SVP-05-001, Summary Report of Changes, Tests, and Experiments Completed

Summary Report of Changes, Tests, and Experiments Completed

ML050120131
Person / Time
Site:	Quad Cities
Issue date:	01/03/2005
From:	Tulon T Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SVP-05-001
Download: ML050120131 (8)

Text

Exeloin.

Exelon Generation Company, LLC www.exeloncorp.com Nuclear Quad Cities Nuclear Power Station 22710 206" Avenue North Cordova, IL 61242-9740 SVP-05-001 January 3, 2005 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Reference:

Letter from T. J. Tulon (Exelon Generation Company) to U. S. NRC, "10 CFR 50.59 Summary Report," dated January 3, 2003

Subject:

10 CFR 50.59 Summary Report In accordance with 10 CFR 50.59, "Changes, tests, and experiments," subpart (d)(2), we are forwarding a summary report of completed changes, tests, and experiments. This submittal contains a summary of 10 CFR 50.59 evaluations completed between January 1,2003 and December 31, 2004.

Should you have any questions concerning this letter, please contact Mr. W. J. Beck at (309) 227-2800.

Respectfully, Timothy J. Tulon Site Vice President Quad Cities Nuclear Power Station

Attachment:

Summary Report of Changes, Tests, and Experiments Completed cc: Regional Administrator- NRC Region IlIl NRC Senior Resident Inspector - Quad Cities Nuclear Power Station 7KCf

I ATTACHMENT Summary Report of Completed Changes, Tests, and Experiments EC Tracking Number: SE-99-049 Change Document(s)

Unit: Common DCP 9900122; UFSAR-03-R8-004 Activity Description The activity involves a plant modification which installs an improved seismograph in place of existing equipment. The new instrument uses the same power supply as the existing equipment and improves data retrieval capabilities.

Impact of Activity The impact of the activity is minimal. The new seismograph will continue to be powered from MCC 16-2 with no additional system load. Consistent with current practices, the seismograph will be checked routinely by operations. No other plant interactions exist.

Bases for Not Requiring NRC Prior Approval The improved seismograph does not adversely impact the system design function. No new failure modes or adverse plant interactions were identified. As such, the probability of occurrence or the consequences of an accident or transient is not increased. Similarly, the likelihood of occurrence or the consequences of a malfunction of a SSC important to safety is not increased.

These changes do not create the possibility of an accident of a different type or a malfunction of a SSC important to safety with a different result. No fission product barriers or methods of evaluation described in the UFSAR are impacted by the change. Therefore, this change may be implemented without prior NRC approval.

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ATTACHMENT Summary Report of Completed Changes, Tests, and Experiments 2 Trackinq Number: QC-E-2003-001 Change Document(s)

Unit: Unit 1 EC334506 Activity Description The activity involves a plant modification. The existing Main Control Room (MCR) Toxic Gas monitoring system is being upgraded to a solid-state design. The change improves the overall reliability of the existing toxic monitoring system by providing a more reliable design and redundant sensors for each HVAC intake.

Impact of Activity The Toxic Gas Analyzer design function remains unchanged, which is to isolate the MCR on detected high concentrations of ammonia. However, the new system increases the Toxic Gas system response time due to the type of ammonia sensors used. The new response time will continue to provide adequate protection to control room personnel. The new design will provide a reliable system design that is versatile while providing sufficient operator protection, including a local monitor in the MCR to inform the operator of actual ammonia concentration.

Bases for Not Requiring NRC Prior Approval While the new ammonia toxic gas analyzer system will have an increased response time, the resulting ammonia concentration is within the regulatory guidelines provided Regulatory Guide 1.78, "Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release." Therefore, the new system continues to provide adequate protection to control room personnel. As such, the probability of occurrence or the consequences of an accident or transient is not increased. Similarly, the likelihood of occurrence or the consequences of a malfunction of a SSC important to safety is not increased. These changes do not create the possibility of an accident of a different type or a malfunction of a SSC important to safety with a different result. No fission product barriers or methods of evaluation described in the UFSAR are impacted by the change. Therefore, this change may be implemented without prior NRC approval.

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ATTACHMENT Summary Report of Completed Changes, Tests, and Experiments 3 Trackinq Number: OC-E-2003-002 Change Document(s)

Unit: Common UFSAR-03-R8-022 Activity Description The proposed activity is a UFSAR change. The change incorporates an analysis performed by Global Nuclear Fuels (GNF) concerning operation with the electrohydraulic control (EHC) pressure regulator out-of-service (PROOS). The change is also reflected in site procedures as necessary.

Impact of Activity This change is being performed to allow additional operational flexibility to operate with one EHC pressure regulator inoperable at greater than 25% core thermal power. The EHC system is equipped with two pressure regulators. If the "in control" regulator should fail downscale, the back-up regulator will take control limiting the reactor pressure increase. If the back-up regulator is unavailable or out of service when the "incontrol" regulator fails, then the turbine control valves would close resulting in a plant transient. GNF has evaluated this transient and determined that it is bounded by existing analysis. The UFSAR will be changed to reflect that this transient event is bounded by an existing analysis. This activity will allow operation with a pressure regulator inoperable by implementing the appropriate thermal limits contained in the Core Operating Limits Report (COLR) for TCV slow closure (i.e., the bounding event).

Bases for Not Requiring NRC Prior Approval The thermal limits in the COLR protect the fuel cladding fission product barrier during transient conditions. Since the fuel cladding would not be operated beyond the currently analyzed limits there is no adverse impact on plant safety. The UFSAR and the COLR recognize that various equipment out-of-service combinations may be utilized to improve operational flexibility provided that the appropriate thermal limits are implemented to protect the Safety Limit Minimum Critical Power Ration (MCPR) and other thermal limits. These analyses are performed using NRC approved methodologies. This change simply adds an additional equipment out-of-service option.

The consequences of this option are encompassed by an existing operating mode described in the COLR (TCV slow closure). As such, the probability of occurrence or the consequences of an accident or transient is not increased. Similarly, the likelihood of occurrence or the consequences of a malfunction of a SSC important to safety is not increased. These changes do not create the possibility of an accident of a different type or a malfunction of a SSC important to safety with a different result. No fission product barriers or methods of evaluation described in the UFSAR are impacted by the change. Therefore, this change may be implemented without prior NRC approval.

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ATTACHMENT Summary Report of Completed Changes, Tests, and Experiments

] Tracking Number: QC-E-2003-003 Change Document(s)

Unit: Unit 1 EC 343934 Activity Description The proposed activity is a temporary design change. The Temporary Configuration Change Package (TCCP) will bypass a Main Steam Line Tunnel High Temperature Switch (TS 1-0261-18A) by installing a jumper across the switch contact. The switch contact is normally closed and opens on high temperature. The switch has been identified as faulty by troubleshooting. The installation of the jumper will ensure the suspect temperature switch will not spuriously trip and subject the plant to a transient. Technical Specifications Section 3.3.6.1 indicates that 2 channels (i.e., switches) per string are required for the Main Steam Line isolation logic function (i.e., Group 1) or the channel must be placed in trip within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Therefore the Technical Specification Required Action will be entered and the temporary jumpers must be removed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of being installed, or the channel will be placed in trip.

Impact of Activity This TCCP will render TS 1-0261-18A, Main Steam Line Tunnel High Temperature Switch, inoperable. However, TS 1-0261-15A, TS 1-0261-16A, and TS 1-0261-17A Main Steam Line Tunnel High Temperature Switches are in the same string and will remain operable. The appropriate Technical Specifications Required Action will be entered. Consequently, no other systems, structures, or components are affected by this temporary change. Therefore, this change has no impact on equipment reliability or on existing design evaluations, analysis, or methodology. No new failure modes are introduced.

Bases for Not Requiring NRC Prior Approval The installation of the temporary jumpers is being performed in accordance with the Technical Specifications Required Actions. The design basis trip function will be maintained by the remaining switches while the jumper is installed. As such, the probability of occurrence or the consequences of an accident or transient is not increased. Similarly, the likelihood of occurrence or the consequences of a malfunction of a SSC important to safety is not increased. These changes do not create the possibility of an accident of a different type or a malfunction of a SSC important to safety with a different result. No fission product barriers or methods of evaluation described in the UFSAR are impacted by the change. Therefore, this change may be implemented without prior NRC approval.

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I ATTACHMENT Summary Report of Completed Changes, Tests, and Experiments 5 Tracking Number: QC-E-2003-004 Change Document(s)

Unit: Unit 1 EC 22082 Activity Description The activity is a plant design change. This activity replaces the existing analog control logic for the Reactor Recirculation Control System (RRCS) with a digital logic system. This includes the Main Control Room (MCR) control stations for pump speed control and master control, and certain transmitters and sensors necessary to interface with the new equipment. Additionally, the activity will replace the present analog jet pump instrumentation with a digital system for compiling and transmitting system information to operator displays. The replacement microprocessor used for the new control system is known as the ADVANT Controller System, supplied by Westinghouse Atom ABB. The ADVANT Control system has redundant, parallel operating processors with the primary normally in service, while the backup system can assume control if the primary fails. The reason for this activity is to replace the existing RRCS and jet pump instrumentation with a modern, fault tolerant, digital-based control system to support plant operation through the remainder of plant life.

Impact of Activity The RRCS controls the operation of the variable speed recirculation pumps. This system also includes the flow instrumentation for the recirculation loops and the jet pumps. The digital hardware used by this activity is very similar to the Feedwater Level Control design that has been used in Exelon and European BWRs for many years. This system has proven to be reliable and does not impact other equipment. Any unique human-machine interfaces have been evaluated for potential challenges to operation personnel and found acceptable.

Bases for Not Requiring NRC Prior APproval The use of the upgraded RRCS system will provide improved control of reactor power during power operations. Safety related protective features are external to the RRCS and not adversely impacted by this design change. As such, the probability of occurrence or the consequences of an accident or transient is not increased. Similarly, the likelihood of occurrence or the consequences of a malfunction of a SSC important to safety is not increased. These changes do not create the possibility of an accident of a different type or a malfunction of a SSC important to safety with a different result. No fission product barriers or methods of evaluation described in the UFSAR are impacted by the change. Therefore, this change may be implemented without prior NRC approval.

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ATTACHMENT Summary Report of Completed Changes, Tests, and Experiments p6 Trackinq Number: QC-E-2003-005 Change Document(s)

Unit: Common EC 23374; EC 23375 Activity Description The activity is a plant design change. The change will abandon the Economic Generation Control (EGC) console from the Unit 1 and Unit 2 plant controls. The EGC system interfaced with the Electro-Hydraulic Control (EHC) and Reactor Recirculation Control (RRCS) systems to provide the system load dispatcher the necessary controls for automatic load control. The EGC function is no longer used for plant operations at Quad Cities Nuclear Power Station.

Impact of Activity The EGC system will be electrically disconnected and abandoned. This will have no impact on normal main turbine operation or reactor pressure control. Neither the EHC or RRCS master controls will be operating in a mode that uses inputs from the EGC system. Plant operating procedures will be revised to eliminate use of EGC.

Bases for Not Requirinq NRC Prior Approval Removal of the EGC equipment in no way impacts safe operation of the plant. The use of EGC has not been used for plant operation in over a decade. EGC functions were not credited to provide any mitigation to the design basis accidents, transients or malfunctions. No new interfaces and no new equipment were added by this change. Plant operation is fully supported without EGC operation. As such, the probability of occurrence or the consequences of an accident or transient is not increased. Similarly, the likelihood of occurrence or the consequences of a malfunction of a SSC important to safety is not increased. These changes do not create the possibility of an accident of a different type or a malfunction of a SSC important to safety with a different result. No fission product barriers or methods of evaluation described in the UFSAR are impacted by the change. Therefore, this change may be implemented without prior NRC approval.

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ATTACHMENT Summary Report of Completed Changes, Tests, and Experiments Trackingq Number: OC-E-2004-001 Change Document(s)

Unit: Unit 2 EC 343933; UFSAR-03-R8-037; QC-BAS-04-001 Activity Description The activity is a plant modification. The four Unit 2 Reactor Pressure Vessel (RPV) relief valves will be replaced with Electromatic Relief Valves (ERVs) manufactured by Industrial Valve Operations/Dresser. The existing relief valves are Power Operated Relief Valves (PORVs) manufactured by Target Rock Corporation. The existing PORVs were installed as replacements to the original Dresser ERVs as an upgrade to Unit 2 in the 1995 timeframe. The Unit 2 RPV relief valves will now revert back to their original design, and match those currently in use in Quad Cities Unit 1 and Dresden Units 2 and 3. Note that the existing Target Rock safety/relief valve is not replaced by this change. The four existing Unit 2 PORVs are susceptible to self-actuation when a limited amount of pilot valve leakage exists. Inadvertent actuation of a relief valve represents a significant plant transient involving an uncontrolled reactor cooldown, manual scram, and elevated suppression pool temperatures.

Impact of Activity The ERV design can withstand greater pilot leakage than the PORV prior to self-actuation. In addition, PORV pilot leakage cannot be differentiated from main seat leakage. ERV pilot and main seat leakage can easily be differentiated. ERVs are commonly used as RPV relief valves and ADS valves at several different stations, whereas the PORV design is unique to Quad Cities Unit 2. The main effect of this activity is to restore the Unit 2 RPV relief valves to their original design configuration. The Dresser ERVs offer a greater tolerance to pilot valve leakage, and permit the pilot valve leakage to be monitored separately from main seat leakage.

Bases for Not Requiring NRC Prior Approval The proposed ERVs are functionally identical to the existing PORVs. The replacement of PORVs with ERVs does not change the purpose or design function of the RPV relief valves. The design conditions remain the same for the new ERVs. The ERV design allows pilot valve and main seat leakage to be easily differentiated. The Environmental Qualification (EQ) Evaluation for the ERV solenoid actuator employs an alternate evaluation methodology because they are qualified to NUREG 0588 Category II (instead of Category I). Re-installation of Unit 2 Dresser ERVs is appropriately viewed as a rescission of the original replacement decision and reversion to the original design. This change was evaluated and was determined to be acceptable based on the guidance provided in Regulatory Guide 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants." As such, the probability of occurrence or the consequences of an accident or transient is not increased. Similarly, the likelihood of occurrence or the consequences of a malfunction of a SSC important to safety is not increased.

These changes do not create the possibility of an accident of a different type or a malfunction of a SSC important to safety with a different result. No fission product barriers or methods of evaluation described in the UFSAR are impacted by the change. Therefore, this change may be implemented without prior NRC approval.

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