ML18270A238
| ML18270A238 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/25/2018 |
| From: | Indiana Michigan Power Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18270A194 | List:
|
| References | |
| AEP-NRC-2018-17 | |
| Download: ML18270A238 (6) | |
Text
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
Page 1 of 6 Changes, Tests, and Experiments As required by 10 CFR 50.59(d)(2), the following report contains brief descriptions of changes made to the facility and/or associated documentation, and summaries of the associated 50.59 evaluations.
SS-SE-2015-0354-01 Unit 1/Unit 2 Technical Requirements Manual Activity
Description:
The proposed activity creates a new Technical Requirement for Operation (TRO) 8.1.3, Boration System - HOT SHUTDOWN which will require only a single Centrifugal Charging Pump flowpath for the addition of borated water (Borated water supplied by either an OPERABLE reactor water storage tank (RWST) OR a flowpath from an OPERABLE Boric Acid Storage Tank) while in MODE 4, HOT SHUTDOWN. The change also impacts TRO 8.1.1, Boration System - Operating changing the MODE of Applicability to MODES 1, 2, and 3 from MODES 1, 2, 3, and 4. Therefore, TRO 8.1.1 will continue to require two separate flowpaths (including separate borated water sources) for Boric Acid; however, the Mode of Applicability will be MODES 1 through 3, instead of MODES I through 4.
Summary of the Evaluation:
There are no accident analyses impacted by the reduction in the number of required boration flowpaths from two to one in Mode 4. The Uncontrolled Boron Dilution accident described in the Updated Final Safety Analysis Report (UFSAR) does not credit boration; it credits operator action. In addition, there are no other Mode 4 accidents that are impacted. There are no discussions, descriptions, or requirements in the UFSAR for having two trains of boration available in Mode 4, and reducing the number of required boration flowpaths from two to one in Mode 4 would be reducing excess redundancy to the level credited in the UFSAR. In addition, plant procedures require that Shutdown Margin be verified prior to changing plant Modes; therefore, the plant procedures require that the chemical and volume control system (CVCS) perform its design function of meeting Shutdown Margin in Mode 4 prior to actually entering Mode 4. This procedural requirement will continue to exist, and will continue to mitigate the risk of the CVCS not performing its design function of meeting shutdown margin (SDM) in all Modes, including Mode 4.
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
Page 2 of 6 SS-SE-2016-0139-00 Implementation of Unit 2 LOCA-Containment Integrity Analysis Using WCOBRA/TRAC Mass and Energy Releases (WCAP-17721-P-A)
This activity is bounded by 10 CFR 50.59 Evaluation SS-SE-2015-0322-00, previously submitted to the U. S. Nuclear Regulatory Commission (NRC). For convenience, the Summary of the Evaluation of SS-SE-2015-0322-00 is replicated below:
Activity
Description:
An Engineering Change was used to replace the Current Licensing Basis methodology for the Loss of Coolant Accident (LOCA) Mass and Energy (M&E) Releases from WCAP-10325-P-A Revision 1 to WCAP-17721-P-A Revision 0 for Donald C. Cook Nuclear Plant (CNP) Unit 1.
The new M&E Releases were then used as input to a new Unit 1-specific LOCA Containment Response analysis to demonstrate continued compliance with Containment Integrity requirements. [Previous Containment Integrity analysis bounded both Units 1 and 2.] In addition to the LOCA M&E Release methodology change, some revised input parameters were used in the new Unit 1-specific analysis. The Engineering Change did not involve any physical changes to the plant, changes to design limits or other system performance parameters, or changes to equipment operation or maintenance.
Summary of the Evaluation:
The Evaluation assessed the impacts of the revised input parameters on existing containment integrity requirements and the acceptability of replacing the LOCA M&E Release methodology.
Input parameter changes were accepted based on the new analysis results that demonstrated continued compliance with Containment Integrity requirements.
Use of WCAP-1772 I-P-A Revision 0 was accepted for Unit 1 LOCA M&E Release methodology based on findings that 1) the LOCA M&E Release methodology is described in the UFSAR and is used in Unit 1's design bases or safety analyses, 2) the NRC has approved use of WCAP-17721-PA for large break LOCA M&E Release analysis for ice condenser plants, 3) the new analysis adequately documents that the Limitations and the Condition identified in the NRC Safety Evaluation of WCAP-1772 I-P-A are met, 4) the Westinghouse personnel performing the new analysis and their supporting organization are qualified to implement the WCAP-17721-P-A methodology, and 5) the WCAP-17721-P methodology is being incorporated into the Unit 1 safety analyses en toto.
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
Page 3 of 6 SS-SE-2016-0148-01 Remove Bellows on 1-CPN-85 for NESW to 1-HV-CUV-3; Redesign the Penetration; and Abandon One NESW Line Activity
Description:
Following wall thinning examination of safety related Non-Essential Service Water (NESW) piping that supplies cooling water to one of four upper containment ventilation units (the subject piping is part of a containment penetration), it was determined that the piping should be replaced. Subsequently, installation issues necessitated a temporary redesign of the piping penetration. Permanent restoration will be implemented during the Unit 1 Cycle 28 refueling outage. The changes involve: removal of the bellows welded to both the involved NESW supply line and penetration I-CPN-85 (on the Auxiliary Building side of the penetration); removal of the protective cover associated with Containment Penetration 1-CPN-85; and removal of two sections of piping in the NESW supply line. Removal of the pipe sections resulted in redesign of the penetration (via welded cap/plug) in order to meet the double isolation requirement for process piping. As a result of the removal of the piping and the change to the penetration, non-essential service water (NESW) flow to one of four upper containment ventilation units will be taken out of service. The containment isolation valves will be closed to isolate the NESW supply lines to and from the affected containment ventilation unit. This temporary modification has since been removed.
All aspects of the change screened out with the exception of the elimination of cooling water supply to one upper containment cooling unit.
Summary of the Evaluation:
Supplying cooling water to the upper containment ventilation units is cited for normal operations and shutdown to permit personnel access. This aspect does not have an accident mitigation or prevention function and is not required for safe shutdown. However, it does provide a support function to ensure that Technical Specification 3.6.5, Containment Air Temperature, is met.
Therefore, the UFSAR described design function of interest is the limiting of upper containment compartment temperature to less than 100ºF. From UFSAR Section 5.5.3, Upper Compartment Ventilation System: "The upper compartment ventilation system consists of four free standing recirculating ventilating units (3 for normal operation, 1 standby)...
The water for the cooling coils is supplied by the non-essential service water system. Any three of the four units have sufficient cooling capacity to maintain the temperature below 100°F during design summer conditions." It was determined that the change constitutes a reduction in redundancy/diversity as previously any combination of three cooling units could be utilized to provide sufficient cooling, whereas this change limits it to a single configuration.
As such, the change is considered to have an adverse effect on a UFSAR described design function/method of performing or controlling a UFSAR described design function. Typically a reduction in diversity of an systems, structures, and components (SSC) important to safety would result in a "more than minimal" likelihood of occurrence of a malfunction and require prior NRC approval.
However, the following clarification/exception was provided in Question E. 12 of NEI Questions and Answers on 10 CFR 50.59 and NEI 96-07, Revision 1, Update 4 - April 2001: "Licensees may, however, without prior NRC approval, reduce excess redundancy, diversity, separation or independence, if any, to the level credited in the UFSAR." Therefore, as the UFSAR clearly indicates that any three units of upper containment ventilation provide sufficient cooling capacity, and neither NESW nor upper containment ventilation is credited for accident prevention or mitigation (upper containment temperature is only an initial condition input
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
Page 4 of 6 assumption to the safety analyses), it was determined that prior NRC approval of the change is not required.
SS-SE-2016-0278-01 Unit 2 BVM Testing of RS56R Last Stage Blades at Reduced Vacuum Activity
Description:
The activity is the performance of test procedure EC-0000054708-TP-001, Unit 2 blade vibration monitoring (BVM) Testing of RS56R Last Stage Blades at Reduced Vacuum, which instructs an operator, utilizing test equipment installed by EC-0000054708 and 2-TM-16-27-RO, to allow a controlled amount of air intrusion into the A, B, and C Main Condensers reducing condenser vacuum during power ascension. The purpose of this test is to record and compare the low pressure turbine last stage blade actual BVM test results to the variable condenser vacuum curve provided by GE/Alstom for the new low pressure (LP) Turbine RS56R Last Stage Blades (LSB).
This comparison will ensure that there is substantial margin between actual LSB vibrations versus the vibration limits that GE/Alstom provided based on design allowable fatigue stress limits for long term safe operation.
Summary of the Evaluation:
This 50.59 Evaluation was performed in response to question II.D of the 50.59 Screen, where it was concluded that the performance of BVM procedure EC-0000054708-TP-0O1 is a test or experiment not described in the UFSAR that will operate or control the main condenser and main feed pump turbine condensers in a manner that is outside the reference bounds of the main turbine and the main feed pump turbines by operating or controlling these UFSAR described SSCs at a reduced condenser vacuum.
As per the 10 CFR 50.59 Reference Manual Appendix 7.1, the Loss of condenser vacuum and other events resulting in a turbine trip is a pressurized water reactor (PWR) Category 11 Event, and the plant is designed to prevent a loss of condenser vacuum to preclude a turbine trip and/or reactor trip. The purposeful introduction of a loss of vacuum under controlled conditions in the performance of the BVM test is inconsistent with how the plant is designed, and will bring the main condenser and Main Feed Pump (MFP) condensers closer to main turbine and main feed pump trip setpoints, increasing the likelihood (albeit small) that the main turbine or MFP will trip, and whenever the turbine generator is above approximately 31% power (above P-8) the reactor also trips.
However, this 50.59 Evaluation concludes that the performance of this test does not require prior NRC approval because the test procedure maintains adequate margin from main turbine trips, MFP trips, and protection of the LP turbine last stage blades, preventing blade failure and turbine missiles. The operation of all associated SSCs are maintained as designed.
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
Page 5 of 6 SS-SE-2017-0012-00 NSAL 14-5 WRB-1 and WRB-2 DNB Correlation Non-Conservatism Activity
Description:
Lower than expected critical heat flux (CHF) results were obtained from 5x5 rod bundle tests simulating the Westinghouse 14 foot 17x17 Robust Fuel Assembly (RFA) design without intermediate flow mixer (IFM) grids. The test data showed lower than expected CHF results from the 5x5 rod bundle tests for a subset of conditions that were previously untested, resulting in non-conservative predictions by the WRB-2M CHF correlation, which is applicable only to 17x17 RFA-type fuel. While the new test data are not directly applicable to the WRB-1 and WRB-2 correlations used at CNP, the new test data for the 17x17 RFA fuel without IFMs were also analyzed using the WRB-1 and WRB-2 correlations.
This issue is resolved by applying a conservative penalty if the fluid conditions are in the potentially non-conservative sub-region of conditions. If the results of a future plant departure from nucleate boiling (DNB) safety analysis show that the local quality at the location of the Minimum Departure from Nucleate Boiling Ratio (MDNBR) does exceed the applicable quality threshold, the plant DNB safety analysis is assumed to be in the potentially non-conservative sub-region of the local fluid parameters of the current WRB-1 or WRB-2 correlation, as applicable.
If the local quality at the location of the MDNBR exceeds the applicable quality threshold value from the technical basis documents, a conservative margin reduction to the CHF correlation prediction (penalty) is applied which effectively uses up all of the DNBR margin to the approved 95/95 DNBR limit (assumes fuel rod failure has occurred due to DNB).
Summary of the Evaluation:
The proposed activity incorporates a conservative change to the approved DNB methodology as described in the UFSAR. The approved DNB correlations are described in Section 3 of the UFSAR.
The proposed activity (methodology change) modifies the approved correlation methodology by adding a conservative penalty to be applied if the fluid conditions are in the potentially non-conservative sub-region of conditions for the affected correlations. As such, the proposed change, which would only be utilized if certain conditions described above are met, reduces margin to applicable safety analysis limits and thus, is considered conservative.
A change to an element of methodology does not require prior NRC approval if it produces conservative results.
Donald C. Cook Nuclear Plant Report of Changes, Test, Experiments Pursuant to 10 CFR 50.59(d)(2)
Page 6 of 6 SS-SE-2017-0120-00 UCR for Feedwater Malfunction Evaluation Activity
Description:
The proposed change is an alteration to Section 14.1.10 of Unit 1 and Unit 2 UFSAR for the Feedwater Malfunction (FWM) event. The FWM event is performed assuming 200% nominal feedwater flow due to the event. The proposed change incorporates an evaluation into the UFSAR that demonstrates that acceptable FWM analysis results are obtained at 240% of nominal feedwater flow.
Summary of the Evaluation:
The FWM event is analyzed to verify compliance with the DNB design basis. For the Feedwater Malfunction (FWM) event the acceptance criterion in the Analysis of Record (AOR) is that DNB, and consequent fuel damage, does not occur. The evaluation demonstrated that with the higher feedwater flow assumed the DNB acceptance criterion was still met for all cases at Hot Zero Power (HZP) and Hot Full Power (HFP) in both units, and no fuel damage occurs at the higher assumed feedwater flowrate. Since no fuel damage occurs in the AOR or the Westinghouse Letter AEP-17-5 evaluation of 240% feedwater flow, the proposed activity does not result in any change in the consequences of an accident or malfunction previously evaluated in the UFSAR and prior NRC approval was not required.