-29A and to determine the correct de:,ignation of valves NBI-V-57, NBI-V-58, NBI-CV-23BCV, and NBI-CV-24BCV.
SAFETY ANALYSIS:
This STP did not permanently change the op: ration of any system or permanently alter any parameters. STP 95-066 removed several reactor ussel level and reactor vessel pressure instruments from service which provide inputs to RilR, ADS, RPS, PCIS, liPCI, RCIC, ARl/RPT, Main Turbine, and Reactor Feedwater systems. Ilowever, this STP was performed with the plant in a cold shutdown condition. With the plant in this condition, the reactor was shutdown and the Main Turbine and RFP turbines were secured. ADS, IIPCI, RCIC, and ARI/RPT were not required to be operable and Primary Containment integrity was not required. The LPCI system could have been required to be operable, but the 2/3 core coverage instrumentation affected by the STP does not prevent operation of the LPCI mode. The removal ofreactor vessel level and pressure instruments from service reduces the margin of safety normally provided by the redundancy ofinstrumentation included in the plant design. Ilowever, with the plant in a cold shutdown condition, operability requirements are limited to instrumentation associated with RPS, LPCI Subsystem B, and Primary Containment surveillance instrumentation. Therefore, STP 95-066 was able to be performed within the requirements of Technical Specifications. This STP removed instrumentation from service and restored instrumentation to service in the same manner as routine surveillance procedures.
STP 95-123 1
TITLE:
Impact of11V-FAN-(BF-C-1 B) on Control Room Emergency Filter System DESCRIPTION: The purpose of this STP was to gather the data necessary to determine the impact of non-operation of Control Room Recirculation Fan 11V-FAN-(BF-C-1B) on the onerability of the Control Room Emergency Filter System (CREFS). He data was used to determine it non-operation of the fan will result in violation of Technical Specification or administrative limits. The STP testing verified the concem that operating the CREFS with BF-C-1B shut down would result in filter system flow exceeding the Technical Specification limit of < 990 cfm. Plant Temporaty Modification 95-21 was implemented during the 1995 outage to install an alternate power supply for BF-C-1B, which is normally powered by Division I, so that it could also be operated by Division II. An Engineering Work Request has been initiated to permanently install the PTM configuration. Operating Procedure 2.2.84 provides appropriate restraints to address this concern during operation.
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l ANALYSIS:
This STP did not permanently change the operation of any system or permanently alter any l
parameters. The CREFS is safety-related; however, it is used for mitigation of an ai cident only.
I Therefore, this STP did not increase the probability of occurrence of an accident. This STP created the potential for the CREFS flow rate to exceed the Technical Specification limit due h operation of the system without the control room HVAC recirculation fan; however, the STP was performed while in an LCO for an operable CREFS. Therefore, the margin of safety of the CREFS as defined in the basis of the Technical Specifications was not reduced. The STP also included limitations and contingencies to terminate the test and return the system to the normal analyzed configuration upon detection of a design basis accident. The STP operated the control room ventilation systems in an abnormal configuration; however, the components of those systems were not operated outside their design parameters. Positive pressurization with filtered air was maintained throughout the STP.
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REPORTABLE DESIGN CHANGES (DCs)/ MINOR MODIFICATION PACKAGES (MMPs)
DC 86-008A TITLE:
Seventh Condensate Filter Demineralizer Unit (Phase !!)
- DESCRIPTION: This Design Change installed a seventh condensate filter demineralizer in order to increase system efficiency. The original DC 86-008 (Phase 1) installed piping tie-ins, up to and including the isolation valves. This DC installed the remaining piping, pipe supports, valves, control cabinet, pump, instrumentation, electrical cable and conduit, and instrument air tubing and sensing lines.
' SAFETY ANALYSIS:
No safety-related equipment was installed or modified by this DC. The piping and supports installed were nonessential and meet the original codes of design and construction. This DC did not increase the probability of occurrence or consequences of an accident or malfunction of safety related equipment. This DC did not involve any equipment as defined in the Technical Specifications.
Service Water Pump Backup Gland Water Flow Time Requirements DESCRIPTION: His DC amendment increased the allowed time that a Service Water (S #) pump can operate without gland water injection during a specific Appendix R fire scenario. Amendment 2 allowed SW pump l
operation without gland water flow for up to ten seconds. Amendment 3 allows pump operation without gland injection for up to twenty seconds.
SAFETY ANALYSIS:
his amendment ensures an adequate supply of gland water to the SW pumps within the time required by the pump manufacturer. It does not degrade the overall operation of any system or equipment used to mitigate the consequences of an accident and serves to reduce the possibility of equipment malfunction. This amendment did not create the possibility for a previously unidentified accident because its potential failure modes would not create conditions in excess of those previously analyzed.
The affected systems maintain their same functions as described in the Technical Specifications. The margin of safety as defined by those Technical Specifications was not decreased.
M MP 90-273 TITLE:
- Steam Jet Air Ejector (SJAE) Radiation Monitor Replacement DESCRIPTION: This Minor Modification replaced existing S.IAE radiation monitors with General Electric NUMAC radiation monitors. The existing monitors were high maintenance items and experienced sensitivity to electromagnetic interference. The replacement monitors are microprocessor based and not as susceptible to the sensitivity problems of the existing units.
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-SAFETY j
ANALYSIS:
The design specifications of the replacement monitors are equivalent to the existing radiation
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monitors and the function is not changed. The monitors have no accident initiation or mitigation j
capability. The microprocessor design of the replacement monitors has greater reliability based on CNS experience with the same type of monitors for the Main Steam Line. De replacement monitors r
satisfy the same performance requirements and margins as discussed in the Technical Specification l
bases. The number of channels, trip points, and surveillance requirements discussed in Technical Specifications were not changed.
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MMP 91-011 A
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TITLE:
Augmented Off-Gas (AOG) Panel Recorder and 112 Indicator Replacements 1
1 DESCRIPTION: This Minor Modification provided enhancements to the AOG system control panel, as follows:
- 1) replacement of three high maintenance recorders with new recorders, and 2) replacement of two 112 indicators for H2 analyzers with higher resolution indicators. The replacement recorders will I
provide better reliability, accuracy, programmability and performance. The new H2 indicators will allow calibration of the H2 analyzers to tighter tolerances and minimize spurious downscale alarms.
SAFETY l
ANALYSIS:
This activity did not affect any equipment whose malfunction is postulated in the USAR to initiate
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an accident or prevent an accident from occurring. This activity did not interact with any structure, i
system, or component important to safety. No new hazards were created that can be postulated to cause an accident or malfunction of equipment different than previously analyzed in the USAR. This activity did not change the design basis, function or operation of any equipment important to safety and Technical Specification requirements were not afTected; therefore, this activity did not reduce the margin of safety as defined in the basis for the Technical Specifications.
MMP 91-Ol l B TITLE:
Augmented Off-Gas (AOG) System Vent / Drain Piping Modification DESCRIPTION: This Minor Modification implemented improvements to the drain and vent piping for various AOG subsystems and equipment, as follows: 1) installation of permanent tubing and a vent valve for each of the four +34 Degree Glycol System vent valves; 2) installation of a two-inch pipe header in the charcoal bed room to connect the ten vent valves for the -30 Degree Glycol System to a common drain to the glycol tank; 3) installation of a valve to provide a bypass line around preheater I A and 1B shell side drain traps; and 4) installation of drain lines and supports for AOG-FT-2001 drain pots.
These modifications were implemented to enhance personnel safety, improve operational convenience, allow smoother operation of the preheater during startup, and facilitate draining and prevent leaks in the drain lines.
SAFETY ANALYSIS:
This modification did not affect any equipment whose malfunction is postulated in the USAR to initiate an accident or prevent an accident from occurring. This activity did not create any interaction with any structure, system, or component important to safety. No hazards were created that could cause an accident or malfunction of equipment different than previously evaluated in the USAR. This activity did not change the design basis, function or operation of any equipment important to safety and did not affect Technical Specification requirements; therefore, this modification did not reduce the margin of safety as defined in the basis for the Technical Specifications.
M MP 91-011C TITLE:
Augmented Off Gas (AOG) Panel Upgrades DESCRIPTION: 'Ihis Minor Modification implemented enhancements to the AOG system control panel, as follows:
- 1) replacement of AOG system temperature recorder AOG TR-103; 2) separation of AOG common annunciator window 2/6-3 into two separate windows; 3) addition of cooling louvers to rear panel doors on AOG control panel, and 4) removal of abandoned devices from AOG control panels. These modifications were implemented to replace obsolete equipment, prevent nuisance alanns and enhance overall operation of the AOG system, reduce negative pressure buildup inside the panel, and improve housekeeping.
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SAFETY ANALYSIS:
This modification did not affect any equipment whose malfunction is postu!ated in the USAR to initiate an accident or prevent an accident from occurring. This activity did not create any interaction with any structure, system, or component important to safety. No hazards were created that could cause an accident or malfunction of equipment different than previously evaluated in the USAR. This activity did not reduce the margin of safety as defined in the basis for the Technical Specifications.
The only AOG system parameter controlled by the Technical Specifications is hydrogen monitoring; the subject modifications are not associated with hydrogen monitoring.
MMP 91-011D TITLE:
AOG-FR-101 Recorder Replacement DESCRIPTION: This Minor Modification replaced the AOG system recorder AOG-FR-101 which had become obsolete. Due to complications of the present location in the AOG Control Room Panel, the recorder was moved to an adjacent AOG Analyzer Control Panel and mounted in place of a spare recorder.
SAFETY ANALYSIS:
This modification did not affect any equipment whose malfunction is postulated in the USAR to initiate an accident or prevent an accident from occurring. His activity did not create any interaction with any structure, system, or component important to safety. No hazards were created that could cause an accident or malfunction ofequipment different than previously evaluated in the USAR. His activity did not reduce the margin of safety as defined in the basis for the Technical Specifications.
He only AOG system parameter controlled by the Technical Specifications is hydrogen monitoring; this recorder is not associated with hydrogen monitoring.
MMP 91-056 i
TITLE:
Replacement of Motor Generator (MG) Exhaust Fan Dampers 1
DESCRIPTION: This Minor Modification replaced the existing MG Exhaust Fan dampers with dampers of a heavier design. De existing dampers were worn beyond repair and did not provide complete air shutoff. It also modified the control air to the air operators to increase the supply air and installed new supports j
due to the additional weight of the new dampers. This modification was performed during the 1995 outage when Secondary Containment was not required to allow access for transporting materials into j
and out of the MG Exhaust Fan Room.
I SAFETY ANALYSIS:
The MG Set ventilation is important for power operation but loss ofit will not initiate an accident.
The discharge dampers are not required in order to mitigate the consequences of an accident. The operation of the MG Set ventilation did not change per this modification. The new components perform the same functions as the previously existing ones and the new dampers will increase the reliability of the system by eliminating operational problems. Therefore, this modification did not increase the probability o consequences of an accident or malfunction of equipment important to safety as previously evaluated in the USAR. There is no margin of safety defined in the Technical Specifications for the MG Set ventilation.
i DC 91-077. DC 91-077-1 TITLE:
Low Level Radwaste Storage Facility DESCRIPTION: This Design Change upgraded the Low Level Radwaste (LLRW) storage capacities and facilities at CNS. A concrete storage pad was constructed outside of the protected area to temporarily store LLRW in high integrity containers in Temporary Storage Modules. To facilitate storage of Dry Active Waste (DAW) in the Multi-Purpose Facility (MPF), a 10-inch concrete wall was constructed to provide active shielding to areas adjacent to the storage area. This DC also analyzed the existing 8
4 Augmented Radwaste Building (ARB) truck loading area to determine if the structural slab has adequate capacity to carry loads caused by the loading of LLRW storage modules onto trailers; the new loading was shown to be within the design loading used in the original ARB design.
SAFETY ANALYSIS:
Equip nent looted in the plant was not modified by this DC. Tne 10-inch shielding wall added in the MPF was shown not to affect the existing structure. Therefore, the probability of occurrence and consequences of previously analyzed accidents were not affected. Postulated failure modes do not impact the cperation, safety, or reliability of the plant. The possibility of an accident of a different type than previously evaluated in the USAR is not introduced. The Main Steam line break and fuel handling accident are bounding when compared to the analyzed accident exposures from a LLRW accident. This modification did not affect any of the Limiting Safety System setpoints or system safety settings. Therefore, the margins of safety or design bases as outlined in the Technical Specification are unaffected.
DC 91-144 TITLE:
RilR lleat Exchanger Tube Plugging Margin DESCRIPTION: This purpose of this Design Change was to provide the technicaljustification for increasing the tube plugging margin of RiiR lleat Exchangers A and B from the existing limit of 4% to 15%, as an alternative to replacing the heat exchangers. At the time of this study, the percentage of tubes plugged in RilR 11 eat Exchangers A and B was 3.7% and 3.2%, respectively. The main factor in increasing the tube plugging margin was the use of the ANS 5.1 decay heat model. Ileat transfer calculations were performed which verified temperature limits would be maintained with a conservative tube plugging margin of 15% (analytic limit was 23.1%). An analysis of the tube plugging history estimates that the heat exchangers will not reach the 15% limit before the year 2008. No physical modifications were involved with this DC.
SAFETY ANALYSIS:
The function and operation of the RilR heat exchangers were not altered by this DC. The reliability of the heat exchangers was not degraded as a result ofincreasing the tube plugging margin. Analyses performed for the DBA-LOCA and limiting transients show that the long-term, peak containment pressures and temperatures are still below acceptance limits. Long-term operability of the RiiR and Core Spray pumps is maintained since adequate net positive suction head (NPSil) margin exists and pump seal temperature limits will not be exceeded. Therefore, containment integrity will not be compromised and this DC will not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety. There will be no adverse affect on the Service Water system resulting from the small increase in pressure drop due to the increased tube plugging margin. De margin of safety as defined in the basis for any Technical Specification is not reduced.
DC 92-159 TITLE:
Saw Deck Renovation DESCRIPTION: His modification removed the existing Machine Shop Decon llood Exhaust System in it's entirety and also removed the exhaust fan serving the southwest corner of the Weld Shop. An existing exhaust fan on the 903' level was reducted to serve the southwest corner of the Weld Shop. He space previously occupied by this equipment on the 928' level was utilized to construct new office space for the Maintenance Depanment.
SAFETY ANALYSIS:
The only USAR related items in this non-essential modification,1-EF-MS-1 A/lB, are not considered in any accident scenario and do not affect any equipment important to safety. These fans and their associated filters and ductwork were removed, ne former Machine Shop Decon Exhaust liood System is not considered in any accident scenario or mitigation activity and does not interface with 9
any equipment important to safety. This modification cannot create an accident scenario that would be nuclear safety related. Only power from non-essential switchboard MS-B was milized. Thus, there was no impact on any Technical Specification margin of safety.
Diesel Generator (DG) Upgrades DESCRIPTION: The purpose of this Design Change was to improve the reliability, availability, and maintainability of the Diesel Generators. he major modifications included in this DC were: 1) replacement of t! e Fuel Oil Booster Pump, Pre / Post Lube Oil Pump, and Lube Oil Bypass Pump assemblies, along with associated piping modifications; 2) replacement of the Woodward EGA governing system with a new design system that incorporates various improvements; 3) replacement of existing AJAX overspeed govemor and AMOT trip system with a Woodward overspeed governor and a new trip system; and
- 4) reversal of the logic of the Control Air system (part of the Starting Air system) from an " air to run" configuration to an " air to stop" configuration. Amendment I was subsequently issued to correct:
- 1) wiring discrepancies that were discovered during DC implementation; 2) an Appendix R design deficiency; and 3) an as-building required change.
SAFETY ANALYSIS:
The subject DG modifications do not impact the probability of occurrence of an accident or an operational transient. The consequences of an accident can be impacted by the reliability of the DGs in providing emergency power as designed; however, this modification improves DG reliability and availability. The DGs supply emergency power to ESF systems used for accident mitigation; however, since this modification does not affect any of the DG loads, it will not change the way that i
the mitigation systems perform their function. Diesel Generators No. I and No. 2 retain safety shutdown features and emergency operation functions as specified in the CNS USAR.
Implementation of this DC serves to increase reliability and availability of the DGs by replacing obsolete equipment with improved and upgraded components, eliminating the need for control air pressure at the engine for DG operation in the emergency mode, and adding several new features which will enhance CNS's ability to meet regulatory guidance concerning DGs. The design, materials, and construction standards specified by this DC meet or exceed the standards originally established for the system and components affected. This DC does not create any new DG failure modes during emergency operation. A previously existing failure mode, failure of the DGs due to loss of control air pressure, was eliminated by this DC. Another existing failure mode, failure of DG No. 2 due to an Appendix R fire, was eliminated by Amendment I to this DC. This modification does not create the possibility of a different type of accident or malfunct.an than those previously evaluated in the USAR and does not reduce the margin of safety as defined in the Technical Specifications.
j DC 93-050. DC 93-050-1 DC 43-050-2 TITLE:
Appendix J Testing in the Accident Direction DESCRIPTION: This Design Change (DC) was implemented to enable Local Leak Rate Testing (LLRT) of twenty valves in the accident direction. Valves affected were in the Main Steam (MS), Reactor Core Isolation Cooling (RCIC), Iligh Pressure Coolant Injection (HPCI), Residual lleat Removal (RHR),
and Primary Containment (PC) systems. Modifications consisted of the installation of test connections, vent valves, and manual block valves. This DC also authorized the drilling of one-founh inch holes in the reactor side of the valve discs of two Core Spray (CS) motor-operated valves to alleviate pressure Scking concerns. Two PC valves which were made obsolete bv DC 89-272 and had been kept in a locked open status since that time were also remove / ' y DC 93-050.
Amendment I to this DC authorized the drilling of a one-fourth inch hole in the setor side of the valve disc for two additional valves (one llPCI valve, one RCIC valve) in order to alleviate pressure locking concerns. Amendment 2 to this DC authorized replacement of four heavy weight PC drain valves with lighter weight needle valves at four locations on the Standby Nitrogen Injection System.
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t SAFETY ANALYSIS:
This DC enhances the capability of the affected systems by allowing local leak rate testing of the valves in the accident direction per 10CFR50, Appendix J. These changes increase the assurance of leak tightness of several containment isolation valves. All piping and component additions were evaluated to ensure that piping configurations meet appropriate design codes and standards. The J
venting of the CS, llPCI, and RCIC valves ensures proper operation of these valves under accident conditions by eliminating the potential for pressure binding. Appropriate Technical Specification LCOs were implemented to ensure adequate core and containment cooling requirements were maintained during installation of this DC. The changes made by this DC do not negatively impact the function of any system or component important to safety. Potential failure modes were analyzed and determined not to create a condition in excess of that previously analyzed. The margin of safety i
as defined in the basis of the Technical Specifications is not reduced.
I i
Removal of Non-Essential RPS MG Set Trips DESCRIPTION: This DC removed thiee nonessential trip functions (undervoltage, overvoltage, and underfrequency) from the Reactor Protection System Motor Generator (RPS MG) Set output breaker. These trip functions had a history of unnecessarily tripping the RPS MG Set output breaker thus causing half scrams and group isolations. Class 1E Electrical Protection Assemblies (EPAs) are installed on the load side of the RPS MG Set output breaker that have identical trip functions as the nonessential RPS MG Set trips; however, their trip setpoints are more conservative. He EPAs have proven to be more reliable than the RPS MG Set trips; therefore, removal of the nonessential trips increases the reliability of the RPS MG Sets as a source of power.
SAFETY ANALYSIS:
ne redundant EPAs located downstream of the RPS MG Sets have the same trip functions as those removed from the RPS MG Sets and will continue to provide the essential trip functions. The trip.
functions of the EPAs were not modified nor were the EPA setpoints changed. The EPA's continue to provide protective functions such that the analysis of the loss of a RPS MG Set as addressed in the USAR remains valid. Removal of the nonessential trips reduces the number of challenges to the RPS components. The probability or consequences of an accident or malfunction of equipment are not increased and the margin of safety as defined in the basis for any Technical Specification is not reduced.
DC 93-120 TITLE:
Scram Discharge Volume (SDV) Pressure Regulating Valve Modification DESCRIPTION: Ris DC installed a check valve and a relief valve in a bypass line around each of the two pressure regulators which are integral parts of the North and South Scram Discharge Instrument Volume Drain Isolation Valve Air Actuators. This modification was implemented due to problems experienced in meeting the Technical Specification maximum limit of 30 seconds closing time for the North and South Scram Discharge Volume Drain Isolation Valves. Implementation of this DC reduced the closing time for the subject valves. Also, in order to place the SDVs in a more testable configuration during inservice Testing, this DC installed a manual isolation valve and check valve to bypass the SDV vent and drain pilot valves.
SAFETY ANALYSIS:
This DC did not alter the basic function of the Scram Discharge Volume (SDV), but removed existing design deficiencies and, therefore, improved reliability. Material compatibility was maintained to current requirements. His modification did not introduce any new failure modes to the SDV or the Control Rod Drive (CRD) system. He CRD and SDV retain their design functions during both normal plant operation and reactor scrams. The overall performance of the SDV subsystem is i
I l
11 l
l
(
improved by decreasing the closing time for the drain valves. Implementation of this DC did not i
reduce the margin of safety as defined in the basis of any Technical Specification because it does not i
increase the required SDV maximum closing time of 30 seconds.
i l
DC 93-150 I
TITLE:
Replacement of SW-MOV-36MV 1
DESCRIPTION: His DC replaced Motor Operated Valve SW-36MV with a new, but similar valve. SW-36MV is a 24"llenry Pratt butterfly valve. This modification was implemented to provide the necessary shaft exposure on the MOV to suppon diagnostic testing required by Generic Letter 89-10. A support modification was also requbed due to the increased weight of the replacement valve. The original scope of this DC included replacement of SW-37MV; however, replacement of this valve was i
postponed to a future outage.
l SAFETY I
ANALYSIS:
His DC enhanced the structural design margin available for SW-36MV. Since the primary function of this modification is to allow GL 89-10 testing, reliability of the valve is enhanced. He replacement valve was evaluated for the design basis functions of the valve and meets the design i
requirements. His valve's accident mitigating functions or ability to perform its design basis safety
{
related functions were not reduced in any way. His DC does not alter operating conditions or syA:
- parameters or introduce any new failure modes. One division of Service Water remained in service during the installation.
DC 93-150-1 TITLE:
Replacement of Motor Operator and Gearbox on Valve SW-MOV-37MV DESCRIPTION: This DC araendment provided for replacement of the motor operator assembly for valve SW-MOV.
37MV without the replacement of the valve. Replacement of the valve was originally within the scope of DC 93-150; however, valve replacement was deferred to a future outage. Replacement of the motor operator assembly was required to assure seismic qualification of the assembly. The DC amendment used the existing motor and replaced the motor operator and gear box. This amendment increased the motor gearing output capability of valve SW-37MV.
SAFETY ANALYSIS:
This DC amendment enhanced the design margin available for SW-37MV and did not change the function of any equipment. Since the primary function of this modification was to provide for qualification of the operator assembly, reliability of the valve was enhanced. The replacement operator assembly was evaluated for the design basis functions of the valve and the replacement operator assembly meets these design requirements. This valve's accident mitigating functions or j
ability to perform its design basis functions were not reduced in any way. This DC amendment did not alter operating conditions or system parameters or introduce any new failure modes. One division of Service Water remained in service during implementation.
DC 03-151 TITLE:
RR MO-MO53A/B Upgrade DESCRIPTION: This DC modified several features of RR-MOV-MO53 A&B in order to increase the design margin available for closing these valves under design basis conditions by increasing the available output torque and thrust. The 100 ft-Ib,250 VDC motors on Reactor Recirculation discharge valve motor operators, RR-MO-53A&B, were replaced with 150 ft-lb,250VDC motors. Motor overload relays and fuses in the starters for these MOVs were also replaced, as well as various associated feeder cables.
12
l i
SAFETY
- ANALYSIS:
RR-MOV-MO53A/B has an essential safety function to close to support LPCI injection during a postulated recirculation system pump suction line break. He larger motor and cabling size increases the available output torque of this motor. This provides an increase in the design margin available j
for this valve to complete its safety function. This DC does not modify or otherwise affect the designed function of RR-MOV-MO53A/B. Since no essential function has been affected and the design margin for a credited safety function has been enhanced, this DC does not increase the probability or consequences of an accident or malfunction of equipment previously evaluated in the USAR, nor create the possibility of a different type of accident. This DC involved the replacement of semcted plant equipment with equipment of similar design, configuration, and function. It did not result in the introduction of component failure modes or consequences that are not already bounded by that for the previously installed equipment. There is no reduction in the margin of safety as defined in the basis of any Technical Specification.
DC 93-151 A TITLE:
RCIC-MOIS Operator Upgrade DESCRIPTION: This DC replaced the existing SMB-000 motor operator for RCIC-MOIS with a SMB-00 motor operator, and replaced the existing motor with a 10 ft-lb motor. This modification was required to enhance the margin between the required opening / closing thrust / torque and the maximum allowed opening / closing thrust / torque. He thrust / torque margin for this MOV is required by Generic Letter 89-10.
SAFETY ANALYSIS:
This DC does not alter the design basis of RCIC-MOIS and increases the available thrust / torque margin to this valve during performance of its safety related fonction. This DC does not alter operating conditions or system parameters or introduce any new failure modes. His valve's accident mitigating functions or ability to perform its design basis safety related function has not been reduced in any way. Replacement con;ponents were evaluated for the design basis functions of this valve and meet the design requirements. The probability of occurrence or consequences of an accident or malfunction of equipment has not been increased as a result of this modification. The margin of safety as defined in the basis for any Technical Specification is not reduced.
DC 93-153 TITLE:
RIIR-MO39A and RilR-MO39B Operator Replacements DESCRIPTION: This DC replaced existing Limitorque SMB-1 motor operators on RHR-MO39A and RHR-MO39B with Limitorque SMB-2 units. The purpose of this DC was to increase the available torque / thrust margin for these valves during performance of their safety related function.
SAFETY ANALYSIS:
His DC enhances design margin available for these valves and does not change the function of any equipment. Replacement components were evaluated for the design basis function of these valves and meet the design requirements. Rese valves' accident mitigating functions or ability to perform their design basis safety related functions have not been reduced in any way. This modification results in a slight increase in stroke time; however, the change in stroke times for these valves was evaluated by an Engineering Judgement which documents the acceptability of the change in stroke time and concludes that the safety function of the RilR system is not affected. No specific limits for stroke times for these valves, except the limits set up per the IST program which is based on actual performance of these valves, exist in the Technical Specifications.
13
i l
DC 93-182 I
l TITLE:
Diesel Generator Voltage Permissive Relay Replacement 1
DESCRIPTION: This DC replaced existing voltage permissive relays, DG-REL-DGl(59) and DG-REL-DG2(59), in the closing logic of Diesel Generator output breakers EGI and EG2. The existing instantaneous voltage relays were replaced with overvoltage relays which are of a solid state construction. This modification was implemented as a result ofprevious out-of-calibration occurrences with these relays.
SAFETY j
ANALYSIS:
This modification did not change the closing control logic; however, the reliability of the breakers has been increased since the new relays are less susceptible to setpoint drift This DC does not change the ability of the Diesel Generators and the 125 VDC system to perform their safety function in an 1
accident. The new relays meet applicable seismic and quality criteria. Breaker / fuse coordination and electrical load analysis show that relay failure will not propagate in failure to the 125 VDC system.
Although the new relays require 125 VDC power whereas the existing relays did not, a loss of 125 VDC power is already an analyzed event. The margin of safety as defined in the basis for any Technical Specification is not reduced because the new relays actually provide improved reliability.
M MP 93-234 TITLE:
Protection Against Malevolent Use of Vehicles at CNS - Passive Barriers DES CRIPTION: His Minor Modification documented the installation of passive barriers around the perimeter of the protected area to protect against malevolent use of vehicles at CNS. The passive barriers consist of l
cable / bollards and bollard systems. Ileat tracing cable installed for removable bollards is fed from the 12.5 KV system. The installation of the passive barrier system, along with the installation of the active barrier system per MMP 93-234.1, brings NPPD into compliance with the new requirements of 10CFR73.55.
SAFETY ANALYSIS:
The portion of the security system affected by this Minor Modification is located outside the protected area and affects no system that is important to safe shutdown. This MMP affects the loading profile of the 12.5 KV system only. The load being added will not cause the 12.5 KV system or equipment to exceed their design capacity. In addition, a fuse was installed to prevent any fault from affecting i
the remainder of the 12.5 KV system. Due to the isolation of equipment faults provided by the aforementioned fuse, the possibility of an accident or malfunction of equipment of a different type than paviously evaluated in the USAR was not created.
MMP 93-2342 TITLE:
Protection Against Malevolent Use of Vehicles at CNS - Active Barriers DESCRIPTION: This Minor Modification documented the installation of two active barriers to protect CNS against the use of malevolent vehicles. This modification installed one barrier by the main Security Entrance and one by the East Warehouse. Both active barriers require AC power for operation. This power is supplied by the 12.5 KV and non-essential plant systems. The installation of the active barriers, in conjunction with the passive barriers installed by MMP 93-234, brings NPPD into compliance with the new requirements of 10CFR73.55.
SAFETY ANALYSIS:
De 12.5 KV system is not safety related and the new electrical equipment added is adequately sized and properly protected so that any localized fault will not adversely affect the 12.5 KV rystem or any other safety related system. A breaker loading and coordination evaluation was performed which showed that an electrical failure will not propagate to equipment important to safety. The non-safety related equipment installed by this modification will not directly or indirectly cause degradation of any safety related system or components.
14
i I
Control Room Emergency Bypass Fan Upgrade DESCRIPTION: This DC was implemented to enhance the ability of the Control Room Emergency Filter System j
(CREFS) to maintain a positive pressure in the Control Room during isolation. This was I
accomplished by replacing the bypass fan with a higher capacity fan, returning the filter unit back to its design rating by reinstalling the third charcoal tray, and removing a restricting orifice from the CREFS. These changes enable CREFS to develop up to 1000 CFM of filtered make-up air for j
Control Room Envelope Pressurization, compared to an original capacity of about 350 CFM. License j
Amendment 167 was issued to revise the CNS Technical Specifications to refwet the changes made by this DC prior to declaring the CREFS operable.
j SAFETY l
ANALYSIS:
This DC increases the performance and reliability of the Control Room flabitability System by increasing the level of poshive pressurization in the Control Room Envelope during emergency j
operation. By ensuring consistent and reliable positive pressurization, the possibility of radioactive air infiltration into the Control Room is reduced and the potential for unfiltered inleakage is minimized. Personnel dose rates will be increased slightly because of the higher flow rates, but calculations indicate they will remain within regulatory limits and industry guidelines. All affected j
equipment continues to operate within design limits. No new failure modes have been created and the existing failure modes remain unchanged. The probability of occurrence of an accident or i
nWfunction of equipment have not been increased. This DC does not reduce the margin of safety denned in the Technical Specifications for Control Room pressurization and filter performance.
DC 94-041 TITLE:
Reload 16 Analysis DESCRIPTION: This DC provided an analysis to address the core changes associated with nuclear fuel Reload 16 (Cycle 17) of Cooper Nuclear Station. The Reload 16 core design has been reviewed and found to be acceptable with respect to nuclear safety design bases and the Technical Specifications. The Standby Liquid Control (SLC) tank level was revised to provide a minimum boron concentration of 660 ppm in the reactor pressure vessel, post-injection (reference License Amendment 173).
SAFETY ANALYSIS:
The core and fuel design have a direct impact on the plant response to a number of transients and accidents evaluated in the USAR. However, Reload 16 core changes themselves do not modify any of the equipment malfunctions or procedural errors that are analyzed as accident initiators in the USAR. The Reload 16 core changes do not adversely impact the integrity of the radiation release barriers (fuel cladding, reactor vessel, and reactor coolant system). 'ihe Reload 16 core design was analyzed in accordance with NRC approved methods (described in GESTAR II). Analysis of the limiting over pressurization event demonstrates that the peak calculated pressures in the reactor vessel and reactor coolant system are less than those allowed by the ASME Code. Based on the above, it was concluded that the Reload 16 core changes do not increase the probability of occurrence or consequences of an accident or equipment malfunction. Margins of safety potentially affected by the Reload 16 changes are related to fuel limits (MCPR, MAPLilGR) and to the reactor vessel / reactor l
coolant system (maximum pressure). Analyses oflimiting USAR transients for Reload 16 establish operating limit MCPR values that ensure that the safety limit MCPR is not violated. Thus, the margin of safety to fuel cladding failure is not reduced. For the Loss of Coolant Accident, the MAPLHGR limits are developed to ensure that the limits of 10CFR50.46 are met. By meeting these limits, the consequences of a LOCA are not increased and the margin of safety is not reduced.
15
The change to the minimum level in the SLC tank was made as a result of a change to the Technical Specifications. It will ensure a larger volume of sodium pentaborate solution is available for injection into the reactor vessel in the event of a failure of some or all of the control rods to insert. His change does not affect the chemical concentration of the liquid solution and does not change any other physical configuration of the plant. By increasing the amount of sodium pentaborate solution injected, the shutdown margin is increased, ensuring that the minimum required shutdown margin is maintaicd. The probability or consequences of an accident or malfunction of equipment are not increased. His change has no efTect upon those SLC parameters which maintain conformance with the ATWS rule.
MMP 94-060 TITLE:
SF6 Injection Valves Documentationfrube Routing DESCRIPTION: This Minor Modification documented as a permanent modification the performance of Plant Temporary Modification 93-58 to facilitate the injection of sulfur hexaflouride gas as a tracer for on-line Main Condensate waterbox leak detection. His modification also installed the necessary tubing, check valves, and isolation valves needed to remove the test point from the high radiation area of the condenser area to a lower radiation area in the Turbine Building.
SAFETY l
ANALYSIS:
The probability of an accident was not increased due to the leak detection components meeting or exceeding the original piping material and strength requirements. Logic functions related to Main Condenser Vacuum, Steam Jet Air Ejectors, and vacuum pumps remain unchanged. There are no accidents related to the Main Condenser described in the USAR. There is an indirect reactor scram and containment isolation logic input based on inadequate Main Condenser vacuum; however, this function is unaffected by this change. The Main Condenser leak detection techniques have no safety basis, accidents remain bounding and unaffected, and operational events are unaffected. Enhanced leak detection techniques should reduce the potential for chemical intrusion. The logic of the protective features described in the Technical Specifications for operational event mitigation have not been altered by this Minor Modification.
DC 94-075B TITLE:
Appendix R Safe Shutdown Lighting Addition DESCRIPTION: This DC installed additional Appendix R lighting units in the Reactor Building, Control Building, Turbine Building, Intake Structure, Critical Switchgear Rooms, and Diesel Generator Rooms. Dese lights are required to ensure adequate lighting for operator actions and access and egress routes to meet Appendix R requirements.
SAFETY ANALYSIS:
This DC provides additional illumination for operator action or access / egress routes dunng scation blackout conditions and is, therefore, considered an enhancement to the existing emergency lighting system design basis. All lighting and associated hardware were installed to Class IS and IIS cr.teria.
The new safe shutdown lights are the same as the existing safe shutdown lights. Failure of a lighting l
unit will not result in a postulated accident or introduce any new failure modes. No Technical Specification basis is affected by this DC.
MMP 94-138
)
TITLE:
liigh Pressure Coolant injection (IIPCI) System Startup Transient Improvement DESCRIPTION: This Minor Modification incorporated the improvements to the 11PCI system recommended by General Electric Service Information Letter 480. This modification reduced the initial rate of11PCI turbine acceleration which in turn reduced the effect of the startup associated water slug on applicable 16
exhaust piping to acceptable levels. Engineering analysis indicated that the effect of water accumulation in the steam exhaust piping for the llPCI system had the potential to exceed code and pipe support operability limits of related piping due to water slug effects during IIPCI startup. As a j
result of the analysis, the IIPCI system was declared inoperable.
SAFETY ANALYSIS:
This modification reduced the initial speed ramp of the liPCI turbine af ter system initiation. It did not affect any other plant system nor increase the probability of an accident. There are no new electrical components installed as a result of this modification and, therefore, no type of electrical fault propagation. The only new equipment which could potentially add a new failure mechanism is a check valve in the hydraulic system. However, the failure of an essential check valve is considered a passive failure which is outside of the CNS Licensing Basis. Since the modification improved the initial startup transient of the llPCI turbine, it lowered the probability of an initial system trip and loss of the liPCI system. This modification, in conjunction with Minor Modifications95-090 and 95-100, allowed the liPCI system to be returned to an operable status.
M MP o4-169 TITLE:
Control Rod Drive (CRD) Removal Chute lloist Replacement DESCRIPTION: This Minor Modification replaced the existing stationary hoist with a new portable hoist. He existing hoist did not have sufficient power to raise the CRD removal chute. He new hoist is a one-ton chain fall hoist with a i 15V electrical plug-in adaptor attached for a power supply. This modification also removed an existing structu*al brace that was no longer being used.
SAFETY ANALYSIS:
He hoist that was replaced is a subcomponent of the CRD Equipment llandling Platform and its sole purpose is to raise and lower the CRD removal chute ramp. The hoist and CRD, when removed from the housing, are stand alone pieces of equipment and do not affect the function of any other equipment. Therefore, this modification did not increase the probability or consequences of a malfunction of equipment important to safety. When removing CRDs from under-vessel, they are removed from their housing and placed on the removal cart. At that time they are considered out of service and no longer affect the CRD system. Berefore, this activity did not increase the probability of an accident. He existing hoist was replaced by a larger and improved design hoist. If the hoist were to fail and cause a CRD or the cart to come loose, it could only come in contact with the CRD Equipment Platform, which is non-safety related. The CRD removal chute hoist is not defined in the basis for any Technical Specification.
DC 94-250 TITLE:
Emergency Core Cooling Systems (ECCS) Leakage Monitoring Program and RiiR lieat Exchanger "A" Flange Leak Connection DESCRIPTION: This DC established the basis for the limits used in the ECCS Leakage Monitoring Program. The analyzed limit for ECCS leakage is 602 ml/ min. It also documented the acceptability of an existing drip tray and drain line configuration which captures leakage from the RilR Heat Exchanger A shell-to-tubesheet fiange.
SAFETY ANALYSIS:
Inspection for ECCS leaks and monitoring of the identified leaks provides for early indication of the degradation of ECCS piping components in the Reactor Building. The program could therefore decrease the probability of gross component failure. ECCS leakage less than or equal to 602 ml/ min will have a negligible effect on the calculated radiological accident doses for a LOCA. The RiiR lleat Exchanger has been seismically restrained and so will not adversely affect the operation and reliability of the RilR system or any other safety-related system. The drip trJy and tubing 17
i components are passive in nature and do not create any new failure modes. A limit for ECCS leakage into secondary containment is not discussed in the Technical Specifications; however, the established limit is well within the acceptance criteria as stated in the NRC SER for CNS.
DC 04-373-1 TITLE:
Intake Structure Guide Wall Modification (Amendment 1) j l
DESCRIPTION: A hole had been cut in the north end of the intake Structure guide wall in accordance with the original j
DC 94-373 in order to assure Service Water system operability during design basis low river level conditions. Because of the Corps of Engineers' control of the river level during the navigation season, the hole in the guide wall is not necessary during the navigation season. DC 94-373 included installation of a steel gate that can be used to close the hole. This DC amendment installed the guide wall gate for the 1995 navigation season and provided a scheduling mechanism for the gate's future removal and installation. The gate will minimize the amount of silt entering the intake Structure, ne amendment also installed a catwalk with handrail and a gate storage frame on the intake Structure guide wall.
SAFETY ANALYSIS:
Perfomiance of this amendment does not increase the probability of occurrence or consequences of an accident or malfunction of equipment. This amendment assures that a malfunction of the SW system does not occur due to low river levels. The consequences of failure ofimportant to safety equipment which uses the intake Structure's heat sink remains bounded by the USAR described analysis. Performance of this amendment does not reduce the margin of safety as defined in the basis for any Technical Specification.
DC 95-033 TITLE:
Sump Z Modifications for Standby Gas Treatment (SGT) Operability DESCRIPTION: This Design Change modified the existing Z Sump controls and power supply to provide added assurance that the sump system will remain operable post-accident in support of SGT system operation. This DC consisted of the following activities: 1) installation of a new welding receptacle to provide the ability to temporarily route a new power cable from the Off Gas Building to the Emergency Diesel Generator Building; 2) installation of new level switches on each Z Sump Pump; and 3) replacement of existing heat trace line and insulation on the liquid radwaste discharge line near the Z sump to ensure freeze protection.
SAFETY ANALYSIS:
he probability of occurrence or consequences of an accident previously evaluated in the US.AR was not affected by this change. This modification was performed with the plant in cold shutdown when no ERP discharge was required. This design provides added assurance that Z Sump water removal required in support of SGT will occur post-accident. He probability of equipment malfunction will be reduced by providing added reliability to support system components that may be required post-accident for SGT operation. This DC does not change the performance characteristics or failure consequences of the SGT system or any other component important to safety as described in the USAR. This DC does not reduce the margin of safety as defined in the Technical Specifications since the DC provides added assurance that the SGT system will operate as required to support the margin of safety defined in Technical Specifications.
18
DC 95-036 l
TITLE:
Service Water (SW) Sequential Start Timer Setting Change DESCRIPTION: His Design Change revised the time delay setting of the SW load sequencing time delay relays from 15 seconds to 13 seconds. It also revised existing surveillance procedures and other associated documentation to reflect that the maximum allowable diesel generator (DG) start time is now 14 seconds instead of 16 seconds. This DG acceptable start time change did not require any physical modifications. The purpose of this DC was to bring the plant to within the design basis bounds established by the CNS SW Water llammer Analysis for SW pump start time and 001 4 analysis (reload accident analysis) for Core Spray (CS) full flow time.
SAFETY ANALYSIS:
This DC does not affect the safety-related functions of systems and components evaluated in the USAR for design basis accidents. The changing of the SW load sequencing time reduces the potential for SW water hammer and increases the availability of the SW system for support of the Diesel Generators. By reducing the maximum allowable DG start time, this ensures that the CS system reaches full flow within analyzed times and prevents fuel cladding failure. Hus, this DC ensures the ECCS response to postulated Design Basis Accidents will be within analyzed bounds. ne existing DG transient calculation already assumes a concurrent start of the CS and SW pump motors so changing the SW load sequencing time delay does not increase a DG failure probability. This DC does not change the control logic of the SW pump motors or the Diesel Generators. This time delay relay setting change, alor"; with the new DG start time of 14 seconds, results in a SW pump start within 31 seconds of a LOOP event, thus ensuring the criteria in the SW Water Hammer Analysis is satisfied. There is no reduction in the margin of safety as defined in the basis of any Technical Specification.
DC 95-037 TITLE:
Primary Containment (PC) and Traversing incore Probe (TIP) Valve Isolation Reset Modification DESCRIPTION: his Design Change modified the control logic for the PC 1300 series and Neutron Monitoring-TIP (NMT) 104 series primary containment valves. The purpose of this modification was to prevent these valves from automatically opening if their control switch is in the open position and the primary containment isolation signal is reset. This modification brings the control logic of these valves into compliance with the USAR.
SAFETY ANALYSIS:
This DC modified the control logic of valves that are used to mitigate the consequences of an accident; it specifically ensures there is not an inadvertent opening of a primary containment penetration by resetting of the primary containment isolation signal. Therefore, this DC ensures the probability or consequences of an accident are not increased. In fact, the primary containment function is enhanced. The modifications perfonned by this DC do not affect or alter the valves' safety function. This DC added components in the existing control logic for the affected valves. However, failure of these components results in the same effect as prior to this DC. The PC and NMT systems are designed with second barriers to take into account failure of one of the barriers. The final result of this DC is to ensure the valves continue to provide a leak-tight boundary to prevent unmonitored radiation releases from occurring post-accident at their respective penetrations. Since this DC brings the affected valves into compliance with stricter requirements of the US AR, the margin of safety as defined in the Technical Specifications is not reduced.
19 1
I DC 95-037A TITLE:
Group 6 Isolation Reset Modification DESCRIPTION: This DC modified the circuitry of both divisions for the Reactor Building Vent liigh Radiation (Group 6) Relay systems. His modification was required due to a Condition Report evaluation which determined that the Standby Gas Treatment system and Reactor Building IIVAC systems realign to the normal operating mode when the isolation signals for these systems are reset. NRC Bulletin 80-06 l
specifies that systems which receive isolation signals should not return to normal positions after the i
isolation signal is reset. His DC corrected that deficiency by adding a seal-in contact to the circuitry for relays PC-REL-RMAX and PC-REL-RMBX, which after the isolation is reset, will remain deenergized and prevent the applicable systems from retuming to their normal status / positions. Since these relays are required to be energized to also energize the Group 6 relays, they will also not l
become deenergized. A manual pushbutton action will be required to allow the Group 6 relays and
)
their applicable systems to return to their normal status / positions.
SAFETY ANALYSIS:
This DC did not modify or otherwise affect the safety-related functions of any structures, systems, or components evaluated in the USAR for design basis accidents. Consequently, this DC did not increase the probability or consequences of an accident or malfunction of equipment previously evaluated in the USAR. Containment integrity requirements were enhanced by ensuring that SGT
)
and Reactor Building IIVAC remain properly aligned afler the applicable Group 2 isolation is reset.
Implementation of this DC was performed during cold shutdown when containment integrity was not required. This change in control logic did not create the possibility of a different type of accident or malfunction of equipment. The margin of safety as defined in the Technical Specifications was not reduced as this DC did not affect the basis of any Technical Specification.
DC 95-044. DC 95-044-1. DC 95-044-1 Rev.1. DC 95-044-2 TITLE:
Replacement of Safety Relief Valve (SRV) Tailpipe Vacuum Breakers DESCRIPTION: The original DC replaced two SRV tailpipe vacuum breakers with valves dcigned by another manufacturer, Anderson-Greenwood, in order to provide increased service life m the demanding application they are required to perform in. He original valves were not able to withstand the severe forces generated from their rapid opening in response to SRV closure. Amendment I to this DC was generated to replace 14 additional SRV tailpipe vacuum breakers with Anderson-Greenwood valves and also upgrade the two vacuum breakers that were originally replaced with the same type of valve to make them interchangeable with the other valves and spares. Revision 1 to Amendment I was issued to address the addition of 15 pipe support modifications due to the increased weight of the replacement valves, and to document the acceptability of relocating any of the vacuum breakers along the tailpipes ifinterferences prevented replacing the valves at their current locations and to relocate electrical conduit and instrument lines to avoid interferences. On-the-spot changes were subsequently implemented to relocate 9 of the 16 vacuum breakers. Amendment 2 to this DC authorized an additional pipe support modification to accommodate revised loads.
SAFETY ANALYSIS:
All materials used met or exceeded those quality standards originally established for the affected systems. The Anderson-Greenwood valves were evaluated to meet or exceed all design requirements for the SRV actuation transient. The accident mitigating functions and margins of any components, equipment, structures, or systems remain unaffected. Piping support changes were implemented to ensure that the piping remains within design limits. Valve relocations were supported by calculations to ensure that with the vacuum breaker relocations, the system will continue to function within the existing design basis. Conduit and instrument line relocations have been designed to ensure that the function of the associated components is not affected. This activity does not create the possibility of a new and unanalyzed failure mode. All margins of safety as defined in the Technical Specifications remain unaffected.
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i DC 95-045 TITLE:
Gear Box Upgrade on REC-712MV and REC-713MV DESCRIPTION: This Design Change upgraded the gear boxes on Motor Operated Valves REC-712MV and REC-713MV, which are both 12" butterfly valves. The gear boxes were changed from Limitorque type 110BC to 111 BC. This change was implemented to increase the available torque margin for these valves during performance of their safety related function. This DC also replaced a damaged flex conduit on REC-712MV.
ANALYSIS:
This Design Ucmge did not alter operating conditions or system parameters or introduce any new failure modes. The:e va!ves' accident mitigating functions or ability to perform their design basis safety related functione were ne reduced in any way. 7his DC enhanced the design margin available for these valves. Replacement components were evaluated and meet design requirements. The marg;.1 of safety as defined in the basis for the Technical Specifications was not reduced.
M M P 95-061 TITLE:
West Warel,ouse Trailer Power Addition DESCRIPTION: This Minor Modification provided electrical power for trailers which r.re positioned west of the West Warehouse. While the placement of the trailers is temporary in nature, installing the required power safely requires that the electrical installation be made permanent. This modification utilized some of the spare capacity which exists on the 12.5 KV system, transformed this voltage to the voltage required, and provided twenty disconnect switches for trailer connections.
SAFETY ANALYSIS:
This modification afTected the loading profile of the 12.5 KV system only. The loss of the 12.5 KV system is not an accident evaluated in the USAR; therefore, this modification did not increase the probability or consequences of an accident previously evaluated in the USAR. The load ac.bd will not cause the 12.5 KV system or equipment to exceed design capacity. In addition, a fuse was installed on the primary side of the new transformer to prevent any fault from affecting the remainder of the 12.5 KV system. Therefore, this modification did not increase the probability or consequences of a malfunction of equipment important to safety, nor create the possibility of a different type of accident or malfunction. There is no margin of safety defined in the Technical Specifications regarding the 12.5 KV system.
MMP 05-072 TITLE:
North Yard Trailer Power Additions DESCRIPTION: This Minor Modification provided electrical power for trailers which are located in the northeast corner of the North Yard. While the placement of the trailers is temporary in nature, installing the required power safely required that the electrical installation be made permanent. This modification atilized some of the spare capacity on the 12.5 KV system, transformed this voltage to the voltage i
required, and provided eight disconnect switches for trailer connections.
SAFETY ANALYSIS:
This modification affected the loading profile of the 12.5 KV system only. The loss of the 12.5 KV system is not an accident evaluated in the USAR; therefore, this modification did not increase the probability or consequences of an accident previously evaluated in the USAR. The load added will not cause the 12.5 KV system or equipment to exceed design capacity. In addition, a fuse was installed on the primary side of the new transfonner to prevent any fault from affecting the remainder i
l 21 l
of the 12.5 KV system. Therefore, this modification did not increase the probability or consequences of a malfunction of equipment important to safety, nor create the possibility of a different type of accident or malfunction. There is no margin of safety defined in the Technical Specifications regarding the 12.5 KV system.
M MP 95-073 TITLE:
SW-MOV-MO89A,B Throttling Trim Modification DESCRIPTION: This Minor Modification replaced degraded valve trim with new valve trim on the Residual Heat Removal Service Wt.cr (RilR SW)lleat Exchanger throttle valves (SW-MOV-MO89A,B). The new trim has been designed to reduce erosion on the trim and on the valves.
SAFETY ANALYSIS:
The modification involved near like-for-like replacement of valve internals which are equivalent in form, fit, and function to those originally supplied and are held to the same Code (ANSI B31.1 -
1973Property "ANSI code" (as page type) with input value "ANSI B31.1 -</br></br>1973" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process.) allowables. The modification did not introduce new failure modes, nor did it change the operation of the RilR SW Booster system. All previous accident analyses as documented in the USAR remain bounding, and no unreviewed safety question was created.
MMP 95-089 TITLE:
Turbine Generator Fluid (TGF) System Unloader Valve and Pump Modification DESCRIPTION: 'Ihis Minor Modification removed the unloader valves from the Electro-llydraulic (Ell) System and also made Special Test Procedure (STP)91-066, installation and Evaluation of Constant Pressure Pump Assembly, a permanent modification. The EH System had been experiencing an event similar to a water hammer which was being caused by the unloader valves. The vendor, Westinghouse, arrived at a new design which involved replacing the constant velocity vane pumps and the associated I
unloaders with a constant pressure variable displacement pump. This design was implemented / tested by STP 91-066 and proved satisfactory during a six-month trial period. Therefore, this Minor Modification made the STP a permanent modification, and since the unloader valves no longer serve a useful function to the system, they were removed.
SAFETY ANALYSIS:
This modification was implemented to improve the reliability of the Turbine Generator Fluid System.
By removing the vane pumps and the unloader valves from the system, the possibility of the valve sticking open due to excessive pressure from the old pump was eliminated. A system relief valve remains in place; therefore, this minor modification does not increase the consequences of an accident. Removal of the unloader valves and replacement of the vane pumps did not increase the q
probability of an Eli System failure nor increase the consequences of equipment malfunction. The subject valves and pumps are not defined in the basis for any Technical Specification; therefore, this modification will not reduce the margin of safety in the Technical Specifications.
MMP 95-090 TITLE:
Replacement ofiiPCI Snubber BS-S2A i
DESCRIPTION: This Minor Modification replaced snubber BS-S2A on the High Pressure Coolant Injection (llPCI)
Turbine Exhaust piping line to the torus with a new, larger load capacity snubber. It was implemented to meet dynamic piping loads (i.e., water slug) which could occur as a result of water accumulation in the Turbine Exhaust lines. Engineering analysis indicated that the effect of water accumulation j
in the steam exhnust piping for the HPCI system had the potential to exceed code and operability l
limits of piping a nd related supports due to water slug effects during HPCI startup. As a result of the analysis, the 11PCI system was declared inoperable.
22
SAFETY ANALYSIS:
The installation of the new snubber ensures that the llPCI Turbine Exhaust piping is within operability limits under all loading conditions including a water slug event during IIPCI restart. The original snubber remained installed until the installation of the new snubber was completed to ensure the llPCI Turbine Exhaust piping met seismic Class IS piping loads and thus ensured Primary Containment operability during the installation time period. Replacement of the existing snubber did not create the possibility of an accident not previously evaluated in the USAR, nor created any additional failure modes. This modification, in conjunction with Minor Modifications95-100 and 94-138, allowed the llPCI system to be returned to an operable status.
M MP 95-100 TITLE:
Modification of Support MS-Ill55A DESCRIPTION: This Minor Modification added cover plates on support MS-Ill55A located on the llPCI Turbine Exhaust piping. His modification was implemented to meet dynamic piping load.s which could occur
]
(i.e., water slug) as a result of water accumulation in the Turbine Exhaust lines. Engineering analysis i
indicated that the effect of water accumulation in the steam exhaust piping for the IIPCI system had the potential to exceed code and operability limits of piping and related pipe supports due to water slug effects during IIPCI startup. As a result of this analysis, the HPCI system was declared inoperable.
SAFETY ANALYSIS:
The installation of the cover plates onto the support ensures that the llPCI Turbine Exhaust piping is within operability limits under all loading conditions including a water slug event during 11PCI restart. The modification only increased the capacity of an existing analyzed support. This activity did not increase tN probability or consequences of an accident. This modification, in conjunction with Minor Modi % cations94-138 and 95-090, allowed the llPCI system to be returned to an operable status.
DC 95-101 TITLE:
liPCI Vacuum Breaker Modification DESCRIPTION: This purpose of this DC was to modify the llPCI Turbine Exhaust Vacuum Breaker System to obtain the highest practical perforn ;ance margin, without involving a major reroute of the exhaust line. This was accomplished by enlarjing the vacuum breaker system's line from 2 to 3-inch nominal size and replacing the existing lifhheck valves with swing check valves. The basic routing and configuration of the system remained unchanged. In addition, a weld-o-let on the 11PCI Turbine Exhaust line within the torus was removed and replaced with another fitting using a fully qualified attachment weld.
SAFETY ANALYSIS:
This DC does not decrease the performance or reliability of the liPCI system because the new valves and hardware were selected, sized, located, and integrated into the plant to prevent any interactions during all modes of operation. Reliability, as well as containment integrity, is improved as a result of this modification. Increasing the nominal piping size of the system, in conjunction with installation of the swing check valves, results in a significant increase in net system flow capacity. Potential failure modes were analyzed and it was determined that no new failure modes were created. This activity did not increase the probability or consequences of an accident or malfunction of equipment.
All margins of safety as defined the Technical Specifications remain unaffected.
23
MM P 95-104
)
TITLE:
REC-71IMV & 714MV, REC-1329MV, RCIC-MO27, RIIR-MOl6B, and RilR-MO27B Gear Changes and Spring Pack Replacements DESCRIPTION: The CNS MOV Program requires that the program MOVs be insitu differential pressure (DP) tested j
to verify operability under near design basis conditions if practicable. This modification enhanced the design margin available for the subject valves. The motor pinion and worm shaft gear sets on
)
REC-711MV, REC-714MV, RCIC-MO27 and R11R-MO27B were replaced to provide additional motor gearing capability and thus enhance margin. The spring packs were changed on REC-
)
1329MV, RliR-MOl6B, REC-711MV, and REC-714MV to better meet the capability requirements of these MOVs and to optimize proper setup with the new configuration. The benefits of performing this modification were eliminating the performance of DP tests and enhancing available margin for these MOVs.
SAFETY
- ANALYSIS:
The new motor operator configurations did not increase the probability or consequences of an accident or malfunction of equipment. This modification enhanced the design margin available for l
these valves and did not change the function of any equipment. Reliability of these MOVs has not been degraded in anyway. Replacement components have been evaluated for the design basis l
functions of these valves and meet the design requirements. These valves' accident mitigating functions or ability to perfonn their design basis safety related functions were not reduced.
MMP 95-108 TITLE:
Engineering Building Construction Power DESCRIPTION: This Minor Modification provided temporary electrical power for use during construction of the Engineering Building located to the north of the CNS Learning Center. While this construction power is temporary, safe and reliable installation requires that the pole mounted, integrally fused transformer be made permanent. This Minor Modification utilized some of the spare capacity on the 12.5 KV system to provide the appropriate power required.
SAFETY ANALYSIS:
This modification alTected the loading profile of the 12.5 KV system only. The loss of the 12.5 KV system is not an accident evaluated in the USAR; therefore, this modification did not increase the probability or consequences of an accident previously evaluated in the USAR. The load added will not cause the 12.5 KV system or equipment to exceed design capacity. In addition, a fuse was installed on the primary side of the new transformer to prevent any fault from affecting the remainder of the 12.5 KV system. Therefore, this modification did not increase the probability or consequences of a malfunction of equipment important to safety, nor create the possibility of a different type of accident or malfunction.
MMP 95-113 TITLE:
Diesel Oil Storage Tank Manhole Covers Modification DESCRIPTION: The purpose of this Minor Modification was to document the as-built configuration of the Diesel Generator oil storage tank exterior manhole covers in order to address tornado generated missile protection criteria required for the vital equipment located in the tank. These covers had been previously modified under a Minor Design Change and Maintenance Work Request, both of which failed to properly address the licensing basis for tornado generated missiles. This MMP serves as the design medium for documenting the Safety Evaluation performed to support the resolution of the wind generated missile concern and to serve as the design basis document for the current configuration.
24
SAFETY EVALUATION: The existing manway configuration has been qualified to provide protection for the vital equipment located within the storage tanks in the event of a tornado. A calculation was performed which analyzed the configuration of the manways and the characteristics of tornado generated missiles and concluded the equipment was protected even if the exterior manhole covers were removed. Leaving the existing manways in their existing analyzed condition does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety. No new features or equipment are added by this MMP which would require a new accident analysis. The existing manway configuration does not reduce the margin of safety as defined in the basis of any Technical Specification.
MMP 95-115 TITLE:
East Warehouse Air Conditioning Unit Replacement DESCRIPTION: His Minor Modification replaced the 12-ton heat pump roof unit in the East Warehouse. The unit was in need of repair and it was determined to be more cost efficient to replace the unit than repair it. The heat pump was resized to a 15-ton unit and fuses were resized from 125 amps to 110 amps.
SAFETY ANALYSIS:
The 125A fuses were replaced by new 110A fuses per vendor recommendation. This action added additional safety margin. He heat pump does not interface with or affect any equipment important to safety and is not considered in any accident scenario or mitigation activity. This modification did not affect the margin of safety as defined in the basis for any Technical Specification.
MMP 95-11 R.1 TITLE:
Addition of Load for the New Engineering Building to the 12.5 KV System DESCRIPTIO N: This Minor Modification documented the acceptability of the addition and connection of the electrical loads for the new Engineering Building to the existing 12.5 KV ring bus. This installation increased the loading of the system by approximately 500KVA which is within the existing capacity of the system.
SAFETY ANALYSIS:
The 12.5 KV ring bus provides power to non-safety related miscellaneous plant services; therefore, any change to the system does not increase the probability of an accident previously evaluated in the USAR. The electrical equipment added is properly sized and protected so that any localized fault will not adversely afTect the 12.5 KV system or any other safety related system. This modification did not i
involve any safety -related equipment; therefore, it did not create the possibility of a different type of accident or malfunction of equipment.
MMP 95-124 1
TITLE:
Diesel Generator Bottom Skirt Oil Ring and Piston Pin End Cap Removal DESCRIPTION: This modification implemented the recommendations of Cooper Bessemer Bulletin 752 which provided infonnation that it was permissible to remove the piston end caps and bottom skirt rings on any KSV engine still so equipped. His change was implemented to provide better oil flow to aid in flushing wear particles from the skirt so they do not become embedded in the liner pores or skirt surface, nis should assist in maintaining a lubricating film on the skirt and thus reduce the possibility of scuffing and overheating.
25
n
)
SAFETY ANALYSIS:
This change did not have any effect on any accident or accident precursor previously evaluated in the USAR and did not affect the ability of the diesel generator to fulfill its design and safety-related functions. It did not increase the probability or consequences of a malfunction of equipment I
important to safety previously evaluated in the USAR, nor create the possibility of a different type of accident or malfunction of equipment. The margin of safety as defined in the basis for any Technical 3pecification was not reduced since this modification did not affect the ability of the diesel generators to perform their design function.
MMP 95-126 TITLE:
Removal of Sample Points 55/56 Sample Valves i
DESCRIPTION: This Minor Modification authorized the removal of the isolation and sample flow control valves for Sample Points 55 and 56 in the steam tunnel. These valves had been abandoned in place by i
MDC 77-29 in 1977. They were marginally supported and were removed as a housekeeping concem.
SAFETY ANALYSIS:
A 10CFR50.59 review could not be located for th: original DC; therefore, the 10CFR50.59 reportability analysis for this Minor Modification accounted for disconnection of the valves from the system as well as removal from the steam tunnel. Removal and capping of the sample points, tubing, and air line removed a primary containment boundary valve. This reduced the likelihood of a feed line rupture inside the containment boundary and eliminated one potential leakage path; therefore, 1
the probability of occurrence or consequences of an accident previously evaluated in the USAR were not increased. 'this modification removed the potential for sample probe failure which has impacted j
the function of other valves in the system. These sample points were located in the Reactor Building and, therefore, could not be used during an accident scenario. The sample points were used fo-General Electric fuel warranty monitoring only. Sample Point 40 has proved an acceptable subst>te.
Removal of the sample points did not create the possibility of an accident or malfunction of equipment of a different type than previously evaluated in the USAR. The sample points were not used for monitoring any Technical Specification parameters. Therefore, their removal cannot reduce the margin of safety as defined in the basis for any Technical Specifications.
MMP 95-127 TITLE:
1995 Main Turbine Modifications DESCRIPTION: This Minor Modification documented three changes to the design of the main turbine, as follows: 1) the last three rows of the stationary blade in each of the low pressure turbines were replaced with a better design; 2) the horizontaljoint on the high pressure turbine was modified in accordance with Westinghouse recommendations; and 3) the design of the cylinder heating supply line connection to the high pressure turbine was modified as recommended by Westinghouse. All three modifications were implemented to improve the performance and reliability of the main turbine.
SAFETY ANALYSIS:
The turbine provides no safety function and does not affect any accident as described in the USAR.
The USAR describes the potential for turbine overspeed in excess of the design overspeed that may l
result in the fracture ar d release of discs from the turbine casing which can act as missiles. These l
modifications had no effect on overspeed control or rotor disc design and do not affect the probability 26
of a turbine missile incident. The modifications are simple and deviate very little from the original design. There is no reasonable possibility of an accident or malfunction of equipment of a different type than previously evaluated. The USAR analysis is bounding and represents the worst possible situation. These modifications do not affect any margin of safety as defined in the basis for any Technical Specification.
I i
MMP 95-131 1
1 TITLE:
Installation of Turbine Building Jib Crane and New Rigging Points in Reactor Building DESCRIPTION: This Minor Modification installed a new jib crane at elevation 903'6" in the Turbine Buading to provide a permanent means of lifting equipment from the 882'6" level. It also installed a new baseplate above the angle valve room of the Reactor Building to lift RHR-MO-MO25B for maintenance and a new anchor point in the concrete beam above HPCI-MO-MO25 to provide an approved rigging point for maintenance.
SAFETY ANALYSIS:
These activities involved the addition of equipment for maintenance activities. This equipment does not adversely afTect any structures or components and does not increase the probability or consequences of an accident evaluated in the USAR. This equipment does not interact with any equipment important to safety. The equipment added was designed with large safety factors on their intended use; therefore, it did not create the possibility of an accident or malfunction of equipment not previously evaluated in the USAR. The use of approved rigging points for maintenance increases j
the safety of such work. The margin of safety as defined in the Technical Specifications was not i
affected by these activities.
MMP 95-132 TITLE:
Replacement of DGSA-RV-20RV and 21RV DESCRIPTION: This Minor Modification replaced DGSA-RV-20RV and DGSA-RV-21RV with higher capacity relief valves. A setpoint change had been initiated to change the :Stpoint of the subject valves from 150 psi to 125 psi. The new setpoint would have reduced the capacity of the valves to 180 scfm, not 210 scfm as required by the applicable Design Calculation. Therefore, this modification replaced the existing J
valves with new model valves with a capacity of 296 scfm at 125 psi. The new valves were tested and set to the desired setpoint by a vendor prior to installation.
SAFETY ANALYSIS:
The Diesel Generators and associated components are not accident precursors; therefore, the probability of occurrence of a previously evaluated accident or malfunction of equipment is not increased. The creation of a different type of accident or malfunction is also not a concern. The basic design of the replacement valves is comparable to the prior design, the replacement valves utilize similar materials, and the modification does not alter the function of the valves. Therefore, the consequences of an accident or malfunction are not changed by this modification. The purpose of the valves is to protect downstream piping and associated components from damage from overpressure.. Reducing the setpoint is conservative, and the new valves provide the capacity required at the lower set pressure. The margin of safety is not affected.
27
l 1
MMP 95-163 TITLE:
Replacement of MS-A086C Packing with ARGO Packing / Optional Stellite Seating Material DESCRIPTION: This Minor Modification replaced the existing packing in MS-A086C with ARGO type packing and installed a new stem fabricated from a different alloy. This MMP authorized the use of stellite 6 or 21 for repairs to the seating surfaces of Main Steam Isolation Valves (MSIVs). ARGO packing is expected to provide longer packing life, less leakage, and lower stem friction. The new stem material
~
is expected to provide increased wear durability and resistance to galling.
SAFETY ANALYSIS:
his MMP did not alter the functionality or operating parameters of any systems or components, nor affect the integrity, reliability, or qualification of the MSIVs. Therefore, it did not increase the probability of an accident or affect the margin of safety as defined in the Technical Specifications.
i It did not alter the mitigating capabilities or margins of any systems or components required for accident mitigation, nor increase the severity of events related to a postulated accident scenario. The new materials meet or exceed design requirements and the new packing / stem alloy combination is expected to provide greater reliability by minimizing stem friction and galling.
MMP 95-167 TITLE:
Modification of the Steam Tunnel's Fiberglass Sealing Cover i
j DESCRIPTION: A fiberglass barrier was installed at the east end of the steam tunnel in 1985 to reduce secondary containment inleakage through the wall. This barrier was also inappropriately applied over portions of the wall designed to blow out in case of a steam line break within the tunnel. This previous installation constituted a station modification and introduced an unreviewed safety question; however, no design document was prepared to support the installation. This Minor Modification provided formal documentation of this installation and also provided the necessary instructions to remove those portions of the barrier impeding the design basis operation of the blowout panels. The panels are needed to prevent structural damage in case of a steam line rupture.
SAFETY ANALYSIS:
This activity did not adversely affect the integrity, reliability, or qualification of the secondary containment boundary or the blowout panels; therefore, it did not increase the probability of an accident. It is intended to ensure that the steam tunnel's blowout panels perform their design function. The consequences of a blowout panel failure remain unchanged, whether or not the malfunction was caused by the fiberglass barrier. This modification did not create the possibility of an accident of a difTerent type than previously evaluated in the USAR and did not create any new and unanalyzed failure modes. Existing margins of safety were not affected by this modification.
MMP 95-179 TITLE:
CRD-SOll8 Solenoid Valve Upgrade DESCRIPTION: This Minor Modification replaced the exhaust diaphragm end caps on the flydraulic Control Unit CRD-sol 18 series scram pilot valves with the latest design from the valve's manufacturer. New exhaust diaphragms and small body passage gaskets were also installed on these valves. This installation was part of an interim fix recommended by General Electric to resolve an adverse trend in scram times noted in other facilities.
l 28
SAFETY ANALYSIS:
The modification did not adversely affect the integrity, reliability, or qualification of the Control Rod Drive (CRD) system or its liydraulic Control Units, nor alter the mitigating capabilities or design margins of the CRD system. Therefore, this activity did not increase the probability or consequences of an accident. The design of the new end cap has been fully qualified by the va!ye's manufacturer through rigorous testing. This testing ensures that the new arrangement does not iwease the probability of a malfunction. The consequences of a scram pilot valve failure remain unchanged; therefore, this activity did not create new and unanalyzed failure modes. No possibility of an accident or malfunction of a different type than previously evaluated in the USAR was created and the margin of safety was not reduced.
MMP 95-184 TITLE:
Position Modulator Removal from SW-MOV-89A,B Control Logic DESCRIPTION: nis Minor Modification removed the ability of the position modulator to control valve position and installed a new control switch for manually throttling SW-MO-89A,B valves. This modification was implemented as the result of a failure of a potentiometer which provides valve position signal to the position modulator. Due to difficulty in reclassifying the potentiometers as Environmentally Qualified, this modification removed the potentiometers from the control circuit, thus deleting their EQ/ essential function.
SAFETY ANALYSIS:
This modification affected the control of SW-MO-89A,B only. Valve operation continues to be a remote manual function from the Control Room per approved procedure. The new control circuit components were installed as essential equipment which meet all applicable quality and design standards for the respective valves. The new essential switches are more reliable than the previous configuration with nonessential components. The failure mode and/or failure position of SW MO-89A,B was not changed as a result of this modification. The modification provides additional assurance that the Service Water system will be available as assumed by the USAR. This modification did not increase the probability of occurrence or consequences of an accident, create the possibility of a new accident, or decrease the margin of safety as defined in the basis for any Technical Specification.
MM P 96-003 TITLE:
Turbine Generator (TG) Auto Stop Oil Relief Valve Modification DESCRIPTION: This Minor Modification replaced existing failed Teledyne Republic TG auto stop oil system relief valves with Fisher relief valves. Existing valves were not maintaining header pressure below the design pressure of the auto stop oil to Eli interface valve diaphragm.
SAFETY ANALYSIS:
This Minor Modification did not modify components which have an accident mitigation function, turbine generator trip function, or imponant to safety function. The TG was off-line during implementation of this Minor Modification. The post-modification configuration is functionally equivalent and conforms to the same design, material, and construction standards. No unreviewed safety question was created as a result of this modification.
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PLANT TEMPORARY MODIFICATIONS (PTMs)
PTM 95-01 TITLE:
Residual lleat Removal (RilR) Pump A Insulation Removal DESCRIPTION: This Plant Temporary Modification documented the removal ofinsulation from RilR Pump A for casing water leak deterraination. The insulation was reinstalled after pump / motor maintenance during the fall 1995 outage.
ANALYSIS:
The removal ofinsulation from RilR Pump A had a negligible effect on RiiR process temperature and the Northwest Quad temperature remained below the design temperature for all normal rand post-accident conditions. The functional capability of the pump was not affected. Therefore, this modification did not increase the probability or consequences of an accident or malfunction of equipment important to safety, nor introduce any new failure modes.
PTM 95-03 TITLE:
Leak Repair of RF-AOV-DRV98 DESCRIPTION: This Plant Temporary Modification was issued to repair a steam leak from the flange / gasket area of a Reactor Feed Pump stop valve drain. A flange clamp was installed and filled with sealant to stop the leak. This PTM was not totally successful in stopping the leak. The valve was subsequently replaced during the fall 1995 outage.
SAFETY ANALYSIS:
The probability of an accident was not increased due to the flange clamp and leak repair capnuts meeting or exceeding the original piping material and strength requirements. The leak sealant was compatible with Main Steam piping material so the probability of a steam line leak was not affected.
The bounding accident remained the steam line break. The existing margin of safety was not changed.
PTM 95-04 TITLE:
Leak Repair of Moisture Separator Calorimeter Flange Steam Leak DESCRIPTION: This Plant Temporary Modification was issued to repair a steam leak from the flange / gasket area of a Moisture Separator calorimeter flange. A flange clamp was installed and filled with sealant to stop the leak. This PTM remained in place until the flange was repaired during the 1995 fall outage.
SAFETY ANALYSIS:
The probability of an accident was not increased due to the flange clamp and leak repair capnuts meeting or exceeding the original piping material and strength requirements. The leak sealant was compatible with Main Steam Line piping rncrir.!i.o the probability of a steam line leak was not affected. The repair was bounded by a Twain Steam Line break downstream of the outboard MSIVs.
The existing margin of safety remained unchanged.
PTM 95-05 l
TITLE:
Installation of Monitoring Equipment on Digital Electro-llydraulic (Dell) Computer DESCRIPTION: This Plant Temporary Modification authorized the installation of temporary monitoring equipment on the digital side of the Dell computer to troubleshoot the cause of a loss of the digital control loop by monitoring for voltage anomalies.
1 30
SAFETY ANALYSIS:
During this activity the analog portion of the computer functioned to maintain pressure control and turbine operation while the digital portion was monitored. It did not increase the probability of a turbine trip without bypass or generator load rejection without bypass. Controls were in place to ensure turbine overspeed above 108% would not occur. Dell does not impact the consequences of an accident or affect the operation of safety equipment. The margin of safety as defined in the Technical Specifications was not impacted.
PTM 95-06 TITLE:
Erection of ScalTolding in if PCI Room DESCRIPTION: This Plant Temporary Modification authorized the erection of temporary scaffolding in the llPCI Room to support inspection of smoke detectors located on the ceiling.
SAFETY ANALYSIS:
All applicable code criteria remained satisfied for plant components which provided vertical and lateral support to the scaffolding. All tie-off points were evaluated to ensure that no seismic class 11 over I concerns were created. Erection of the scaffolding did not interfere with personnel access to equipment nor with actions taken to mitigate the consequences of equipment malfunction. An Engineering Judgement was performed which ensures the scaffolding will not allow the bounds of a USAR analyzed accident to be exceeded and ensures that applicable code criteria will continue to be met. The margin of safety of affected equipment as defined in the Technical Specifications was not reduced.
PTM 95-09 TITLE:
Removal of Filter Element from OG-F-B DESCRIPTION: This PTM authorized the removal of a filter element from Off Gas Filter B until a positive source of water entrainment into the filter unit could be identified and corrected. He filter in the A train filter unit remained installed. This PTM was released aller a new filter was installed.
SAFETY ANALYSIS:
The off gas filters only provide mitigation of radioactivity release and do not affect the probability of occurrence of an accident or malfunction of equipment important to safety. The off gas system is monitored and control!ed to ensure release limits are not exceeded. This monitoring and control was not afTected by this PTM. The filter is isolated in an accident and thus cannot create the possibility of a different type of accident. His PTM did not affect any Technical Specification related equipment and did not affect any margin of safety as defined in the basis for any Technical Specification.
PTM 95-10 TITLE:
Bypass of Average Power Range Monitor (APRM) Upscale and Rod Block Alarms DESCRIPTION: his Plant Temporary Modification authorized bypassing of the APRM upscale and rod block alarms in the Control Room. This PTM was initiated because operating rear the rod block causes the inherent noise of the APRM system to bring in numerous spurious rod block nuisance alarms.
SAFETY ANALYSIS:
Bypassing the APRM upscale and'or rod block alarm (s) does not affect the function of the APRM, rod block, or any other equipment. Bypassing these alarms cannot initiate an accident. The alam1(s) only function to draw the operator's attention to the event. All other trip / protection functions of the APRM and rod block systems remain fully functional with the alarm (s) bypassed. No credit is taken for the alarm (s) in mitigating the consequences of an accident previously evaluated in the USAR. He alarms are not a part of the basis of any Technical Specification.
31
i i
i PTM 95-12 TITLE:
Disabling of Downscale Annunciator for RMA-RA-3 DESCRIPTION: This Plant Temporary Modification disabled the downscale annunciator for the New Fuel Storage Vault (NFSV) area radiation monitor. The NFSV is not used. The detector is located inside of the vault which exhibits a background less than the minimum possible setpoint ofits radiation monitor, thus a downscale alann is produced. Since the NFSV is not used, the alarm provided a nuisance to the Control Room. Administrative entrols are in place to verify the operability of RMA-RA-3 prior to putting fuel into the NFSV. An Operations Instruction was subsequently issued which provided controls for disabling of annunciators and allowed this PTM to be restored.
i SAFETY ANALYSIS:
RMA-RA-3 is a nonessential monitor with alarm only functions. It only provides an alarm for the measurement of activity within the NFSV, Area radiation monitors do not impact safety related equipment and are not covered in the fechnical Specifications. His PTM did not increase the probability or consequences of an accident or malfunction of equipment important to safety.
)
l'lht9h.ll TITLE:
Repair of Leak on Moisture Separator "A" Piping Manway DESCRIPTION: This Plant Temporary Modification was issued as a temporary repair of a steam leak on the Moisture i
Separator "A" piping manway. Special leak repair nuts were installed and injected with sealant to stop the leak. This PTM remained in place until the fall 1995 outage when the leak was permanently repaired.
SAFETY ANALYSIS:
The probability of an accident was not increased due to the leak repair nut and sealant meeting or exceeding the original piping material and strength requirements. The leak sealant was compatible i
with the Moisture Separator piping material, so the probability of a steam line break was not affected.
The bounding accident remained the steam 15e leak. The Moisture Separator and associated piping have no safety design basis. The existing margin of safety as defined in the Technical Specifications remained unchanged.
PTM 95-14 TITLE:
Relocatioc of Oflice on the Turbire Deck DESCRIPTION: This Plant Temporary Modification relocated the Westinghouse office on the turbine deck closer to the area of work in order to maximize turbine outage ef5ciency, it also installed gaitronics in the relocated office. An Engineering Project Request has been initiated to make this modification permanent.
SAFETY ANALYSIS:
The relocation of this office does not afTect any equipment required for the safe shutdown of the plant.
The change does not alter the design floor loadir.g of this area and does not alter the seismic class of i
the building. The office is not located near any equipment important to safety. The installation of the gaitronics handset does not create the possibility of a communication system failure, so current communications analyses are bounding. All margins of safety remain unchanged.
l 32
l l
PTM 95-15(1)
TITLE:
Installation of Temporary Pipe Support for Control Rod Drive (CRD) Equalizer Line DESCRIPTION: His Plant Temporary Modification installed a temporary pipe support for the equalizing line between the A and B CRD Pumps. He line was uncoupled from the A CRD pump, which was replaced. ne remaining piping, connected to the B CRD pump, required temporary support.
SAFETY ANALYSIS:
This modification was required to support the piping connected to the B CRD pump to prevent the pipe from failing. It could only affect the B CRD pump. No failure of this pump could increase the probability or consequences of an accident previously evaluated in the USAR. The design of the CRD system anticipates the loss of CRD pump pressure / flow by providing for accumulators or reactor pressure for the scram function. The support provided at least equivalent support to the support removed for maintenance. The CRD system is not relied upon to mitigate system leakage.
PTM 95-15(M TITLE:
Closure of TGF-V-S I, Governor Valve (GV) No. 2 liigh Pressure Fluid Shutoff DESCRIPTION: This Plant Temporary Modification authorized closing the GV 2 liigh Pressure Fluid shutoff to isolate the hydraulic actuator on GV-2 from Turbine Generator Fluid system pressure and reduce oil leakage from the actuator. Due to end of cycle coastdown and sequential valve positioning, GV-2 was in the closed position and not required for reactor pressure control.
SAFETY ANALYSIS:
The probability of an accident was not increased due to GV 2 being closed. GV failure is bounded by the main steam line break accident and load reject without bypass transient. Isolation of GV-2 in the closed (fail safe) position does not increase the probabihty of a turbine overspeed or missile generation. Main turbine govemor valves are not used to mitigate the consequences of any previously analyzed design basis accidents. ne safety lin>it as described in Technical Specifications for vessel over prestarization was not altered by isolating GV-2 in the closed position. Existing pressure margins fo. the high pressure scrun, relief and safety valve setpoints remained the same.
PTM 05-16 TITLE:
Installnion of Temporary 24V Battery Charger DESCRIPTION: This Plant Temporary Modification authorized installation of a temporary 24V battery charger in Reactor Protection System Room B in place of the B train 24V battery charger, EE-CrfG-24(IBI),
'vhile EE-CIIG-24(IBI) was being repaired. Even though the temporary charger was rated less than the pennanent charger (20 amps at 24 volts vs. 25 amps at 24 volts), the load measured off of battery iB1 was 3.5 amps which was well within the 20 amp rating of the temporary battery charger.
SAFETY ANALYSIS:
The temporary charger satisfied the loads typically found on 24V IBl bus. The loads on the 1B124V bus are small; therefore, the reduction in battery charger capacity from 25 amps to 20 amps was inconsequential. The 24V battery system is not an accident mitigation system. According to the USAR, the loss of one 24V bus will not impact plant safety. The installation of tnis PTM ensures that none of the 24V busses will be lost. The 24V busses me not included in any Technical Specifications; equipment powered off of the 24V batteries is meationed, but this equipment is fail safe.
33
PTM 95-18 and PTM 95-20 TITLE:
Connection of Data Accorder to Reactor Recirculation Motor Generator (RRMG) Flow Control Circuitry DESCRIPTION: This Plant Temporary ModificJion authorized connection of a data recorder to the RRMG flow control circuitry to gather data to dearmine the reason for unexpected changes in recirculation pump i
speed. RRMG-B was previously lockehut because of the transients experienced. He scoop tubes j
of both RRMG sets were also locked out hmporarily while the data recorder was connected and disconnected. During the data-taking process,the scoop tubes were unlocked. 'Ihe Safety Evaluation addressed the safety significance of operating w& the scoop tubes locked out, and operating RRMG-B with the scoop tube unlocked with the knowleQe that a speed transient could occur. This modification was performed two different times approximately ten days apart. The problem was found to be a connector which was subsequently replaced during the fall 1995 outage.
SAFETY ANALYSIS:
His activity was initiated to gather operating data only. The data recorder has a high impedance so connecting it to the electrical circuit had no effect on the circuit and could not cause the control system to malfunction. Connecting the recorder to components of the Recirculation Flow Control System did not change the function or method of performing the function of the Recirculation Flow Control System. Locking out the scoop tubes affects only the runback function of the Recirculation Flow Control System. Scoop tube lockout does not affect any recirculation pump trip function.
Recirculation pump runback is not required to mitigate any accident analyzed in the USAR nor can the failure to run back cause an accident. Therefore, there were no safety concerns with operating with either or both scoop tubes locked out. RRMG-B scoop tube was unlocked during the data gathering period to monitor any transient. Operation with this scoop tube unlocked, even with the known problems with the flow control system, did not change the probability of an accident previously evaluated in the USAR. To provide the operator with early warning of a recirculation speed transient, a temporary computer point alarm was installed during the data-taking process to monitor RRMG speed. This activity did not affect the basis of any Technical Specification or any other equipment relied upon to provide a margin of safety in the basis of any Technical Specification.
PTM 95-10 TITLE:
Installation of Temporary Pipe Patch in Water Treatment System DESCRIPTION: This Plant Temporary Modification installed a temporary pipe patch on Water Treatment System dilute acid piping which developed a through-wall pinhole leak. A temporary patch is required to allow ion exchanger regeneration until a replacement spoolpiece can be obtained.
SAFETY ANALYSIS:
The plant makeup water system is not a contributor, either in initiation or mitigation, to any accident or transient considered in the USAR. The plant makeup water system is entirely contained within the Water Treatment area of the Turbine Building and is not in close proximity to any equipment imponant to safety. This modification does not increase the probability or consequences of an accident or malfunction of equipment to safety, nor create the possibility of a different type of 4
accident.
PTM 95-21 TITLE:
Alternative Power to HV-MOT-(BF-C-1B)
DESCRIPTION: This Plant Temporary Modification documented the installation of temporary power for ilV-MOT-(BF-C-1B), which is normally powered by Division I, so that it can be operated by Dis ision 11 power during the Division i power outage during the 1995 refueling outage. A Condition Report documented a condition in which the Control Room envelope could be compromised upon 34
l 1
i a loss of Division I power and subsequent inoperability of 11V-FAN-(BF-C-1B). This PTM l
established alternate power for the fan motor through the transfer switch that powers BF-C-1 A fan.
l.
This PTM resolved the concern ofloss of operability due to Division I power being potentially l
unavailable during a DGl outage and permitted the movement of fuel during the DG l power outage.
I nis PTM was removed prior to plant startup. An Engineering Work Request has been initiated to make this PTM configuration a permanent modification.
SAFETY ANALYSIS:
The afTected equipment safety function ensures Control Room pressurizatica to ensure post-accident habitability to allow operator post-accident response. This PTM provided alternative power feed to safety-related equipment in the event of a divisional power failure, thus adding reliability to safety-related equipment to ensure that the margin of safety was not compromised. The actual performance levels of the safety related equipment were not affected, but the equipment merely powered by an alternate power source. This PTM did not increase the probability or consequences of an accident or malfunction of equipment important to safety and did not reduce the Technical Specification margin of safety.
PTM 95-26 TITLE:
Replacement of Door Locksets on 11 eater Bay Area Doors DESCRIPTION: This Plant Temporary Modification changed the locksets on IIcater Bay Area doors T114 and Tl17 to support RE16 work. These doors are normally locked to prevent personnel access to high radiation areas. Ilowever, during periods when the plant is shutdown, the lleater Bay for which these doors provide access is not a high radiation area; therefore, the doors are not required to be locked. The existing locksets are designed such that a key is always required on the outside knob to open the door.
Since significant work was being performed in the 11 eater Bay Area, the existing locksets were temporarily replaced with locksets which unlock in order to support this work.
SAFETY ANALYSIS:
The door locks serve a personnel radiation control function to prevent personnel from entering a high radiation area during plant operation. The door locks are not a precursor to any accident or event evaluated in the USAR, therefore cannot affect the probability of an accident or malfunction of equipment important to safety previously evaluated in the USAR. The door locks do not provide any function in the response to or mitigation of any accident or equipment malfunction. When the plant is shutdown, the high radiation areas in the Heater Bay are not present and the locks are not required by Technical Specifications.
PTM 95-28 TITLE:
Temporary Power for Outage Loads in Reactor Building and General Electric Trailers DESCRIPTION: His Plant Temporary Modification installed temporary electrical power for outage loads inside the Reactor Building and inside of General Electric trailers to accommodate the Inservice inspection program inspections. Power was supplied by spare circuits on the 12.5 KV electrical distribution 1
system. This PTM also temporarily changed the configuration of fire seals being altered from their spare configuration to a configuration containing cabling. The secondary containment sealing method was also temporarily changed from a steel pipe cap to an essential RTV silicone caulk. These penetrations were restored to their original configuration following cable removal with the necessary fire seals and essential seais.
SAFETY ANALYSIS:
The probability of an accident or fire was not increased by these changes. The electrical loads were supplied from nonessential 12.5KV power which has no impact on accident analysis, and the fire seal j
and secondary containment seal are for accident mitigation and do not create the possibility of an l
1 35 l
1 1
A accident. He fire barrier and secondary containment seals were demonstrated to provide an equivalent level of protection to the openings. Applicable Technical Specification Limiting Conditions for Operation were observed during the installation of the PTM and operability was verified following installation.
PTM 95-32 TITLE:
Demineralized Water Supply to Condensate Pressure Maintenance DESCRIPTION: This Plant Temporary Modification installed a temporary cross-connect between the Demineralized l
l Water system and the Condensate Makeup system to maintain a pressure source to ECCS system pressure maintenance piping. The normal sources of water for pressure maintenance were out of i
service due to Main Condensate system shutdown and a planned 480V Bus IF outage.
i SAFETY ANALYSIS:
The ECCS systems will still perform their design basis functions without pressure maintenance.
i Utilizing a different pressure source for the system will not affect alarm or operational functions described in the USAR. The pressure source being changed from the nonessential Condensate Makeup system to the nonessential Demineralized Water system did not increase the probability of occurrence of a malfunction of equipment important to safety. The existing isolation check valves remained in the flow path, maintaining the safety related pressure boundary intact. The use of a hose as a cross-connect does not create the possibility of an accident or malfunction of a different type than any previously evaluated in the USAR. Since the PTM was intended to maintain pressure maintenance in operation, the margin of safety as defined in the basis for the applicable Technical Specifications was maintained.
j PTM 95-33 TITLE:
Temporary Power for Turbine Repairs DESCRIPTION: This Plant Temporary Modification provided temporary power for contractor equipment used to repair low pressure turbine horizontaljoint erosion. Equipment used to perform the repair was connected to spare breakers on LRP-PNL-TDT 2.
SAFETY ANALYSIS:
This PTM did not adversely impact cable load carrying capacity and provided adequate electrical load protection for the equipment that was connected. It did not alter the arrangement such that a single failure would prevent or impair the operation ofessential station safety functions. The equipment was located in a seismic class 11 area and was not located near any safety related equipment. Installation of this equipment could not cause a failure of equipment important to safety nor cause a failure of other equipment powered from EE-PNL-TDT 2. The logic and protective features of the electrical system remain unchanged by this PTM.
PTM 95-34 TITLE:
Temporary Power to Instrument Rack 25-2 DESCRIPTION: This Plant Temporary Modification provided temporary power to instrument rack 25-2 to replace the power normally fed from CPP-2, which was deenergized when the 4160 VAC Bus 1G was deenergized for maintenance. This supplied power to RWCU-TIS-99, RWCU-SOV-SPV1247, REC-TIS-89A, and REC-TIS-898. 'ihis modification was necessary to keep Reactor Water Cleanup l
(RWCU)in service during the 4160 VAC Bus 1G outage.
j SAFETY ANALYSIS:
A loss of this temporary power would cause RWCU-MOV-MOIS and RWCU-MOV-MOIS to isolate, the RWCU pump to trip, and the loss of RWCU blowdown to Radwaste or the Hotwell.
j RWCU is required to maintain a containment integrity pressure boundary and to isolate on a primary 36
containment group 3 isolation signal. Neither of these functions were adversely affected by the addition of the temporary power source; therefore, there was no reduction in the margin of safety.
The isolation logic for RWCU-MOV-MOIS and RWCU-MOV-MOI 8 is a fail safe logic. An isolation was required during the installation of the temporary power, but was a planned evolution and had only a very short term, minimal impact on reactor coolant water chemistry. No safety related equipment was impacted by the addition of this temporary power. The temporary power was added to a non-safety related ponion of the RWCU and Reactor Equipment Cooling (REC) systems and did not introduce the potential for any new accident scenarios or potential safety related malfunctions.
PTM 95-37 TlTLE:
Removal of Control Rod Drive (CRD) Pump ilatch Cover DESCRIPTION: his Plant Temporary Modification authorizcd the removal of the hatch cover on the 903' level of the Reactor Building over the CRD pump in order to facilitate removal of the CRD pump for repair.
SAFETY ANALYSIS:
%e only essential equipment which the plug removal could have affected is Core Spray Pump B in the Southeast Quad. The effect on ECCS equipment functions as a result of the removal of a plug on the 903' level was previously analyzed when DC 93-062 was implemented to permanently remove the plugs above the Northwest and Southwest Quads which house the RHR pumps. The one exception with this PTM is that a dam was not crected to prevent flooding. A dam was not required for the ECCS system to maintain the plant in a safe shutdown condition following a break in the feedwater line in the steam tunnel. Since Core Spray Pump IB was available to fulfill its safety function for those events for which it is required, it was considered operable during the time period of this PTM. This PTM did not increase the probability of an accident since the work did not increase the probability of a LOOP, a loss of SDC, nor an inadvertent ECCS initiation because there is no interaction with any of these systems. %is activity did not violate any Technical Specification since j
all safety equipment would be available when required by Technical Specifications.
PTM 95-38 TITLE:
Leak Repair of MS-V-186 DESCRIPTION: This Plant Temporary Modification authorized injection of leak repair compound as a temporary repair to eliminate packing leakage from MS-V-186, the Main Steam Line C Vent. This PTM installed a special leak repair injection fixture in the outlet plug of the valve which was then injected with leak repair compound.
SAFETY ANALYSIS:
The probability of an accident was not increased due to the leak repair valve and sealant material meeting or exceeding the original piping material and strength requirements. The leak scalant is compatible with the Main Steam piping material so the probability of a steam line break was not affected. he valve and associated piping are outside primary containment and do not affect operation of the Main Steam isolation Valves or steam line flow restrictors. De bounding accident remained the steam line break. The existing margin of safety was not changed.
PTM 96-01 TITLE:
Leak Repair of MS-AOV-DRV4 DESCRIPTION: This Plant Temporary Modification authorized injection ofleak repair compound as a temporary repair of MS-AOV-DRV4 inlet flange. This PTM installed a special leak repair flange which was injected with sealing compound.
37
I 1
SAFETY i
ANALYSIS:
ne probability of an accident was not increased due to the leak repair fiange and sealant meeting or exceeding the original piping material and strength requirements. De leak sealant is compatible with the Main Steam piping material, so the probability of a steam line break is not affected. De valve and associated piping is outside primary containment and does not affect operation of the Main Steam Isolation Valves or steam line flow restrictors. The bounding accident remained the steam line break.
The existing margin of safety was not changed.
PTM 96-02 TITLE:
Leak Repair of Low Pressure Turbine Intercept Valve Flange s
DESCRIPTION: His Plant Temporary Miification authorized injection ofleak repair compound as a temporary repair on the left side Low Pressure Turbine #2 intercept valve #4 downstream flange. His PTM installed a special leak repair flange which was injected with leak sealing compound.
SAFETY ANALYSIS:
ne probability of an accident was not increased due to the leak repair flange and sealant meeting or exceeding the original piping material and strength requirements. The leak sealant is compatible with the Main Steam piping material, so the probability of a steam line break is not affected. The flange and associated piping is outside primary containment and does not affect operation of the Main Steam isolation Valves or steam line flow restrictors. The bounding accident remained the steam line break.
The existing margin of safety was not changed.
PTM 96-04 j
TITLE:
Leak Repair ofliigh Pressure Turbine Inlet Flange DESCRIPTION: This Pl.mt Temporary Modification authorized injection of leak repair compound as a temporary repair of the right side high pressure turbine steam inlet flange. This PTM installed a special leak repair flange which was injected with leak repair compound.
SAFETY ANALYSIS:
The probability of an accident was not increased due to P leak repair flange and sealant meeting or exceeding the original piping material and strength Mmms. The leak sealant is compatible with the Main Steam piping material, so the probabilk %gu, tine break is not affected. The flange and associated piping is outside primary contaimnat and dow i.)t affect operation of the Main Steam Isolation Valves or steam line flow restrictor1 Ths ? mdin ; < ident remained the steam line b ::
The existing margin of safety was not changed.
The following PTM's were implemented in 1993 and 1994, but inadverte4 not reported at that time:
PTM 93-58 TITLE:
Main Condenser Waterbox Leak Detection DESCRIPTION: This Plant Temporary Modification was issued to facilitate the injection of sulfur hexaflouride gas as a tracer for on-line Main Condenser waterbox leak detection. This PTM installed piping and performed a hot tap on each of the Circulating Water waterbox inlet lines. This PTM subsequently became a permanent modification as documented by MMP 94-060.
38
SAFETY
- ANALYSIS:
The probability of an accident is not increased due to the leak detection components meeting or exceeding the original piping material and strength requirements. Also, logic functions related to Main Condenser vacuum, Steam Jet Air Ejectors, and vacuum pumps remain unchanged. There are no accidents related to the Main Condenser in the USAR. Loss of the Main Condenser is bounded by the Turbine Trip Arialysis in the USAR, which is an Abnormal Operational Event. The Main Condenser leak detection techniques have no safety design basis. Enhanced leak detection techniques j
should reduce the potential for chemical intrusion. The existing margin of safety remained j
unchanged.
PTM 93 74
)
TITLE:
Steam Leak Repair on liigh Pressure (IIP) Turbine j
DESCRIPTION: This Plant Temporary Modification was issued to facilitate the injection ofleak repair compound as a temporary repair on the llP Turbine manway. This PTM installed four leak repair capnuts with
]
injection valves on the IIP Turbine manway to allow injection ofleak repair compound if a steam leak
]
developed.
SAFETY I
ANALYSIS:
The probability of an accident was not increased due to the leak repair capnuts meeting or exceeding the original piping material and strength requirements. De leak sealant is compatible with the Main Steam Piping material, so the probability of a steam line break was not affected. The bounding accident remained the steam line break. The IIP Turbine manways have no safety design basis. The margin of safety defined in the basis for any Technical Specification was not reduced.
PTM 94-07 and PTM 94-17 TITLE:
Monitoring of Digital Electro-llydraulic (Dell) Analog Circuitry
^
DESCRIPTION: Plant Temporary Modification 94-07 was issued to facilitate the monitoring of Deli analog circuitry, specifically governor valve demand signals associated with abnormal governor valve movement, by attaching two recorders. Plant Temporary Modification 94-17 authorized further monitoring of the drift identified by PTM 94-07.
SAFETY ANALYSIS:
The Deli system has no safety design basis and cannot initiate an accident that is not bounded by the 4
USAR. This PTM did not introduce any new failure modes. The probability of an accident or malfunction was not increased due to the existence of isolation resistors at the test connections to preclude a fault in the recorders from propagating back to the Deli system. The chart recorders do not generate any signals as input to the Deli system. The logic of the various protestive actions associated with operational events was not changed. Administrative controls were implemented to reduce the potential for physical contact with the temporary equipment and the Dell cabinet internals while the cabinet panel was removed. The panel is not subject to fire protection water spray. The existing margin of safety was unchanged.
PTM 94-13 TITLE:
Bypassing of Smoke Detector DESCRIPTION: his PTM installed jumpers and lifled a lead to bypass the smoke detector trip of two Control Room supply fans and four fire / smoke dampers. The Control Room 11VAC system was reclassified as essential, llowever, the smoke detector in the Control Room supply fan and fire / smoke dampers control logic could not be reclassified in the required time frame. This PTM temporarily bypassed the smoke detector so a non-essential component would not have the possibility of spuriousiv tripping an essential component. DC 94-262 subsequently replaced the subject smoke detector.
39
I SAFETY ANALYSl; Dis PTM affected only a smoke detector in the Control Room supply fan and fire / smoke isolation damper logic. It did not affect any components or systems involved in the USAR design basis accident analysis. A fire in the Cable Spreading Room is already analyzed in the USAR and Fire llazards Analysis with the alternate shutdown room the backup for the resulting Control Room uninhabitability. His PTM did not affect the Control Room fire / smoke dampers from isolating due to temperature. It did not affect the annunciation capability of the subject smoke detector and a Temporary Procedure Change Notice was issued to procedurally trip or close the Control Room supply fans and fire / smoke dampers upon a smoke detector annunciation. His PTM bypassed a non-safety related component to prevent malfunction of the Control Room IIVAC system. Affected components are no' identified in the Technical Specifications.
t PTM 94-20 TITLE:
Installation of Differential Pressure Indicators on Various Gland Steam Exhaust Lines DESCRIPTION: This Plant Temporary Modification was issued to facilitate quantifying flow through 10" and 12" gland steam exhaust lines and stem leak-off from the bypass valves. This data was intended to be used to assess gland steam condenser performance and possible causes for excess moisture flowing to the Elevated Release Point (ERP). The PTM utilized existing permanently installed annubars and installed tubing and two differential pressure indicators on a temporary basis. The data gathered proved not to be useful and data taking was discontinued. Ilowever, the PTM cannot be removed until RE17 due to ALARA concerns.
SAFETY ANALYSIS:
The probability of an accident or malfunction was not increased due to the instrumentation connections and tubing meeting or exceeding original piping material and strength requirements, ne Main Condenser Gas Removal, Turbine Sealing and Turbir.e Bypass systems have no safety design basis described in the USAR. The instrumentation involved cannot initiate an accident that is not bounded by the USAR and the PTM did not introduce any new failure modes. Operational transients are bounded in the USAR and the logic of the protective features described in the Technical Specifications for operational event mitigation were not altered by this PTM. The existing margin of safety remained unchanged.
PTM 94-21 TITLE:
Installation of Differential Pressure Indicators on Main Steam Bypass Valve Exhaust Lines DESCRIPTION: This Plant Temporary Modification was issued to facilitate quantifying flow through bypass valve exhaust lines. This data was intended to be used to assess bypass valve leakage. The PTM utilized existing permanently installed elbow flow meters and installed tubing and two differential pressure indicators on a temporary basis. The data gathered proved not to be useful and data taking was discontinued. liowever, the PTM cannot be removed until RE17 due to ALARA concerns.
SAFETY ANALYSIS:
The probability of an accident was not increased due to instrumentation components meeting or exceeding the original piping material and strength requirements. There are no accidents related to the bypass valves described in the USAR and the Turbine Bypass System has no safety design basis.
The instrumentation involved cannot initiate an accident that is not bounded by the USAR and the PTM does not introduce any new failure modes. De logic of the protective features described in the Technical Specifications for operational event mitigation was not altered by this PTM. The margin of safety was not reduced.
40
PTM 94-23 TITLE:
Replacement of Reactor Building Floor Drain Caps DESCRIPTION: This Plant Temporary Modification replaced two existing pneumatic pipe plugs (caps) which were subject to deterioration with a gasketed plate cap on Floor Drains No. 6 and 7 which are located in the Reactor Building. The purpose of this PTM was to enhance maintaining secondary containment 1
boundary requirements. A Minor Modification is planned to make this installation permanent.
SAFETY ANALYSIS:
There are no accidents related to these Reactor Building floor drains. The gasket floor drain cap is functionally equivalent to the pneumatic floor drain plug. The PTM was performed while secondary containment was not required to provide assurance that the PTM cannot affect the consequences of an accident.
PTM 94-24 TITLE:
Computer Room Pressure Sensing Line Installation DESCRIPTION: This Plant Temporary Modification was issued to facilitate the performance of Surveillance Procedure 63.17.18, Control Room Envelope Pressurization Test. The PTM installed a temporary pressure sensing line through an existing Computer Room wall conduit. This conduit does not penetrate the Control Room envelope, but it is an Appendix A fire barrier penetration fire seal. This modification was subsequently made permanent per an On-The-Spot-Change to DC 93-257.
SAFETY ANALYSIS:
There are no accidents related to this fire barrier penetration seal nor the equipment passing through it described in the USAR. No important to safety equipment passes through the affected conduit. Fire seal components meet or exceed applicable requirements. A continuous fire watch was posted during PTM installation which ensured that the PTM would not affect the consequences of an accident.
Except for the affected fire seal, this PTM had no interaction with equipment important to safety. The margin of safety remained unchanged.
PTM 94-25 TITLE:
Cable Spreading Room Pressure Sensing Line Installation DESCRIPTION: This Plant Temporary Modification was issued to facilitate the performance of Surveillance Procedure 63.17.18, Control Room Envelope Pressuri7ation Test. The PTM installed a temporary pressure sensing line through an existing Cable Spreading Room wall conduit. This conduit penetrates the Control Room Envelope and is an Appendix R fire barrier penetration fire seal. A Minor Modification is planned to make this installatioa permanent.
SAFETY ANALYSIS:
There are no accidents related to this spare fire barrier penetration seal described in the USAR. No important to safety equipment passes through the affected conduit. Fire seal components meet or exceed applicable requirements. A continuous fire watch was maintained during PTM installation which ensured that the PTM would not affect the consequences of an accident. Except for the fire seal, this PTM had no interaction with equipment important to safety. The PTM was performed when the Control Room Emergency Filtration System was not required to maintain a pressurized Control Room envelope.
1 41
PTM 94-32 i
TITLE:
Installation of Metals Analyzers in Main Condensate System DESCRIPTION: This Plant Temporary Modification installed Condensate Filter /Demineralizer influent and efiluent metals analyzers. The analyvers were connected to the Chemical Addition Tank Condensate lleader Root Valves, MC-V-745 and MC-V-746. The purpose of the installation is to allow sampling for metals on the Condensate Filter /Demineralizer inlet and outlet on a periodic basis. An Engineering Work Request has been initiated to install the analyzers permanently either in this location or a j
different location.
SAFETY ANALYSIS:
All components installed by this PTM are rated for pressure and temperature in excess of design piping requirements and all material used is compatible with reactor coolant service. The part of the Condensate system affected by this PTM is nonessential ar d ths Condensate system provides no safety function. The possibility exists that Core Standby Coolleg System pressure maintenance could be affected; however, the Reactor B.:ilding Auxiliary Condensate Booster Pump provides a backup supply for the nonessential pres:;ure maintenance. The only fenction of the metals analyzers is to remove a small sample stream from the condensate header at two locations. This sample stream is discarded, not returned to the system. Therefore, the possibility of a different type of accident or malfunction is not introduced. The margin of safety is not affected.
PTM 94-34 TITLE:
RilR lleat Exchanger Temporary Drip Pan DESCRIPTION: This Plant Temporary Modification was written to document the acceptability of an existing
)
temporary drip pan attached to RilR lleat Exchanger A. The drip tray captures a small amount of RilR lleat Exchanger A shell-to-tubesheet flange leakage and channels it to a Reactor Building sump.
i This installation was subsequently documented as a permanent modification per DC 94-250.
SAFETY ANALYSIS:
The temporary drip pan for flange leakage did not adversely affect the operation and reliability of the i
RilR system or any other safety-related system / components. The RilR Heat Exchanger flange leakage had a negligible effect on onsite and offsite radiological accident doses and posed no internal flooding concerns. The drip pan is constructed oflight gauge sheet metal which has no adverse effect on the RilR Ileat Exchanger due to the small amount of weight added. It is also securely attached and cannot become dislodged; therefore, seismic criteria have been met. Flange leakage is not expected to grow over time.
PTM 94-37 TITLE:
Replacement Diesel Generator Exhaust Bypass Valves DESCRIPTION: This Plant Temporary Modification authorized temporary installation of replacement valves for DG-AOV-MBI and DG-AOV-MB2. The replacement valves have bolt holes drilled for a 25 lb.
flange instead of a 125 lb. Flange, which means that four bolts / studs cannot be installed. An Engineering Work Request (EWR 95-085) has been initiated to approve the installed valve design and configuration.
SAFETY ANALYSIS:
The effect of the four missing studs / bolts is insignificant because the reduction in joint strength is less than 15% and these joints do not support any significant pressure loading. These aspects ensure a sufficient margin of strength remains for this configuration. This nonstandard joint configuration j
does not create any new failure modes, nor increase the probability of existing failure modes. It does 1
not significantly reduce any margins of safety, nor affect system or component performance.
I 42
i 1
TEMPORARY DESIGN CII ANGES (TDCs)
TDC 93-059 TITLE:
Temporary Turbine Equipment Cooling (TEC) Pump Suction Strainers DESCRIPTION: As a result ofInformation Notice 85-96 and Inspection Report 93-06, the temporary suction strainers for TEC pumps A and C were to be removed. This Temporary Design Change documented that the strainers would remain in place until such time as they were removed during normal maintenance activities. It also provided instructions for the subsequent permanent removal of the strainers from the piping system.
SAFETY i
ANALYSIS:
This TDC only afTected nonessential equipment located in the Turbine Building. Therefore, the
{
probability of occurrence or consequences of an accident or malfunction of equipment important to safety were not increased, and the margin of safety as defined in the Technical Specifications was not reduced.
TDC 44-224
)
TITLE:
Automatic Closure Modifications for CS-MOV-MOSA and CS-MOV-MOSB and Diesel Generator liVAC Fuse Replacement DESCRIPTION: This Temporary Design Change added a time delay relay to the automatic closure function of CS-MOV-MOS A and CS-MOV-MO5B in order to prevent the minimum flow isolation valves from closing on spurious high flow signals. This TDC also increased the fuse size for the Diesel Generator exhaust fan control circuit to increase the margin in the circuit to withstand sustained low voltage dips which may occur during the starting of the Core Spray (CS) pump motors. An Engineering Project Request has been initiated to make these modifications permanent.
SAFETY ANALYSIS:
This TDC afTects various safety related equipment and temporarily disables safety functions during l
installation and testing. No modifications or testing perforued directly increases the likelihood of an accident occurring. Modifications to the CS minimum flow bypass valves results in them remaining open longer during CS operation. The resulting impact on delivered CS flow has been analyzed to show that CS performance will still meet Technical Specification requirements and there is negligible effect on the CNS LOCA analysis. Changing of the Diesel Generator ilVAC fuses has no impact on the operational characteristics of the system. Therefore, this TDC does not increase the consequences of an accident or reduce the margin of safety in the bases of any Technical Specifications. Modifications performed for this TDC do not introduce any new failure modes for the associated components or systems.
TDC 44-364 TITLE:
liPCI Check Valve Disc / Stud Repair DESCRIPTION: This Temporary Design Change repaired the disc / stud assembly for swing check valve llPCI-CV-15CV since indications were found in the existing stud and replacement parts were unavailable. This TDC was removed during the 1995 refueling outage when internal check valve components were replaced with new components.
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43
r
. SAFETY
~ ANALYSIS:
Repair of the subject check valve restored the stmetural integrity of the component and did not affect its safety function. Failure of the valve was not identined as an initiating event for any accidents evaluated in the USAR. The check valve repair did not affect the stresses in the piping system and l
the containment isolation function was not impaired, ModiGcations per this TDC did not introduce any new failure modes for the affected component or system; therefore, the probability of occurrence or consequences of an accident or equipment malfunction were not increased. 'Ihe design philosophy utilized by this TDC complied with all established design requirements, thus ensuring that margins of safety were not reduced.
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l PROCEDURE CllANGE NOTICES (PCNs) 1 Precedure Chance Notice (PCN) 0.5 (Revision 4)
TITLE:
Condition Reporting DESCRIPTION: This PCN provides clarification to the Operability Assessment process and shifts existing requirements into difTerent procedures. It does not affect the operability determination as described in the Technical Specifications or USAR. The new process has a licensed Shift Supervisor and the Condition Review Group review operability determinations instead of the Station Operations Review l
Committee (SORC).
SAFETY ANALYSIS:
The PCN changes the review process for operability determinations, but the criteria for determining operability remain unchanged. Operability Assessments will still be made within the bounds of Technical Specifications for related equipment, thus USAR assumptions remain bounding. This change climinates the Operability Determination and Operability Evaluation documents that were i
previously reviewed by SORC under their respective procedures. Ilowever, this PCN maintains SORC review of " station operation to detect potential nuclear safety hazards," as required by Technical Specifications and the USAR, by the requirement for reporting of significant Condition Reports to SORC. No reduction in the margin of safety is introduced.
Procedure Chance Notice (PCN) 0.23 (Revision 12)
Procedure Chance Notice (PCN) 5.4.1 (Revision 28)
TITLE:
CNS Fire Protection Plan (0.23)
General Fire Procedure (5.4.1)
DESCRIPTION: These PCNs eliminate reference to the Reserve Fire Brigade. The Reserve Fire Brigade has been eliminated due to its duplication under the Emergency Response Organization. Also, responsibility for fire drills and brigade training has been assigned to Training; the Maintenance Supervisor is no longer assigned this responsibility.
SAFETY ANALYSIS:
The probability of fire occurrence is not increased by this change. The personnel formerly on the reserve fire brigade will respond to a fire as part of the Emergency Response Organization for fires of sufficient severity to require personnel beyond Technical Specification minimums for fire brigade staffing. The consequences of equipment malfunction during a fire event have been evaluated in response to Appendix R. The elimination of the Reserve Fire Brigade has no impact on these consequences since all equipment important to safety is assumed lost in a given fire zone. The reserve fire brigade is not described in the Technical Specifice.tions.
Procedure Chance Notice (PCN) 0.26 (Revision 13)
TITLE:
Surveillance Program DESCRIPTION: This PCN instituted additional administrative controls on the scheduling and performance of routine surveillance testing. The majority of the changes under this PCN represent either an alternative method of satisfying USAR requirements or are more restrictive methods than the previous revision of this procedure. Major changes were the establishment of administrative Allowed Out-of-Service Times (AOTs) and the establishment of divisional testing.
45
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l SAFETY ANALYSIS:
The implementation of the new Surveillance Program administrative controls of AOTs and divisional testing are consistent with the intent and letter of the regulatiors, CNS Technical Specifications, and the USAR. This new approach balances the safety and regulatory interests of verifying operability and maximizing safety system availability for coping with operational transients or design basis accidents, ne probability or consequences of an accident or malfunction are not increased as a result of these changes. There are no formal Bases for the Surveillance Testing Program administrative controls; however, the change is consistent with Technical Specification provisions. As a result, there is not a reduction in the margin of safety.
Procedure Chance Notice (PCN) 0.39 (Revision 8)
TITLE:
Fire Watches DESCRIPTION: This procedure was revised to define and clarify several issues related to hot work permits, fire protection impairment permits, and fire watch duties. It specifically defined the tolerance on the frequency of the fire watch patrol as 1.25 times the interval as specified in the fire protection impairment permit after establishment of the initial patrol within 60 minutes.
{
SAFETY ANALYSIS:
This PCN does not increase the probability of occurrence or consequences of a fire. Fire watch patrol frequency does not affect equipment malfunction or create the possibility of an accident of a different type than previously evaluated; it serves as a compensatory measure when equipment fails. The basis for the fire watch patrol interval is given in NUREG 1433, Example 1.3-6, Completion Times. His g
change adopts standard industry practice.
Procedure Chance Notice (PCN) 0.40 (Revision 21 Procedure Chance Notice (PCN) 7.0.1.2 (Revision 5)
Procedure Chance Notice (PCN) 7.01.5 (Revision 0) j Procedure Chance Notice (PCN) 7.0.1.7 (Revision 0)
TITLE:
Work Control Program (0.40)
Maintenance Work Request Planning (7.0.1.2)
Special Instructions (7.0.1.5)
Troubleshooting Plant Equipment (7.0.1.7)
DESCRIPTION: Procedure 0.40 was revised to identify the administrative requirements to generate and control work activities at CNS. It added administrative requirements for troubleshooting equipment to determine corrective maintenance when required, added administrative requirements for Maintenance Planning i
to provide work instructions on the Maintenance Work Request (MWR) without writing Special i
instructions, and added administrative requirements for revising MWRs. Procedure 7.0.1.2 added requirements for Maintenance Planning to provide written work instructions from approved vendor manuals, drawings, etc., and removed the section for Special Instructions. Procedure 7.0.1.5 will only be used to provide Special Instructions if nuclear safety could be affected when performing routine maintenance. Procedure 7.0.1.7 is a new procedure to perform troubleshooting on inoperable equipment when corrective action is uncertain or cannot be identified for normal routine maintenance activities.
SAFETY ANALYSIS:
The new maintenance controls simplify the administrative process. Administrative requirements remain in place for maintenance and troubleshooting such that system availability and operability are controlled consistent with USAR assumptions. Post-maintenance testing confirms that equipment functions as designed prior to being considered operable. Administrative requirements limit the equipment that can be out of service during maintenance and troubleshooting such that the probability of equipment malfunction is not increased. Rese changes are consistent with Technical Specification provisions and do not reduce the margin of safety defined in the basis for any Technical Specification.
46
Procedure Chance Notice (PCN) 2.L1 (Revision 70)
TITLE:
Startup Procedure DESCRIPTION: Most of the changes made by this PCN were administrative in nature. Changes were made related to the Scram Discharge Volume level switches and Average Power Range Monitor calibration settings. Neither of these changes affected the USAR. A change was made to the Daily Jet Pump Operability Check in Startup which changed thejet pump operability monitoring program which is included in the USAR. This change deleted the procedural requirement to perform a third check which was determined to be inadequate to comply with Technical Specification surveillance l
requirements.
SAFETY ANALYSIS:
Jet pumps are not initiators of accidents; jet pump integrity is required for accident mitigation. Jet pump operability is assured by the surveillance required by Technical Specification 4.6.E. Although a third check in the procedure was determined to be inadequate to demonstrate Technical Specification compliance and has been removed from the procedure, there are two other checks that are adequa* to ensurejet pump operability. Herefore, this procedure change does not increase the consequences of an accident or malfunction of equipment. The margin of safety in the Technical Specifications is increased by precluding jet pump operability from being based on the satisfactory result of a test which does not adequately address the Technical Specification requirement.
Procedure Chance Notice (PCN) 2.1 11 (Revision 66)
Procedure Chance Notice (PCN) 6 3.10.14 (Revision 61 Procedure Chance Notice (PCN) 13.1 (Revision 01 TITLE:
Station Operators Tour (2.1.11)
ECCS Leakage Walkdown (6.3.10.14)
ECCS Leakage Evaluation (13.1)
DESCRIPTION: De revision to Procedure 2.1.11 implemented a more comptbensive gathering and evaluation of ECCS leakage data. DC 94-250 established the basis for an ECCS leakage limit and outlined an ECCS leakage monitoring program. Procedure 2.1.11 identifies the inspections to be performed as part of the operator rounds to inspect for leakage in the Reactor Building during normal, standby operations. This procedure checks for gross loss ofintegrity. Leakage information collected during this procedure is provided to CNS Engineering for evaluatica as part of Procedure 13.1 which was developed to evaluate ECCS leakage to ensure that design basis assumptions are met. Procedure 6.3.10.14 provides for a walkdown of ECCS components involved in post-LOCA recirculation.
SAFETY ANALYSIS:
Le subject tests are a supplementary inspection that is performed in parallel with existing plant test procedures and do not require any intrusion into plant systems or structures. They are intended to verify the ability of the plant to respond to a design basis accident within the licensing basis assumptions. These inspections do not increase the probability or consequences of an accident or malfunction of equipment. There is no specific reference to leak tightness criteria for the ECCS j
systems in the CNS Technical Specifications. These ECCS leakage inspections do not reduce the j
margin of safety as defined in the basis for any Technical Specifications.
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I Procedure Chance Notice (PCN) 2.1.20 (Revision 24)
Procedure Chance Notice (PCN) 7.4.3 (Revision 19)
Procedure Chance Notice (PCN) 7.4.4 (Revision 23)
T!TLE:
Reactor Pressure Vessel Refueling Preparation (2.1.20)
Reactor Vessel Top flead insulation Removal (7.4.3)
Reactor Pressure Vessel llead Removal (7.4.4)
DESCRIPTION: These procedures were revised to reflect a modified method of Reactor Pressure Vessel (RPV) head cooldown and removal of the head and associated components. Reactor vessel water level will be raised 3 to 4 feet into the vessel head during plant cooldown to increase the rate of vessel head cooldown. This will allow work on the vessel to commence sooner than normal. In order to free the main turbine for maintenance as soon as possible afler reactor nutdown, the main condenser system j
will not be utilized for venting. Instead, the Main Steam Iso. ion Valves will be closed, the vessel depressurized and vented through the vessel head vent. When the head has cooled sufTiciently, the vessel head vent will be opened and a llEPA filter unit connected to the reactor vessel head vent.
Gasses removed from the vessel will be filtered and routed via the Spent Fuel Pool Ventilation to the Reactor Building Ventilation System for monitoring and release.
SAFETY ANALYSIS:
The 11 EPA filter unit will be installed after reaching cold shutdown when reactor pressure is essentially atmospheric. 'the llEPA filter unit will be disconnected and the head vent returned to its original configuration prior to reinstalling the vessel head after refueling. The temporary connection will have no effect on any refueling accident and does not interact with any other equipment in such a way as to cause an accident or change the probability of occurrence or consequences of an accident or malfunction of equipment. Technical Specification 3.6 that describes the requirements for the reactor pressure boundary is not atTected by this activity. Technical Specification 3/4.21.C, Gaseous EfTluents, provides limits for offsite doses from radioactive gas releases. The effluent from the llEPA filter unit will be sent to the Reactor Building Vent Monitor which, in conjunction with grab samples, assures that the margin of safety is not reduced.
Procedure Chance Notice (PCN12.2.22 (Revision 31)
TITLE:
VitalInstrument Power System DESCRIPTION: This procedure change adds steps to prevent the loss of shutdown cooling during the transfer of power for the Reactor Protection System (RPS) from the RPS Motor Generator (MG) Set to the alternate power source. This power transfer results in a momentary power loss to the RPS which results in an isolation signal to the RilR valves used for loss of shutdown cooling. To prevent the loss of shutdown cooling, this PCN deenergizes the motor and control power to the affected motor operated valves for the brief period during which the RPS power transfer is being made. Personnel are stationed at the valve breakers and are in continuous communication with the Control Room. Should i
reactor cavity level rapidly decrease during this evolution, the operators are directed to close the applicable breakers.
SAFETY ANALYSIS:
This power supply transfer will be performed only when the reactor cavity is flooded and equalized with the fuel pool level. Although the automatic isolation capability of the valves will be disabled for a short period of time during the power transfer, an Operator will be stationed at the applicable motor control center / starter to reclose the disconnect switch in the event a low water level signal is received. This action will occur within the time evaluated for any loss of coolant and associated containment isolation function evaluated in the USAR. Inadvertent loss of shutdown cooling is a malfunction analyzed in the USAR; however, this change will reduce the potential for this malfunction during power supply transfer. The probability of occurrence or consequences of an accident or malftmetion are not increased by this change and the margin of safety is not reduced.
48
Procedurt Chance Notice (PCN) 2.2 31 (Revision i1)
Procedure Chance Notice (PCN16.1.27.1 (Revision 3)
TITLE:
Fuel llandling - Refueling Platform (2.2.31)
Daily Refueling Equipment Check (6.1.27.1)
DESCRIPTION: nese procedures were revised as a result ofimplementation of Minor Modification 95-093, Refuel Bridge Air Compressor Replacement. The procedures were revised to reflect the loading and unloading pressures of the new compressor.
SAFETY ANALYSIS:
The subject compressor supplies air to the grapple on the refuel bridge. If the air compressor were to fail, the grapple would not be able to perform any opening or closing function; however, this activity is not considered in any accident scenario or mitigation activity. The air compressor was replaced with a comparable component; therefore, the probability or consequences of a malfunction of equipment to safety were not increased. The refuel bridge does not affect the function of any equipment important to safety. The air compressor is not defined in the bases for any Technical Specification; therefore, the margin of safety is not affected.
Procedure Chance Notice (PCN12.2.65 (Revision 33)
Procedure Chance Notice (PCN) 2.2.65 A (Revision 12)
TITLE:
Reactor Equipment Cooling (REC) Water System (2.2.65)
REC Valve Checklist (2.2.65A)
DESCRIPTION: These procedure revisions specify that the Drywell throttling valves for the REC system are maintained in the full open position and that the indicated flows are suggested minimums. The Drywell REC valves are maintained full open in order to ensure that cooling water flow can be maintained above recommended minimums during plant operation (while the Drywell throttling valves are inaccessible).
SAFETY ANALYSIS:
De Drywell cooling loop of the REC system does not provide any safety function. This change will not affect the ability of the REC system to perform its safety function as specified in the USAR or affect the design of the system. This change does not affect any of the precursors for accidents previously evaluated in the USAR. Maintaining these valves in the open position is in accordance with the tested configuration of the system. Therefore, this change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR, nor create the possibility of a different type of accident or malfunction. The Bases of the Technical Specifications state that each subsystem is capable of supplying the cooling requirements of the essential services following design accident conditions with only one pump in either subsystem. His change does not affect the ability of the REC system to meet these conditions; therefore, the margin of safety is not reduced.
Procedure Chance Notice (PCN) 2.2.68 (Revision 40)
Procedure Chance Notice (PCN) 2.2.68.1 (Revision 13)
Procedure Chance Notice (PCN31013 (Revision 25)
TITLE:
Reactor Recirculation System (2.2.68)
Reactor Recirculation System Operations (2.2.68.1)
Control Rod Sequence and Movement Control (10.13) l l
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DESCRIPTION: These PCNs deleted information concerning Preconditioning Interim Operating Recommendations (PCIOMR) because its use has been discontinued. Preconditioning of fuel was required to reduce the potential for pellet clad interaction (PCI) and fuel failure. Use of barrier fuel interrupts the PCI process, thereby eliminating PCI type fuel failures. This change was also incorporated into the USAR.
SAFETY ANALYSIS:
PCIOMR does not affect accidents as described in the USAR. PCIOMR implemented a guideline to reduce PCI related fuel failures. Fuel failures from a LOCA or other accidents bound any possible PCI fuel failures. Herefore, there is no change in the probability or consequences of an accident.
PCIOMR does not affect any plant equipment important to safety. Equipment important to safety previously evaluated in the USAR did not use PCIOMR; thus elimination of PCIOMR cannot increase the probability or change the consequences of a malfunction of equipment. PCIOMR is not used in the basis for any Technical Specifications.
Procedure Chance Notice (PCN,)].j&. ision 41)
TITLE:
Reactor Recirculation (RR) System DESCRIPTION: This procedure change specifies that the valves for the REC flow to the Recirculation Pump Seal Cooler and to the Pump Motor Upper and Lower Bearing Coolers are maintained in de full open position, and that 48 gpm is the expected maximum through the Seal Coolers. This change also includes a check of the RR pump motor bearing cooler outlet valves.
SAFETY ANALYSIS:
This change does not affect the ability of the RR pumps or the REC system to perform their safety functions as specified in the USAR. His change does not afTect the design of the systems and does not afTect any of the precursors for accidents previously evaluated in the USAR. Maintaining these valves in the open position is in accordance with the tested configuration of the system. Therefore, this change does not increase the probability of occurence or consequences of an accident or malfunction of equipment previously evaluated in the US AR, nor create the possibility of a different type of accident or malfunction. This change does nc4 affect the capability of the RP, pumps or the REC system to meet the conditions specified in the basis of the Technical Specifications; therefore, the margin of safety is not reduced.
Procedure Chunce Notice (PCN) 2.2.69.2 (Revision 16)
TITLE:
Residual lleat Removal (RilR) System Shutdown Operations DESCRIPTION: This change proceduralized the use of two RHR pumps in a loop in Shutdown Cooling (SDC) for Reactor Pressure Vessel heatup. This mode of operation is desired to use the mechanical work of two pumps to increase the heatup rate of the reactor over one pump operation.
SAFETY ANALYSIS:
Reactor coolant pressure is limited to the RilR SDC mode pressure, so accidents as evaluated in the USAR are not credible. USAR Apperdix G events bound two pump operation in that reactor level, pressure, and temperature are controlled. The established limits for two pump operation are adequate to ensure proper Net Positive Suction llead (NPSH), heat exchanger flow control, and protection of incore instrumentation. He consequences of accidents remain unaffected. Two pump operation does not introduce any operational characteristics that create the possibility of new accidents or equipment malfunction. Limitations in the procedure ensure that the margin of safety as defined in the basis of the Technical Specifications is not reduced.
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Procedure Chance Notice (PCN) 2.2.77.1 (Revision 1)
TITLE:
Digital Electro-hydraulic (Deli) Control System DESCRIPTION: The intent of this change is to increase cycle efficiency by increasing reactor vessel dome pressure to a pressure slightly less than rated reactor operating pressure. This change allows Operations personnel to adjust the pressure setpoint during full power operation and end of cycle coastdown to maintain pressure < 1004 psig.
SAFETY
)
ANALYSIS:
This activity permits operation within rated parameters without compromising previously established margins of safety for vessel overpressurization and does not increase the probability of a turbine trip l
without bypass, which is the bounding pressure transient. Therefore, the probability of occurrence i
or consequences of a previously evaluated accident are not increased. Operating conditions created by this activity will not exceed normal operating conditions of the reactor coolant system. This does not increase the possibility of a different type of accident or malfunction of equipment.
Preestablished margins of safety as defined in the basis for any Technical Specifications are not reduced.
Procedure Chance Notice (PCN) 2.2.97 (Revision 0)
TITLE:
Torus Drain and Refill Operation DESCRIPTION: This new procedure was developed to provide guidance for draining the suppression pool using the
)
installed torus drain pump and also for refilling the torus after draining. The USAR was also revised accordingly.
SAFETY ANALYSIS:
No accident initiators are associated with the torus drain subsystem. This system is used only during cold shutdown conditions and with sufficient / redundant available ECCS cooling from an alternate source as allowed by Technical Specifications. The system is returned to normal configuration prior to startup. Failure of this equipment will not affect cooling water supply to any other important to safety equipment. This procedure imposes all applicable Technical Specification and USAR j
requirements.
-l Procedure Chance Notice (PCN) 2.3.2.37 (Revision 17)
TITLE:
Fire Protection - Annunciator 1 DESCRIPTION: This PCN added steps to perform manual actions to override the fire detector interlocks with the Reactor Building Sump inlet isolation valves. The purpose of this action is to prevent torus area contamination when a fire detector inadvertently alarms and closes the Reactor Building Sump isolation valves.
SAFETY ANALYSIS:
The Reactor Building Floor Drain Sump liigh-liigh Level Switch and fire detector interlocks with the Reactor Building Sump inlet isolation valves are not credited by the flooding analysis or credited to prevent flooding from fire suppression activities. The automatic interlocks are not required to perform any safety function. Therefore, taking manual actions to override the interlocks is acceptable from a safety function perspective and does not increase or affect the probability of occurrence or consequences of an accident or malfunction of equipment. The interlocks are not described in the basis for any Technical Specification; therefore, the margin of safety is not reduced.
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Procedure Chance Notice (PCN) 3.4.5 (Revision 5)
TITLE:
Minor Modifications DESCRIPTION: This procedure for Minor Modifications replaced the Equipment Specification Change procedure.
The new Minor Modification procedure provides a more efficient means of completing simple modifications than the Design Change procedure.
SAFETY ANALYSIS:
This procedure provides the necessary control to ensure that the requirements of 10CFR50.59 and 10CFR50, Appendix B will continue to be satisfied. It does not increase the probability or consequences of an accident or malfunction of equipment and does not reduce the margin of safety j
in the basis for any Technical Specification, i
Procedure Chance Notice (PCN) 4.6.1 (Revision 18)
Procedure Chance Notice (PCN) 4.6.1 A (Revision 31 TITLE:
Reactor Vessel Water Level Indication (4.6.1)
Reactor Vessel Water Level Indication Component Checklist (4.6.l A)
DESCRIPTION: These procedures were revised to reflect Core Spray system valve lineup changes intended to preclude excessive wear on Reference Leg Injection system valves NBI-SOV-SSV738 and NBI-SOV-SSV739. Leakage of air through these SOVs had previously resulted in reactor water level indication fluctuations. This valve lineup change maintains isolation valves CS-V-147A and CS-V-1478 closed below a reactor pressure of approximately 310 psig for normal stanup and shutdown conditions.
SAFETY ANALYSIS:
This valve lineup change does not affect the safety objective of the Reference Leg Injection system.
The system will still be fully functional during normal operation and capable of remote manual initiation when required. Maintaining CS-V-147A and B closed at vessel pressures below 310 psig enhances the reliability of the affected NBI SOVs, thereby minimizing the potential of air / water leakage to the reference leg and avoiding spurious water level indications when the vessel is depressurized. The CS/NBI system interface will be unchanged during normal operation and is available to respond to postulated events for which it was designed. Any single failure of the CS isolation valves to open would affect only one train of reactor water level indication. Any failure of components due to this valve sequencing change is envcloped by existing accident evaluations with loss of one train of CS and one train of reactor water level indication. The margin of safety as defined
)
in the Technical Specifications is not reduced.
j Procedure Chance Notice (PCN) 5.1.3 (Revision 20)
TITLE:
Flood DESCRIPTION: This PCN was revised to initiate monitoring of Z sump operability when flood levels exceed 890' elevation, instead of 895' elevation. A Condition Report evaluation determined that Z sump is required to function to maintain Standby Gas Treatment and secondary containment operability and consequently required the Z sump pumps and level control components to be classified as essential.
The Z sump components, although contained in a sealed sump, are vulnerable to flooding above 890' ekvation.
SAFETY ANALYSIS:
A flood affecting Z sump and potentially affecting SGT does not increase the probability of an accident occurring since SGT is only used to mitigate the consequences of an accident. Ifflooding causes a failure of Z sump, SGT will be declared inoperable and appropriate actions taken to place the plant in a safe condition. Increased monitoring of Z sump will ensure that the effects of flooding will be noted and appropriate action taken. Section 3.13 of Technical Specifications specifies 52
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entering the flooding procedures at 895' and placing the plant in a cold shutdown condition at 902' elevation. This is based upon the locations and elevations for safe-shutdown components. Z sump and the SGT system are not required for safe-shutdown, so the margin of safety in the Technical Specifications is not reduced by lowering the flood initiation level to 890'.
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Procedure Chance Notice (PCN) 5.8 (Revision 8)
J TITLE:
Emergency Operating Procedures (EOPs)
DESCRIPTION: This revision to Emergency Opeiating Procedure 5.8 incorporated revisions to various EOP i
flowcharts. The most significant revision was the incorporation of changes resulting from DC 94-04I which initiated changes to the SLC tank levels and concentrations.
SAFETY ANALYSIS:
CNS EOPs accommoda% A Agies for events beyond the licensed design basis of the plant. The NRC, in its Safety Eval.sation Report (SER) on the Boiling Water Reactor Owners Group (BWROG)
Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPGs and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised EOP does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
His revised EOP carmot increase the probability of occurrence of any event analyzed in the USAR j
because the procedure will only be used after the event has commenced. The implementation of this revised EOP does not create a possibility for an accident or malfunction of a different type than any previously evaluated in the USAR because the revised EOP does not modify the operation or design basis of the plant. For plant conditions which already exceed the licens;ng design basis, the question of a reduced margin of safety is not meaningful.
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Procedure Chance Notice (PCN) 5.8.5 (Revision 4)
TITLE:
Injection Subsystems (Table 5)
DESCRIPTION: This revision to EOP Support Procedure (ESP) 5.8.5 made the following changes: 1) incorporated changes from DC 94-332 which changed the normal position of the RHR minimum flow valves from closed to open; 2) revised procedure sequence to attempt to use LPCI A and then LPCI B rather than attempt to line up both systems concurrently; 3) added EOP Plant Temporary Modifications to allow remote closure of the RHR heat exchanger SW outlet valves after the SW Booster Pumps have been started; and 4) changed the NBl Coritinuous Backfill isolation valve used for RPV injection with the a
CRD system from CRD-15 to CRD-63 to utilize a valve that is easier to use and access.
SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design basis of the plant. De NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG)
Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPGs and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised ESP does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
This revised ESP cannot increase the probability of occurrence of any event analyzed in the USAR because the procedure will only be used after the event has commenced. The implementation of this revised ESP does not create a possibility for an accident or malfunction of a different type than any previously evaluated in the USAR because the revised ESP does not modify the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
53
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Procedure Chance Notice (PCN) 5.8.8 (Revision 3) l TITLE:
Alternate Boron Injection and Preparation DESCRIPTION: This Emergency Support Procedure (ESP) was revised to incorporate changes to the Standby Liquid l
1 Control (SLC) Tank level due to the implementation of DC 94-04 L Chemical concentration in the SLC tank was also revised to reflect the normal concentration of the SLC tank.
SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design ba is of the plant. He
)
NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG)
Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPGs and EOPs, j
rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised ESP does not increase the probability of occurrence or consequences i
of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
This revised ESP cannot increase the probability of occurrence of any event analyzed in the t'SAR because the procedure will only be used after the event has commenced. The implementation of this revised ESP does not create a possibility for an accident or malfunction of a different type than any previously evaluated in the USAR because the revised ESP does not modify the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
Procedure Chance Notice (PCN) 5.8.9 (Revision 3)
TITLE:
Average Suppression Pool Temperature Calculation DESCRIPTION: This Emergency Support Procedure (ESP) revision incorporated changes implemented by DC 94-212A, which removed PC-TE-20B. Steps were added to provide for use of a different recorder, RilR-TR-131, as a last recourse for measurement of suppression pool temperatures. This recorder has a scale of zero to six-hundred degrees, which is more than adequate for suppression pool temperature information.
SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG)
Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPGs and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised ESP does not increase the probability of occurrence or consequences j
of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
i This revised ESP cannot increase the probability of occurrence of any event analyzed in the USAR j
because the procedure will only be used after the event has commenced. The implementation of this
)
revised ESP does not create a possibility for an accident or malfunction of a different type than any previously evaluated in the USAR because the revised ESP does not modify the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question
)
of a reduced margin of safety la not meaningful.
Procedure Chance Notice (PCN) 5.8.11 (Revision 3)
TITLE:
Emergency Opert ting Procedure (EOP)- RPV Venting During Primary Containment Flooding DESCRIPTION: This revision to EC P Support Procedure (ESP) 5.8.11 moved the step for removal of PTM 47 so that the PTM is removed prior to closing IIPCI-MO-15 because the valve cannot be closed while the PTM is installed. This fact is identified earlier in the procedure in the form of a caution just prior to installation of the PW.
54
1 SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed basis of the plant. The NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG)
Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPG and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised ESP does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
This revised ESP cannot increase the probability of occurrence of any event analyzed in the USAR because the procedure will only be used after the event has commenced. Implementation of this revised ESP does not create a possibility for an accident or malfunction of a different type than previously evaluated in the USAR because the revised ESP does not modify the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
Procedure Chance Notice (PCN) 5.R.13 (RevisionJ)
TITLE:
Emergency Operating Procedure (EOP)- Outside Shroud Injection Systems (Failure to Scram)
DESCRIPTION: This revision to EOP Support PrNedure (ESP) 5.8.13 incorporated changes from DC 94-332, which changed the normal position of the RHR minimum flow valves from closed to open. It also adds a new step to secure the RCIC gland seal vacuum pump after the RCIC turbine is secured and a caution that primary containment oxygen levels may rise during RCIC gland seal vacuum pump operation.
In addition, it adds cautions regarding RHR SWBP amps to confonn to system opcrating procedures.
SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG)
Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the f J and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised ESP does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
This revised ESP cannot increase the probability of occurrence of any event analyzed in the USAR because the procedure will only be used after the event has commenced. Implementation of this revised ESP does not create a possibility for an accident or malfunction of a different type than previously evaluated in the l' because the revised ESP does not modify the operation or design basis of the plant. For plant (
.ons which already exceed the licensing design basis, the question of a reduced margin of safet.
aot meaningful.
Procedure Chance Notice (PCN) 5.8.16 (Revision 4)
TITLE:
Emergency Operating Procedure (EOP)- Outside Shroud Flooding Systems (Failure to Scram)
DESCRIPTION: This revision to EOP Support Procedure (ESP) 5.8.16 incorporated changes from DC 94-332, which changed the normal position of the RHR minimum flow valves from closed to open. It also changed the NBl continuous backfill isolation used for injection with the CRD system from CRD-15 to CRD-63 to utilize a valve that is easier to use and access.
SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG)
Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPG and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised ESP does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
l This revised ESP cannot increase the probability of occurrence of any event analyzed in the USAR l
because the procedure will only be used after the event has commenced. Implementation of this 55 l
l J
t i
revised ESP does not create a possibility for an accident or malfunction of a different type than
.j previously evaluated in the USAR because the revised ESP does not modify the operation or design l
basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
1 Procedure Chance Notice (PCN) 5.8.17 (Revision 3)
\\
1 i
TITLE:
Emergency Operating Procedure (EOP)- Primary Containment Venting DESCRIPTION: This revision to EOP Support Procedure (ESP) 5.8.17 provided clarification to direct monitoring of Elevated Release Point (ERP) effluent radiation, versus monitoring only the ERP cffluent radiation monitors. This provides flexibility to the operator to be able to use other means to monitor ERP radiation.
1 SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG) 3 Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPG and EOPs, J
rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised ESP does not increase the probability of occurrence or consequences of an ac:ident or malfunction of equipment important to safety previously evaluated in the USAR.
This revised ESP cannot increase the probability of occurrence of any event analyzed in the USA 3 because the procedure will only be used after the event has commenced. Implementation of this revised ESP does not create a possibility for an accident or malfunction of a different type than previously evaluated in the USAR because the revised ESP does not modify the operation er dnic.n basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
Procedure Chance Notice (PCN) S 8.18 (Revision 4)
TITLE:
Emergency Operating Procedure (EOP) Primary Containment Venting for Primary Containment Pressure Limit DESCRIPTION: This revision to EOP Support Procedure (ESP) 5.8.18 incorporated the modified control logic implemented per DC 95-037A. It added steps to jumper the Group VI Reactor Building high radiation signal and depress the new reset pushbuttons.
SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG) 1 Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPG and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. De implementation of this revised ESP does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
{
Dis revised ESP cannot increase the probability of occurrence of any event analyzed in the USAR because the procedure will only be used after the event has commenced. Implementation of this revised ESP does not create a possibility for an accident or malfunction of a different type than previously evaluated in the USAR because the revised ESP does not modify the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
j 56
i Procedure Chance Notice (PCN) 5.819 (Revision 3)
TITLE:
Emergency Operating Procedure (EOP)- Reference Leg injection DESCRIPTION: This revision to EOP Support Procedure (ESP) 5.8.19 provided additional guidance to ensure personnel safety while performing approved EOP actions outside the Control Room. The guidance
{
improves personnel safety awareness and does not alter or create new EOP actions. Instructions were also added to cover the possibility that the reference leg isolation valve is closed when entering the EOP.
SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG) i Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPG and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. 'Ihe implementation of this revised ESP does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
This revised ESP cannot increase the probability of occurrence of any event analyzed in the USAR because the procedure will only be used after the event has commenced. Implementation of this revised ESP does not create a possibility for an accident or malfunction of a different type than previously evaluated in the USAR because the revised ESP does not modify the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
Procedure Chance Notice (PCN) 5.8.20 (Revision 3 and 4)
TITLE:
Emerg,ency Operating Procedure (EOP)- EOP Plant Temporary Modifications (PTMs)
DESCRIPTION: Revision 3 to EOP Support Procedure (ESP) 5.8.20 provided a new step to place all closed Main i
Steam isolation Valve control switches to the CLOSE position to prevent inadvertent opening due to l
the installation of EOP PTMs 57 through 60. New steps were also added to provide guidance for the operator to transfer llPCI suction to the Emergency Condensate Storage Tank ifit had transferred to the torus prior to installation of EOP PTMs 51 and 52. Changes were made to the EOP PTM Table of Contents to reflect changes made in other EOPs; however, this revision did not make any PTM changes itself. Revision 4 to ESP 5.8.20 incorporated various editorial, administrative and format changes due to an EOP Verification and Validation that was performed in October 1995.
SAFETY ANALYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG)
Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPG and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised ESP does not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
This revised ESP cannot increase the probability of occurrence of any event analyzed in the USAR because the procedure will only be used afler the event has commenced. Implementation of this j
revised ESP does not create a possibility for an accident or malfunction of a different type than previously evaluated in the USAR because the revised ESP does not modify the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
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Procedure Chance Notice (PCN) 5.8.22 (Revision 4)
TITLE:
Emergency Operating Procedure (EOP) - Primary Containment (PC) Venting and liydrogen Control (Greater Than Combustible Limits)
DESCRIPTION: This revision to EOP Support Procedure (ESP) 5.8.22 made the following changes: 1) added steps for resetting the Group VI isolation logic due to installation of DC 95-037A; 2) provided clarification to direct monitoring of ERP effluent radiation versus monitoring only the ERP effluent radiation monitors; 3) deleted steps to close PC-MO-232, PC-AO-238, PC-MO-233, and PC-AO-237 because these steps were determined to be unnecessary; 4) revised step to provide instructions to close OG-AO-254 via the Off Gas Timer control switch versus the valve control switch; and 5) revised sequence of steps so that PC-MO-305 and PC-MO-306 are closed first and then the override switches are placed to normal.
SAFETY AN.kYSIS:
CNS EOPs accommodate strategies for events beyond the licensed design basis of the plant. The NRC, in its Safety Evaluation Report (SER) on the Boiling Water Reactor Owners Group (BWROG)
Emergency Procedure Guidelines (EPGs) found the use of the limits specified in the EPG and EOPs, rather than those specified in the licensed design basis, acceptable during degraded conditions. The implementation of this revised ESP does not increase the probability of occurreuce or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR.
This revised ESP cannot increase the probability of occurrence of any event analyzed in the USAR because the procedure will only be used after the event has commenced. Implementation of this revised ESP does not create a possibility for an accident or malfunction of a different type than previously evaluated in the USAR because the revised ESP does not modify the operation or design basis of the plant. For plant conditions which already exceed the licensing design basis, the question of a reduced margin of safety is not meaningful.
Procedure Chance Notice (PCN) 6.PC.312 (Revision 0)
TITLE:
PC-243AV/244AV Accumulator Functional Test DESCRIPTION: This is a new procedure to provide a means of testing Reactor Building to Torus vacuum breaker accumulators PC-ACC-243AV and PC-ACC-244AV. The test will determine if the subject accumulators will retain sufficient pressure to actuate PC-AOV-243AV and PC-AOV-244AV one hour after a loss of instrument air. This test will also determine if the valves will cycle at an established minimum pressure.
SAFETY ANALYSIS:
This procedure will not result in any permanent modifications to the plant and will not affect long-term plant configuration. Restoration of temporary modifications will be verified by post-procedure testing. This procedure will only be run during cold shutdown when Primary Containment is not required. Periodic testing will improve or enhance the reliability of the tested valves. This procedure does not add or alter any equipment capable of creating a different type of accident or malfunction of equipment, nor does it detrimentally alter any existing equipment's function, operating parameters, service conditions, or accident modes.
Procedure Chance Notice (PCN) 6.PC.509 (Revision 1)
TITLE:
Core Spray (CS) Local Leak Rate Test DESCRIPTION: This PCN deleted Local Leak Rate Testing for certain CS va!ces (CS-MO5A/B, CS-MO26A/B, CS-MO7A/B) which are on lines which terminate below minimum torus water level and are considered water sealed. Water sealed valves are not subject to 10CFR50, Appendix J. Therefore, these valves were deleted from this procedure.
58 l
f i
SAFETY ANALYSIS:
The affected valves, which are containment iseution barriers, are located on lines which are water scaled. They are the single containment isolation burier on lines which penetrate the torus below the minimum water level. The water seal provides the second containment isolation barrier, as allowed by the USAR. The leak tightness of the torus, and thus the water seal, is demonstrated on a continuous basis since the torus water level and the torus area is monitored at least once per shift. This change did not involve any physical modification to the plant nor change the safety classification of any system or component. Removing these valves has no detrimental effect on the Primary Containment system and does not create an unreviewed safety question.
Procedure Chance Notice (PCN) 6.RPS 301 (Revision 0)
TITLE:
Mode Switch in Shutdown, SDV Valve Timing, and Manual Scram Functional Test DESCRIPTION: This PCN provides for timing Agastat time delay relays RPS-REL-Kl7A and RPS-REL-K17B separately. This procedure change requires a temporary change to the facility by temporarily liftmg and relanding leads in Panels 9-15 and 9-17.
SAFETY ANALYSIS:
This Surveillance Procedure is performed with the reactor shutdown. All control rods are full-in and the change only separates the testing of the Mode Switch Shutdown position scram into its two divisions. The Shutdown position scram is not considered a protective function because it is not j
required to protect the fuel or nuclear system process barrier. The temporarily lifted leads will be j
independently verified when relanded. The leads being lifted provide annunciation only. This PCN does not increase the probability of occurrence or consequences of an accident or malfunction of equipment previously evaluated in the USAR. Splitting of the testing by lifting annunciator leads does not disable the Mode Switch in Shutdown scram.
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Procedure Chance Notice (PCN) 6.1RPS 309 (Revision 1)
Procedure Chance Notice (PCN) 6.2RPS.309 (Revision 11
'9 Reactor Protection System (RPS) Channel Test Switch Functional Test - Division I (6.l RPS.309)
Reactor Protection System (RPS) Channel Test Switch Functional Test - Division 2 (6.2RPS.309)
DESCRIPTION: These PCNs added steps to test the 3-4 contacts on RPS-REL-K14A, IL C, D, E, F, G, and 11. The 3-4 contacts on these relays input into the backup scram valve logic and are in parallel with each other. The backup scram function is disabled during this testing. This change reflects testing performed by Special Procedure 94-208D.
SAFETY ANALYSIS:
The backup scram valves are normally deenergized and energize to vent the scram air header. Pulling the identified fuses does not change the normal state of the valves. Tha irerequisites of the procedure require the reactor to be in cold shutdown,inode switch in REFUEL, c. shutdown margin to be met.
j With the specified plant functions, resctivity is controlled such that backup scran functions are not i
needed during the test. The potential for a single operator error due to leaving the backup scram function defeated is eliminated since the fuses are independently verified to be installed by a second i
Operator. The automatic scram functions are unafTected by the performance of this procedure. There is no credit taken for the backup scram function in any Technical Specification basis.
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r Procedure Chance Notice (PCN) 6.1 APRM.302 (Revision 1 Procedure Chance Notice (PCN) 6.2APRM.302 (Revision 1)
Procedure Chance Notice (PCN) 6.1 APRM.304 (Revision 1)
Procedure Chance Notice (PCN) 6.2 APRM.304 (Revision 1)
TITLE:
APRM System 15% liigh Flux and Inop Trip Functional Test - Division 1 (6.l APRM.302)
APRM System 15% liigh Flux and inop Trip Functional Test - Division 2 (6.2APRM.302) l APRM System (Flow Bias and Startup) Functional Test (Mode Switch Not in Run)- Division 1 (6.l APRM.304) j APRM System (Flow Bias and Startup) Functional Test (Mode Switch Not in Run)- Division 2 (6.2APRM.304) 1 DESCRIPTION: nese procedures were revised to enhance the ability to test the APRM instrument channels during conditions when the SRMs are downscale due to refueling conditions. With SRMs downscale, rod block and alarm functions of the particular SRM channel being tested are masked by the downscale function of the APRMs. These procedure changes installjumpers to defeat the downscale trip function of the SRMs.
SAFETY ANALYSIS:
This change does not affect the ability of the Neutron Monitoring Instruments or the Reactor Protection System to perform their safety and design functions. His portion of the subject procedures will only be run during times when the reactor is shutdown and fuel is removed from the vessel. A single operator errcr which would cause the downscale trip functions of the SRMs to bejumpered out would not affect the high flux and rate of power increase functions of the APRMs or the Reactor Protection System. Therefore, these changes will not increase the probability of occurrence or i
consequences of an accident or malfunction of equipment previously evaluated in the USAR. With no fuel in the vessel, the Neutron Monitoring System does not perform a safety function for the protection of the fuel. The system will be retumed to its normal configuration prior to the loading of fuel into the reactor vessel. Failure of the Neutron Monitoring System and associated protective functions has been evaluated in the USAR, and the activities implemented by these procedures do not create the possibility of a different type of equipment malfunction than any previously evaluated in the USAR. His change does not reduce the margin of safety as defined in the basis for any Technical
)
Specification.
Procedure Chance Notice (PCN) 611RM.301 (Revision 1)
Procedure Channe Notice (PCN) 6.21RM 301 (Revision 1)
TITLE:
IRM Functional Test (Mode Switch Not in Run)- Division 1 (6.llRM.301)
IRM Functional Test (Mode Switch Not in Run). Division 2 (6.2iRM.301)
DESCRIPTION: These procedures were revised to enhance the ability to test the IRM instrument channels during conditions when the SRMs are downscale due to refueling conditions. With SRMs downscale, rod block and alarm functions of the particular SRM channel being tested are masked by the downscale function of the IRMs. nese procedure changes installjumpers to defeat the downscale trip functions of the SRMs.
SAFETY ANALYSIS:
This change does not affect the ability of the Neutron Monitoring Instruments or the Reactor Protection System to perform their safety and design functions. This portion of the subject procedures will only be run during times when the reactor is shutdown and fuel is removed from the vessel. A single operator error which would cause the downscale trip functions of the SRMs to be jumpered out would not affect the high flux and rate of power increase functions of the IRMs or the Reactor Protectica System. Herefore, these changes will not increase the probability of occurrence or consequences of an accident or malfunction of equipment previously evaluated in the USAR. With no fuel in the vessel, the Neutron Monitoring System does not perform a safety function for the protection of the fuel. The system will be retumed to its normal configuration prior to the loading of 60 l
I
1 fuel into the reactor vessel. Failure of the Neutron Monitoring System and associated protective functions has been evaluated in the USAR, and the activities implemented by these procedures do not create the possibility of a different type of equipment malfunction than any previously evaluated in the USAR. His change does not reduce the margin of safety as defined in the basis for any Technical Specification.
Procedure Chance Notice (PCN) 6. I SRM.303 (Revision 1)
Procedure Chance Notice (PCN) 6.2SRM 303 (Revision 1)
)
I TITLE:
SRM Fur.ctiona!(Reactor Not in Run with Shorting Link Switches Open)- Division 1 (6.lSRM.303) i SRM Functional (Reactor Not in Run with Shorting Link Switches Open)- Division 2 (6.2SRM.303)
DESCRIPTION: These procedures were revised to enhance the ability to test individual SRM instrument channels i
during conditions when the SRMs are downscale due to refueling conditions. With SRMs downscale, rod block and alarm functions of the particular SRM channel being tested are masked by the downscale function of the other SRMs. These procedure changes install jumpers to defeat the downscale trip functions of the SRMs not being tested.
SAFETY ANALYSIS:
This change only affects the performance of the subject procedures during the refuel mode when there is no fuel in the reactor vessel. During this operational state, the Neutron Monitoring System does not perfonn any function to prevent or mitigate the consequences of an accident, nor perform a safety i
function for the protection of the fuel. Therefore, these changes do not increase the probability of occurrence or consequences of an accident or malfunction of equipment previously evaluated in the USAR. He system will be returned to its normal configuration prior to the loading of fuel into the j
reactor vessel. Failure of the Neutron Monitoring System and associated protective functions has been evaluated in the USAR, and the activities implemented by these procedures do not create the possibility of a difTerent type of equipment malfunction than any previously evaluated in the USAR.
This change does not reduce the number of Neutron Monitoring channels available, and does not affect the ability of the system to perform its design and safety functions. Therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.
l Procedure Chance Notice (PCN) 6 LOG.601 (Revision 4)
TITLE:
Daily Survei: lance Log (Technical Specifications)
DESCRIPTION: nis procedure change affected the jet pump operability program. It deleted the check forjet pump differential pressure to loop average differential pressure as it was determined that this test did not i
comply with Technical Specification requirements and did not provide useful information aboutjet pump operability. It also deleted the non-Technical Specification check ofloop flow / speed vs. speed.
It added a new check of individual jet pump differential pressure to its characteristic performance curve, which was determined to be a reliable and sensitive test of jet pump performance and operability.
SAFETY ANALYSIS:
Jet pumps are not initiators of accidents; jet pump integrity is required for accident mitigation. Jet pump operability is assured by the surveillance required by Technical Specification 4.6.E, which requires that only one of three checks perfonned be satisfactory for operability. No changes are being made to the other two operability checks required by Technical Specifications. The new more reliable check ofjet pump difTerential pressure to its characteristic performance curve also enhances the surveillance program. Therefore, the probability of occurrence or consequences of an accident or malfunction of equipment are not increased; this change actually provides a greater assurance that jet pump failures will bc detected. His change increases the margin of safety in the Technical Specifications by precluding jet pump operability from being based on the satisfactory result of an unreliable test.
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1 Procedure Chance Notice (PCN) 7.0.1.6 (Revision 1)
TITLE:
Maintenance Work Request (MWR)- Minor Maimenance i
DESCRIPTION: This procedure revision changed / clarified the conditions which must be met to allow Minor Maintenance. It also added criteria for allowing Minor Maintenance on Essential equipment.
SAFETY ANALYSIS:
Minor Maintenance i.; contro!!ed within the scope of the MWR process. Minor Maintenance is limited by definition to activities which do not have the potential to affect nuclear safety. Therefore, 1
Minor Maintenance on Essential equipment rr mains within the licensing bases described in the Technical Specifications and USAR.
Procedure Chance Notice (PCN) 7.0.7 (Revision 01 TITLE:
Scaffolding Construction and Control DESCRIPTION: This new procedure was written to provide for the construction and control of scaffolding. It contains provisions for pre-walkdowns and post-walkdowns ofscaffolding installations. This procedure would allow scaffolding to be attached to an operable component described in the USAR.
SAFETY ANALYSIS:
his procedure ensures that applicable code criteria will remain satisfied for plant components which provide support to scaffolding and that scaffolding cannot affect the USAR described safety design function of s:ructures, systems, or components relied upon to mitigate the consequences of an accident. Any potential increase in the probability of a malfunction due to scaffolding i
erection / disassembly /use is offset by the reduction in malfunction probability resulting from:
i
- 1) performing maintenance / inspection activities with ladders or other less stable arrangements and i
- 2) not performing the maintenance activity and thereby increasing the potential for equipment malfunction. This procedure ensures that scaffolding will not interfere with personnel access to equipment nor with actions taken to mitigate the consequences of equipment malfunction. Adherence to this procedure ensures that Technical Specification requirements for affected equipment are not violated. Ei.gineering evaluation of plant equipment will ensure that applicable code criteria will continue to be met and thereby not reduce an affected component's margin of safety.
i Procedure Chance Notice (PCN) 7.013 (Revision 0)
TITLE:
Control of Insulation Removal and Installation DESCRIPTION: This new procedure was written to control removal and reinstallation of insulation for piping, equipment, and HVAC. It only provides general instruction to reinstall insulation as this is a skill-of-the-craft activity. The Topical Reference Information Manual (TRIM) 001 document specifies the type ofinsulation to be installed. If a different type ofinsulation than is specified in the TRIM-001 document is to be installed, an Engineering approved document is required.
SAFETY ANALYSIS:
Thermal insulation is a passive plant system. Seismic, radiological tolerance, and chemical reactivity requirements for thermal insulation are not changed. TRIM-001, Appendix H, Safety Evaluation for Thermal Insulation, addresses the effect of adding or removing insulation. This procedure requires Engineering evaluation for removal ofinsulation on operable systems to ensure equipment important to safety is maintained in an operable status. It requires Engineering evaluation to ensure that the environmental effect and the effect on an operable system does not render systems that are important to safety in an unanalyzed condition. The increase in area temperatures during normal, off-normal, and emergency conditions is bounded by the analyses contained in the Equipment Qualification Data Package for Environmental Conditions. The Technical Specification Bases do not specifically address insulation and the implications ofinsulation removal.
62 1
Procedure Chance Notice (PCN) 7.2.57.1 (Revision 01 TITLE:
Pipe Support Removal and Reinstallation DESCRIPTION: This new procedure was written to provide control of the removal and reinstallation of pipe hangers / supports. It requires an Engineering evaluation to ensure support removal will not affect system operability.
SAFETY ANALYSIS:
This procedure ensures that applicable code criteria will remain satisfied for plant components when supports require removal for maintenance activities. Procedure instructions will ensure that no USAR analyzed accident could be initiated and the USAR described safety design function of structures, systems, or components relied upon to mitigate the consequences of an accident will not be affected by support remoul. Administrative requirements remain in pisce during this activity such that system availability and operability are controlled consistent witi.'SAR assumptions. Maintenance controls are consistent with Technical Specification and USAR provisions. Adherence to this procedure ensures that Technical Specification requirements for affected equipment are not violated.
Engineering evaluation of plant equipment will ensure that applicable code criteria will continue to be met and thereby not reduce the margin of safety of affected equipment.
Procedure Chance Notices (PCNs) 9 1.1.1 (Revision 161 9.1.1.1.1 (Revision 31 9.1.1.2 (Revision 51 9.1.1.3 (Revision 3619.1.1.4 (Revision 2319.1.1.5 (Revision 1319.1.2.2 (Revision 1719.1.2.4 (Revision 91 9.1.3 (Revision 201 9.1.4 (Revision 141 9.1.5 (Revision 301 9.1.5.1 (Revision 41 9.1.5.2 (Revision 41 9.1.7 (Revision I11 9.1.7.1 (Revision 319.1.7.2 (Revision 31 o.2.3 (Revision 1319.3.1.4.2 (Revision 101 9.3.2.3 (Revision 31 9.3.4.2 (Revision 31 9.3.4.4 (Revision 11 9.5.1.1 (Revision 21 9.5.2 (Revision t il 9.5.3.3 (Revision 1319.5 3.4 (Revision 12L 9.5.3.6 (Revision 119.5.3.7 (Revision 119.5.3.8 (Revision 11 9.5.3.9 (Revision 119.5.3.11 (Revision 119.5.3.12 (Revision 119.8.1 (Revision 619.8.3 (Revision 6)
TITLE:
Radiation Protection at CNS (9.1.1.1)
Contract Senior Radiological Protection Technician Qualification (9.1.1.1.1)
CNS 110t Particle Program (9.1.1.2)
Personnel Dosimeter Program (9.1.1.3)
Special Work Permit (9.1.1.4)
Radiography (9.1.1.5)
Area Posting - Radiological (9.1.2.2)
Access Control - Radiological (9.1.2.4)
Radiation Safety Standards and Limits (9.1.3)
Protective Clothing (Anti-C) (9.1.4) l Radiological Respiratory Protection Program (9.1.5)
Self-Contained Breathing Apparatus (9.1.5.1) f Respirator Fit Test Program (9.1.5.2) i CNS Laundry Operations (9.1.7) i Ludlum Model 329-1 Laundry Monitor (9.1.7.1) d Eberline Model ACM-300 Laundry Monitor (9.1.7.2)
Contamination Surveys (9.2.3) lon Chamber Survey Instrument Eberline Models RO-2, RO-2A, and RO-20 (9.3.1.4.2)
Tennelec LB-4100 Drawer Smear Counter (9.3.2.3)
SAIC Model PD-1 Electronic Dosimeter (9.3.4.2) l SAIC Model PD-4/PDE-4 Electronic Dosimetry System (9.3.4.4) l Survey of New Nuclear Fuel Shipments (9.5.1.1)
Radioactive Sources Control and Accountability (9.5.2)
Condensate Cleanup Waste Resins, Spent Resins, and Waste Sludge Classification and Listing (9.5.3.3)
RWCU Waste Resins Classification and Listing (9.5.3.4) 63 l
l
Control of On-site Storage of RWCU and Condensate Resins and Wastes - Transfer into Storage (9.5.3.6)
Completion of Chem-Nuclear's Waste Manifest Instructions (9.5.3.7)
Control of On-site Storage of RWCU and Condensate Resins and Wastes - Transfer Out of Storage (9.5.3.8)
Control of On-site Dry Active Waste Storage (9.5.3.9)
Inspection of On-site LLRW Stcrage (9.5.3.1 I)
Filling Containers with Waste / Radioactive Material (9.5.3.12)
Method and index of Records and Data Forms (9.8.1)
Instrument llistory Records (9.8.3)
DESCRIPTION: The above procedures were revised to reflect a change in terminology by the Radiological Department. He term " Health Physics" was changed to " Radiological Protection" and the term "HP" was changed to "RP" throughout the procedures. In addition, title changes due to the reorganization of the Radiological Department were incorporated.
SAFETY ANALYSIS:
Although the title changes represent an alternative method of identifyiag the Health Physics organization as discussed in the USAR, the title changes are an administrative change only and do not reflect a change in intent. These title changes have no impact on the probability or consequences of an accident or malfunction of equipment important to safety. There are no formal Technical Specification bases for the subject titles.
Procedure Chance Notice (PCN) 9.5.3.6 (Revision 01 Procedure Chance Notice (PCN) 9.5.3.8 (Revision 01 Procedure Chance Notice (PCN) 9.5.3.0 (Revision 01 Procedure Chance Notice (PCN) 9.5.3.11 (Revision 01 TITLE:
Control of On-Site Storage of RWCU and Condensate Resins and Wastes - Transfer into Storage (9.5.3.6)
Control of On-Site Storage of RWCU and Condensate Resins and Wastes - Transfer Out of Storage (9.5.3.8)
Control of On-Site Dry Active Waste Storage (9.5.3.9)
Inspection of On-Site LLRW Storage (9.5.3.11)
DESCRIPTION: Procedures 9.5.3.6 and 9.5.3.8 provide instructions for the placement and removal of Reactor Water Cleanup (RWCU) and Condensate Resins / Wastes, as well as other forms of Low Level Radioactive Waste (LLRW), on and from the LLRW Storage Facility installed by DC 91-077.
Procedure 9.5.3.9 provides instructions for the placement of Dry Active Waste containers in the Multi-Purpose Facility (MPF) Storage Area created by DC 91-077. Procedure 9.5.3.11 provides instructions for the quarterly monitoring of the RWCU and Condensate Resins / Wastes ad other forms of LLRW stored at the LLRW Storage Facility and of Dry Active Waste containers stored in the MPF.
SAFETY ANALYSIS:
The storage of LLRW inside and outside of the plant does not alTect the probability of occurrence or consequences of an accident or malfunction of equipment previously evaluated in the USAR, nor create the possibility of a different type of accident or malfunction of equipment important to safety. Equipment located within the plant has not been modified by DC 91-077. A ten-inch shield wall was installed in the MPF; however, no equipment important to safety is located in the MPF. Removal of LLRW from the storage pad, storage of Dry Active Waste in the MPF, and monitoring of LLRW storage will also not affect the probability of occurrence or consequences of an accident or malfunction of equipment. Postulated failure modes do not impact the operation, safety, or reliability of the plant.
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Procedure Chance Notice (PCN) 10.22 (Revision 7)
Procedure Chance Notice (PCN) 10.23 (Revision 9)
TITLE:
Receiving New Fuel (10.22)
New Fuel Inspection, Channeling and Control Blade inspection (10.23)
DESCRIPTION: These procedure changes allow temporary storage of new fuel in metal shipping containers on the refueling floor, versus storing new fuel in a dry storage vault adjacent to the spent fuel pool area in the Reactor Building.
SAFETY ANALYSIS:
The temporary storage of new fuel in metal shipping boxes on 1001' level of the Reactor Building does not increase the probability of occurrence or consequences of an accident or malfunction of
{
equipment. The floor is capable of supporting the fuel in the unlikely event of an earthquake.
Should the stack of fuel shift during a seismic event, the floor loading will be less severe than for j
previously analyzed heavy load drops. Even if the stacked boxes were to shift during the unlikely event of an earthquake, there is no equipment important to safety in proximity to the storage area.
The new fuel in the metal shipping box is not likely to be damaged should the box fall from the stack. However, since the new fuel is not irradiated and has negligible gap activity, the i
consequences of fuel damage would be less severe than for irradiated fuel. This temporary i
storage of new fuel causes no interaction with any equipment important to safety. There are no i
Technical Specifications concerning this temporary storage.
Procedure Chance Notice (PCN) 14.15 3 (Revision 8)
TITLE:
Reactor Vessel Open llead Level Monitor System l
DESCRIPTION: This PCN added sections to allow using the upper pressure sensor of the reactor core support plate differential pressure instrument to monitor water level in the vessel during refueling. This will provide an additional indication of water level to the operator in the Control Room.
SAFETY ANALYSIS:
The proposed activity will temporarily disable the reactor core support plate ditTerential pressure instrument. This instrument is not required when the reactor is in cold shutdown. The normal function of this instrument will be restored prior to startup. The core support plate differential pressure instrument is not an accident initiator and is not used to mitigate the consequences of an accident or malfunction of equipment important to safety. It is used as an independent verification of core flow and does not provide any margin of safety in any Technical Specifications.
65
I i
OTIIER REPORTABLE ACTIVITIES 1
Setnoint Chance Reauest 94-26 TITLE:
Reactor Recirculation Pump A/B Low Suction Pressure to RilR Interlock DISCUSSION: This Setpoint Change Request was generated to document the setpoints for RR-PS-128A/B as l
determined in Nuclear Engineering Department Calculation (NEDC) 92-050AN. The purpose of these switches is to provide an interlock to prevent shutdown cooling flow unless the reactor pressure vessel is below 75 psig. The setpoint margin was increased and the elevation correction decreased, both of which provide a conservative change to the previously existing setpoint.
SAFETY ANALYSIS:
This Setpoint Change does not alter the operation of the Reactor Recirculation pump low suction pressure to Residual IIcat Removal interlock function in a nonconservative way, nor result in any physical changes to the instruments. No new failure effects or failure modes are introduced. Analysis in NEDC 92-050AN shows the setpoint is adequately conservative to ensure the Technical l
Specification limits will not be exceeded. The result of the calculation changes the actual setpoint in the conservative direction; therefore, the margin of safety as defined in the basis of any Technical Specification is not reduced.
Setnoint Chance Reauest 94-27 TITLE:
RCIC-FIS-57; Pump Discharge Low Flow Alarm and Minimum Flow Control l
DESCRIPTION: This Setpoint Change Request was generated to document the setpoint for RCIC-FIS-57 as determined by Nuclear Engineering Department Calculation (NEDC) 92-050AS. RCIC-FIS-57 senses RCIC system flow. This instrument has two switches with independent setpoints. Switch #1 provides a signal to open the minimum flow bypass valve on decreasing flow. Switch #2 provides a signal to close the minimum flow bypass valve on increasing flow. NEDC 92-050AS calculated a new, more conservative instrument setpoint for Switch #1.
SAFETY ANALYSIS:
'Ihis Setpoint Change does not significantly alter the function or operation of the RCIC system or the minimum flow bypass function. No new failure effects or failure modes are introduced. During decreasing flow conditions, the new setpoint will result in the minimum flow bypass valve opening earlier and diverting flow through the minimum flow line and away from the vessel. Analysis performed in NEDC 92-050AS shows the setpoint is adequately conservative to ensure Technical Specification limits will not be exceeded. The calculation results in a setpoint which is more conservative with respect to the Technical Specification limit; therefore, the margin of safety as defined in the basis of any Technical Specification is not reduced.
Setnoint Chance Reauest 94-28 TITLE:
Core Spray and Residual lleat Removal Pump Discharge Pressure Automatic Depressurization System (ADS) Interlock DESCRIPTION: This Setpoint Change Request was generated to document the setpoints for CS-PS-37A/B, CS-PS-44A/B, RiiR-PS-105A/B/C/D, and RilR-PS-120A/B/C/D as determined in Nuclear Engineering Department Calculation (NEDC) 92-050AK and NEDC 92-050AL. These switches sense Core Spray and Residuallleat Removal discharge pressure. Upon sensing that the discharge pressure has reached a predetermined setpoint, these switches provide an interlock with the ADS system to enable ADS operation. The trip setpoint was increased by 5 psig to ensure instruments reset within the Technical Specification limit.
66
SAFETY ANALYSIS:
This Setpoint Change does not significantly alter the function or operation of the Core Spray and Residual lleat Removal discharge pressure ADS permissive. No new failure effects or failure modes are introduced. This change results in a new setpoint which is within the Technical Specification limits. During pump start, the new setpoint will result in the ADS permissive occurring slightly later than with the existing setpoint; however, the change is small ($ psig). Analysis performed in NEDC 92-050AK and NEDC 92-050AL shows the setpoints are adequately conservative to ensure the Technical Specification limits will not be exceeded; therefore, the margin of safety as defined in the basis of any Technical Specification is not reduced.
Maintenance Work Reauests (MWRs) 04-3537 and 94-3801 TITLE:
Reference Leg injection Solenoid Valves DESCRIPTION: Special Instructions were developed for these MWRs to isolate valves NBI-SOV-SSV738 and NBI-SOV-SSV739 from their respectia loops of Core Spray, NBl instrumentation, and reference leg to check for leakage. The valves were isolated by two manual valves and leakage was measured from a drain located between the manual isolation valves. De leakage rate for both vah es was acceptable.
SAFETY ANALYSIS:
This activity was performed with the reactor in cold shutdown; therefore, reference leg injection capabilities were not required. De reference leg serves no accident mitigation function. The pressure boundary function of the injection line remained intact, the injection function was not required during cold shutdown, and reactor coolant level indication was provided by alternate means. The probability of occurrence or consequences of an accident or equipment malfunction were unaffected by this activity. Safety margins defined in the Technical Specifications were not reduced.
Safety Evaluation for Thermal Insulation TITLE:
Safety Evaluation for Thermal Insulation at Cooper Nuclear Station DESCRIPTION: This Safety Evaluation was performed to determine the effect on safety due to changes made to the thermal insulation configuration at CNS.
SAFETY ANALYSIS:
The probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the USAR was not increased by changes to plant thermal insulation.
Dermal insulation is a passive plant system. Seismic, radiological tolerance and chemical reactivity requirements for thermal insulation are not changed. Area heat loads are maintained within calculated or specified limits. He hazards from pipe break have been bounded by the pipe break analysis.
Documentation and evaluation of area heat load when insulation is found to be different from specification provides assurance that the HV system can maintain area temperatures within specified normal, abnormal and accident conditions. The possibility of a different type of malfunction of equipment important to safety is not increased by the subject changes to plant thennal insulation. The margin of safety as defined in the basis of any Technical Specification is not reduced.
Revision to Fire Hazards Analysis TITLE:
Evaluation of Combustible Loading Revisions Incorporated into the Fire Hazards Analysis DESCRIPTION: His evaluation analyzed the impact of combustible load changes which were incorporated into the 1990,1992, and 1994 revisions to the CNS Fire Hazards Analysis. Since the preparation of the 1989 Fire llazards Analysis, several changes have occurred that impact on the calculated combustible loading values identified for several fire zones at CNS. These combustible loading changes resulted from either plant design changes, room reconfiguration, changes to transient combustible storage controls, or other plant walkdown and combustible loading calculation verification reviews. Based 67
on Fire Protection Engineering Evaluation 941 Rev. O, the changes in combustible loadings and equivalent fire severities do not adversely impact the technical bases for associated fire barrier ratings, exemption requests, or Generic Letter 86-10 evaluations.
SAFETY ANALYSIS:
The probability of fire occurrence is not affected by changes in combustible loading. Ignition frequency data is unchanged. The changes in combustible loading do not exceed the capabilities of existing fire protection features except as previously identified and accepted by the NRC. The Safe Shutdown Analysis assumes loss of all equipment within a fire area; the probability of this loss is i
unchanged by changes to combustible loading. No new type of failure is introduced by these changes.
Combustible loading does not form the basis for any Technical Specification.
Revision to Fire 11azards Analysis i
TITLE:
Evaluation of Maintaining Primary Containment Barriers During Plant Outages and Clarification of Appendix A Fire Barriers DESCRIPTION: These evaluations were performed to analyze the deletion of specific Appendix A fire barriers from the 1992 CNS Fire llazards Analysis. The specific barriers deleted were; l) Primary Containment (when the reactor is shutdown and containment breached),2) elevator shaft walls, and 3) the Office Building south stairway. Fire Protection Engineering Evaluations 94-5 Rev. I and 90-4 Rev. 2 4
provide the basis and justification for deleting these barriers from the fire barrier control program.
SAFETY ANALYSIS:
The probability of occurrence of a fire is not increased by deleting the referenced fire barriers from the CNS Fire Bar ier Control Program. The consequences of a fire are not increased because the safe shutdown capability of CNS is maintained independent of the referenced barriers. No physical plant changes are introduced. CNS conformance to Appendix A to Branch Technical Position (BTP) 9.5-1 is maintained. CNS has provided adequate fire protection capability in the Reactor Building to combat drywell fires during shutdown and refueling periods. In addition, CNS Pre-Fire Plans and Brigade drills are practiced to ensure manual fire fighting access and egress capability.
Reclassification of barriers will not result in any new accidents. The Fire Barrier Penetration Fire Seals Technical Specification does not define a margin of safety. NRC regulations and guidelines in both 10CFR50 Appendix R and Appendix A to BTP 9.5-1 are satisfied and not adversely impacted by these f're barrier revisions.
Revision to Fire llazards Analysis TITLE:
Fire Door Evaluation - Fire Door 11102 DESCRIPTION: Fire Protection Engineering Evaluation 85-4 Rev. I was performed to document the acceptability of the fire rating assigned to Door 11102, which forms a part of an Appendix R fire barrier. This door is a 1 % hour fire door that separates the Control Building corridor on 903' from RPS Room 1 B. The fire rating of the subject door was determined to be acceptable due to the large margin of safety that the door provides (1000%).
SAFETY l
ANALYSIS:
The probability of occurrence or consequences of a fire are unchanged by the results of this analysis.
The barrier rating is still adequate to mitigate the hazards associated with the area and prevent the spread of fire. Fire barrier rating and combustible loading do not define the margin of safety for any Technical Specification basis.
68
Revision to Fire llazards Analysis TITLE:
Enhancement of Appendix R Exemption Discussions DESCRIPTION. 2.ppendix B to the CNS Fire Hazards Analysis provides a summary of all 10CFR50 Appendix R exemptions. The introduction information associated with the individual exemptions required updating to reflect current plant configurations.
SAFETY ANALYSIS:
The probability of occurrence or consequences of an accident or equipment malfunction are not affected by this revision to descriptive information in the Fila exemption request summaries.
Existing consequences of fire are bounded in the safe shutdown analysis. Changes to FHA descriptive information have no impact on results or conclusions. This information is not part of any Technical Specification basis; therefore, the margin of safety defined in the Technical Specifications is not reduced.
Revision to Fire Hazards Anahsig TITLE:
10CFR50 Appendix R Emergency Battery Lighting Unit Adequacy DESCRIPTION: Fire Protection Engineering Evaluation 95-01 Rev. 0 was performed to assess the adequacy of the current configuration of emergency battery lighting units relied on for 10CFR50 Appendix R compliance. It addresses the actions taken to resolve identified issues and interim measures to ensure that the plant is safe and meets the safety intent of Appendix R.
SAFETY ANALYSIS:
This evaluation concludes that the safety intent of 10CFR50 Appendix R is maintained. The probability of occurrence or consequences of a fire are not increased. Plant modification and equipment upgrades have been performed to improve emergency battery lighting (EBL) unit operation in critical plant areas. Interim measures are provided to ensure that potentially unreliable g
EBL units do not hinder safe shutdown actions. The probability of EBL malfimetion is reduced due to the installation of new/ upgraded EBL units in critical plant areas. The margin of safety for performing critical safe and alternative shutdown actions is improved via EBL equipment upgrades.
I icense Chance Reauest (LCR) 94-0056. LCR 95-0073 TITLE:
USAR Change - Containment Boundary DESCRIPTION: These LCRs support moving the outer containment boundary for penetrations X-13 A/B from valves RHR-MO27A/B to valves RHR-MO25A/B. This requires check valves RIIR-26CV/27CV and bypass valves RHR-MO274A/B to function as the containment isolation valves for these penetrations.
RIIR-MO274 A/B are deenergized above 212 degrees Fahrenheit.
SAFETY ANALYSIS:
The new containment isolation valves (CIVs) meet all of the requirements for containment isolation.
There are no changes to the containment leakage requirements since the new CIVs are incorporated in the Appendix J requirements. Therefore, the probability of occurrence or consequences of an accident or equipment malfunction are not increased. No physical changes were made to the plant as a result of this change. Related procedure changes ensure that there are no safety concerns as a result of this change in containment isolation. By removing valves RHR 274 A/B from power above 212 degrees Fahrenheit, the potential for maloperation of these valves has been eliminated. There is no reduction in the margin of safety as all of the containment isolation requirements in the Technical Specifications are fulfilled by the new arrangement.
69
l l
License Chance Reauest (I CR) 94-0084 TITLE:
USAR Change - Core Spray Discharge Header Pressure Switches DESCRIPTION: This LCR changed the instrument range for the Core Spray Discharge lleader Pressure Switches (CS-PS-37A/B and CS-PS-44A/B). These pressure switches provide an Automatic Depressurization System (ADS) permissive when a Core Spray (CS) pump is running and were replaced under ESC 86-049. The pressure switch range previously provided in USAR Table Vil-4-3 was in error.
SAFETY ANALYSIS:
This change corrects an error pertaining to the instrument range for the CS discharge header pressure
)
switches. It does not change the function, operation, or reliability of the switches. Therefore, the
{
probability or radiological consequences of an accident or equipment malfunction are not increased.
The switches will be able to perfcrm their safety-related functions. This change does not affect the setting limits given in the Technical Specifications; therefore, the margin of safety is not reduced.
Ljcense Chance Reauest (1 CR) 44-0121 TITLE:
USAR Change - Deletion of Tables VI-7-1 and Vll-4-5 DESCRIPTION: This LCR deletes Table VI-7-1, Core Standby Cooling Systems Operation Requirements, and Table VII-4-5, Core Standby Cooling Systems (Logic) Operational Requirements, and all associated references from the USAR. The subject tables are based on analyses performed at the time of original I
license submittal. Current infonnation is contained in the Technical Specifications. Since the subject tables are not current, there is a concern that users may be using this information unaware that the current LCO and surveillance requirements are contained in the Technical Specifications. The FSAR contains the subject tables from a historical perspective. If updated to reflect current requirements, the tables would be redundant to the Technical Specifications; therefore, they are being removed from the USAR.
SAFETY ANALYSIS:
As discussed above, the information contained in the subject tables is not current. Consequently, removal will have no impact on the probability of occurrence or consequences of an accident or malfunction of equipment important to safety and will have no impact on the margin of safety for any Technical Specification.
License Chance Reauest 94-0125 TITLE:
USAR Change - Scram Discharge Valve Leakage Testing DESCRIPTION: This LCR revised the USAR to better reflect the current method of detecting leakage past the 11ydraulic Control Unit (liCU) scram discharge valves. Outlet scram valve leakage can be detected by elevated scram discharge riser temperatures for each IICU.
SAFETY ANALYSIS:
Testing the scram riser for elevated temperatures in the manner now being described does not require any system configurational changes. It decreases the probability of an ATWS since the Scram Discharge Volume vent and drain valves remain open. This change does not increase the probability or consequences of an accident or malunction of equipment and does not impact the Technical Specification margin of safety.
l i
70
License Chance Reauest (LCR) 95-0013 TITLE:
USAR Change - Ilatch Covering on the 1001' Level of the Reactor Building DESCRIPTION: A Safety Evaluation was performed to determine whether the hatch and plastic covering over the Reactor Building Crane Access liatch constitutes an unreviewed safety question. This evaluation was prompted by the generation of Condition Report 94-0345 USAR Figures XII-2-6 and V-3-1 were revised to show the hatch covering.
SAFETY ANALYSIS:
The design of the hatch and its cover and the materials used in its construction do not adversely affect any systems or components in such a way to increase the probability of an accident previously evaluated in the USAR. The hatch and its covering do not dectease the reliability or alter the operating characteristics of the llV system nor does it affect the ability of other systems and components to perform their intended safety functions. The construction of the hatch cover is such that the SGT system can perform its intended safety function. The hatch would lift due to a differential pressure well below that postulated to occur due to a high energy line break inside the drywell. His ensures that temperature and pressure analyses are unaffected and, therefore, operation of equipment required to mitigate an accident is unaffected. Failure of the ha*ch or cover does not create a new accident or change a previously evaluated accident. The margin ofsafety as defined in the bases for any Technical Specification is not reduced.
License Chance Reauest (LCR) 95-0014 TITLE:
USAR Change - Revision to Description of Service Water (SW) System DESCRIPTION: This LCR revised the USAR general description of the SW system to clarify the number of SW pumps operating during particular operating modes, and the isolation pressure setpoint for the nonessential header.
SAFETY ANALYSIS:
The SW system is designed to perform its required Design Basis Accident (DBA) safety function with one operating SW pump degraded to a minimum allowable performance level. An increase in the number of operating pumps in this condition will only serve to increase the safety margin. Per the Design Criteria Document for the SW and RilRSW system, operation of the SW system during planned operations is acceptable with one to four operating SW pumps. A Nuclear Engineering Department Calculation (NEDC) evaluated the adequacy of the existing setpoints for the SW system non-critical loop isolation on low pressure and found them to be satisfactory for the most limiting design basis condition. His change does not increase the probability of occurrence or consequences of an accident or malfunction of equipment.
License Chance Reauest (LCR) 95-0043 TITLE:
USAR Change - Reactor Coolant Pressure Boundary DESCRIPTION: The USAR was revised to clarify that the redundant seals and restrictive flow areas in the CRDs provide the Reactor Coolant Pressure Boundary isolation and the CRD system hydraulic lines are not part of the Reactor Coolant Pressure Boundary. The USAR was revised to reflect the positions documented in Generic Letter 86-01, BWROG-8420, and NEDO-24342.
71
SAFETY ANALYSIS:
Per General Electric, the CRDs are designed and fabricated to ASME Class 1 and the CRO bydraulic lines to ASME Class 2. This LCR clarified the existing design in the USAR for the CRDs and associated hydraulic lines, but did not change the design requirements. Therefore, the probability of an accident or equipment malfunction was not increased. This LCR did not affbet any accident analysis or radiological releases so the consequences of an accident or equipment malfunction were not increased. This clarification does not conflict with any Technical Specification bases and does not reduce the margin of safety.
License Chance Reauest (LCR) 95-0045 TITLE:
USAR Chaage - Flow Rate for Core Spray Pumps DESCRIPTION: This LCR revised USAR Table VI-3-1 to change the flow rate for the Core Spray Pumps from 4500 gpm to 4720 gpm to be consistent with CNS Technical Specifications, CNS Vendor Manual No. 31, and other sect ocs of the USAR. ne origin of the 4500 gpm in USAR Table VI-3-1 could i
not be determined, but suce it was not supported by any other reference and was in conflict with the Technical Specifications, Vendor Manual, and cther USAR sections, it was changed to be consistent.
SAFETY ANALYSIS:
This revision to the USAR identifies the corrd Core Spray Pump flow rate. Providing this consistency does not increase the probability of aa accident or equipment malfunction. This revision involved no physical change to the Core Spray system and did not affect actual performance or reliability of the Core Spray Pumps or any other equipment. It did not affect the accident analysis which assumes an available Core Spray system flow of 4720 gpm to the reactor vessel. This LCR changed the USAR to agree with Technical Specifications. The equipment requirements specified in the LCOs and the margin of safety as defined in the bases for the Technical Specifications were not affected.
License Chance Reauest (LCR) 95-0046 TITLE:
USAR Change - Closure Capability of Main Steam Isolation Valves (MSIVs)
DESCRIPTION: nis LCR revised USAR Section VII to make it consistent with USAR Section IV and with General Electric (G.E.) Analysis PED-67-1288 to clarify the closure capability of the MSIVs. This G.E.
Analysis concluded that the combination of spring force and locally swed air pressure was adequate to close the MSIVs, but that the springs alone may not be able to clvse the inside MSIVs during a LOCA. Following performance of this G.E. Analysic., USAR Section IV was revised to reflect its conclusions, but Section Vil was not. This USAR reusion corrected this inconsistency.
SAFETY ANALYSIS:
The subject G.E. analysis is specific to the MSIVs and does not affect any USAR accident analysis.
Consistently incorporating the conclusions of this G.E. analysis can not increase the probability of occurrence or consequences of an accident or equipment malfunction. The need to have pneumatic pressure assist the springs in valve closure is assumed as demonstrated by the inclusion of accumulators in the design. Herefore, MSIV response to a DB A (LOCA) remains unchanged. He MSIVs will continue to function as designed, closing on receipt of a MSIV closure signal to prevent a
the release of radiation to the environment. The MSIVs are not the source of any postulated accident.
There is no Technical Specification requirement for spring-only MSIV operation. This USAR revision does not reduce the margin of safety.
72
License Chance Recuest (LCR) 95-0047 TITLE:
USAR Change - Core Flooding Capability of RilRSW Booster System DESCRIPTION: his LCR revised the USAR to clarify that the capability of the RHRSW System to flood the core is not an analyzed safety function, but rather a design feature provided for conservatism for beyond design basis events. Flooding the core using the RilRSW Booster System would only be required after multiple failures of ESF equipment which are well beyond the scope of credible accidents.
SAFETY ANALYSIS:
This LCR provided a USAR clarification only. The capability to flood the core from the RilRSW Booster System is not part of the Safety Design Basis. The existing accident analysis does not take credit for this capability and it is not affected by this clarification. This capability will not be used unless there are multiple failures in both loops of Core Spray and LPCI as well as a failure of the Standby Liquid Control System. Providing this clarification in the USAR does not affect the reliability of these systems. No equipment important to safety relies upon the core flooding capability of the RHRSW Booster System to mitigate its failure consequences. Providing this clarification does j
not affect the operation or reliability of any plant equipment. This capability is not required by the Technical Specifications License Chance Recuest (LCR) 95-0048 TITLE:
USAR Change - Heating, Ventilation and Air Conditioning Systems DESCRIPTION: This LCR revised the USAR to reflect the as-built plant configuration for the following non-essential l
systems: 1) Station Heating System,2) Reactor Building Heating and Ventilating System, and
- 3) Off-Gas Building fleating and Ventilating. His LCR corrected a discrepancy that existed between the original as-configured design and the SAR.
SAFETY ANALYSIS:
nese heating and pressure control system USAR revisions describe non-essential systems which do not affect the function or operation of any system or component important to safety. None of the affected systems can initiate an accident. The Reactor Building pressure control system which was revised is not used to control the building pressure under accident conditions. No new failure modes have been introduced. Technical Specification requirements are not affected and the margin of safety as defined in the basis for any Technical Specification is not reduced.
I icense Chance Recuest (LCR) 95-0057 TITLE:
USAR Change - Deletion of Descriptive Requirement for " Essential Circuit" Cable Splicing DESCRIPTION: This LCR deleted an unnecessary descriptive requirement from the USAR which stated that splicing of " essential circuit" cables would be analyzed on a cese-by-case basis by Engineering and documented via a Work item on Design Change. A new cable splice to a USAR described system / component already requires a 10CFR50.59 safety evaluation. A replacement splice is a maintenance activity already controlled via approved maintenance procedures which are required to address 10CFR50.59 for approval. Herefore, the requirement for a case-by-case analysis by Engineering is not necessary.
SAFETY ANALYSIS:
This change does not remove 10CFR50.59 requirements to analyze whether a new splice (installed via an approved design modification process) or a generic replacement splice procedure could have an adverse effect upon the USAR described facility or reduce the margin of safety as defined in the basis for any Technical Specification. By remaining within the bounds of the applicable splicing 73
1 procedure, the replacement of a splice on any " essential circuit" cable does not require a case-by-case analysis by Engineering to prevent an adverse effect to nuclear safety. Appropriate analysis of an
" essential circuit" cable splice addition is assured by procedural conformance with existing 10CFR50.59 requirements.
Ligrnse Chance Reauest (LCR) 95-0061 TITLE:
USAR Change - Reorganization of Engineering Department DESCRIPTION: This purpose of this LCR was to revise the USAR to incorporate organizational changes in the Engineering Department. The changes reflect the new reporting alignment of the Engineering organization.
i SAFETY ANALYSIS:
ne changes in the organizational reporting alignments do not affect the design or operation of any system, structure or component. The changes are simply in the reporting alignment of the Nuclea-Power Group organization. These changes will continue to provide for qualified and responsible j
individuals in the management positions; personnel qualification requirements are not affected. Rese changes are considered administrative changes that will not affect the performance of the plant to respond to transients or emergencies. All responsibilities described in the Technical Specificat ons 3
will continue to be performed by qualified individuals.
License Chance Reauest (LCR) 95-0063 TITLE:
USAR Change - Containment Oxygen Monitoring DESCRIPTION: This LCR removed the instrumentation accuracy range for containment oxygen monitoring from USAR Sections V and Vil. Instrumentation setpoints are controlled via CNS Procedure 3.26,
" Instrument Setpoint Control", as supplemented by CNS Procedure 3.26.3," Instrument Setpoint and Channel Error Calculation Methodology." Therefore, the instrument accuracy discussion was removed from these USAR sections.
SAFETY ANALYSIS:
Primary containment oxygen is monitored to ensure it is kept below a 4% flammability limit to provide post-acc~ dent combustible gas control; this is a post accident monitoring function. Since this change does not impact any plant equipment, change any mode of plant operation, alter the control of the 112/02 oxygen concentration setpoint, nor affect any analysis relied upon in the mitigation of an accident, it cannot increase the probability or consequences of an accident or malfunction of equipment important to safety. Control of the instrument setpoint and associated accuracy will be maintained in accordance with plant procedures for the setpoint control program. This change does not affect any analysis relied upon to support the bases for any Technical Specification.
l License Chance Reauest (I CR) 95-0064 TITLE:
USAR Change - Manning of Emergency Response Facilitites DESCRIPTION: his LCR revised the USAR to reflect a higher level of emergency respense. De USAR previously stated that the Emergency Operations Facility (EOF) would be manned after the declaration of a SITE AREA EMERGENCY or GENERAL EMERGENCY. In accordance with Revision 30 of the CNS Emergency Plan, the USAR was revised to state that the EOF and Operation Support Center (OSC) will be manned and operational after declaration of an ALERT er higher level incident.
74
SAFETY ANALYSIS:
his change does not increase the probability of an accident; activation of the OSC and EOF occurs after accident initiation. Activation of the OSC and EOF at the ALERT stage versus the previous SITE AREA EMERGENCY classification enhances the ability of the site to mitigate the consequences cf an accident. He 10CFR50.54(q) review of Emergency Plan Revision 30 indicated that the EOF ci ange did not decrease the effectiveness of the plan. He OSC activation is additional information thn was previously in the Emergency Plan but not in the USAR. Operational command and control reniains under the cogni ace of Control Room personnel. The Emergency Plan is not credited either directly or indirectly in the Bases of any Technical Specification.
Qcemg C6nce Reauest (LCR) 95-0068 TITLE:
USAR Change - Primary Containment Isolation Signal Bypasses DESCRIPTION: In the section of the USAR describing the safety design basis of the Primary Containment and Reactor Vessel ! solation Control System, it states "if the ability to trip some essential part of the system has been bypassed, this fact shall be continuously indicated in the control room." The Safety Evaluation states "because no manual bypasses are provided in the isolation control system, safety design basis
.. is met." llowever, a nonconformance report identified there are several manual bypasses in the isolation control system and there is no continuous indication when the bypass switches are in the bypass position. An Engineering Work Request has been initiated for the 16 identified valves to bring the plant into compliance with the safety design basis. This LCR updates the USAR to reflect the present configuration until full compliance with the safety design basis can be met.
SAFETY ANALYSIS:
The lack of bypass indication cannot cause an accident and would have no impact on the consequences of an accident. The indication would serve as an operator aid only. The bypass switches for six of the valves are keylock switches, under Shift Supervisor control, and operation of the other ten valves (as well as the keylocked switch valves) is controlled by strtion procedures and EOPs. Restoration is independently verified. The addition of the bypass function to the control logic was previously evaluated by the applicable design change documents that installed the bypass switches. The bypass indication does not perform or impact a safety related function. The absence or subsequent addition of bypass indication would not increase the consequences of an accident, nor create the possibility of a different type of accident or malfunction of equipment. The lack of bypass indication does not introduce any new failure modes to the PC valve control logic. This indication is not relied upon in the accident analysis.
License Chance Recuest (LCR) 95-0075 TITLE:
USAR Change - IIPCI Control Governor DESCRIPTION: The purpose of this LCR was to correct an error in the USAR pertaining to the pre-startup position of the llPCI control governor. His LCR changed the USAR to reflect that the liPCI control governor is always in the low speed stop position prior to startup.
SAFETY ANALYSIS:
This change was implemented to make the USAR accurately reflect the facility. The pre-startup position of the control governor cannot increase the probability of occurrence or consequences of an accident or malfunction of equipment. In addition, at the low speed stop, the amount of available mect.anisms which contribute to startup related pipe stresses is reduced. The ability of the liPCI system to provide the required startup response is not affected and has been successfully demonstrated. The pre-startup position of the control governor does not adversely impact any I
margins of safety.
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License Chance Reauest (I CR) 95-0082 TITLE:
USAR Change - Mark I Containment Program DESCRIPTION: This LCR affected portions of USAR Section V and Appendix C that describe the Mark I Containment Program. This change allows a less conservative methodology to be used for determining the structural adequacy of torus attached piping and supports when they are subjected to dynamic loading. The new methodology for combining dynamic loads removes the 1.1 factor from the Square-Root-of-the-Sum-of-Squares (SRSS) dynamic load combination and allows all dynamic loads to be combined via SRSS instead of only allowing the combination of two loads. This new methodology was accepted by the NRC on a generic basis for use in Mark I plants.
SAFETY ANALYSIS:
The SRSS methodology was developed generically by General Electric and documented in NEDE-24632, which was subsequently reviewed and accepted by the NRC with an SER on a generic basis for use in Mark I plants. The CNS Mark 1 Containment System has no unique design features / functions that would render the conclusions of the NRC's SER inapplicable for CNS. The accidents and malfunctions that are applicable to Nis USAR change are those that involve piping failures. He new methodology for combining the loads used to evaluate piping attached to the torus does not increase the probability of occurrence or consequences of a piping failure and r,ubsequent LOCA. The new methodology does not reduce any design margin for the affected piping because the acceptance criteria for allowable piping stress is not changed. Automatic and manual responses to a piping failure remain unchanged. He margin of safety that is relevant for this USAR change is that which is applied to piping which is part of the reactor coolant boundary. Since the new load combination methodology does not change the previously defined allowable piping stress, the margin of safety is not reduced.
License Chance Reauest (LCR) 95-0084 TITLE:
USAR Change - Spent Fuel Pool Makeup Source DESCRIPTION: This LCR revised the USAR to delete the Reactor Equipment Cooling (REC)lleat Exchanger Service Water (SW) drain connections as a source of spent fuel pool makeup and adjusts the makeup rate accordingly. It was determined that insufficient head exists to supply makeup from the SW drains on the REC Ileat Exchangers. Nuclear Engineering Department Calculation (NEDC)95-062 was performed to determine the flow rate that can be supplied to the fuel pool by a fire hose connected to the SW drain of a RilR heat exchanger. Reference to this calculation was added to the UMR.
SAFETY ANALYSIS:
Loss of Spent Fuel Pool Cooling or Service Water systems cannot initiate an accident as described in the USAR. This change reflects the use of two 1-1/2 inch hoses connected to SW side drains of the RilR lleat Exchangers vice five 1 inch hoses connected to the SW drains of both R11R and REC 11 eat Exchangers. This reduces the demand on the essential SW and RIIR SW Booster Systems and results in an estimated 120 gpm being removed from the SW side of the RIIR lleat Exchangers vice at least 400 gpm under previous assumptions. Berefore, the overall probability of a malfunction of equipment important to safety is reduced by implementing this USAR change. Elimination of hose connections on the SW side of the REC lieat Exchangers removes any consequences involving the REC system. This activity does not introduce any new operational modes not described in the USAR.
NEDC 95-062 conservatively shows that at least 120 gpm can be provided via the SW drain connections of the RliR licat Exchangers. This exceeds the design makeup and calculated boil-off rates ensuring adequate spent fuel pool water level can be maintained.
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I icense Chance Reauest (LCR) 05-0085 TITLE:
USAR Change - Revision of USAR Figure 1-6-1 l
DESCRIPTION: This LCR revised USAR Figure 1-6-1, Reactor Heat Balance - Rated. This figure was changed to reflect that modifications to the Reactor Recirculation pump seals and Reactor Water Cleanup pump seals added seal purge water from the Control Rod Drive (CRD) system and that a portion of the purge water enters the Reactor Coolant System (RCS). The significance of the flow addition to the RCS is that it appears as "feedwater" flow in the reactor thermal heat balance and must be accounted for.
SAFETY ANALYSIS:
The addition of 6 gpm to the RCS through the pump seals cannot increase the probrWy or consequences of an accident or equipment malfunction previously evaluated in the Us 4
.nor increase the probability or consequences of a different type of accident or equipment maLoction.
The purpose of the seal purge is to increase the reliability of the pump seals. liowever, if a seal were to fail, the presence of the purge flow would not affect the consequences of such a failure.
Malfunction of the CRD system could reduce or stop purge; however, the purge flow is not required for the seals to function and cannot cause the seals to fail. Pump seal failure is bounded by accidents such as a LOCA and transients such as a pump seizure. The addition of 6 gpm to the RCS is not used in any Technical Specification and does not affect the margin of safety of any Technical Specification.
License Chance Reauest (LCR) 95-0086 TITLE:
USAR Change - Spent Fuel Pool Couting System DESCRIPTION: This LCR revised the USAR to include a discussion of spent fuel pool thermal hydraulics in the Safety Evaluation section and to remove the existing discussion ofnormal and emergency heat loads.
It also included additional description of the interconnection of the reactor cavity and spent fuel pool during refueling operations, which provides additional decay heat removal capability beyond the Fuel Pool Cooling System. Table X-5-1 was updated to reflect recent calculations that substantiate the heat removal capability of the various spent fuel pool cooling subsystems.
SAFETY ANALYSIS:
The loss of Spent Fuel Pool Cooling is not an accident as described in the CNS USAR. This change does not increase the opportunity for fuel bundles to be dropped on other fuel assemblies and does not change the procedures by which fuel is handled; therefore, this change does not increase the probability of occurrence of a fuel handling accident. This USAR change provides clarification of system! equipment performance criteria, but has no effect on systems or components utilized to mitigate the consequences of previously evaluated accidents or equipment malfunctions. Recognition of the potential for a full-core off-load shortly after restart from a refueling outage does not increase the potential for or consequences of any existing failure modes, nor introduce any new failure modes.
The loss of Spent Fuel Pool Cooling is described in the USAR and the NRC SER for Amendment No.
52 to the CNS license. The mitigation strategies for coping with the loss of Spent Fuel Pool Cooling remain unchanged. This change does not affect the physical configuration of the plant and does not result in a change to the operation of plant systems. Therefore, it does not create the possibility of a different type of accident or malfunction and has no effect on the margin of safety as defined in the basis of any Technical Specification. The Technical Specification Spent Fuel Pool level and time after shutdown requirements are not affected by this change, nor are the specified number of fuel assemblies capable of being stored in the pool altered by this change.
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1 License Chance Reauest (LCR195-0087 TITLE:
USAR Change - Rewrite of Appendix G and Update of Accident Analysis DESCRIPTION: AN -dix G to the USAR, Station Nuclear Safety Operational Analysis, was completely rewritten baseo n the need to update the appendix to be consistent with present plant design and include the entire scope of transients, accidents, and special events considered for CNS. As a result of the Appendix G changes, changes were made to other sections of the USAR to provide terminology consistent with the revised Appendix G. The USAR Accident Analysis was revised to accurately reflect the current plant configuration in order to meet a commitment made to the NRC. His revision ensures that the entire spectrum of transients, accidents, and special events which have been analyzed for CNS are included in the USAR.
SAFETY ANALYSIS:
His change updates the accident analysis descriptions based on the current plant design and analyses.
Changes from the original plant design and analyses, and accident consequences, are supported by i
NRC Safety Evaluation Reports approving changes to the GESTAR and/or Technical Specification j
amendments. The systems involved were already included in the USAR, but the expected plant response with these systems in place needed to be added. The updated information is taken from i
design documents applicable to CNS. Where operator actions have been described, these actions have been reviewed against plant procedures to ensure that no discrepancies exist. These USAR changes do not result in any violation of the Technical Specifications or reduce the margin of safety defined in the basis for any Technical Specification.
License Chance Reauest (LCR195-0000 TITLE:
USAR Change - Inservice Inspection (ISI)
DESCRIPTION: This LCR deleted references to a specific edition of ASME Section XI from the USAR and made the statements in the USAR that refer to ASME XI generic. His will preciude future revisions due to changes in the edition or addenda of the Code used for ISI. It also clarified that the containment spray system is pneumatically tested during plant outages as a Technical Specification flow test, and not an ASME XI pressure test.
SAFETY ANALYSIS:
No credit is assumed in the accident analysis for the ISI program. The probability or consequences of an accident or malfunction of equipment are independent of the nondestructive examination (NDE) techniques and sampling methods used to find a service generated flaw. NDE is non-intrusive and does not affect the operability of any equipment important to safety. Updating to a later approved edition of the Code is considered an enhancement to the ISI program. The Technical Specifications require an ISI program but do not specify the edition or addenda of the Code to be used. The applicable edition of the Code is determined by 10CFR50.55a(g).
License Chance Reauest (LCR) 95-0097 TITLE:
USAR Change - Personnel Radiation Monitoring DESCRIPTION: This LCR revised the USAR to only require personnel radiation monitoring devices for personnel assigned to work within radiologically controlled areas at the station. 10CFR20.1502 requires monitoring for individuals likely to receive doses greater than 10% of the limits in 10CFR20.1201.
An evaluation was performed which concluded that the unmonitored individuals will not exceed this threshold dose.
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f SAFETY ANALYSIS:
Personnel monitoring devices and the radiation monitoring program cannot initiate an accident described in the USAR, nor affect the USAR assumed method of accident mitigation. This change does not affect the design or operation of components which may initiate an accident. His LCR does not affect the Radiologically Controlled Area (RCA) access of personnel who perform functions which may mitigate the consequences of an accident, nor affect personnel access to areas where equipment impodant to safety is located. This change conforms with 10CFR20 and Technical j
Specification provisions for personnel monitoring.
License Chance Reauest (LCR) 96-0002 TITLE:
USAR Change - RWCU-MOV-MO68 DESCRIPTION: This LCR revised the USAR to show that RWCU-MOV-MO68 is not a containment isolation valve.
i His valve is on the return line to the reactor vessel; it has no safety function and does not receive any containment isolation signals.
SAFETY ANALYSIS:
RWCU-MOV-MO68 is located outside the containment isolation valves and does not receive a containment isolation signal. No credit is assumed in the accident analysis for valve RWCU-MOV-j MO68 closing. De valve's method of operation is not changed and it will continue to meet its applicable design, material, and quality standards. The conta'.nment isolation check valves for the RWCU supply line are not affected by this activity. This adivity does not operate the containment isolation valves in a manner beyond their design parameters or in a manner not described in the USAR. He Technical Specifications do not identify valve RWCU-MOV-MO68 as having any safety function. The margin of safety for containment isolation is not affected.
I License Chance Reauest (LCR196-0008 TITLE:
USAR Change - Mitigation of Station Blackout Event DESCRIPTION: This LCR revised the USAR to identify those systems which are credited with mitigating a Station Blackout (SBO) event, and providing reference to the SBO Coping Assessment and NRC Safety Evaluation Repons for SBO. This change is made to increase awareness regarding the role of plant systems in mitigating the licensing basis SBO event and reduce the possibility of making changes to those systems without considering the requirements of the SBO Coping Assessment.
SAFETY ANALYSIS:
This USAR change does not alter the design, function, or method of performing the function of a USAR described structure, system, or component. Operation of USAR described structures, systems, or components is not changed in order to provide mitigation for the SBO event. The updated information is taken from the SBO Coping Assessment applicable to CNS and the NRC SERs which accepted the CNS method of complying to the SBO Rule,10CFR50.63. The systems involved were already included in the USAR, but their function in mitigating the SBO event was added. These changes do not reduce the margin of safety defined in the basis for any Technical Specification.
License Chance Reouest (I CR) 96 nQll TITLE:
USAR Change "Windmilling" of the Residual lleat Removal Service Water Booster Pump (RHR SWBP)
DESCRIPTION: This LCR revised the USAR to include a description of the method presently used for shutdown cooling (SDC) whereby the RHR SWBP is secured and the SW pumps are used to supply cooling of the RHR Heat Exchanger (windmilling of the RHR SWBP). It also includes restrictions imposed on the use of the RHR SWBP so as to maintain the plant within the safety analysis.
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SAFETY ANALYSIS:
Windmilling of the RHR SWBP while in the SDC mode can have radiological effects, both during normal SDC operation and under accident conditions. The SW coolant flow from the RHR Heat Exchangers is monitored by a radiation detector which alarms on a high radiation level in the SW effluent. Even with the RHR SWBP being windmilled while in the SDC mode, the SW discharge to -
the environs is only a small fraction of the 10CFR20 limits, provided that the SW radiation monitor is operable and at least one circulating water pump is operating. Station operating procedures in lude these restrictions. Two events which could potentially be affected by windmilling of the RHR Sk 3P
- 1) a critical crack of a single tube of the RHR Heat Exchanger, and 2) a refueling accident.
are:
However, calculations were performed which show that the increase in the release of radioactive materials to the environs for both of these events is negligible. Previous Bums & Roe analysis demonstrated that windmilling of the RHR SWBP has no detrimental effect on any equipment in the SW system. The existing USAR analyses bound the possible types of malfunctions. Windmilling of the RHR SWBP during plant operation in the SDC mode is within the CNS de:tign basis and therefore does not involve an unreviewed safety question.
License Chance Reauest (LCR196-0014 TITLE:
USAR Change - Secondary Containment Penetrations DESCRIPTION: This LCR revised the USAR to clarify that piping that penetrates secondary containment need not meet Seismic IS requirements. Bis was previously implied by NPPD responses to AEC questions but not explicitly stated in the USAR.
SAFETY ANALYSIS:
This USAR revision does not change the licensing basis under which the accident analysis was initially performed and the accident analysis remains unchanged. His revision does not change any equipment specifications nor make any physical modifications; therefore, the probability of occurrence or consequences of failure are not increased from that previously evaluated in the USAR.
The secondary containment will continue to function as specified in the Technical Specification and its Bases.
USAR Chanee Reauests (UCRs)96-006. 96-008.96-009. 96-011. and 96-012: License Chance Reauest ( LCR) 967.0017 TITLE:
Editorial USAR Changes DESCRIPTION: These change requests made various editorial changes to the USAR. Included in these editorial changes were: 1) various corrections to the Table of Contents, List of Figures, and List of Tables to make them match the USAR; 2) correction of a figure reference; 3) removal of a figure which was to be removed in a previous revision but was inadvertently left off the filing instructions; 4) addition of a word which was inadvertently deleted from the end of a page; 5) addition of a reference to a following section for clarification; 6) correction of a typographical error; and 7) relocation of a figure to the proper section of the USAR.
SAFETY ANALYSIS:
By definition, editorial changes have no consequential effect on the impacted sections /pages.
Accordingly, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety are unchanged. These changes have no effect on the basis for any Technical Specification.
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USAR Chance Reauest (UCR)96-007 TITLE:
USAR Change - Steam Tunnel Blowout Panels DESCRIPTION: This USAR Change Request incorporated details on the function and location of the steam tunnel blowout panels, as part of the committed corrective action for an NRC Enforcemat Action. He hck of sufficient details may have contributed to the inappropriate installation of fiberglass material over the blowout panels. This change also included discussion of the calculations used to determine the time variation of steam pressure and temperature following a steam line break in the main steam line tunnel. The peak pressure value for the steam tunnel and the value of the yield stress in the reinforcing steel were also updated to reflect Amendment 25 of the FSAR.
SAFETY ANALYSIS:
This change does not involve any physical change to the facility. It enhances the USAR's description of the function and location of a safety feature, reflects the results of newly referenced calculations, and makes corrections to pressure and stress values previously documented in Amendment 25 of the FSAR. This change is not associated with nor can it cause any malfunction of equipment. The only postulated accidents associated with this change are high energy line breaks. These changes cannot increase the probability or consequences of an accident or equipment malfunction. No margins of safety are affected by this USAR change.
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l USAR Chance Reauest (UCR196-010 TITLE:
USAR Change - Organizational Changes DESCRIPTION: This UCR made changes to USAR Section Xill to reflect the current management organization.
Principal changes included: 1) deletion of Nuclear Safety Support Manager position; 2) creation of SORC/SRAB Administrator position; 3) creation of Security Manager position; 4) correction of duties of Events Analysis Manager; 5) correction of reporting relationship of Senior Manager of Safety Assessment; and 6) revision of organization charts.
SAFETY ANALYSIS:
Organizational and staffing requirements essential to the safe operation of CNS are established by the Technical Specifications, ANSI N-18.1, Regulatory Guide 1.8, the QA Plan, the Safeguards Plan, and j
the Emergency Plan. The consequences of changes that do not affect these documents, or that have been submitted to the NRC via the appropriate requirements, either do not impact safety or have been previously evaluated. Therefore, organizational changes covered by this revision do not increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety. These changes do not reduce the margin of safety as defined in the basis for any Technical Specification.
USAR Chance Reauest (UCR)96-013 TITLE:
Editorial Changes / Minor Corrections to USAR Revision XIV DESCRIPTION: his UCR incorporated several editorial changes and minor corrections / changes that resulted from Management review of USAR Revision XIV. These changes provide improved consistency and clarity in the subject USAR revision.
SAFETY ANALYSIS:
These changes are editorial or minor enhancements to improve consistency and clarity. Accordingly, the probability of occurrence or consequences of an accident or malfunction of equipment important to safety are unchanged. These changes have no effect on the basis for any Technical Specification.
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1 US AR Chance Reauests (UCRs)96-014. 96-016.96-017. 96-022.96-026. 96-028.96-029. 96-030.96-040. 96-047.96-050. and 96-056 TITLE:
USAR Drawing Changes - Inservice inspection (ISI) Boundaries DESCRIPTION: The following USAR figures were revised as a result of drawing changes to add or revise the locations of ISI boundaries: Figure 1119-1; Figure IV-3-3, Sheet 1; Figure IV-3-3, Sheet 2; Figure IV-71; Figure IV-9-1, Sheet 1; Figure V-2-13; Figure VII-4-4; Figure VII-4-11; Figure X-8-6; Figure XI-6-1, Sheet 1; and Figure XI-6-1, Sheet 4.
SAFETY ANALYSIS:
The boundary classification is for inspection purposes only and does not affect the ability of any
]
component to perform its safety function. The boundary classification does not alter the physical or i
operating characteristic of any component. Therefore, the probability of occurrence or consequences of an accident or equipment malfunction are not increased and the margin of safety is not reduced.
US AR Chance Reauest (UCR)96-020 2
TITLE:
USAR Drawing Change - Figure X-10-lb I
DESCRIPTION: Figure X-10-lb was updated as a result of two drawing changes. One change added a section of piping that was not shown on the drawing. A section of piping from a line upstream of valve DW-V-459 to a line upstream of valve DW-V-335 exists in the field, but was not shown on the drawing. A second change deleted valve AS-V-235 from the drawing because it is not installed in the plant.
ANALYSIS:
The piping addition in the Demineralized Water system does not alter or add any precursor to any transient or accident described in the USAR, nor alter any equipment used to mitigate the consequences of any transient or accident described in the USAR. The added section of piping is not located near any safety-related equipment and the piping system is non-essential. This change does not impact the containment boundary and has no impact on any radiological concems for any accident addressed in the USAR. It hu no impact on any equipment that is required to support the safe j
operation of the plant. The Demineralized Water system is not a part of any system addressed by any Technical Specification; therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.
Valve AS-V-235 is part of the non-essential Auxiliary Steam system and is not located near any safety-related equipment. Removal of this valve does not alter or add any precursor to eny transient or accident described in the USAR, nor alter any equipment used to mitigate the consequences of any transient or accident described in the USAR. His change does not impact the containment boundary and has no impact on any radiological concems for any accident addressed in the USAR. His change will not add any open/ unrestricted flow paths to the Auxiliary Steam system. This valve, as shown on the drawing, is not located near any equipment important to the safe operation of the plant.
AS-V-235 is not a part of any system addressed by any Technical Specification; therefore, the margin of safety as defined in the basis for any Technical Specification is not reduced.
USAR Chance Reauest (UCR)96-023 TITLE:
USAR Drawing Change - Figure X-12-1 DESCRIPTION: Figure X-12-1 was updated to reflect a drawing change that corrected an error that showed temperature switch SA-TS-203C connected to the compressor high temperature trip circuit. This error was identified as a result ofinvestigation of a Design Criteria Document Open Item.
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i SAFETY ANALYSIS:
This change does not add any new or existing equipment to the drawing. It alters a control signal that was incorrectly indicated on the drawing. Based on a review of other existing related plant drawings, this change is the correction of a drawing error. His change does not alter the design, function, or method of performing the function of any structure, system, or component described in the USAR. There is no change in the manner in which this equipment is operated based on this drawing change. Therefore, this change to Figure X-12-1 does not increase the probability of occurrence or consequences of an accident or equipment malfunction, and does not reduce the margin of safety as defined in the basis for any Technical Specification.
j USAR Chance Reauest (UCR)96-031 TITLE:
USAR Drawing Change - Figure X-14-2 l
DESCRIPTION: Figure X-14-2 was revised to reflect various configuration changes to the Radioactive Waste i
Drains / Vents system as a result of a drawing / system verification and an Engineering Work
-)
Request.
SAFETY ANALYSIS:
These changes to the floor drain system do not alter or add any precursor to any transient or i
accident described in the USAR, nor alter any equipment used to mitigate the consequences of any transient or accident described in the USAR. These changes do not alter the functional relationship of the floor drain system with any supponed or supporting system or its safety design l
basis. These changes do not impact the containment boundary or compromise the use ofloop seals to maintain Reactor Building integrity. No new sources were added to the floor drain system l
that could exceed the capacity of the Liquid Radwaste system or the Augmented Liquid Treatment l
system. These changes do not alter the conclusion of the Safety Evaluation or the nuclear safety l
l op; rational requirements. The conservatism of the CNS Technical Specifications as described in the USAR is not altered.
I USAR Chance Reauest (UCR)96-032 I
TITLE:
USAR Drawing Change - Figure X-14-1 l
f DESCRIPTION: Figure X-14-1 was revised to reflect various configuration changes to the Reactor Building l
Equipment Drain System as a result of a drawing / system verification.
l SAFETY l
ANALYSIS:
These changes to the equipment drain system do not alter or add any precursor to any transient or accident described in the USAR, nor alter any equipment used to mitigate the consequences of any transient or accident described in the USAR. These changes do not alter the functional
(
relationship of the equipment drain system with any supported or supporting system or its safety design basis. These changes do not impact the containment boundary or compromise the use of l
loop seals to maintain Reactor Building integrity. No new sources were added to the equipment drain system that could exceed the capacity of the Liquid Radwaste system or the Augmented l
Liquid Treatment system. These changes do not alter the conclusion of the Safety Evaluation or the nuclear safety operational requirements. He conservatism of the CNS Technical Specifications as described in the USAR is not altered.
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USAR Chance Reauest (UCR)96-046 TITLE:
USAR Drawing Change - Figure IV-9-1, Sheet 1 l
DESCRIPTION: Figure IV-9-1 was updated to reflect a drawing change that added a pipe cap at the end of the drain line downstream of valve RWCU-V-172. The pipe cap was part of the original plant design and installation; however it was inadvertently deleted by a 1984 Design Change which replaced the valve and drain line. The pipe cap was reinstalled as a minor maintenance item; therefore, this drawing was revised to reflect the addition of the pipe cap.
SAFETY ANALYSIS:
The drain line downstream of valve RWCU-V-172 is not involved in any accident s:quence analyzed in the USAR. In addition, the drain line is not involved as the initiator of a ny analyzed event. It does not perform any accident mitigating function, nor is it an essential part of any system which performs an accident mitigating function. The drain line does not interface nor interact with any equipment important to safety as currently analyzed in the USAR. Placing a cap on the drain line returns the system to a condition which respresents its currently analyzed state.
This drain line has no function which could affect the basis of any Technical Specification.
USAR Chance Recuest (UCR)96-051 TITLE:
USAR Drawing Change - Figure Vill-4-2 DESCRIPTION: Figure Vill-4-2 was updated to reflect a drawing change that added a load onto Motor Control Center (MCC)"C"(compartment 6ER) and revised a cable designation. The load is a 480 VAC receptacle in the Hot Shop and the cable supplies a weld:ng receptacle. This load is addressed in the MCC "C" load study. The load supplied by compartment 6ER is presently listed in the 480 VAC Power Checklist in Operating Procedure 2.2.19A. Based on the application, this load is considered to be intermittent. The change to the cable designation was a change to the drawing only; no changes were required in the plant.
SAFETY ANALYSIS:
This change does not relocate or add any continuous load to the MCC and, therefore, will have no impact on the MCC loading. The subject receptacle is protected by a fused disconnect such that a fault at the receptacle will be interrupted before it can impact the MCC. This load is not connected to any safety-related power sources and, therefore, will have no impact on standby AC power sources. This load is not located near any safety-related equipment and its malfunction cannot adversely impact any safety-related equipment. This receptacle is not required to support any operations in the plant. MCC "C" is not required to be used in any accident scenario or to minimize the impact of any abnormal operational transients. For these reasons this change does not increase the probability of occurrence or consequences of an accident or equipment malfunction. This receptacle has no impact on any Technical Specification related equipment.
USAR Chance Reauest (UCR)96-053 TITLE:
USAR Drawing Change - Figure X-10-ta DESCRIPTION: Figure X-10-t a was updated to reflect a drawing change that revised the valve type for Auxiliary Condensate system valves ACD-V-392 and ACD-V-393 from globe to gate. The valve type change was made to update the drawing to reflect as-built conditions as specified in a Component Evaluation Package.
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SAFETY ANALYSIS:
The valves associated with this change are non-essential manual valves and the manner in which they are operated has not changed. They are not required to mitigate the consequences of any accident and are not required for safe shutdown of the plant. The valve type change is not the result of any equipment changes in the plant, but is based on updating the plant equipment files to match the installed conditions. Therefore, this change does not increase the probability of occurrence or consequences of an accident or equipment malfunction. These valves are not identified in the basis for any Technical Specification.
USAR Chance Reauest (UCR)96-062 TITLE:
USAR Drawing Change - Figure XI-6-1 DESCRIPTION: Figure XI-6-1 was revised to add a 1" drain valve (CW-V-299) to strainer C and a 1" drain valve (CW-V-301) to strainer D of the Circulating Screen Wash and Service Water System. This installation was determined to be an undocumented plant modification and has been documented in the corrective action program. Review of the vendor manual for the strainers revealed that the vendor recommended that the drain plugs be removed and drain valves be installed. The Screen Wash system is classified as non-essential.
SAFETY ANALYSIS:
Replacing a drain plug with a drain valve on the screen wash strainer in the non-essential Screen Wash system is not associated with any accident or malfunction of equipment important to safety previously evaluated in the USAR. It also does not create the possibility of a different type of accident or malfunction than previously evaluated. This activity is not associated with the basis for any Technical Specification and does not reduce the margin of safety.
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