ML20308A476
| ML20308A476 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 11/02/2020 |
| From: | Hunnewell S Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML20308A476 (38) | |
Text
Tennessee Valley Authority, Sequoyah Nuclear Plant, P.O. Box 2000, Soddy Daisy, Tennessee 37384 November 2, 2020 10 CFR 50.59 10 CFR 72.48 10 CFR 50.71 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327, 50-328, and 72-034
Subject:
10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report and Commitment Summary Report
Reference:
TVA letter to NRC, 10 CFR 50.59 and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report and Commitment Summary Report, dated June 5, 2019 In accordance with 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2), Enclosure 1 is the Sequoyah Nuclear Plant (SQN), Units 1 and 2, Summary Report regarding the implemented changes, tests, and experiments for which evaluations were performed in accordance with 10 CFR 50.59(c) and 10 CFR 72.48(c). The summarized evaluations provided in the enclosure were implemented since the Reference Letter through October 23, 2020. During review of 10 CFR 50.59 evaluations for submission, it was identified that an evaluation summary for a SQN Unit 2 design change notice 22499, had not been submitted during a previous reporting period wherein the similar Unit 1 design change notice had been submitted. This issue has been captured in our corrective action program.
Since last reported in the Reference letter, SQN has revised a regulatory commitment in accordance with NEI 99-04, the Nuclear Energy Institute's "Guidelines for Managing NRC
[Nuclear Regulatory Commission] Commitment Changes," as endorsed in NRC Regulatory Issue Summary 2000-17. The commitment change summary is provided in the Enclosure 2.
U.S. Nuclear Regulatory Commission Page 2 November 2, 2020 There are no new commitments contained in this letter. If you have any questions concerning this submittal, please contact Mr. Andrew McNeil, SQN Licensing Manager (Acting) at (423) 843-8098.
Scott Hunnewell Interim Site Vice President Sequoyah Nuclear Plant Enclosures 1. 10 CFR 50.59, and 10 CFR 72.48 Changes, Tests, and Experiments Summary Report
- 2. Commitment Change Report cc (Enclosures):
NRC Regional Administrator- Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant Director, Division of Fuel 'Management, Office of Nuclear Material Safety and Safeguards
ENCLOSURE 1 SEQUOYAH NUCLEAR PLANT 10 CFR 50.59 AND 10 CFR 72.48 CHANGES, TESTS, AND EXPERIMENTS
SUMMARY
REPORT
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES 22501 Rev. A Design Change Notice (DCN) 22501 re-gears the actuators for Containment The re-gearing of Containment Sump Isolation Evaluation Sump Isolation Valves 2-FCV-63-72 and 2-FCV-63-73. This will allow each Valves 2-FCV-63-72 and 2-FCV-63-73 changes Rev 0 valve to develop additional torque/thrust which is needed to overcome the the maximum allowable stroke time of the expected differential pressure across the valve. This re-gearing results in a valves from 45 seconds to 120 seconds.
longer opening stroke time for these valves. The spring packs will be modified Analysis shows that this is acceptable and will or replaced to preclude any potential issues associated with hydraulic lock as a have no adverse effect on ECCS operations.
result of grease infiltrating the spring pack as well as in support of developing Sufficient net positive suction head (NPSH) will the additional torque/thrust. DCN 22501 will ensure compliance to the Nuclear still be available to the RHR pumps during the Regulatory Commission (NRC) Generic Letter (GL) 89-10/96-05 program switchover from ECCS injection to ECCS including the new Joint Owner Group (JOG) Motor Operated Valve (MOV) recirculation. For the limiting loss of coolant program requirements. accident (LOCA), the ECCS sump water inventory present during the switchover from Valves 2-FCV-63-72 and 2-FCV-63-73 are part of the Safety Injection System injection to recirculation mode, at the time (SIS) and part of the Emergency Core Cooling System (ECCS). These valves suction is first taken from the sump, will provide the communication with the Containment Sump for suction to the continue to conservatively meet all ECCS flow Residual Heat Removal (RHR) pumps. These valves also serve as the requirements and will preclude unacceptable containment isolation valve for containment penetration X-19A and X-198. vortexing. The time to accomplish operator actions to complete the remainder of the Valves 2-FCV-63-72 and 2-FCV-63-73 are normally closed and are interlocked switchover to the recirculation phase of ECCS with normally open valves 2-FCV-74-3 and 2-FCV-74-21. Valves 2-FCV-74-3 core cooling is not affected by this change.
and 2-FCV-74-21 provide ECCS flow to the RHR pumps from the Refueling There will be no increase in the likelihood of Water Storage Tank (RWST) during the injection phase of core cooling. Valves accidents, malfunctions, or consequences due 2-FCV-63-72 and 2-FCV-63-73 go to the open position and valves 2-FCV-74-3 to this change; therefore, this activity does not and 2-FCV-74-21 go to the closed position automatically during the switchover require NRC approval.
from ECCS injection to ECCS recirculation. The switchover occurs when the RWST low level setpoint is reached and the sump high level setpoint is reached. This switchover allows the RHR pumps to take suction from the Containment Sump to provide uninterrupted ECCS flow during the transition to the recirculation phase of core cooling to the Reactor Coolant System (RCS) and to isolate the emptied RWST from the RHR pump suction. The flow is uninterrupted since 2-FCV-63-72 and 2-FCV-63-73 travel to 5 percent open before valves 2-FCV-74-3 and 2-FCV-74-21 start to close which ensures a source of water supply to the RHR pumps. During the injection phase, all ECCS pumps take suction directly from the RWST. Following the swapover, the RHR E1-1 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES pumps take suction directly from the ECCS sump and the Charging and SIS pumps take suction from RHR discharge downstream of the RHR heat exchangers.
22703 Rev. 2 The existing SQN Unit 1 Turbine Driven Auxiliary Feedwater (TDAFW) speed This evaluation has determined that the TDAFW Evaluation governor system will be replaced in its entirety by this modification. The system will continue to meet its design and Rev. 0 replacement governor system will replicate the functionality of the existing licensing bases requirements following the speed control and electrical overspeed trip components of the existing system, implementation of the proposed modification as well as incorporating the flow control function of the existing Yokogawa that converts to a digital governor control single-loop process controller (SLPC) located in panel 1-L-381. Incorporating method for the system.
the flow control functions within the new governor system will simplify the control system and reduce the number of active components required to Since the new TDAFW System components are perform the TDAFW design function, as well as allowing the use of a standard more reliable than the existing components and replacement governor design across the Tennessee Valley Authority (TVA) no new system level failure mode effects are nuclear fleet. introduced, the proposed modification does not result in more than a minimum increase in the The new governor system is configured to accept the existing flow signal and frequency of occurrence of an accident will perform the flow control function of the existing 1-FIC-46-57 Yokogawa previously evaluated in the Sequoyah Nuclear controller as a cascade control function within the new Woodward 505 Plant Updated Final Safety Analysis Report governor. The Yokogawa controller, as well as the associated power supply (UFSAR).
and status relay will be deleted from panel 1-L-381.
The new equipment being installed will not A new governor panel will be installed adjacent to the existing panel initiate any new system malfunctions. Credit is immediately outside the TDAFW room, and a new actuator positioner panel, to taken for the TDAFW System for the successful house the actuator positioner (controller) and associated components, will be mitigation of the following transients, special installed on a common support on the rear of the new governor panel. The events, and accidents. The Auxiliary Feedwater existing governor panel, located in the TDAFW room, will be modified to serve (AFW) system supplies, in the event of a loss of as a junction box. A new panel will also be installed to house the non-safety the main feedwater supply, sufficient feedwater related Integrated Computer System (ICS) interface programmable logic to the steam generators (SGs) to remove controller (PLC) and associated components. primary system stored and residual core energy.
It may also be required in some other The new governor panel, positioner panel, and governor valve actuator will be circumstances such as the evacuation of the powered from the existing governor 125 V Battery Board III breaker 321 feed, MCR, cooldown after a LOCA for a small break, with the existing alternate feed from 125V Battery Board IV breaker 321, maintaining a water head in the SGs following a identical to the existing TDAFW governor. The new governor system will result LOCA, a flood above plant grade, Anticipated in an additional load on this circuit. Transient Without Scram (ATWS) event, and 10 E1-2 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES CFR 50, Appendix R Fires. The TDAFW The existing two magnetic speed pickups on the Auxiliary Feedwater (AFW) System will not adversely impact any of the turbine will be replaced with similar magnetic pickups and vendor supplied systems that have a dynamic interface with the cables to the existing TDAFW skid mounted terminal blocks. New speed TDAFW System. Namely: Condensate Storage sensor cabling and raceway from existing panel to the new governor panel will Tanks; Essential Raw Cooling Water (ERCW);
be installed for the two speed pickup signals. Main Steam; Feedwater; and 125 Volt DC Power Systems. Therefore, the modification The existing TDAFW Main Control Room (MCR) speed indicator and speed does not result in more than a minimum potentiometer position indicator on panel 1-M-3 will be replaced with new increase in the likelihood of occurrence of a indicators to accept a 4-20mA signal from the replacement system. The speed malfunction of a structure, system, or potentiometer position indicator will be re-labeled as a speed setpoint component (SSC) important to safety previously indication, and scaling will be unchanged. evaluated in the UFSAR.
The existing TDAFW MCR pump flow controller output indicator on panel 1-M-3 Performance requirements associated with core will be re-labeled and re-scaled. This indicator will now display the governor cooling are unaltered such that fuel integrity will flow setpoint and will be rescaled as 0-1000 GPM versus the existing 0-100%. be maintained and the UFSAR analysis of radiological consequences remains bounding.
The existing local governor demand signal isolator located in panel 1-L-381 is The TDAFW Systems ability to mitigate any deleted by this modification as its function is no longer required with the revised postulated design basis accidents will not be governor controls. decreased. The new equipment will not initiate any new accidents. The modification will not The existing TDAFW MCR flow indicator on panel 1-M-3 and Auxiliary Control impair or prevent the ECCS from mitigating the Room (ACR) flow indicator on panel 1-L-10 are unaffected by this modification. consequences of any design basis accidents.
Therefore, this activity does not result in more The flow control function of the existing Yokogawa controller will be than a minimum increase in the consequence of incorporated into the new TDAFW governor. The Yokogawa controller on panel an accident previously evaluated in the UFSAR.
1 L-381 will be replaced with a new standard Electroswitch W2 module. This new local control switch will be similar to the existing control switch located in Failure or malfunction of the new equipment will MCR panel 1-M-3. The new governor flow control setpoint will be adjustable not prevent or affect the ability of safety related locally and provide control functionally similar to the existing Yokogawa systems or systems important to safety to controller. respond to the accidents described in the UFSAR. Therefore, implementation of the The existing hand switch, 1-HC-46-57-S, located in MCR panel 1-M-3 will be proposed modification does not result in more replaced with a functionally similar switch. In MCR panel 1-M-3 the than a minimal increase in the consequences of replacement Electroswitch control switch will have Pull-to-Manual operation for a malfunction of an SSC important to safety E1-3 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES manual speed control, and an accident signal reset function in the pushed previously evaluated in the UFSAR. The position, identical to the existing control switch. A momentary spring return to potential malfunctions of the modified center Raise and Lower position while pulled will be used to provide for manual equipment are bounded at a system level in the control of the new governor speed setpoint when in the pulled position. The UFSAR. Therefore, the possibility for an existing accident signal reset functionality is maintained on the new hand switch unanalyzed malfunction of an SSC important to by either the left (Lower) or right (Raise) position while pushed similar to the safety or an accident of a different type than any existing hand switch. previously evaluated in the UFSAR are not created.
A similar Electroswitch control switch will be used in the local control panel.
This will allow for adjustment of the governor flow control setpoint while in Auto, As described in the UFSAR accident analysis, while the MCR control switch will only allow manual control of governor speed, no malfunction of the AFW System can cause a which replicates the existing functionality of the existing controls. Both the transient sufficient to damage the fuel barrier or MCR and local control switch will incorporate Red and Green indicating lights exceed the nuclear limits as required by the for Auto and Manual status indication, identical to the existing MCR control safety design basis. No new failure modes are station. Panel modifications to the local panel will be required to mount the new created by replacement of the TDAFW governor control switch. valve controller. The proposed modification does not adversely impact the technical The existing servo amplifier and vertically mounted hydraulic actuator, attributes supporting this conclusion. Therefore, associated oil tubing, and linkage assembly on governor valve, 1-FCV-1-52, will the possibility for an unanalyzed malfunction of be removed. The replacement actuator will be a horizontally mounted electrical an SSC important to safety or an accident of a linear actuator. The installation of this electrical actuator will require different type than any previously evaluated in replacement of the existing governor valve stem and the installation of an the UFSAR are not created.
actuator mounting bracket on the governor valve bonnet.
The new digital equipment does not necessitate The electronic over speed trip function currently provided by a tachometer is a revision or replacement of any currently used replaced by an overspeed trip output from the Woodward 505 governor. The evaluation methodology for the TDAFW System.
existing electronic overspeed trip annunciator in the MCR will be relabeled and The modification does not result in a departure repurposed to alarm on either an electronic overspeed trip or a generic TDAFW from the method of evaluation described in the control system trouble alarm. The existing 4300 revolutions per minute (RPM) UFSAR in establishing the design bases or in electrical overspeed trip setpoint is not changed by this modification. The the safety analyses.
existing mechanical overspeed trip and associated annunciator are unaffected by this modification. Guidance for evaluation of digital upgrades is contained in Nuclear Energy Institute (NEI) 01-01, Guideline on Licensing Digital Upgrades, March 2002. NEI discusses the use of a Failure E1-4 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES The existing function to transfer (force) the TDAFW flow control loop to Auto Modes and Effects Analysis (FMEA) and in (flow control) on either an Accident Signal, TDAFW pump high flow, or Main accordance with the graded approach Feed Pump Turbine trip is maintained with the replacement system. referenced in NEI 01-01, a detailed component level FMEA was developed to identify those The existing flow switch nominal setpoint of 985 gallon-per-minute (gpm) will be malfunctions important to safety which could be revised to 970 gpm to ensure the flow switch actuates within the flow loop caused by the digital controls.
range of 0-1000 gpm. The existing switch reset value of 1% (10 gpm) will be revised to be 2.5% (25 gpm) to provide for margin between the high flow This evaluation concludes that implementation setpoint and the nominal return to manual point. This revised logic will continue of the modification does not require a Technical to provide the required pump runout protection function in manual control. The Specification change, does not require a increased flow switch dead band will reduce the frequency of manual-to-auto License Amendment, and therefore may swaps on decreasing SG pressure. proceed without NRC approval.
A new ICS data acquisition device, as well as an associated 24VDC power supply and fuses will be installed on elevation 669 of the Auxiliary Building in a dedicated junction box. This new data acquisition device will be connected to the existing ICS Data Acquisition (DAQ) network. The new equipment will be powered from an existing 120VAC Vital Power Board 1-III/1-VI breaker through a new 1E isolation fuse to provide a 1E/Non-1E class break.
A new interposing relay will be installed to provide for voltage level segregation of 125VDC and 24VDC to the existing control room control switch, to avoid a new cable pull to the MCR. This relay will allow the use of existing 125VDC cable for the hand control Auto/Manual switch function, which is then input to the 24VDC circuitry via normally open (auto) contacts from the new relay. This relay will be energized for manual operation of the TDAFW governor.
New class 1E signal isolators will be installed to provide for 1E/non-1E isolation of analog signals from the replacement system to the ICS. These signals will be connected to the new ICS data acquisition device.
Spare relay contacts in the new governor panel will be utilized to provide for digital status signals from the replacement system to the ICS. Coil-to-contact isolation will be used to provide non-1E status signals. These signals will also be connected to the new ICS data acquisition device.
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DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES The new TDAFW control system is provided with local manual control capability at the TDAFW turbine requiring no electrical power (AC or DC) available by manual operation of the new governor valve actuator jacking screw. Although available, this Black Start capability is not credited for any design basis events. The governor valve can be manually positioned to control AFW speed with the actuator jacking screw, or the existing Black Start methodology of leaving the governor valve in the normal full open state and controlling AFW turbine speed by manually positioning the trip & throttle valve can be utilized.
Terminal blocks will be installed in the existing governor panel to allow connection of new cabling to the new governor panel for the two existing speed probes, the Trip and Throttle Valve trip solenoid, the ACR/MCR speed indication loop, and the Governor valve ramp initiate limit switch.
The existing junction box (JB) 5088 and associated raceway that contains the power supply voltage dropping resistor will be deleted by this change.
Two AFW indicators and associated cabling, raceway, and housing will be removed and their function incorporated into the new governor panel.
The functionality of the existing operator overspeed test controls will be incorporated into the new governor panel controls. The new actuator positioner panel will have no external indications or controls. Positioner demand and status signals will be connected back to the governor panel. A trip and throttle valve position signal is provided to the governor controller via the LS1 limit switch. As the governor valve unseats (not full closed) and engages the limit switch, the contact is made up and initiates the TDAFW governor start sequence. The output of this circuit will initiate a controlled startup of the governor control valve, which will minimize the potential for a possible electrical or mechanical overspeed trip. This limit switch signal, in parallel with a greater than 100 RPM speed contact from the 505 governor, will be used to enable the new actuator positioner. Enabling the actuator positioner will stroke the governor valve from its normally open/fail state of full open to the position demanded by the governor.
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DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES New cabling and raceway will be installed from and to certain panels.
The new TDAFW control panel and associated components are installed in a similar location as the existing TDAFW equipment which ensures that the separation requirements between the TDAFW and Motor Driven AFW systems are maintained per the stations design bases documented in the UFSAR and design criteria. The cables routed between the TDAFW skid and the new TDAFW control panels are routed in existing and new conduits between the two end points. This also ensures the new control panel remains in the same fire zones to maintain separation required for a 10 CFR 50 Appendix R fire event.
23085 Rev. A This design change removes the existing Kirk Key interlocking scheme from the Due to the response on question 2 regarding Evaluation feeder breakers and tie breakers for Essential Raw Cooling Water (ERCW) system diversity/separation, prior approval from Rev. 2 Motor Control Center (MCC) 1A-A, 1 B-B, 2A-A and 2B-B and replaces the the NRC is required for DCN D23085A. This existing exterior rotary style breaker handle and breaker mechanism on the modification is safe with respect to nuclear main and alternate feeds with a model that will properly function with the safety and is in compliance with SQN design replacement breakers. The existing ERCW Feeder Breakers are obsolete. The basis.
replacement breakers have been evaluated through the equivalency process, but the replacement breaker's physical footprint has slight variations which prevents the existing Kirk Key interlocking scheme to be mounted onto the breaker. This DCN is staged to replace the rotary handle and breaker mechanism as well as remove the Kirk-Key interlock on the ERCW MCC Feeder Breakers and tie breakers.
23638 Rev. A This DCN installs a new digital Distributed Control System (DCS) that will The new digital DCS system replaces existing Evaluation replace the existing obsolete analog control systems. The analog control analog components for Balance of Plant (BOP)
Rev. 2 components will be replaced with a digital control system and is also termed control systems and reduces many single point "DCS" for Distributed Control System. The new system will not only correct the failure vulnerabilities existing in the current obsolescence issue, but will also eliminate a multitude of single point failures analog system. System reliability is improved (SPFs) which will improve the reliability of the major control systems on through the use of automatic signal selection Sequoyah Nuclear Plant (SQN) Unit 1. The use of redundant power supplies, from multiple control signal inputs. The new redundant signal processors, and redundant signal paths make the system system provides redundant inputs, redundant more reliable from a hardware standpoint. The division of software processing processors, networks, and power supplies. The tasks across multiple control processors maintains functional diversity within the new system is designated as "Quality Related" E1-7 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES software. This DCN will replace balance of plant (BOP) analog controls and is designed to meet Quality Related associated with pressurizer pressure and level, rod control, pressurizer power requirements. The reliability of the DCS is operated relief valve (PORV) controls, Volume Control Tank (VCT) level superior to the old analog system. The control, low temperature over-pressure (LTOP), Boric Acid Blender control, modification does not negatively impact any Steam Generators (SGs) Atmospheric Relief Valve (ARV) pressure controllers, system, structure, or component (SSC) that is Steam Dump pressure and temperature controllers, Emergency Core Cooling important to safety nor does it adversely impact System (ECCS) Cold Leg Accumulator (CLA) Nitrogen Vent controller, RHR the consequences or the frequency of a heat exchanger flow controllers, SG Blowdown flow controller, Hotwell level malfunction. The new DCS does not create a dumpback and make-up controllers, Generator Hydrogen heat exchanger new type of malfunction or accident. The new temperature controller, and Main Turbine Oil Tank (MTOT) temperature control DCS reduces the likelihood of failures and their systems and makes other changes to eliminate many existing SPFs present in consequences by providing a more reliable and the existing analog system. redundant control system. In addition, this modification provides the capability to reduce In addition to the DCS installation, the Cold Overpressure Mitigation System manual operator actions and adds greater (COMS) settings implemented by this DCN are the setpoints for the opportunity for assessment, monitoring, and replacement Pressurizer PORVs that will be installed by a separate DCN. response.
The upgrade to DCS results in overall improvement in the plant and the ability to function with individual devices out of service.
The DCS provides for use of additional input signals for control. The DCS will continue to maintain function with the loss of a single input for control loops with multiple inputs. In the case of failure of a single input, the last good value prior to the failure will be used. The DCS will provide an alarm on the DCS Visual Display Unit (VDU) for failure of any input.
The DCS is powered from redundant power sources, one being from battery-backed Vital power boards, thus for loss of any single power source, the DCS will continue to maintain control.
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DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES The signal outputs to plant control devices, such as valves, use redundant Field Bus Modules (FBM) such that should one FBM fail the other FBM maintains control of the device. Reliability data based on the operational history of these systems in service throughout a wide range of industrial applications is extremely high. A SPF analysis was performed to evaluate the failure modes and effects of the new components including software failures. Many of the SPFs existing in the analog system were eliminated by the DCS implementation. Because the two redundant processors in a control group have the same software in common, the software is considered a SPF. As previously discussed, the software is designed to limit the likelihood of software errors that could result in common cause failure. The SPF analysis concluded the software SPF is acceptable based on thorough testing of the software, software control by an approved QA program, and an extensive operating history.
Functional diversity and independence are provided through the segmentation of control functions in different control groups and processor pairs. For example, the SG ARV controls are segmented such that each SG ARV is on a separate DCS processor pair.
Segmentation of control functions also limits the impact of software common cause failures. A segmentation analysis has been prepared to further analyze the separation of signal processing and control. The segmentation analysis concluded that the design of the DCS E1-9 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES does not introduce control system failures which could result in events not bounded by the UFSAR safety analysis or an event not analyzed in the UFSAR.
Therefore this modification may be implemented without obtaining a License Amendment.
10069 Rev. 0 The existing SQN Unit 2 Turbine Driven Auxiliary Feedwater (TDAFW) speed This evaluation has determined that the TDAFW Evaluation governor system will be replaced in its entirety by this modification. The system will continue to meet its design and Rev. 0 replacement governor system will replicate the functionality of the existing licensing bases requirements following the speed control and electrical overspeed trip components of the existing system, implementation of the proposed modification as well as incorporating the flow control function of the existing Yokogawa that converts to a digital governor control single-loop process controller (SLPC) located in panel 2-L-381. Incorporating method for the system.
the flow controls functions within the new governor system will simplify the control system and reduce the number of active components required to Since the new TDAFW System components are perform the TDAFW design function, as well as allowing the use of a standard more reliable than the existing components and replacement governor design across the Tennessee Valley Authority (TVA) no new system level failure mode effects are nuclear fleet. introduced, the proposed modification does not result in more than a minimum increase in the The new governor system is configured to accept the existing flow signal and frequency of occurrence of an accident will perform the flow control function of the existing 2-FIC-46-57 Yokogawa previously evaluated in the SQN UFSAR.
controller as a cascade control function within the new Woodward 505 governor. The Yokogawa controller, as well as the associated power supply The new equipment being installed will not and status relay will be deleted from panel 2-L-381. initiate any new system malfunctions. Credit is taken for the TDAFW System for the successful A new governor panel will be installed adjacent to the existing panel mitigation of the following UFSAR analyzed immediately outside the TDAFW room, and a new actuator positioner panel to events. The AFW system supplies, in the event house the actuator positioner (controller) and associated components will be of a loss of the main feedwater supply, sufficient installed on a common support on the rear of the new governor panel. The feedwater to the SGs to remove primary system existing governor panel located in the TDAFW room will be modified to serve as stored and residual core energy. It may also be a junction box (JB). A new panel will also be installed on the common support required in some other circumstances such as with to house the non-safety related Integrated Computer System (ICS) the evacuation of the MCR, cooldown after a interface programmable logic controller (PLC) and associated components. LOCA for a small break, maintaining a water The new governor panel, positioner panel, and governor valve actuator will be head in the SGs following a LOCA, a flood powered from the existing governor 125 V Battery Board I breaker 321 feed, above plant grade, ATWS event, and 10 CFR E1-10 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES with the existing alternate feed from 125V Battery Board II breaker 321, 50, Appendix R, Fires. The TDAFW System will identical to the existing TDAFW governor. The new governor system will result not adversely impact any of the systems that in an additional load on this circuit. have a dynamic interface with the TDAFW System. Namely: Condensate Storage Tanks; The existing two magnetic speed pickups on the Auxiliary Feedwater (AFW) ERCW; Main Steam; Feedwater; and 125 Volt turbine will be replaced with similar magnetic pickups and vendor supplied DC Power Systems. Therefore, the modification cables to the existing TDAFW skid mounted terminal blocks. New speed does not result in more than a minimum sensor cabling and raceway from the existing panel to the new governor panel increase in the likelihood of occurrence of a will be installed for the two speed pickup signals. malfunction of an SSC important to safety previously evaluated in the UFSAR.
The existing TDAFW Main Control Room (MCR) speed indicator and speed potentiometer position indicator on panel 2-M-3 will be replaced with new Performance requirements associated with core indicators to accept a 4-20mA signal from the replacement system. The speed cooling are unaltered such that fuel integrity will potentiometer position indicator will be re-labeled as a speed setpoint be maintained and the UFSAR analysis of indication, and scaling will be unchanged. radiological consequences remains bounding.
The TDAFW Systems ability to mitigate any The existing TDAFW MCR pump flow controller output indicator on panel 2-M-3 postulated design basis accidents will not be will be re-labeled and re-scaled. This indicator will now display the governor decreased. The new equipment will not initiate flow setpoint and will be rescaled as 0-1000 GPM versus the existing 0-100%. any new accidents. The modification will not impair or prevent the ECCS from mitigating the The existing local governor demand signal isolator located in panel 2-L-381 is consequences of any design basis accidents.
deleted by this modification as its function is no longer required with the revised Therefore, this activity does not result in more governor controls. than a minimum increase in the consequence of an accident previously evaluated in the UFSAR.
The existing TDAFW MCR flow indicator on panel 2-M-3 and Auxiliary Control Room (ACR) flow indicator on panel 2-L-10 are unaffected by this modification. Failure or malfunction of the new equipment will not prevent or affect the ability of safety related The flow control function of the existing Yokogawa controller will be systems or systems important to safety to incorporated into the new TDAFW governor. The Yokogawa controller on panel respond to the accidents described in the 2-L-381 will be replaced with a new standard Electroswitch W2 module. This UFSAR. Therefore, implementation of the new local control switch will be similar to the existing control switch located in proposed modification does not result in more MCR panel 2-M-3. The new governor flow control setpoint will be adjustable than a minimal increase in the consequences of locally and provide control functionally similar to the existing Yokogawa a malfunction of an SSC important to safety controller. previously evaluated in the UFSAR. The potential malfunctions of the modified E1-11 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES The existing hand switch, 2-HC-46-57-S, located in MCR panel 2-M-3 will be equipment are bounded at a system level in the replaced with a functionally similar switch. In MCR panel 2-M-3 the UFSAR. Therefore, the possibility for an replacement Electroswitch control switch will have Pull-to-Manual operation for unanalyzed malfunction of an SSC important to manual speed control, and an accident signal reset function in the pushed safety or an accident of a different type than any position, identical to the existing control switch. A momentary spring return to previously evaluated in the UFSAR are not center Raise and Lower position while pulled will be used to provide for manual created.
control of the new governor speed setpoint when in the pulled position. The existing accident signal reset functionality is maintained on the new hand switch As described in the UFSAR accident analysis, by either the left (Lower) or right (Raise) position while pushed similar to the no malfunction of the AFW System can cause a existing hand switch. transient sufficient to damage the fuel barrier or exceed the nuclear limits as required by the A similar Electroswitch control switch will be used in the local control panel. safety design basis. No new failure modes are This will allow for adjustment of the governor flow control setpoint while in Auto, created by replacement of TDAFW governor while the MCR control switch will only allow manual control of governor speed, valve controller. The proposed modification which replicates the existing functionality of the existing controls. Both the does not adversely impact the technical MCR and local control switch will incorporate Red and Green indicating lights attributes supporting this conclusion. Therefore, for Auto and Manual status indication, identical to the existing MCR control the possibility for an unanalyzed malfunction of station. Panel modifications to the local panel will be required to mount the new an SSC important to safety or an accident of a control switch. different type than any previously evaluated in the UFSAR are not created.
The existing servo amplifier and vertically mounted hydraulic actuator, associated oil tubing, and linkage assembly on governor valve 2-FCV-1-52 will The new digital equipment does not necessitate be removed. The replacement actuator will be a horizontally mounted electrical a revision or replacement of any currently used linear actuator. The installation of this electrical actuator will require evaluation methodology for the TDAFW System.
replacement of the existing governor valve stem and the installation of an The modification does not result in a departure actuator mounting bracket on the governor valve bonnet. from the method of evaluation described in the UFSAR in establishing the design bases or in The electronic over speed trip function currently provided by a tachometer is the safety analyses.
replaced by an overspeed trip output from the Woodward 505 governor. The existing electronic overspeed trip annunciator in the MCR will be relabeled and Guidance for evaluation of digital upgrades is repurposed to alarm on either an electronic overspeed trip or a generic TDAFW contained in NEI 01-01. NEI discusses the use control system trouble alarm. The existing 4300 RPM electrical overspeed trip of a FMEA and in accordance with the graded setpoint is not changed by this modification. The existing mechanical approach referenced in NEI 01-01, a detailed overspeed trip and associated annunciator are unaffected by this modification. component level FMEA was developed to E1-12 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES The existing function to transfer (force) the TDAFW flow control loop to Auto identify those malfunctions important to safety (flow control) on either an Accident Signal, TDAFW pump high flow, or Main which could be caused by the digital controls.
Feed Pump Turbine trip is maintained with the replacement system.
This evaluation concludes that implementation The existing flow switch nominal setpoint of 985 gpm will be revised to 970 gpm of the modification does not require a Technical to ensure the flow switch actuates within the flow loop range of 0-1000 gpm. Specification change, does not require a The existing switch reset value of 1% (10 gpm) will be revised to be 2.5% (25 License Amendment, and therefore may gpm) to provide for margin between the high flow setpoint and the nominal proceed without NRC approval.
return to manual point. This revised logic will continue to provide the required pump runout protection function in manual control. The increased flow switch dead band will reduce the frequency of manual-to-auto swaps on decreasing Steam Generator pressure.
A new ICS data acquisition device, as well as an associated 24VDC power supply and fuses will be installed on elevation 669 of the Auxiliary Building (AB) in a dedicated junction box. This new data acquisition device will be connected to the existing ICS Data Acquisition (DAQ) network. The new equipment will be powered from an existing 120VAC Vital Power BD 2-I/2-II breaker through a new 1E isolation fuse to provide a 1E/Non-1E class break.
A new interposing relay will be installed to provide for voltage level segregation of 125VDC and 24VDC to the existing control room control switch. This is necessary to avoid a new cable pull to the MCR. This relay will allow the use of existing 125VDC cable for the hand control Auto/Manual switch function, which is then input to the 24VDC circuitry via normally open (auto) contacts from the new relay. This relay will be energized for manual operation of the TDAFW governor.
New class 1E signal isolators will be installed to provide for 1E/non-1E isolation of analog signals from the replacement system to the integrated computer system. These signals will be connected to the new ICS data acquisition device.
Spare relay contacts in the new governor panel will be utilized to provide for digital status signals from the replacement system to the integrated computer E1-13 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES system. Coil-to-contact isolation will be used to provide non-1E status signals.
These signals will also be connected to the new ICS data acquisition device.
The new TDAFW control system is provided with local manual control capability at the TDAFW turbine requiring no electrical power (AC or DC) available by manual operation of the new governor valve actuator jacking screw. Although available, this Black Start capability is not credited for any design basis events. The governor valve can be manually positioned to control AFW speed with the actuator jacking screw, or the existing Black Start methodology of leaving the governor valve in the normal full open state and controlling AFW turbine speed by manually positioning the trip & throttle valve can be utilized.
Terminal blocks will be installed in the existing governor panel to allow connection of new cabling to the new governor panel for the two existing speed probes, the Trip and Throttle Valve trip solenoid, the ACR/MCR speed indication loop, and the Governor valve ramp initiate limit switch.
The existing JB 4848 and associated raceway that contains the power supply voltage dropping resistor will be deleted by this change.
Two AFW indicators and associated cabling, raceway, and housing will be removed and their function incorporated into the new governor panel.
The functionality of the existing operator overspeed test controls will be incorporated into the new governor panel controls. The new actuator positioner panel will have no external indications or controls. Positioner demand and status signals will be connected back to the governor panel. A trip and throttle valve position signal is provided to the governor controller via the LS1 limit switch. As the governor valve unseats (not full closed) and engages the limit switch, the contact is made up and initiates the TDAFW governor start sequence. The output of this circuit will initiate a controlled startup of the governor control valve, which will minimize the potential for a possible electrical or mechanical overspeed trip. This limit switch signal, in parallel with a greater than 100 RPM speed contact from the 505 governor, will be used to enable the new actuator positioner. Enabling the actuator positioner will stroke the E1-14 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES governor valve from its normally open/fail state of full open to the position demanded by the governor.
New cabling and raceway will be installed from and to certain panels.
The new TDAFW control panel and associated components are installed in a similar location as the existing TDAFW equipment which ensures that the separation requirements between the TDAFW and Motor Driven AFW systems are maintained per the stations design bases documented in the UFSAR and design criteria. The cables routed between the TDAFW skid and the new TDAFW control panels are routed in existing and new conduits between the two end points. This also ensures the new control panel remains in the same fire zones to maintain separation required for a 10 CFR 50 Appendix R fire event.
100096 Rev. 2 This engineering change package (ECP) installs a time delay relay (TDR) in the The C-2 interlock is designed to limit rod Evaluation actuation circuit for annunciator XA-55-4B window E-3, NC-46B Nuclear withdrawal and although not credited in the Rev. 2 Instrumentation System (NIS) Power Range Channel Deviation, to reduce analysis of the Uncontrolled Rod Cluster Control nuisance alarms. This change also increases the Power Range High Rod Stop Assembly Bank Withdrawal events described in setpoint (C-2) from 103% Reactor Thermal Power (RTP) to 105% RTP to UFSAR Sections 15.2.1 and 15.2.2 it is reduce nuisance alarms on XA-55-4B window D-3, IPRS NIS Power Range considered as an additional assurance that Overpower Rod Withdrawal Stop. In addition, this change also replaces reactor power will not exceed the analytical limit.
existing obsolete Quadrant Power Tilt Ratio (QPTR) time delay relays RLY-92-1 This modification increases the NIS Power and RLY-92-2 associated with annunciator XA-55-4B windows B-3 and C-3 with Range C-2 setpoint and results in a decrease in equivalent currently available TDRs. All three TDRs (new and two existing) will the margin to the NIS Power Range High Power be non-digital socket type allowing easy replacement. The TDRs will be Reactor Trip (109% RTP). However, based on energized and the normally open contact will be closed during normal the as-found tolerance for the associated operation. The contact will open to alarm. The design will cause an alarm if the bistables, sufficient margin remains between the power is removed. This arrangement is the same as the original contacts in the C-2 setpoint and the NIS Power Range high NIS equipment. The relays are non-safety related devices fed from a Class 1E power trip setpoint. This margin was further safety related board and are isolated by qualified fuses selectively coordinated determined to be acceptable by evaluation of with the upstream safety related breaker. The fuses and fuse holders are bistable test data which concluded that over a seismically qualified. data set of 143 setpoint checks, the setpoint was consistently found within the 0.6% RTP as-This ECP changes the NIS power range signal select strategy from high select left tolerance.
(highest of the four channels) to high median select (second highest signal).
The four channels of NIS power range signals are input to the plant DCS. The E1-15 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES DCS validates each of the inputs independently, then inputs the four power Based on this analysis and the responses to range signals to the signal select software block. The signal select block Questions 1 through 8 of this Evaluation, it is determines which signal to select, and outputs that signal to the DCS power concluded that an increase in the C-2 mismatch logic to determine the automatic rod control outputs (speed and interlock/alarm setpoint is within the design direction). This ECP modifies the DCS signal select software block to change basis of the plant and is acceptable.
the signal selection strategy from high select to high median select. This signal Additionally, vendor review of the impact of select strategy change does not affect NIS power range input validation or increasing the C-2 interlock/alarm setpoint otherwise affect the Automatic Rod Control System (ARCS). concluded that increasing the C-2 setpoint would not affect the accident analysis.
Based on the analysis detailed in this evaluation, the ARCS NIS power range signal select strategy change from high select to high median select is bounded by the current plant accident analysis, and may be implemented without prior NRC approval.
100117 Rev. 3 Design Equivalent Change (DEC) 100117 replaces the Auxiliary Building High The new digital Yokogawa DX2048 Advanced Evaluation Energy Line Break Temperature (AUX BLDG HELB TEMP) Recorder 1-TR series recorder provides the same design Rev. 1 810-A, which is installed on Main Control Room (MCR) Panel (PNL) 1-M-23A. functions as the existing recorder, the The existing recorder is a Yokogawa HR2400 digital recorder with a paper redundant, Safety Related, AUX BLDG HELB chart. The new recorder is a Yokogawa DX2048 Advanced series recorder. TEMP Recorder, 1-TR-30-820-B, is also a The new recorder is configured with a Liquid Crystal Display (LCD) and a similar digital Yokogawa DX2020 series Compact Flash Memory Card (CF card) for data storage and data retrieval. recorder and is located within the same PNL 1-M-23A. Therefore, this modification was This recorder receives 5 Temperature Element (TE) inputs from Resistance evaluated for the potential introduction of a Temperature Detectors (RTDs) located in RHR Pump Rooms 1A-A & 1B-B, hardware and/or software Common-Cause RHR Heat Exchanger Rooms 1A & 1B, and the U1 Chemical and Volume Failure (CCF). However, this evaluation Control System (CVCS) Letdown Heat Exchanger Room. The TEs are demonstrated that the possibility of a CCF is Environmentally Qualified (EQ) to operate 30 minutes after a Design Basis sufficiently low. Also, a failure of this recorder is Accident (DBA) but are not impacted by this DEC. detectable and diverse indication for HELB conditions within RHR Pump Rooms 1A-A & 1 The RTD input option for the DX2048 Advanced series recorder did not pass B-B, RHR Heat Exchanger Rooms 1A & 1 B, Electromagnetic Interference/Radio Frequency Interference (EMI/RFI) and the U1 CVCS Letdown Heat Exchanger requirements; therefore, each of the 5 RTD signals are converted to a 4-20mA Room exist.
signal by 5 new, analog RTD transmitters (XMTRs). The 4-20mA signal is then E1-16 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES processed through a 2500 dropping resistor at the input terminals of the Therefore, this modification may be recorder to generate a 1-5Vdc input to each qualified DC input channel of the implemented without obtaining a License recorder. Amendment.
This recorder initiates alarm 1-XA-55-6D Window 29 "AUX BLDG HIGH ENERGY LINE BREAK" when temperature setpoints are reached that trigger Operator Actions to isolate the RHR and/or CVCS due to a potential pipe rupture event. Alarm Response Procedures provide instructions to enter the appropriate abnormal operating procedure (AOP) to implement applicable isolation actions.
22499 Rev. A The Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 96-05 The screening review determined that the Evaluation to establish a periodic verification (PV) program to provide confidence in the increase in maximum stroke time limit portion of Rev. 1 long term capability of Motor Operated Valves (MOVs) to perform their design the proposed change is potentially adverse to basis safety functions. The final GL 96-05 program is based on compliance with the UFSAR described design function of the the Joint Owners Group (JOG) Topical Report MPR 2524-A and the NRC subject valves and requires evaluation under Safety Evaluation (SE) endorsing Report MPR 2524-A. Implementation of GL 10 CFR 50.59.
96-05 is required in accordance with issuance of the NRC Safety Evaluation (SE), with licensees required to notify the NRC of deviations in the program or It is has been determined that the increase in schedule. the maximum stroke time of the proposed changes do not result in the possibility of new DCN 22499A performs the following modifications on the subject MOVs in order accidents or malfunctions, and do not result in to comply with the GL 96-05 program: increased frequency of accidents or malfunctions evaluated in the UFSAR. The 2-FCV-62-91: The Charging Flow Isolation valve is a Velan 3-inch, Class 1500 changes do not result in more than minimal gate valve with a Limitorque SMB-00 actuator; the valve body is of stainless increases of the consequences of an accident steel construction, with butt-weld end connections. This valve has a larger seat or malfunction and do not result in an diameter than similar valves in the plant, which has led to issues with its unacceptable departure from methodologies capability. To address this condition, the double lead stem will be replaced with used to establish the design basis and safety a single lead stem. Additionally, the existing 25 ft-lb, 1800 RPM motor will be analysis. In addition, no fission product barriers replaced with a 15 ft-lb, 3600 RPM motor. With the stem pitch halved and design basis limits are exceeded or altered by motor speed doubled, the stroke time of the valve will remain unchanged. this change, and the Technical Specifications are not affected.
2-FCV-63-25, -26, -39, and -40: The Centrifugal Charging Pump Injection Tank (CCPIT) isolation valves are Anchor Darling 4-inch, Class 1500, Screw and Yoke gate valves with Limitorque SMB-0 actuators; these valves are made of E1-17 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES stainless steel with butt-weld end connections. The internals of these valves will Therefore, this activity may be implemented per be replaced (modified discs, upper and lower wedge set, stem, stem nut, plant procedures without obtaining a License wedge pin and replacement Electric Power Research Institute (EPRI) packing Amendment.
and bonnet gaskets) to ensure the valves' weak link capabilities have sufficient margin based upon the more rigid JOG guidelines. Additionally, the existing 15 ft-lb, 3600 RPM motors (with an Overall Gear Ratio (OAR) of about 89) will each be replaced with 40 ft-lb, 1800 RPM motors (with an OAR of about 39).
As a result, the Maximum Allowable Stroke Time (MAST) will be extended from 10 seconds to 25 seconds.
2-FCV-63-156 and -157: The Safety Injection Pump Isolation valves are Anchor Darling 4-inch, Class 1500, Screw and Yoke gate valves with Limitorque SMB-0 actuators, similar to the CCPIT valves. These valves are potentially subject to pressure-locking conditions, in which the inlet and outlet of the valve disc are depressurized with pressure remaining in the valve bonnet, preventing it from opening. To address the potential of this condition for these valves, this DCN will revise the pressure locking calculation and required thrust calculations to demonstrate that sufficient margin exists to ensure operation of the valves. No parts are replaced on these valves.
DCN SQN Design Change (DC) SQN-19-647 is a documentation only DC that resolves a In summary, the modification performed to the 647 Rev. 0 condition identified in Condition Report (CR) 1507644. This CR documented MFWP Hand Speed Changer impacted the Evaluation that during the Root Cause Analysis (RCA) investigation for Sequoyah's ability of the MFWPs to perform their design Rev. 0 (SQN's) April 14, 2019 Unit 1 trip it had been determined that the cause of the function following a loss of one MFWP. This trip was a modification performed upon the Main Feedwater Pump (MFWP) resulting degradation in pump performance turbine hydraulic control system coupled with an un-validated turbine runback potentially increases the reactor trip frequency set-point. The direct cause was determined to be the MFWPs' operating range by approximately 2.78%, which is less than have insufficient margin to ensure adequate Steam Generator (SG) level make minimal. This activity does not result in more up for the loss of a MFWP. The modification the RCA is referring to was an than a minimal increase in the frequency of increase in the Hand Speed Changer cup valve diameter size from 0.750 occurrence of an accident or likelihood of inches to 0.839 inches performed in the late 1980s. The cup valve diameter occurrence of a malfunction of an SSC change limited each pump's top end performance preventing it from being able important to safety that has been previously to fully perform its design function. As a consequence, the operating range of evaluated in the UFSAR. The activity does not the MFWP turbine was reduced such that it had insufficient margin to ensure result in any increase to the expected radiation adequate SG level make up for the loss of a MFWP. As stated in CR 1507644, releases or operator doses evaluated in the the cup valve modification was performed through a vendor manual change, UFSAR. The cup valve modification does not E1-18 of E1-33
DESIGN DESCRIPTION SAFETY ANALYSIS CHANGES which was not the appropriate process for such a change. As a consequence, result in more than a minimal increase in the the modification was not properly evaluated and not all impacted documents consequences of a malfunction of an SSC were identified and revised. This proposed activity resolves this legacy issue important to safety previously evaluated in the by formally evaluating and approving the modified Hand Speed Changer cup UFSAR. The modification also does not create valve. This screening review / evaluation also supports the revision to impacted the possibility for a new type of event that has Operations procedure AOP-S.01, "Main Feedwater Malfunction." not been previously evaluated in the UFSAR.
This activity does not have any impact upon the Abnormal Operating Procedure AOP-S.01 is revised to add steps for a pre- three fission product barriers described in the emptive reactor trip above a threshold power (95%) determined by Operations. UFSAR. Based on this evaluation, it is The current guidance in AOP-S.01 directs initiating a manual reactor trip if an acceptable for this activity to be performed automatic trip (on low-low SG level) is imminent. This effectively forces without prior NRC approval.
operators to wait until SG level is abnormally low (approaching the trip setpoint) prior to initiating the manual trip. Modifying this guidance to pre-emptively initiate a manual trip above a threshold power level is appropriate based on the following considerations:
- 1. This avoids an unnecessary delay in a (likely) inevitable reactor trip following a loss of one MFWP from full power.
- 2. This avoids a potential unnecessary reliance on the automatic reactor function if operators do not manually trip quickly enough.
- 3. This avoids an unnecessary reduction in SG level prior to the manual trip initiation.
TEMPORARY DESCRIPTION SAFETY ANALYSIS MODIFICATION (TMOD)
SQN-1-2019- The full-length Control Rod Drive (CRD) System maintains a programmed The Rod Urgent Failure Alarm affects the 085-002 Rev. 0 average reactor temperature compensating for reactivity effects associated with plant in Condition II - Faults of Moderate Evaluation scheduled and transient load changes. The shutdown rod banks along with the Frequency and specifically the Rev. 0 full-length control banks are designed to shut down the reactor with adequate Uncontrolled Boron Dilution events. The margin under conditions of normal operation and anticipated operational reactor maintains the function to trip due occurrences. to overpressurization in the unlikely event of an Uncontrolled Boron Dilution event.
On March 11, 2015 and again on August 27, 2019, the H-8 Control Rod in Control While the Rod Urgent Failure Alarm is Bank D unexpectedly dropped to the bottom of the reactor resulting in a negative active, Control Bank D control rods will E1-19 of E1-33
TEMPORARY DESCRIPTION SAFETY ANALYSIS MODIFICATION (TMOD) rate reactor trip. The change being evaluated mitigates the potential for dropping not unexpectedly drop resulting in a H-8 by energizing both the stationary and movable gripper coils of the Control negative rate reactor trip. Disabling the Rod Drive Mechanisms (CRDMs) for Group 2 Control Bank D rods in the 2BD Rod Urgent Failure Alarm will allow the power cabinet. Control Bank D control rods to be placed back in manual or automatic via a simple This change temporarily installs a momentary pushbutton at the terminal block operator action of pressing reset interface between the rod control logic cabinet and 2BD power cabinet. This pushbutton 1-RCAS.
pushbutton electrically shorts an AC voltage being sent from the logic cabinet to the power cabinet movable gripper circuitry. When the pushbutton is pressed and If the pushbutton contact fails closed, the held, closing the contact, the voltage will be shorted and the movable gripper Rod Urgent Failure Alarm will not be able circuitry will demand firing of the movable silicon controlled rectifiers (SCRs) to to be cleared by pressing pushbutton 1-generate full current. This will latch the movable grippers of the currently selected RCAS. This situation does not impact the group of rods into the CRDM driveshafts. Subsequently, a regulation failure will ability of the rods being inserted during a be generated within the 2BD power cabinet due to full current being generated for reactor trip. This change does not longer than allowed resulting in an urgent failure alarm. An urgent failure alarm in increase the frequency of any accident or a Rod Control power cabinet forces reduced current on the movable gripper coils malfunction evaluated in the UFSAR. It of the selected group of rods. Once this takes place, the Group 2 Control Bank D does not increase the consequences of rods will be held by both the stationary and movable grippers with both the any accident or malfunction. It does not stationary and movable gripper coils at reduced current. The pushbutton can create the possibility of a new type of then be released, opening the contact. This condition will prohibit movement of accident or a malfunction with a different Control Bank D in either automatic or manual as long as the urgent failure alarm result than previously evaluated. It does is active, although the rods are still able to drop upon a reactor trip. The urgent not alter the design basis limit for any failure alarm is capable of being reset by Operator Action by the Rod Urgent fission product barrier.
Failure Alarm reset pushbutton, 1-RCAS, on panel 1-M-4 in the Main Control Room (MCR). Once Operator Action reset is completed, the reduced moveable Therefore, this activity may be gripper current is eliminated, and the Rod Control System will again be fully implemented per plant procedures without capable of automatic or manual rod motion. The addition of the pushbutton adds obtaining a License Amendment.
a new failure mode of the contact failing closed when the pushbutton is released after inducing the urgent failure alarm. If the contact remains closed, the urgent failure alarm will not be able to be cleared by pushbutton 1-RCAS. This will prohibit movement of Control Bank D in either automatic or manual. The contact remaining closed would only be identifiable to the operators when pressing 1-RCAS does not clear the urgent failure alarm. This does not create a new failure effect since the purpose of the pushbutton is to simulate a condition that would bring in the urgent failure alarm, energizing the movable grippers along with the E1-20 of E1-33
TEMPORARY DESCRIPTION SAFETY ANALYSIS MODIFICATION (TMOD) stationary grippers, thus providing a double gripper hold. This does not have an impact on the rods being capable of dropping upon opening of the reactor trip breakers and safely shutting down the plant.
SQN-2-2018- This Temporary Modification (TMOD) is being prepared for use in the event of The valves are not able to initiate any 067-001 Rev. 1 inoperability of motor-operated valve (MOV) 2-FCV-67-146. The valve 2-FCV UFSAR accident, thus do not increase the Evaluation 146 regulates Essential Raw Cooling Water (ERCW) discharge from the 2A1 and frequency of an existing UFSAR Rev. 0 2A2 Component Cooling System (CCS) heat exchangers, and affects ERCW flow evaluated accident.
to other ERCW loads, including the 2A Containment Spray (CS) heat exchanger.
For control of CCS HX 2A1 and 2A2 In this 50.59 Screening Review/Evaluation, use of "2-67-146" references MOV 2- ERCW flow this TMOD activity changes FCV-67-146 and "2-67-551" references manual valve 2-VLV-67-551. Valves 2- from a remote manual action (2-67-146)67-146 and 2-67-551 control ERCW flow through the 2A1 and 2A2 CCS heat to a local manual action (2-67-551). This exchangers (HX), and are located in the ERCW outlet piping, prior to Discharge local manual action has been evaluated, Header A. Valves 2-67-146 and 2-67-551 are in series in the common ERCW and does not result in more than a outlet piping section (after the individual HX outlet piping joins). minimal increase in the likelihood of malfunction of a system, structure, or The 2A1 and 2A2 CCS HXs are supplied by ERCW Supply Header 2A (ERCW component (SSC) important to safety.
Train A). Flow from the ERCW Supply Header 2A also supplies the 2A CS HX, This activity does not affect the and other plant cooling loads. During some plant accidents and events, the 2 radiological consequences of any UFSAR 146 valve is throttled to ensure correct ERCW flow balance to the accident unit's evaluated accident or the radiological CCS and CS HXs. consequences of malfunction of an SSC important to safety.
Valve 2-67-146 is an MOV that is controlled from Main Control Room (MCR) handswitch 2-HS-67-146A. The handswitch has four (4) positions: Close, the All possible valve malfunctions (full open, "35" position, the "50" position, and Open. The 2-67-146 valve position is full closed, or midposition) are fully changed by remote manual operator action during some plant events and bounded by the existing UFSAR analysis.
accidents to adjust the balance of ERCW flow between the non-accident and This activity does not create the accident unit. In the event of a Unit 1 accident the 2-67-146 valve is set to "35" possibility of an accident of a different position (ensures adequate flow to the accident unit). In the event of a Unit 2 type than any previously evaluated in the accident the 2-67-146 valve is set to the "50" position (ensures adequate flow to UFSAR or the possibility of a malfunction the 2A1/2A2 CCS HXs and other ERCW loads, including the 2A CS HX). The of an SSC important to safety with a ERCW system flows are modeled and adequate ERCW flow to safety-related different result.
components are determined by calculations. The "35" and "50" positions of 2 E1-21 of E1-33
TEMPORARY DESCRIPTION SAFETY ANALYSIS MODIFICATION (TMOD) 146 are verified, and the valve limit switches are set, in procedure 0-PI-SFT-067- This activity does not affect any fission 006.0, ERCW Performance Testing. product barriers or UFSAR methods of evaluation.
Valve 2-67-551 is a manual valve that is set to a throttling position and functions as a balancing valve for ERCW flow through 2A1 and 2A2 CCS HXs. Valve 2 Therefore, this activity may be 551 is not modulated for any plant events. The throttled position of 2-67-551 is implemented per plant procedures without determined in procedure 0-PI-SFT-067-006.0 and verified in procedure 0-SI-OPS- obtaining a License Amendment.
067-682.M, ERCW Flow Balance Valve Position Verification.
Valves 2-67-146 and 2-67-551 have the same hydraulic performance characteristics. The valves are located approximately twenty feet apart in the same 24" pipe. Therefore the pressure drop and associated ERCW flow rates in the CCS HX 2A1 and 2A2 discharge line remain the same independently of which valve is in the fixed throttle position and which valve is positioned for the plant event.
In the event of valve 2-67-146 being inoperable, the functions of the two valves are switched by this TMOD. The 2-67-146 valve will be set to a fixed throttled position (equivalent to the normal balancing position of 2-67-551) and power will be removed by opening the breaker. The 2-67-551 valve will be manually modulated to the required position for plant events and accidents. Inoperability of the 2-67-146 valve potentially renders the 2A1 and 2A2 CCS HXs inoperable and may cause the 2A CS HX to be inoperable. Inoperability of the 2-67-146 valve also has the potential to affect the operability of the ERCW system.
The 2-67-146 valve is removed from the active valve list, and the 2-67-551 is added to the active valve list.
This activity screens in for evaluation due to the change from remote manual action to control 2-67-146 to local manual action to control 2-67-551. The change to local manual action adversely affects how FSAR described design functions are performed or controlled: additional time is required to operate 2-67-551 locally; operators are potentially exposed to increased dose during the local manual action; potential for human error; isolation of ERCW during an ERCW E1-22 of E1-33
TEMPORARY DESCRIPTION SAFETY ANALYSIS MODIFICATION (TMOD) pipe rupture; and reliance on a local action without adequate permanently-installed lighting.
SQN-2-2019- Temporary modification (TMOD) SQN-2-2019-063-001 installs a temporary The temporary relief line is acceptable 063-001 Rev. 1 continuous pressure relief line from the Safety Injection System (SIS) cold leg from a design, fabrication, and testing, Evaluation discharge piping to the suction side of the Centrifugal Charging Pumps (CCPs). standpoint. The line will be fully qualified Rev. 0 to TVA Class B (ASME B&PV Section Ill This TMOD is needed due to Reactor Coolant System (RCS) primary and Class 2) requirements like the existing secondary check valve leakage pressurizing the SIS cold leg discharge piping. portions of ECCS. The continuous relief This condition was documented by CR# 1472802 following the Unit 2 Cycle 22 line only has the passive function of refueling outage (U2R22). A troubleshooting work order determined that the leak maintaining a pressure boundary in the path is through check valves 2-VLV-63-557 and -563. The negative effect of the event of an accident. During normal SIS discharge piping pressurizing is that the relief valves in the line will lift at 1750 operation it has the functions to maintain psig. Once, the relief valves lift, it is possible for them to be problematic getting pressure boundary, prevent RWST them to reseat. There are also negative effects from having to vent the SIS inventory loss and back flow from CVCS.
discharge piping frequently. Currently, Operators are having to manually vent the No new accident initiating events are piping about every 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> using procedure 2-SO-63-5, Emergency Core created by this change, and there is no Cooling System, when header pressure increases to about 1600 psig. The increase to the frequency of previously venting requires stroking flow control valves (FCVs), 2-FCV-63-23 and -84. There evaluated accidents. There is no is a chance that the frequent stroking of the valves could cause degradation increase in the likelihood of a malfunction resulting in a valve failure. Also, having to vent frequently is a burden to the due to this change. If the line were to fail Operations Department. In the event that one of the FCVs fails, procedure 2-SO- it does not increase the consequences of 63-5 has contingency actions to vent the header using alternate methods. any accident described in Chapter 15 of the UFSAR. This change does not create The relief line will connect to the flange downstream of 2-VLV-63-539 on the SIS the possibility of an accident not already side and the 1" pipe nipple coming from 2-VL V-62-765 on the Chemical and described by the UFSAR. This change Volume Control System (CVCS) side. The relief line will be connected to each will not result in the design basis limit for end of the existing piping with a flex hose to reduce the impact to existing seismic any fission product barrier being qualification. Two 1/4" globe valves will be installed to allow for isolation as exceeded or altered.
needed. A 1/2" check valve will be installed in the line as well. The check valve will serve the dual purpose of preventing the CCP suction flow from being Therefore, it is acceptable to implement diverted and prevent the possibility for RWST water draining into the CCP suction this modification per plant procedures during normal operation. This is due to the RWST head pressure and normal without obtaining a License Amendment.
Volume Control Tank (VCT) pressure being approximately equal. The check valve is a spring loaded piston check with an adjustable cracking pressure. The cracking pressure will be set at its maximum cracking pressure of about 100 psid, E1-23 of E1-33
TEMPORARY DESCRIPTION SAFETY ANALYSIS MODIFICATION (TMOD) per the manufacturer's drawing, to maintain the SIS discharge header at a pressure greater than RWST head pressure. All new material will be TVA Class B (American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section Ill Class 2).
ECCS is impacted by this change due to the fact that the modification will create a new flow path in the SIS. The diverted flow will be a reduction in flow from the SIPs providing core cooling. The modification also impacts the CVCS because it has the ability to cause a negative reactivity event.
WOs 116644794 The full-length Control Rod System maintains a programmed average reactor The Rod Urgent Failure Alarm only affects and 120712590 temperature compensating for reactivity effects associated with scheduled and the plant in the Condition II - Faults of Evaluation Rev. 3 transient load changes. The shutdown rod banks along with the full-length control Moderate Frequency and specifically the banks are designed to shut down the reactor with adequate margin under Uncontrolled Boron Dilution events. Since conditions of normal operation and anticipated operational occurrences. On the reactor will maintain the function to trip March 11, 2015, the H-8 Control Rod in Control Bank D unexpectedly dropped to due to overpressurization, in the unlikely the bottom of the reactor resulting in a negative rate reactor trip. The change event of a Uncontrolled Boron Dilution being evaluated mitigates the potential for dropping H-8 by energizing both the event. Engaging this alarm ensures the stationary and movable gripper coils of the Control Rod Drive Mechanisms Control Bank D control rod will not (CRDMs) for all Group 2 Control Bank D rods in the 2BD power cabinet. This unexpectedly drop resulting in a negative change momentarily installs a jumper at the terminal block interface between the rate reactor trip. Disabling the Rod Urgent rod control logic cabinet and 2BD power cabinet. This jumper electrically shorts failure Alarm will allow the Control bank D an AC voltage being sent from the logic cabinet to the power cabinet movable control rods to be placed back in Manual gripper circuitry; this voltage is approximately 12 VAC during normal operation to or the preferred setting of automatic and is demand zero movable gripper current. When the jumper is installed the voltage a simple operator action.
will be shorted, and the movable gripper circuitry will demand firing of the movable silicon controlled rectifiers (SCRs) to generate full current; this will latch the Based on the response to all Evaluation movable grippers of the currently selected group of rods into the CRDM drive questions being "no", this activity may be shafts. Subsequently, a regulation failure will be generated within the 2BD power implemented per plant procedures without cabinet due to full current being generated for longer than allowed resulting in an obtaining a License Amendment.
urgent alarm. An urgent alarm in a Rod Control power cabinet forces reduced current on the movable gripper coils of the selected group of rods. Once this takes place, the Group 2 Control Bank D rods will be held by both the stationary and movable grippers with both sets of coils at reduced current. The jumper will then be removed. This condition will prohibit movement of Control Bank D in E1-24 of E1-33
TEMPORARY DESCRIPTION SAFETY ANALYSIS MODIFICATION (TMOD) either automatic or manual as long as the urgent alarm is active. The urgent alarm is capable of being reset by Operator Action by the Rod Urgent Failure Alarm reset push button in the Main Control Room. Once Operator Action reset is completed, the reduced movable gripper current is eliminated, and the Rod Control System will again be fully capable of automatic or manual rod motion.
Additionally, including the urgent alarm has no impact on the rods being capable of dropping upon opening of the reactor trip breakers and safely shutting down the plant.
ACCEPT AS-IS DESCRIPTION SAFETY ANALYSIS Accept As-Is for A volume knob was noted to be missing from one of the communications headsets The evaluation concludes that the CR 1560796 in use by the Unit 1 Refuel Floor (RFF) crew in containment. Due to the unknown postulated foreign material left in the Evaluation Rev. location of the knob and time it was lost during refueling (after the fuel was back in Unit 1 Reactor Vessel or Upper 0 the core), this Evaluation will consider the knob is either located somewhere in Containment at the restart from the U1R23 upper containment, the equipment pit, or has made its way into the open reactor. outage will not adversely affect nuclear safety. The knob is expected to break Therefore, the proposed activity is to allow the communications headset knob to apart into pieces such that its size is of no remain inside Unit 1's Upper Containment or Reactor Coolant System (RCS). concern, become trapped somewhere that According to site procedures foreign material in an SSC that is not retrieved is it cannot cause any credible failure such considered an "accept-as-is" disposition of a degraded or nonconforming condition as below the lower core plate or lower tie and must have 10 CFR 50.59 applied. plate, within a heat exchanger or filter, or melt and become diluted into the bulk SSCs that could be affected by this change are the RCS, CVCS, RHR, and the RCS. If the knob remains a solid, it lacks Containment Sump Sub-System of ECCS and components within each system. the volume, mass, or rigidity to cause any This Screening Review/Safety Evaluation (SR/SE) postulated all credible locations physical damage by impacting any for where the foreign material could travel within these systems. This activity was component before becoming trapped.
screened in because the presence of this item in the reactor vessel or Upper There is no postulated location the knob Containment has the potential to change RCS and/or ECCS chemistry, increase risk could become trapped that would prevent of corrosion or damage to any RCS, RHR, CVCS or ECCS component and fuel any design function from being performed.
cladding. The change also has the potential to result in a Steam Generator (SG) or If the knob melts, based on the low RHR Heat Exchanger tube being blocked. The effects of this activity on the SSCs concentration associated with this size of important to safety within these systems were evaluated and whether they could foreign material, no measurable effects E1-25 of E1-33
ACCEPT AS-IS DESCRIPTION SAFETY ANALYSIS perform their intended design functions. Due to the potential to impact the nuclear are expected on RCS water chemistry.
fuel, Framatome was contacted to provide additional input on this aspect of the There will also be no effect on the fuel due activity and effects to reactor coolant chemistry due to potential degradation of the to chemical attack. Therefore this activity knob. does not result in more than a minimal increase in the frequency of occurrence of an accident or likelihood of occurrence of a malfunction of an SSC important to safety that has been previously evaluated in the UFSAR. The activity also does not result in any increase to the expected radiation releases or operator doses evaluated in the UFSAR. This activity does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR. The modification does not create the possibility for a new type of event that has not been previously evaluated in the UFSAR. This activity does not have any impact on the three fission product barriers described in the UFSAR. Question 8 was not required to be addressed. Based on this evaluation, it is acceptable for this activity to be accepted as-is without prior NRC approval.
Accept As-Is for The proposed change is an accept-as-is evaluation for Condition Reports (CRs) This evaluation concludes that it is CRs 1601718 1601718 and 1602267. The condition being evaluated is the possibility of foreign acceptable for the debris items identified in and 1602267 material being left either in the Unit 2 Reactor Vessel or Reactor Cavity at the CRs 1601718 and 1602267 to be left in Evaluation conclusion of Unit 2 Refueling Outage 23. the Unit 2 Reactor Vessel at the restart Rev. 0 from the U2R23 refueling outage. The As documented in CR 1601718, a broken bolt was discovered on the power cord responses to the 10 CFR 50.59(c)(2) outrigger of the Unit 2 refuel machine auxiliary hoist during inspection. One half of determined that there is no increase in the the 1/4"-20 bolt was recovered, but the remaining half of the bolt and attached nut frequency of an accident or likelihood of a were unable to be located. At the time the broken bolt and nut were discovered to malfunction; there is no increase in the be missing, the Reactor Vessel head had been removed with the upper internals consequences of an accident or E1-26 of E1-33
ACCEPT AS-IS DESCRIPTION SAFETY ANALYSIS remaining in place. Thus, the item is either in the equipment pit, the reactor cavity, malfunction; there is no possibility of a on the upper reactor internals, or in the open reactor. On April 15, 2020, a new accident; there is no possiblity of a submarine was deployed in an attempt to locate the broken bolt / nut. The sub malfunction with a different result; and searched the upper cavity, equipment pit, and upender area but was unable to there are no changes to any design basis locate the items. Additional inspections over the course of the refueling outage limits for fission product barriers.
include: visual inspection of upper internals for the debris, four-face inspection of each fuel assembly, inspection of the lower core plate, and inspection below the Therefore, this change may be accepted lower core plate. The broken bolt / nut was not located. as-is without obtaining a License Amendment.
It was documented in CR 1602267 that two paint chips were discovered during a lower core plate inspection. One paint chip was knocked away before determining its location, and the paint chip was not able to be located again. The other paint chip is near core location K-10. It was stated in CR 1602267 that Westinghouse did not have the equipment available at the time to retrieve or attempt to retrieve the paint chips.
PROCEDURE DESCRIPTION SAFETY ANALYSIS ES-1.3 Rev. 24 & E-2 The proposed change consists of revisions to two Emergency Operating The proposed changes modify the EOP Rev. 18 Procedures (EOPs) in response to CR 1496213, which identified a conflict guidance which applies after a MS line Evaluation Rev. 0 between the existing EOP guidance and the UFSAR-described analysis for a break accident has already occurred with main steam line break with a concurrent or consequential rupture of the a concurrent or consequential RWST RWST. UFSAR Section 6.2 states that the RWST basin (moat) ensures that rupture. These changes cannot cause a minimum of 20,000 gallons of borated water is maintained if the tank is another accident and will not increase the ruptured by a tornado. As described in the Prompt Determination of frequency of any accident evaluated in the Operability (PDO) for this CR, the same storage capacity is credited if the UFSAR. The impact of the proposed EOP RWST is ruptured by pipe whip following a main steam line break in the yard changes has been evaluated; these area near the RWST. The PDO clarified that, in addition to a tornado, a changes do not result in a more than seismic event or postulated failure of the non-safety related main steam (MS) minimal increase in the likelihood of a piping outside of containment may result in loss of RWST in parallel with a malfunction in ECCS or any other UFSAR-MS line failure. The problem is that the credited 20,000 gallon water volume described function. The proposed following an RWST rupture is less than the volume corresponding to the low- changes do not increase the radiological low level setpoint (8%), at which existing EOPs require Emergency Core consequences of any accident, including a Cooling System (ECCS) pumps to be stopped to prevent loss of suction. design basis MS line break, nor do they Also, a postulated RWST rupture may result in loss of all the tank level increase the consequences of any instruments due to submergence of the level transmitters when the RWST malfunction to systems, structures, or E1-27 of E1-33
PROCEDURE DESCRIPTION SAFETY ANALYSIS basin floods. Based on the PDO analysis, the indicated RWST level is components (SSCs) important to safety.
expected to fail high when the transmitters are submerged. However, if one The proposed changes cannot create the or more of the RWST level transmitters remains available, operators could possibility for any different type of accident take action to stop all running ECCS pumps in response to low indicated nor any unanalyzed combination of level. This could make the credited 20,000 gallon water volume unusable accidents. The proposed changes do not with the existing EOP guidance. EOP changes are needed to address these create the possibility of a malfunction of plant-specific concerns. ECCS nor any other SSC important to safety with a different result. These EOP ES-1.3, Transfer to Residual Heat Removal (RHR) Containment Sump, changes do not impact any design basis provides instructions for transferring the ECCS and CS system to the sump limits for the RCS, reactor fuel, or recirculation mode. This EOP is entered when RWST level reaches the low containment. These changes have no level setpoint (27%) and is based on the Westinghouse Owners Group impact on evaluation methodologies (WOG) Emergency Response Guideline (ERG), with various deviations described in the UFSAR.
documented in the plant-specific basis document. The steps in this EOP are also based on UFSAR tables which summarize the plant-specific sequence of Therefore, this activity may be actions for sump recirculation. The following changes are incorporated in ES- implemented per plant procedures without 1.3: obtaining a License Amendment.
- Revised Section 1.0 (Purpose) to clarify the applicability of the plant-specific time critical action (TCA) to stop all pumps taking suction from the RWST within 1.5 minutes after RWST level reaches 8%. This section contains an informational table which lists all of the TCAs in ES-1.3. The table row for stopping ECCS and CS pumps within 1.5 minutes is modified to state that this action does not apply during a RWST rupture with the Reactor Coolant System (RCS) intact.
- Revised the Foldout Page (FOP) action which stops the ECCS pumps (with suction aligned to the RWST) and CS pumps when tank level is less than or equal to 8%. The FOP conditional action is modified to designate that it does NOT apply during an RWST rupture with the RCS intact. The existing FOP action is a deviation from the generic WOG ERG, which contains a caution warning operators that any pumps taking suction from the RWST should be stopped if RWST level lowers to setpoint U03; the generic caution was previously converted to a FOP action to improve usability and to comply with the SQN EOP Writer's Guide. The proposed change to the FOP action expands the scope of the existing deviation by exempting this action during E1-28 of E1-33
PROCEDURE DESCRIPTION SAFETY ANALYSIS an RWST rupture with the RCS intact. This deviation is justified based on the plant-specific RWST design vulnerability described above.
- Revised Step 3 (CHECK sump recirculation capability) to incorporate new contingency actions if containment sump level is less than the required setpoint (11%, setpoint T93) due to an RWST rupture with the RCS intact (containment conditions normal). The existing response not obtained (RNO) action directs operators to resume implementation of any applicable Function Restoration Procedures (FRPs) and to transition to Emergency Condition Action Procedure ECA-1.1, Loss of ECCS Sump Recirculation. ECA-1.1 is intended for beyond design basis accidents involving a loss of coolant accident (LOCA) with a loss of both trains of ECCS recirculation capability; ECA-1.1 is not the desired/optimal mitigating strategy for a design basis steam line break with the RWST ruptured because it may delay ECCS termination. If ES-1.3 was entered due to an RWST rupture with the RCS intact (based on containment conditions normal), the revised RNO action directs performing new Appendix B, Contingency Actions for RWST Rupture with RCS Intact. Appendix B provides information and guidance on the potential effects on the RWST of a steam line break in the yard area. This appendix directs ensuring that only one CCP and no more than one Safety Injection (SI) pump are running to conserve the remaining RWST inventory and to reduce the vulnerability to air entrainment. The CCP and SI pump steps are worded differently because it is possible to enter ES-1.3 Appendix B with no SI pump running (if SI pumps were stopped in ES-1.1 prior to the initiation of switchover); therefore, the SI pump step directs stopping one pump if both are running. If indicated RWST level is less than the empty setpoint (U03) or less than two wide range RWST level instruments are available, this appendix conservatively requires monitoring the running ECCS pumps for signs of cavitation and loss of suction.
At least two RWST level channels are required in this step to ensure that operators can confirm a valid tank level indication through comparison of the available indications and considering that Technical Specifications (TS) only allow one of four RWST level instrument to be inoperable. During a steam line break with a ruptured RWST, air entrainment resulting in cavitation or loss of suction is NOT expected based on the analyzed water volume. The E1-29 of E1-33
PROCEDURE DESCRIPTION SAFETY ANALYSIS changes described above are a new deviation from the generic WOG ERG but are justified based on the plant-specific RWST design vulnerability described above.
Condition Report These revisions are required in response to CR 1521279. This CR identified The proposed changes modify the 1521279 - that plant procedures currently direct or allow establishing ERCW flow to the procedural guidance for aligning ERCW EA-67-8 Rev. 5 station air compressors (SACs) by opening motor-operated isolation valve 0- flow to the SACs; these changes cannot EA-32-2 Rev. 5 FCV-67-205 (Train A) or 0-FCV-67-208 (Train B) following a loss of offsite cause another accident and will not 0-SO-32-1 Rev. 106 power (LOOP) or other conditions in which the normal cooling water supply to increase the frequency of any accident 1-SI-OPS-000-002.0 the SACs (from non-safety related Raw Cooling Water [RCW]) is lost. evaluated in the UFSAR. The impact of Rev. 127 Establishing ERCW to the SACs allows restarting the air compressors to the proposed procedure changes has 2-SI-OPS-000-002.2 restore or maintain non-essential control air pressure. Non-essential control been evaluated; these changes do not Rev. 115 air is needed for the CVCS letdown flowpaths and to allow remotely result in an increase in the likelihood of a AOP-M.02 Rev. 25 controlling charging flow to prevent pressurizer overfill. Design documents malfunction of ERCW or any other AOP-M.05 Rev. 11 currently state that the maximum ERCW temperature with flow established to supported UFSAR-described function.
AOP-M.08 Rev. 6 the SACs is 82°F (for Train A) or 85°F (for Train B). These temperature limits The proposed changes do not increase AOP-N.04 Rev. 12 exist because some of the safety-related components (such as Main Control the radiological consequences of any AOP-P.01 Rev. 41 Room [MCR] and Electric Board Room [EBR] chillers) may have inadequate accident, nor do they increase the Evaluation Rev. 0 flow if ERCW flow is established to the air compressors above the applicable consequences of any malfunction to temperature limit. systems, structures, or components (SSCs) important to safety. The proposed Current procedures identify these design output temperature limits but allow changes cannot create the possibility for operators to continue with establishing ERCW flow to the SACs while entering any different type of accident nor any the applicable TS action for the affected ERCW train. Therefore, if off-site unanalyzed combination of accidents. The power or RCW is lost with ERCW temperature above the applicable limit, proposed changes do not create the current procedures may result in operators taking a deliberate action which possibility of a malfunction of ERCW nor makes one train of safety-related systems/components inoperable. This any other SSC important to safety with a condition potentially violates the single failure criterion because a postulated different result. These EOP changes do single failure on one ERCW train (either before or after the operator action to not impact any design basis limits for the align ERCW flow) could result in the second (remaining) train of safety-related RCS, reactor fuel, or containment. These equipment being inoperable. Note that when a TS Limiting Condition for changes have no impact on evaluation Operation (LCO) condition/action which limits the duration of continued plant methodologies described in the UFSAR.
operation is entered due to one train of equipment inoperable, a subsequent single failure on the remaining operable train is normally NOT required to be Therefore, this activity may be postulated. However, suspending the single failure criterion when ERCW is implemented per plant procedures without aligned to SACs is NOT appropriate because the inoperability occurs due to a obtaining a License Amendment.
preplanned action, which applies anytime offsite power or RCW is lost. Also, E1-30 of E1-33
PROCEDURE DESCRIPTION SAFETY ANALYSIS this involves deliberately entering an LCO action for operational convenience (to allow restoring non-safety related CVCS functions), which is contrary to the bases for LCO 3.0.2. This condition was evaluated in the Prompt Determination of Operability (PDO) for CR 1521279. This PDO included a new interim analysis that justified that ERCW flow may be supplied to the SACs from either train up to a maximum ERCW supply temperature of 85°F.
The PDO states that this interim temperature limit is applicable with the ERCW yard header crosstie, the 6 ESF header crosstie, and the 16 Auxiliary Building header crosstie open or closed, in any combination. The PDO requires revising affected procedures to incorporate this modified temperature limit.
These changes have the following potential adverse impacts:
- This interim limit deviates from the current temperature limits specified in design output documents and is designated as being applicable regardless of the status of the ERCW header cross-ties. These changes have potential adverse impacts on ERCW design functions (reduced system cooling capacity).
- The proposed revisions also incorporate more restrictive guidance which may prevent establishing ERCW flow to the SACs if temperature is above the interim temperature limit. This additional restriction may prevent restoring non-essential control air pressure which, subsequently, would prevent restoration of a CVCS letdown flowpath and may challenge control of RCS inventory (pressurizer level).
Core DESCRIPTION SAFETY ANALYSIS Operating Limits Report SQN Unit 1 This Screening Review/Evaluation addresses the Sequoyah (SQN) Unit 1 Cycle 24 The removal of the H-08 RCCA does not Cycle 24 Core (S1C24) reload core design and Core Operating Limits Report (COLR) needed to impact the frequency or likelihood of an Operating support operation in all modes to a core average cycle burnup of 21,214 Megawatt- accident or malfunction because an RCCA Limits Report Days per Metric Ton of Uranium (MWD/MTU). All of the fuel assemblies loaded for is being removed and can no longer fail.
(COLR) S1C24 are Framatome Adv. W17 HTP design. This is unchanged from the previous There are no dose consequences cycle. associated with the removal of an RCCA.
E1-31 of E1-33
Core DESCRIPTION SAFETY ANALYSIS Operating Limits Report Evaluation Rev The accidents analyzed in the UFSAR are 0 The H-08 rod cluster control assembly (RCCA) is being removed for Unit 1 Cycle 24 independent of the number of control rods.
to prevent a dropped rod during the cycle. The removal of the H-08 RCCA The cycle-specific nuclear design adversely impacts the design function of the RCCA by reducing the trip reactivity parameters with the H-08 RCCA removed worth and available shutdown margin. The removal of the H-08 RCCA also were confirmed to be bounded by the increases bypass flow which adversely impacts the thermal and hydraulic design of conservative input values used in the the reactor core. The reactor coolant boundary is not modified with the removal of UFSAR accident analyses. Therefore no the H-08 RCCA. design basis limits were violated, and no malfunctions with a different result are possible.
Therefore, this activity may be implemented per plant procedures without obtaining a License Amendment.
SQN Unit 2 This Screening Review/Evaluation addresses the Sequoyah (SQN) Unit 2 Cycle 24 The core bypass flow increase does not Cycle 24 COLR reload core design and Core Operating Limits Report (COLR) needed to support require obtaining a License Amendment Evaluation Rev operation in all modes to a core average cycle burnup of 21,562 MWD/MTU. based on the responses to the 10 CFR 1 Information in the SQN UFSAR and TSs were reviewed in the preparation of this 50.59(c)(2) evaluation questions. No Screening Review/Evaluation. During SQN Unit 2 Refueling Outage 23 (S2RFO23), UFSAR Chapter 15 accident analysis it was determined that removal of the H-08 RCCA was necessary due to the Control need to be reevaluated to address core Rod Drive Mechanism (CRDM) stationary gripper's wear condition. All of the fuel bypass flow. The activity can be assemblies loaded for Sequoyah Unit 2 Cycle 24 (S2C24) are Framatome Adv. W17 implemented per plant procedures.
HTP design. This is unchanged from the previous cycle.
The removal of the H-08 RCCA does not Sequoyah Unit 1 Cycle 24 reduced the number of operating RCCAs from 53 to 52. impact the frequency or likelihood of an During the evaluation for this change, a bounding bypass flow for both units was accident or malfunction because an RCCA created to apply to both Unit 1 and Unit 2. Based on the increase in core bypass is being removed and can no longer fail.
flow, the UFSAR was evaluated to confirm that no safety/accident analysis, SSCs, There are no dose consequences or downstream additional accidents were impacted due to this change. The associated with the removal of an RCCA.
increase in core bypass flow does not increase the total core flow, but impacts the The accidents analyzed in the UFSAR are departure from nucleate boiling (DNB) analyses as the flow through the guide tubes independent of the number of control rods.
has been increased. The flow restrictor ensures that the thermal hydraulic The cycle-specific nuclear design characteristics are equivalent and therefore will not create any new failure modes of parameters with the H-08 RCCA removed E1-32 of E1-33
Core DESCRIPTION SAFETY ANALYSIS Operating Limits Report other SSCs. The removal of the H-08 RCCA adversely impacts the design function were confirmed to be bounded by the of the RCCA by reducing the trip reactivity worth and available shutdown margin. conservative input values used in the UFSAR accident analyses. Therefore, no design basis limits were violated, and no malfunctions with a different result are possible.
Therefore, this activity may be implemented per plant procedures without obtaining a License Amendment.
DOCUMENT NUMBER/72.48 EVALUATION DESCRIPTION SAFETY ANALYSIS TRACKING NUMBER None E1-33 of E1-33
ENCLOSURE 2 SEQUOYAH NUCLEAR PLANT COMMITMENT CHANGE REPORT
Commitment Evaluation No./ Source Summary of Summary of Basis/Justification Commitment Document Original Commitment Commitment Changes for Changes Tracking No.
NCO0900026001 TVA letter to 1. A corporate standard describing TVA is no longer maintaining this This commitment is redundant to NRC dated TVA's vendor manual control commitment as a Regulatory other programs, processes and February 7, program will be issued by July 31, Commitment. Therefore, TVA tools that are standard today, but 1990 1990. may change the vendor manual were not available in 1990.
control process NPG-SPP-09.20,
- 2. The standard will standardize Vendor Manual Control to not The technological and process efficient and effective ways of meet the past industry guidance. advancements in equipment performing site activities and performance information functions related to vendor exchanges and other operating manuals and associated vendor experience sharing meet the intent information. of GL 83-28, Required Actions Based on Generic Implications of
- 3. The standard will incorporate Salem ATWS Events and GL 90-the guidance supplied in 03, Relaxation of Staff Position in NUTAC/VETIP (INPO 84-010), Generic Letter 83-28, Item 2.2 Part INPO Good Practice 87-009, and 2 "Vendor Interface for Safety-the guidance provided in the Related Components" and have February 1987 letter from NRC. obviated the need for direct periodic contact with some vendors.
E2-1 of E2-1