ML24122C637
| ML24122C637 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 05/01/2024 |
| From: | Ameren Missouri, Union Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML24122C635 | List: |
| References | |
| ULNRC- 06856 | |
| Download: ML24122C637 (1) | |
Text
Enclosure to ULNRC- 06856 UNION ELECTRIC COMPANY (dba AMEREN MISSOURI)
CALLAWAY PLANT DOCKET NOS. 50-483 AND 72-1045 10 CFR 50.59 AND 10 CFR 72.48
SUMMARY
REPORT Report Period: November 24, 2022 to May 1, 2024 Page 1 of9
Enclosure to ULNRC- 06856 EXECUTIVE
SUMMARY
In accordance with 10 CFR 50.59(d)(2) and 10 CFR 72.48(d)(2), a summary report has been prepared which provides summaries of the 1 0 CFR 50.59 and 1 0 CFR 72.48 evaluations of changes, tests, and experiments approved and implemented for activities at Callaway Plant.
This report covers all 1 0 CFR 50.59 evaluations for changes that were implemented from November 24, 2022 to May 1, 2024. During this period there were three changes implemented that required a 10 CFR 50.59 evaluation. For two ofthese changes (10 CfR 50.59 Evaluations 23-01 and 23-02), it was determined per 10 CFR 50.59(c)(1) that NRC approval is not required. for the third change (10 CFR 50.59 Evaluation 22-02), it was determined that NRC approval is required, and consequently, a license amendment was requested. A summary for each of those 1 0 CFR 50.59 evaluations is hereby provided.
Additionally, this report is intended to cover all 10 CFR 72.48 evaluations for changes that were implemented during the identified reporting period for the independent spent fuel storage installation (ISfSI) at the Callaway site. However, from November 24, 2022 to May 1, 2024, there were no changes implemented that required a 1 0 CFR 72.48 evaluation.
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Enclosure to ULNRC- 06856 10 CFR 50.59 EVALUATIONS:
10 CFR 72.48 EVALUATIONS:
Evaluation Activity:
Number:
None MP = Modification Package Evaluation Activity:
Number:
22-02 Allow SW to replace ESW train in MODES 5 and 6 23-01 Operability ofESW during Aux Feed Flushes in Modes 5 and 6 23 02 Removed Reactor Vessel Specimen Access Plugs (per Westinghouse SCP-RVO1 0-GN-L5-00000 1)
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Enclosure to ULNRC- 06856 10 CFR 50.59 Evaluation 22-02: Allow Service Water train to replace Essential Service Water train in MODES 5 and 6 Activity
Description:
The proposed change would allow use of the non-safety related Service Water (SW) system to provide cooling water support (in lieu of support from the associated safety-related Essential Service Water (ESW) train) for one oftwo redundant trains ofTS-required equipment when both equipment trains are required to be Operable during cold shutdown/refueling conditions. The supported equipment/systems affected by the proposed change are the Residual Heat Removal (RHR) system (which is cooled via the intermediary Component Cooling Water (CCW) system) and Control Room Air Conditioning System (CRACS), as applicable during Modes 5 and 6. The applicable/affected TS Limiting Conditions ofOperation (LCOs) are TS LCO 3.4.8, RCS Loops
Mode 5, Loops Not Filled; TS LCO 3.7.11, Control Room Air Conditioning System (CRACS);
and TS LCO 3.9.6, Residual Heat Removal (RHR) and Coolant Circulation Low Water Level.
Summary of Evaluation:
The only FSAR-described accidents of concern in Modes 5 and 6 and during the movement of irradiated fuel are the Fuel Handling Accident (FHA) and the Waste Gas Tank Rupture (WGTR) event. The substitution of one train of ESW with SW as a support system for the CRACS and RHR/CCW systems does not impact the likelihood of occurrence of either of the noted accidents since the CRACS and RHR systems (e.g., their failure or misoperation) do not contribute to either of those accidents as initiators. This change does not impact the availability or capability of any system credited with responding to the FHA, nor does it affect the analyzed WGTR event sequence in any way, since no active systems are credited for mitigation of that event. As such, the consequences ofthese accidents remain unaffected by this proposed change.
The use of the non-safety SW system in lieu of a safety-related ESW train to support one of the trains of the CRACS and RHR systems carries an inherent increase in the likelihood of occurrence of a malfunction ofthe support system. A Yes response was therefore documented for the malfunction question (such that the proposed change involves more than a minimal increase in the likelihood of a malfunction), which is the basis for seeking NRC approval. In the event of a malfunction ofthe operating RHR or CRACS train, however, the alternate train would still be available to fulfill the required system function, resulting in no changes to the consequences of a malfunction.
As noted above, following a failure of either the SW system or ESW train, operability of the opposite train is assured such that one train of equipment remains capable of responding to events that may occur during shutdown conditions (e.g., a seismic event or a severe tornado). This continued capability precludes the loss of decay heat removal or loss of emergency cooling to the control room such that those events or conditions are not a new type of accident (or hazard) not previously evaluated.
The proposed change does not modify any SSC important to safety, and correspondingly, no Page 4 of 9
Enclosure to ULNRC- 06856 changes to assumed malfunctions of safety related S$Cs are realized by this change. A malfunction of one train of the ESW system has been previously evaluated for shutdown conditions, wherein the opposite train of E$W remains available to respond. This malfunction and response is not significantly different than what the response would be following the proposed change, and therefore, the proposed change does not create the possibility of a malfunction with a different result.
As previously explained, allowing one train of the SW system to provide cooling water for one train ofthe CRACS and RHR systems, while in Modes 5 and 6 and during the movement of irradiated fuel, carries an increased risk of a malfunction when compared with use of the ESW system to provide this function (due to the change from a safety-grade system to a commercial-grade system).
Therefore, the proposed change requires prior NRC approval.
A license amendment request (LAR) was prepared for the proposed change and was submitted on December 1, 2022 via Ameren Missouri letter ULNRC-0678 1.
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Enclosure to ULNRC- 06856 10 CFR 50.59 Evaluation 23-01: Operability of E$W during Aux Feed Flushes in Modes 5 and 6 Activity
Description:
The proposed change pertains to the procedures used for flushing the Essential Service Water (ESW) supply piping to the Auxiliary Feedwater (AfW) pumps. The procedures are being revised to address performance of the flushing during Modes 5 and 6 (since these Modes are not currently addressed in the procedures), specifically to clarify that the flushing does not render the affected ESW train (or the ultimate heat sink) inoperable, even while the train is supplying its shutdown loads. This includes specifying and crediting an operator action to isolate the flushing discharge in the event of a break in the flushing discharge line (which is non-safety related).
Summary of Evaluation:
The only applicable accident (i.e., hazard) to consider in Modes 5 and 6 for this proposed change is a seismic event, for which the only requirement is ensuring the plant can be maintained in a safe shutdown condition. for this event, the non-safety related line used to return the ESW flushing flow to the ultimate heat sink (UHS) is assumed to fail per the engineering evaluation performed in support ofthe procedure changes. The flushing procedures are being revised to address the potential for a break/malfunction in the non-safety related discharge pipe connected to the associated ESW line, which is not assumed to be isolated. The likelihood of this break/malfunction, however, may be assumed to be no greater than for any other non-safety line connected to a safety related system, and therefore, the proposed change does not increase the likelihood of such a break occurring in the discharge line during a seismic event.
There are no analyzed dose consequences associated with a seismic event, and the capability to achieve safe shutdown has essentially already been met with the plant already in a shutdown condition. The proposed change to the noted flushing procedures involves recognition of a malfunction that could be caused by a seismic event (such that the consequences of a seismic event could potentially be changed), but it does not introduce a new accident.
During shutdown conditions, a loss of decay heat removal potentially leading to a radiological release is possible, but such an event is assumed to be very remote with respect to the plants licensing basis, as there is no FSAR-described analysis of such an unlikely accident.
- Moreover, the potential for such an event resulting from a seismically induced break in the E$W flushing discharge line is deemed to be extremely remote due to the multiple failures that would have to occur for such an event/accident to occur and since the affected ESW train has been shown to remain Operable in the event ofthe break in the discharge line, based on the engineering evaluation performed in support ofthe change.
The engineering evaluation in support ofthe change includes the provision for an Operator to close the discharge isolation valve(s) in the event of a break/leak in the discharge line during flushing.
This can be accomplished by stationing a dedicated Operator in the area of the flushing activity.
Due to the large margin involved with respect to UHS inventory (relative to the ESW flow exiting Page 6 of 9
Enclosure to ULNRC- 06856 the ESW/UHS system via a break in the flushing discharge line), no time limit was specified for the Operator action, as it is not considered to be time critical. Based on the capability for a malfunction (break/leak) of the non-safety related discharge line to be isolated during the flushing activity performed per the affected procedures (in Modes 5 and 6), the proposed change does not create a possibility for a malfunction of an $SC important to safety with a different result than any previously evaluated in FSAR.
The proposed change does not make any physical changes to the plant or to any of the calculations utilized to determine the design basis limits for fission product barriers. The proposed change only applies when the plant is in a shutdown condition, during which there are no FSAR-described accidents that can significantly challenge the design basis limits for fission product barriers, as far as the reactor or reactor operation is concerned.
Potential adverse effects identified by the 1 0 CFR 50.59 Screen were evaluated and the 1 0 CFR 50.59 Evaluation questions were answered NO, thus indicating that the aspects ofthe proposed activity that are covered under 10 CFR 50.59 can be implemented without prior NRC approval.
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Enclosure to ULNRC- 06856 10 CFR 50.59 Evaluation 23-02: Removed Reactor Vessel Specimen Access Plugs (per Westinghouse SCP-RVO1O-GN-L5-000001)
Activity
Description:
In the Callaway Unit 1 reactor internals, there are six holes in the main support flange of the core barrel. These holes provide necessary access for inspections and for handling the reactor vessel irradiation specimens. There are plugs installed in these holes, which are used to prevent reactor coolant flow from passing through the holes and thereby bypassing the reactor vessel core region.
The plugs are captured (in place) by the installation of the upper internals.
During the Fall 2023 refueling outage, two ofthe specimen access plugs were found to be dislodged and unable to be reinserted. Ameren requested Westinghouse to provide justification for leaving the two specimen access plugs out of service until the next cycle. An engineering evaluation was performed to demonstrate that bypass flow will remain within analyzed parameters if the holes remain. The bypass flow evaluation is applicable to Cycle 27 only.
Summary of Evaluation:
In the Westinghouse analysis performed for Callaway, it is concluded that despite the increased bypass flow through the open holes in the core barrel (due to the removed specimen access plugs),
both the design core bypass flow and the best estimate core bypass flow used in safety analyses for both thimble plugs installed (TPI) and thimble plugs removed (TPR) conditions remain valid. The Reload Safety Evaluation for Callaway Unit 1 Cycle 27 thus remains applicable.
The function ofthe safety injection system will not be impacted since the design core bypass flow is not being violated. Also, even with the specimen plugs removed, the hydraulic resistance associated with the downcomer and lower plenum is much less than the hydraulic resistance associated with the flow entering the upper head. Thus, the vast majority of the flow will still be conducted through the downcomer and into the core. On that basis, the change does not create a malfunction with a different result.
In addition, there are no operability or safety concerns associated with a small increase in flow into the upper head. Hence, no additional testing or monitoring is required. Furthermore, there is no adverse impact on the reactor internals structures and components or any additional potential for loose parts in the vessel, when operating with the specimen access plugs out. Also, there will be no impact to TAVG or TREF scaling. This assessment is acceptable for one cycle of operation, and Westinghouse recommends the specimen access plug be re-installed during the next refueling outage.
The removal of two specimen access plugs does not create any initiator of an accident, nor is it an initiator of any malfunction. The increased bypass flow does not increase the consequences of an accident because with the slight increase in flow, there is still sufficient margin available in the design basis core bypass flow analysis to accommodate this increase. Operation with the specimen access plugs removed at two locations does not affect the integrity of the irradiation specimen guide Page 8 of 9
Enclosure to ULNRC- 06856 baskets or future specimen insertion. Since the resultant core bypass flow has been evaluated with respect to conformance with the accident analyses, the change in core bypass flow does not result in more than a minimal increase in the consequences of an accident previously evaluated in the FSAR (since there is no impact on analyzed fuel performance), nor does it create a new or different accident than any previously evaluated in the FSAR. The calculated bypass flow is still within both the design basis core bypass flow and the best estimate core bypass flow.
Potential adverse effects identified by the 10 CFR 50.59 Screen were evaluated, and all of the 10 CFR 50.59 Evaluation questions were answered No, thus indicating that the aspects of the proposed activity reviewed under 1 0 CFR 50.59 can be implemented without prior NRC approval.
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