ML23348A239
| ML23348A239 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 12/19/2023 |
| From: | Dennis Galvin NRC/NRR/DORL/LPL4 |
| To: | Peters K Vistra Operations Company |
| References | |
| EPID L-2023-LLA-0165 | |
| Download: ML23348A239 (1) | |
Text
December 19, 2023 Mr. Ken J. Peters Senior Vice President and Chief Nuclear Officer Attention: Regulatory Affairs Vistra Operations Company LLC Comanche Peak Nuclear Power Plant 6322 N FM 56 P.O. Box 1002 Glen Rose, TX 76043
SUBJECT:
COMANCHE PEAK NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 - NONACCEPTANCE OF LICENSE AMENDMENT REQUEST TO RELOCATE TECHNICAL SPECIFICATION 3.9.3, NUCLEAR INSTRUMENTATION, TO THE TECHNICAL REQUIREMENTS MANUAL (EPID L-2023-LLA-0165)
Dear Mr. Peters:
By letter dated November 20, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23324A050), Vistra Operations Company LLC (the licensee) submitted a license amendment request (LAR) for Comanche Peak Nuclear Power Plant, Unit Nos. 1 and 2 (Comanche Peak). The proposed amendment would relocate technical specification (TS) 3.9.3, Nuclear Instrumentation, to the technical requirements manual (TRM).
There are four source range neutron flux monitors (SRMs) located external to the reactor vessel that detect neutrons leaking from the core. TS 3.9.3 requires two SRMs to be OPERABLE in Mode 6 (Refueling).
The purpose of this letter is to provide the results of the U.S. Nuclear Regulatory Commission (NRC) staffs acceptance review of this amendment request. The acceptance review was performed to determine if there is sufficient technical information in scope and depth to allow the NRC staff to complete its detailed technical review. The acceptance review is also intended to identify whether the application has any readily apparent information insufficiencies in its characterization of the regulatory requirements or the licensing basis of the plant.
Consistent with Section 50.90, Application for amendment of license, construction permit, or early site permit, of Title 10 of the Code of Federal Regulations (10 CFR), an application for an amendment to a license (including the TSs) must fully describe the changes requested, and following as far as applicable, the form prescribed for original applications. Section 50.34, Contents of applications; technical information, of 10 CFR addresses the content of technical information required. This section stipulates that the submittal address the design and operating characteristics, unusual or novel design features, and principal safety considerations.
On August 7, 2023, a presubmittal meeting was held between the NRC staff and the licensee on the proposal to submit this LAR. The NRC staff identified several concerns with the proposed LAR similar to those discussed below. A summary of the presubmittal meeting was issued on September 18, 2023 (ML23255A075).
Licensees Basis The licensees basis for the LAR is that TS 3.9.3 does not satisfy any of the criteria of 10 CFR 50.36(c)(2)(ii) and can be relocated out of the TS to the TRM, a licensee-controlled document. The licensees current TS Bases and the Westinghouse Standard Technical Specification (STS) Bases (NUREG-1431, Revision 5, Volume 2) state that TS 3.9.3 meets 10 CFR 50.36(c)(2)(ii), Criterion 3, which is:
A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
The licensee states that the nuclear instrumentation covered by TS 3.9.3 does not mitigate any design basis accident in Chapter 15 of the updated final safety analysis report (UFSAR) and therefore does not meet Criterion 3. The licensee discusses three Chapter 15 events in the LAR. UFSAR Section 15.4.6, Chemical and Volume Control System Malfunction that Results in a Decrease in the Boron Concentration in the Reactor Coolant, addresses a boron dilution event, which is mitigated by closing and securing isolation valves for unborated water sources in accordance with TS 3.9.2, Unborated Water Source Isolation Valves. The licensee then cites a reviewers note in the Westinghouse STS Bases (discussed further below), which states that for plants with TS 3.9.2, the boron dilution event does not need to be analyzed in Chapter 15. UFSAR Section 15.4.7, Inadvertent Loading and Operation of a Fuel Assembly into an Improper Position, addresses the inadvertent loading and operation of a fuel assembly in an improper position. The licensee states that the event relies on administrative controls to prevent fuel assembly misloads and flux monitoring during reactor startup with in-core movable detectors to detect potential power distribution anomalies during reactor operation that could lead to fuel damage. The licensee notes that the SRMs are not a primary success path to mitigate fuel assembly misloads. UFSAR Section 15.7.4, Design Basis Fuel Handling Accidents, addresses a design basis fuel handling accident either in the containment building or on the spent fuel storage area floor. The licensee states the SRMs do not mitigate this accident.
The licensee further states that the statement in the TS Bases that describes how TS 3.9.3 relates to the safety analyses and to 10 CFR 50.36, Criterion 3 is not applicable. The applicable Bases statement was:
Two OPERABLE source range neutron flux monitors are required to provide a visual signal to alert the operator to unexpected changes in core reactivity such as with a boron dilution accident... or an improperly loaded fuel assembly...
The licensees basis for its position is (1) a boron dilution event does not need to be considered for plants like Comanche Peak with TS 3.9.2, (2) there is no licensing basis analysis assumption to detect misloaded fuel assemblies during refueling operations with the SRMs, and (3) SRMs cannot detect misloaded fuel assemblies.
The licensee states that the appropriate location for the requirements in TS 3.9.3 is the TRM, which contains selected requirements that do not meet the criteria for inclusion in the TSs but are important to the operation of Comanche Peak. The changes to the TRM are controlled by the 10 CFR 50.59, Changes, tests and experiments, process.
NRC Evaluation The NRC staff reviewed the licensees submittal and has determined that the submittal misapplies 10 CFR 50.36, Criterion 3 and the rationale for how TS 3.9.3 meets Criterion 3 in the TS Bases. Therefore, the SRMs continue to meet Criterion 3 and should remain in the TSs. The staffs reasoning and explanations are provided below.
The prevention of criticality during refueling operations by the detection of unexpected changes in core reactivity is an applicable event.
Although an inadvertent criticality is not an explicit design basis accident, the prevention of criticality during refueling operations is part of the Comanche Peak licensing basis and thus is an implicit applicable event as described in the final policy statement for TSs as published in the Federal Register (FR) (58 FR 39132), effective July 22, 1993:
A safety sequence analysis is a systematic examination of the actions required to mitigate the consequences of events considered in the plants Design Basis Accident and Transient analyses, as presented in Chapters 6 and 15 of the plants FSAR (or equivalent chapters). Such a safety sequence analysis considers all applicable events, whether explicitly or implicitly presented.
There are several Comanche Peak TSs related to the prevention of criticality during refueling operations, and therefore, an inadvertent criticality event during refueling is a part of the Comanche Peak licensing basis and an applicable event:
o TS 3.9.1, Boron Concentration, requires the boron concentration in the core to be maintained at a specified limit to limit the core reactivity during planned refueling operations.
o TS 3.9.2 requires each valve used to isolate unborated water sources to be closed and secured to prevent unplanned boron dilution of the reactor coolant.
o TS 3.9.3 requires two SRMs to be operable. As explained in the Applicable Safety Analyses section of the TS Bases, the SRMs provide a visual signal to alert the operator to unexpected changes in core reactivity such as with a boron dilution accident or an improperly loaded fuel assembly. The unexpected changes in core reactivity would result from unintended events such as unplanned boron dilution from leaking isolation valves or unplanned core alterations which add reactivity.
In the absence of the SRMs required by TS 3.9.3, there is no direct means of monitoring criticality and thus verifying core configuration remains subcritical (Keff 0.95) under the required loading conditions. This opens the potential for a criticality transient/accident occurring during the fuel loading process due to misloads not being detected.
Unexpected changes in core reactivity may occur from boron dilution or unplanned core alterations.
The licensee indicated that the boron dilution event does not need to be considered for plants like Comanche Peak with TS 3.9.2. The NRC staff notes that isolating unborated water sources removes the need for an SRM audible alarm or count rate function but not visual monitoring. The Westinghouse STS Bases includes two TS 3.9.3 reviewers notes.
The reviewers note in the Applicable Safety Analyses Section states that the need for a safety analysis for an uncontrolled boron dilution accident is eliminated by isolating all unborated water sources as required by TS 3.9.2. The reviewers note in the Background section indicates that plants that isolate boron dilution paths (per TS 3.9.2) do not require SRM channels to provide an audible alarm or count rate function, but the SRMs are still expected to provide a visual monitoring function. Therefore, TS 3.9.3 remains applicable to plants with the TS 3.9.2 to isolate all boron dilution paths.
The LAR only partially address the basis for the SRMs (as discussed in TS 3.9.3 Bases) by implying that the function of the SRMs is to detect an improperly loaded fuel assembly. The actual function of the SRMs, from the TS Bases, is to detect unexpected changes in core reactivity, which can result due to events such as a boron dilution accident or a misloaded assembly (and provide a signal to alert the operator). Fuel assembly misloads, broadly understood as a range of unplanned core alterations involving either fuel or reactivity control elements, can be postulated to result in unexpected changes in core reactivity and thus the description and statements in the TS 3.9.3 Bases are still deemed to be accurate and appropriate. The LAR postulates fuel assembly misloads that do not result in unexpected changes in core reactivity and then erroneously asserts that the TS Bases were in error.
Therefore, the LAR provides a flawed and/or incomplete rationale for modifying the plant specific TS Bases.
The SRMs are the primary success path to detect unexpected changes in core reactivity.
The SRMs are the only method identified to monitor core reactivity changes during refueling operations, in that there are no other direct means to perform this function.
UFSAR Section 14.2.10, Initial Fuel Loading, Criticality, and Power Operation, lays out multiple criteria to stop fuel loading operations based on neutron count rates which are detected by the SRMs.
Each subsequent fuel addition will be accompanied by detailed neutron count rate monitoring to determine that the just-loaded fuel assembly does not excessively increase the count rate and that the extrapolated inverse count rate ratio is behaving as expected. These results for each loading step will be evaluated before the next fuel assembly is placed in the active core region. The final, as loaded, core configuration will be subcritical (Keff 0.95) under the required loading conditions.
As indicated in the FSAR description, detailed neutron monitoring during fuel additions is necessary to provide an indication that the core is being loaded as expected and remains subcritical during refueling operations.
An unplanned criticality during refueling operations presents a challenge to the integrity of a fission product barrier.
The impact of an inadvertent criticality during refueling operations on the fuel cladding fission product barrier has not been analyzed and one fission product barrier, the reactor coolant system, is open during refueling operations. The final fission product barrier, the containment, may be open or closed in accordance with TS 3.9.4, Containment Penetrations, which requires various containment penetrations to be closed or capable of being closed. An inadvertent criticality during any plant Mode would generate excessive heat (greater than decay heat) which could lead to departure from nucleate boiling (DNB), and thus adversely impact the fuel cladding.
Conclusion In summary, monitoring the core reactivity condition using the SRMs during refueling operations satisfies Criterion 3 because the system (SRMs) is part of the primary success path (there are no other direct means for monitoring core reactivity during refueling) which functions to mitigate an applicable event (alerts the operator to an unexpected change in core reactivity) that presents a challenge to the integrity of a fission product barrier (inadvertent criticality). In addition, the TS Bases appropriately describes how TS 3.9.3 meets Criterion 3. Therefore, the Comanche Peak TS 3.9.3 satisfies Criterion 3 and is required by 10 CFR 50.36(c)(2)(ii). Thus, the LAR does not provide a valid basis for the request and the LAR is not acceptable for review.
The staff anticipates that the licensee would need to provide an alternative basis to justify relocating TS 3.9.3. However, given the staffs concerns stated above, any supplemental information to the current LAR would represent an essentially new request. Additional consideration for any new request should include, among other things, appropriate analyses to consider criticality events during refueling operations and the impacts of criticality events on fission product barriers. In addition, the NRC has not issued guidance and has not approved methodologies for evaluating criticality events and analyzing their consequences during refueling operations. Therefore, the NRC staff does not anticipate the licensee could produce such analysis as a supplement to the current LAR as an alternative to having SRMs detect unexpected reactivity changes.
Based on the foregoing, NRC staff activities have ceased, and the associated Enterprise Project Identifier number has been closed.
If you have any questions, please contact me at (301) 415-6256 or dennis.galvin@nrc.gov.
Sincerely,
/RA/
Dennis J. Galvin, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-445 and 50-446 cc: Listserv
ML23348A239 NRR-058 OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL3/LA NRR/DSS/SFNB/BC NAME DGalvin ABaxter SKrepel DATE 12/13/2023 12/15/2023 12/13/2023 OFFICE NRR/DSS/SNSB/BC NRR/DSS/STSB/BC (A)
NRR/DORL/LPL4/BC NAME PSahd SMehta (Cashley for)
JRankin DATE 12/13/2023 12/13/2023 12/13/2023 OFFICE OGC NRR/DSS/D NRR/DSS/D NAME SVrahoretis JDonoghue BPham DATE 12/14/2023 12/15/2023 12/15/2023 OFFICE NRR/DORL/LPL4/PM NAME DGalvin DATE 12/19/2023