ML20118D320
| ML20118D320 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 04/27/2020 |
| From: | Xcel Energy, Northern States Power Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| MML20118D173 | List: |
| References | |
| L-PI-20-016 | |
| Download: ML20118D320 (6) | |
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ENCLOSURE 5 PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS 50.59 Evaluation No. 1153, Rev 0 - Turbine Overspeed Protection Test Frequency Changes (1/24/20)
Description of Change The proposed change is to the turbine valve testing frequency of SP 1054/2054 from 6 months to 12 months. The purpose of this testing is to demonstrate the function of the turbine stop; governor, intercept, and reheat stop valves. These valves are key elements of the turbine overspeed protection system due to their use in terminating steam flow to the high pressure (HP) turbine and low pressure (LP) turbines. Prompt and effective termination of steam flow to the main turbine during an overspeed event limits the potential that turbine missiles can be generated as a result of turbine rotor failure. The change in frequency reduces the number of load reductions on both units during the operating cycle.
Summary of 50.59 Evaluations The likelihood of a turbine missile ejection accident for Prairie Island 6 month valve stroke testing interval quotes to a 2.84x10-6 per year turbine missile ejection frequency. A 12 month valve testing interval missile ejection frequency increases to 4.34x10-6 per year. The potential accident frequency remains below the required threshold 1x10-5, as discussed in WCAP 11525 and 16054, and does not result in a more than minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR. As stated in NEI 96-07, if the likelihood of occurrence of a malfunction can be determined to be greater than a factor of 2, NRC approval would be required. The overall change in the reliability of the turbine stop, governor, intercept, and reheat stop valves would not be expected to exceed a factor of 2. The proposed activity does not result in a more than minimal increase in the likelihood of occurrence of a malfunction.
50.59 Evaluation No 1151, Rev. 0 - Revision to Post-Accident EAB, LPZ, and CR Doses - AST (8/9/19)
Description of Change The proposed activity is a revision to several post-accident dose consequence calculations resulting from a revision to the post-LOCA dose consequence calculation GEN-Pl-079 to align with current reactor power limits. Two of the updated calculation outputs from GEN-Pl- 079 - Control Room dose from "External Cloud" and "CR Filter Shine" sources - serve as inputs to other accident dose calculations and have an effect on their respective outcomes. The analyses affected by the change in these inputs include the Locked Rotor Accident [LRA] described in USFAR Section 14.4.8 (GEN-Pl-083), the Main Steam Line Break [MSLB] described in USFAR Section 14.5.5 (GEN-Pl-078), the Steam Generator Tube Rupture [SGTR] described in USFAR Section 14.5.4 (GEN-Pl-081), and the Control Rod Ejection Accident [CREA] described in USFAR Section 14.5.6 (GEN-Pl-082). The revised outputs from GEN-Pl-079 result in a net increase in dose consequences of 8E-4 rem for these downstream calculations (-0.0040 rem from "External Cloud" and +0.0048 rem from the "CR Filter Shine").
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ENCLOSURE 5 PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS Summary of 50.59 Evaluations The LRA, MSLB, SGTR, and CREA accidents are the only accidents that could be affected by the proposed revision to post-accident dose consequence calculations GEN-Pl-078, -081, -082, & -083.
The proposed revision to these dose consequence calculations does not physically alter any equipment, system performance, or operator actions that could affect the respective malfunctions for the applicable accident scenarios. The proposed revision to these dose consequence calculations represents a not more than minimal increase in the dose consequences for these accidents described in the UFSAR as evaluated against the criteria of NEI 96-07. The worst case proposed total dose Js 4.68 rem, as compared to current standard review plan guidelines of 5 rem. The worst case proposed dose increase is approximately 0.25% of the difference in current margin to the SRP guidelines, and is well below the threshold of the 10% increase limitation for acceptable increases in dose consequences.
50.59 Evaluation No 1149, Rev 0 Post HELB Battery Room Cooling (6/21/19)
Description of Change There is no Safety Related (SR) cooling/ventilation for the Safeguards Battery Rooms. In the event of high room temperature the current response is to open the doors to thee battery rooms to promote air circulation and lower the room ambient temperature. The current analysis of record shows that the circulation created by opening of the battery room doors is not enough to keep the room ambient temperature below the established limit following Main Steam (MS) High Energy Line Break (HELB) event. ECR 601000001353 is a commercial change that will procure two 480VAC fans and construct carts which the fans will be placed on, thus making them mobile for Battery Room post HELB cooling.
Summary of 50.59 Evaluations This activity is limited to addition of procedurally controlled manual actions to place and connect portable fans in 11 and 21 Battery Rooms. C18.1 currently has operations personnel monitor Battery Room temperatures. If Battery Room cooling cannot maintain room temperatures below the design limit of 120°F, blocks open the Battery Room doors. The current analysis shows that the air circulation due to opening of the Battery Room doors is not sufficient to keep the room temperature below the established limit following a Main Steam (MS) High Energy Line Break (HELB) event.
50.59 Evaluation No 1145, Rev 0 - Digital Upgrade of FW and AMSAC/DSS Control Systems (12/18/18)
Description of Change Upgrading the feedwater control system (FWCS) and the ATWS mitigation system actuation circuitry /
diverse scram system (AMSAC/DSS). In order to address equipment obsolescence issues, the distributed control system (DCS) used for these two systems will be changed from the existing Westinghouse Distributed Processing Family (WDPF) platform to an Ovation-based platform, as recommended by the original designer of both systems and both platforms, Westinghouse. The upgrade includes functional enhancements to the FWCS and changes to the instrumentation and 2
ENCLOSURE 5 PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS controls on the main control board. Local equipment used to position the feedwater regulating valves and feedwater bypass valves will also be modified.
Summary of 50.59 Evaluations The proposed activity upgrades the FWCS and the AMSAC/DSS control system. Failures in the existing FWCS - such as a failure of a feedwater control valve or feedwater bypass valve, a failure in the feedwater control system, or an operator error - have the potential to increase or decrease feedwater flow to the steam generators. Failures in the existing AMSAC/DSS have the potential to result in spurious actuation of a turbine trip or inadvertent initiation of auxiliary feedwater.
50.59 Evaluation No. 1144, Rev. 0 - Incorporate Supplemental UFSAR Chapter 14.5.6 RCCA Ejection Analysis (10/19/2018)
Description of Change The purpose of this evaluation is to incorporate a supplemental analysis for the Unit 1 and Unit 2 Rupture of a Control Rod Drive Mechanism Housing (RCCA Ejection) analysis, UFSAR Chapter 14.5.6 into the Prairie Island licensing basis. This activity is required because current Prairie Island core designs are using a range of 0 to 8 weight percent gadolinium as a burnable absorber, whereas historical core designs have only used 0 and 8 weight percent Gadolinium. The use of lower Gadolinium enrichments allows for higher uranium enrichment (up to 4.95 weight percent uranium-235 in the non-Gadolinium rods). Westinghouse reviewed the Chapter 14 RCCA ejection analysis and determined that the analysis only considered fuel rods that contained either 0 weight percent gadolinium or 8 weight percent gadolinium. The supplemental analysis will cover a range of gadolinium weight percentages (0 to 8 weight percent inclusive), uranium enrichments (up to 4.95 w/o uranium-235 in the non-Gadolinium rods) as well as a lower power suppression credit (5% versus 20%).
Summary of 50.59 Evaluations The supplemental analysis, including accounting for a range of gadolinium weight percentages, uranium-235 enrichments, and a lower power suppression credit, used the Prairie Island current licensing basis NRC approved methodology and the results showed that the design limits as currently described in the Prairie Island Updated Final Safety Analysis Report are met. Thus, there is no increase to the consequences of an accident or malfunction. In addition, this activity does not impact equipment operations, performance and reliability thus there is no change to the frequency of an accident, likelihood of a malfunction, possibility of a new accident, or possibility of a malfunction with a different result.
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ENCLOSURE 5 PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS 50.59 Evaluation No. 1143, Rev 1- Changes to C18.1 to incorporate manual actions as compensatory measures (9/4/2018)
Description of Change The activity covered by this evaluation is the permanently proceduralized actions taken by Operations to open doors in the Battery Rooms, Auxiliary Feedwater Pump (AFWP) Rooms, Unit 1 Safeguards 4.16 kV Switchgear (Buses 15 & 16) Room, Unit 1 Safeguards 480V Switchgear (Buses 111/121/112/122) Rooms, and Relay Room in order to cool these rooms in the event that normal room cooling is lost. The doors to be opened are doors from the identified room to the Turbine Building. Opening these doors will allow air exchange with the relatively cooler air in the Turbine Building. Opening the doors is currently controlled by operating procedure C18.1, Engineered Safeguards Equipment Support Systems. Engineering analysis demonstrates that these actions are sufficient to maintain room temperatures at acceptable levels such that the safety related equipment located therein remain Operable Summary of 50.59 Evaluations This activity is limited to procedurally controlled manual actions to open various doors to the Turbine Building as alternate means of maintaining adequate room temperatures in areas containing safety related SSCs. The associated rooms are normally cooled by safety related Unit coolers (except the Battery Rooms) which, as shown by ENG-ME-847 rev 0, are capable of maintaining adequate room temperatures during High Energy Line Breaks. Thus, the operator actions covered by this activity are compensatory actions to be taken only when these safety related coolers are nonfunctional and actual room temperatures approach pre-established limits. Specific to the Battery Rooms, ENG-ME-847 rev 0 and ENG-ME-849 rev 1 document the acceptability of opening the Battery Room doors to maintain operability of the SSCs within these rooms following an extended loss of ventilation to these rooms.
Since there are no changes to pre-existing safety related SSCs or other plant equipment, there are no impacts on UFSAR-evaluated accidents (frequency, consequence, or new types), consequences of SSC malfunction, or fission product barriers. Engineering analysis shows that, even under a worst case HELB event, room temperatures can remain at acceptable levels without any room cooling until the HELB conditions in the Turbine Building subside and allow individuals to reach the affected room.
Furthermore, by the time that doors would need to be opened, conditions in the Turbine Building post-HELB will have subsided to the point that it will permit opening the doors.
50.59 Evaluation No. 1142, Rev. 0 - Bypassing Various D5 and D6 EOG trips after an Under voltage Start.(10/1/2018)
Description of Change Management of Change (MOC) 601000000058 is being performed to remove automatic D5 and D6 Emergency Diesel Generator (EOG) trips during an Undervoltage (UV) start condition. The alarm functions will remain. The trips removed with this MOC do not involve conditions considered immediately detrimental to the engine (such as an over-speed condition could). Sufficient time for operators to react to an abnormal condition exists. The change involves updating the UV Start condition trip logic. Currently the diesel generator logic scheme bypasses numerous trips during the Safety Injection start or Emergency Manual start. However, in the event of an UV start, all engine and 4
ENCLOSURE 5 PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS electrical generator trips remain active. Spurious instrument problems create a potential vulnerability to support plant electrical loads in the event of a Loss of Offsite Power (LOOP) condition.
Summary of 50.59 Evaluations The change does not increase the likelihood of occurrence of a malfunction of an SSC important to safety. The intent of the change is to make the Unit 2 EOG's more reliable in the event of a LOOP by bypassing automatic trips that not considered immediately detrimental to the EOG. Bypassing automatic trips does not increase the probability of a component or equipment malfunction detrimental to the EOG design function. The electrical or mechanical component issue that causes the EOG alarm is the initiator and is independent of the rotective EOG circuitry. Additionally, the change has been determined to remain in compliance with design standards and regulatory guidance.
50.59 Evaluation No. 1141, Rev 0- Instrument Air CIV Redundant Solenoid Air Supply (7/16/18)
Description of Change Activity is to install a redundant, Safety-Related solenoid valve in the air supply to the Instrument .Air Containment Isolation Valves (CIV). The purpose of this change is to mitigate a loss of instrument air to containment due to a single soleno.id valve coil short failure.
Summary of 50.59 Evaluations As stated in NEI 96-07, if the likelihood of occurrence of a malfunction can be determined to be greater than a factor of 2, NRC approval would be required. The input for solenoid valve failure to the overall failure rate of the control valve increase is limited to a factor of 2. The impact on the failure rate of the containment isolation control valve would be significantly less than 2. Considering NUREG/CR-6928 the failure rates for the overall impact on the containment isolation control valve estimated failure probability is essentially unchanged, such that, the proposed activity does not result in a more than minimal increase in the likelihood of occurrence of a malfunction.
50.59 Evaluation No. 1140, Rev. 0 - NBFD-65NR Relay Replacement (Unit 1 A and B Train Reactor Protection System) (11/28/2018)
Description of Change ECRs 6EQVENG26249 and 6EQVENG26250 will each replace 44 Westinghouse NBFD-65NR relays with Westinghouse NBFD065NR relays. The changed characteristic which requires evaluation is dropout response time, which increased from 1.450 to 1.480 seconds. As discussed in 50.59 Screenings 5130 and 5131 the replacement of the NBFD065NR relays associated with the OTT and OPT resulted in an adverse impact on the Unit 1 trip time delay. The original time delay for the OTT and OPT functions is 6.0 seconds as documented in Prairie Island USAR page 14.3-4. The replacement NBFD065NR relays increase the logic delay by 0.03 seconds which results in a total delay time of 6.006 seconds. As this new value is greater than the 6.0 seconds listed in USAR page14.3-4, the change is identified as adverse and screens in.
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ENCLOSURE 5 PRAIRIE ISLAND NUCLEAR GENERATING PLANT REPORT OF CHANGES, TESTS, AND EXPERIMENTS Summary of 50.59 Evaluations The evaluation shows that there is no more than a minimal increase in the frequency of occurrence of any accidents or malfunctions previously evaluated in the USAR, nor are there any new types of accidents or malfunctions with different results introduced as a result of the proposed change. The evaluation also determines that there are impacts on the consequences of any accident or malfunction previously evaluated in the USAR. This activity does not involve any Design Basis Limit for Fission Product Barriers DBLFBs). This activity does involve a departure from methods of evaluation described in the USAR. Therefore, NRC approval is not required for the proposed change.
50.59 Evaluation No. 1139, Rev. 0 - EC23208, Unit1 AVR Digital System Upgrade and Software Common Mode Failure (5/2/18)
Description of Change Replacing of the Automatic Voltage Regulator (AVR) in Unit 1.
Summary of 50.59 Evaluations There are no new types of accidents or malfunctions with different results introduced as a result of this proposed change. Since the generator does not directly contribute to accident response and mitigation, there are no impacts on the consequences of an accident or malfunction previously evaluated in the USAR.
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