ML24215A177
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Response to SDAA Audit Question Question Number: A-16-4 Receipt Date: 07/17/2023 Question:
Besides using the terms end of cycle (EOC) and beginning of cycle (BOC), US460 proposed generic TS and Bases may be inconsistent when describing the cyclic period from initial unit startup after completion of a core reload, which is the BOC, until unit shutdown for a refueling outage at EOC.
Page 1.1-3, COLR definition, uses cycle specific parameter limits and reload cycle Page 5.6-2, Specification 5.6.3, uses reload cycle Page 5.6-4, Specification 5.6.3, uses mid-cycle and reload cycle Page B 3.1.2-1, Background section, uses cycle burnup, fuel cycle, and cycle Page B 3.1.2-2, ASA section, uses fuel cycle and reload cycle Page B 3.1.2-3, ASA section, uses during the cycle Page B 3.1.2-5, SRs section, uses operating cycle and fuel cycle Page B 3.1.3-1, Background section, uses fuel cycle Page B 3.1.3-2, Background section, uses fuel cycle Page B 3.1.3-3, ASA section, uses core operating cycle Page B 3.1.3-3, LCO section, uses fuel cycle design Page B 3.1.3-4, Actions section, uses fuel cycle designs Page B 3.1.3-4, SRs section, uses fuel cycle and during fuel cycle operation Page B 3.1.8-3, ASA section, uses reload fuel cycle Page B 3.1.8-4, ASA section, uses fuel cycle Page B 3.2.1-3, SRs section, uses fuel cycle Page B 3.4.1-2, LCO section, uses from cycle to cycle Page B 3.5.4-2, Applicability section, uses fuel cycle NuScale Nonproprietary NuScale Nonproprietary
Response
NuScale revises the GTS Technical Specifications and Bases to ensure consistent use of the term fuel cycle.
Markups of the affected changes, as described in the response, are provided below:
NuScale Nonproprietary NuScale Nonproprietary
Definitions 1.1 NuScale US460 1.1-3 Draft Revision 1 1.1 Definitions CHANNEL RESPONSE TIME The CHANNEL RESPONSE TIME is the time from when the process variable exceeds its setpoint until the output from the channel analog logic reaches the input of the digital portion of the Module Protection System digital logic.
CORE OPERATING LIMITS REPORT (COLR)
The COLR is the unit-specific document that provides fuel cycle specific parameter limits for the current reloadfuel cycle. These fuel cycle specific parameter limits shall be determined for each reloadfuel cycle in accordance with Specification 5.6.3. Module operation within these parameter limits is addressed in individual Specifications.
DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The dose conversion factors used for this calculation shall be those listed in Table 2.1 of EPA Federal Guidance Report No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, EPA-520/1-88-020, September 1988.
DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same effective dose equivalent as the quantity and isotopic mixture of noble gases (Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138) actually present. The dose conversion factors used for this calculation shall be those listed in Table III.1 of EPA Federal Guidance Report No. 12, External Exposure to Radionuclides in Air, Water, and Soil, EPA 402-R-93-081, September 1993.
INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).
Core Reactivity B 3.1.2 NuScale US460 B 3.1.2-1 Draft Revision 1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Core Reactivity BASES BACKGROUND According to GDC 26, GDC 28, and GDC 29 (Ref. 1), reactivity shall be controllable, such that subcriticality is maintained under cold conditions, and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity balance is used as a measure of the predicted versus measured core reactivity during power operation. The periodic confirmation of core reactivity is necessary to ensure that Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity difference could be the result of unanticipated changes in fuel, control rod assembly (CRA) worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3.1.1, SHUTDOWN MARGIN (SDM))
in ensuring the reactor can be brought safely to cold, subcritical conditions.
When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance since parameters are being maintained relatively stable under steady-state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers producing zero net reactivity. Excess reactivity can be inferred from the boron letdown curve (or critical boron curve), which provides an indication of the soluble boron concentration in the reactor coolant system (RCS) versus fuel cycle burnup. Periodic measurement of the RCS boron concentration for comparison with the predicted value with other variables fixed (such as rod height, temperature, pressure, and power), provides a convenient method of ensuring that core reactivity is within design expectations, and that the calculation models used to generate the safety analysis are adequate.
In order to achieve the required fuel cycle energy output, the uranium enrichment, in the new fuel loading and in the fuel remaining from the previous fuel cycle, provides excess positive reactivity beyond that required to sustain steady state operation throughout the fuel cycle.
When the reactor is critical the excess positive reactivity is
Core Reactivity B 3.1.2 NuScale US460 B 3.1.2-2 Draft Revision 1 BASES BACKGROUND (continued) compensated by burnable absorbers (if any), control rods, whatever neutron poisons (mainly xenon and samarium) are present in the fuel, and the RCS boron concentration.
When the core is producing THERMAL POWER, the fuel is being depleted and excess reactivity is decreasing. As the fuel depletes, the RCS boron concentration is reduced to decrease negative reactivity and maintain constant THERMAL POWER. The boron letdown curve is based on steady state operation at RTP. Therefore, deviations from the predicted boron letdown curve may indicate deficiencies in the design analysis, deficiencies in the calculational models, or abnormal core conditions, and must be evaluated.
APPLICABLE SAFETY ANALYSES The acceptance criteria for core reactivity are that the reactivity balance limit ensures plant operation is maintained within the assumptions of the safety analyses.
Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations. Accident evaluations (Ref. 2) are, therefore, dependent upon accurate evaluation of core reactivity. In particular, SDM and reactivity transients, such as CRA withdrawal accidents or CRA ejection accidents, are sensitive to accurate predictions of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks.
Monitoring reactivity balance provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity.
Design calculations and safety analysis are performed for each fuel cycle for the purpose of predetermining reactivity behavior and the RCS boron concentration requirements for reactivity control during fuel depletion.
The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted RCS boron concentrations for identical core conditions at beginning of cycle (BOC) do not agree, then the assumptions used in the reloadfuel cycle design analysis or the calculation models used to predict soluble boron requirements may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured boron concentration.
Core Reactivity B 3.1.2 NuScale US460 B 3.1.2-3 Draft Revision 1 BASES APPLICABLE SAFETY ANALYSES (continued)
Thereafter, any significant deviations in the measured boron concentration from the predicted boron letdown curve that develop during fuel depletion may be an indication that the calculational model is not adequate for core burnups beyond BOC, or that an unexpected change in core conditions has occurred.
The normalization of predicted RCS boron concentration to the measured value is typically performed after reaching RTP following startup from a refueling outage, with the CRAs in their normal positions for power operation. The normalization is performed at BOC conditions so that core reactivity relative to predicted values can be continually monitored and evaluated as core conditions change during the fuel cycle.
Core reactivity satisfies Criterion 2 of 10 CFR 50.36©(2)(ii).
LCO Long term core reactivity behavior is a result of the core physics design and cannot be easily controlled once the core design is fixed.
During operation, therefore, the Conditions of the LCO can only be ensured through measurement and tracking, and appropriate actions taken as necessary. Large differences between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the nuclear design methodology are larger than expected. A limit on the reactivity balance of +/- 1% k/k has been established based on engineering judgment and operating experience. A 1% deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated.
When measured core reactivity is within 1% k/k of the predicted value at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Since deviations from the limit are normally detected by comparing predicted and measured steady state RCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached. These values are well within the uncertainty limits for analysis of boron concentration samples, so that spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely.
Core Reactivity B 3.1.2 NuScale US460 B 3.1.2-5 Draft Revision 1 BASES ACTIONS (continued) acceptable for continued operation, then the boron letdown curve may be renormalized and power operation may continue. If operational restriction or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined.
The required Completion Time of 7 days is adequate for preparing and implementing whatever operating restrictions that may be required to allow continued reactor operation.
B.1 If the core reactivity cannot be restored to within the 1% k/k limit, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 2 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the SDM for MODE 2 is not met, then boration may be required to meet SR 3.1.1.1 prior to entry into MODE 2. The allowed Completion Time is reasonable, for reaching MODE 2 from full power conditions in an orderly manner.
SURVEILLANCE REQUIREMENTS SR 3.1.2.1 Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made considering that other core conditions are fixed or stable, including CRA position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to exceeding 5% RTP as an initial check on core conditions and design calculations at BOC. The Surveillance is performed again prior to exceeding 60 effective full power days (EFPDs) to confirm the core reactivity is responding to reactivity predictions and then periodically thereafter during the fueloperating cycle in accordance with the Surveillance Frequency Control Program.
The SR is modified by a Note indicating that the predicted core reactivity may be adjusted to the measured value provided this normalization is performed prior to exceeding a fuel burnup of 60 EFPDs. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations.
The subsequent Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
MTC B 3.1.3 NuScale US460 B 3.1.3-1 Draft Revision 1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Moderator Temperature Coefficient (MTC)
BASES BACKGROUND According to GDC 11 (Ref. 1), the reactor core and its interaction with the reactor coolant system (RCS) must be designed for inherently stable power operation even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.
The MTC relates a change in core reactivity to a change in reactor coolant temperature (a positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature). The reactor is designed to operate with a non-positive MTC during the majority of fuel cycle operation. Therefore, a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self-limiting, and stable power operation will result. There are times at the beginning of fuel cycle and at less than normal operating temperature the MTC may be slightly positive.
MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by measurements. Both initial and reload cores are designed so that the MTC is less than zero when reactor power is at RTP. The actual value of the MTC is dependent on core characteristics such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons (burnable absorbers) to yield an MTC within the range analyzed in the plant accident analysis.
The end of cycle (EOC) MTC is also limited by the requirements of the accident analysis. Fuel cycles that are designed to achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the EOC limit.
The limitations on MTC are provide to nsure that the value of this coefficient remains within the limiting conditions assumed in the FSAR accident and transient analyses (Ref. 2).
If the LCO limits are not met, the unit response during transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity.
MTC B 3.1.3 NuScale US460 B 3.1.3-3 Draft Revision 1 BASES APPLICABLE SAFETY ANALYSES (continued) one, which is assumed withdrawn. Even if the reactivity increase produces slightly subcritical conditions, a large fraction of core power may be produced through the effects of subcritical neutron multiplication.
MTC values are bounded in reload safety evaluations assuming steady state conditions at core beginning of cycle (BOC) and EOC. A measurement is conducted two-thirds of the way through the fuelcore operating cycle. The measured value may be extrapolated to project the EOC value, in order to confirm reload design predictions.
MTC satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO LCO 3.1.3 requires the MTC to be within specified limits of the COLR to ensure the core operates within the assumptions of the accident analysis. During the reload core safety evaluation, the MTC is analyzed to determine that its values remain within the bounds of the original accident analysis during operation. The limit on a least negative MTC ensures that core overheating accidents will not violate the accident analysis assumptions. The most negative MTC limit for EOC specified in the COLR ensures that core overcooling accidents will not violate the accident analysis assumptions.
MTC is a core physics parameter determined by the fuel and fuel cycle design and cannot be easily controlled once the core design is fixed.
During operation, therefore, the LCO can only be ensured through measurement. The surveillance checks of MTC at BOC and near two-thirds of core burnup provide confirmation that the MTC is behaving as anticipated, so that the acceptance criteria are met.
APPLICABILITY In MODE 1, the upper limit on the MTC must be maintained to ensure that any accident will not violate the design assumptions of the accident analysis. The limits must also be maintained to ensure startup and subcritical accidents, such as the uncontrolled CRA withdrawal, will not violate the assumptions of the accident analysis.
The lower MTC limit must be maintained in MODES 1 and 2 and MODE 3 with any RCS temperature 200 °F, to ensure that cooldown accidents will not violate the assumptions of the accident analysis.
RCS Pressure, Temperature, and Flow Resistance CHF Limits B 3.4.1 NuScale US460 B 3.4.1-2 Draft Revision 1 BASES APPLICABLE SAFETY ANALYSES (continued)
The NSP4 correlation limit is used for comparison to conditions representative of normal operation, operational transients, anticipated operational occurrences, and accidents other than events that are initiated by rapid reductions in primary system inventory. The NSPN-1 correlation is used to evaluate events for which analyses postulate a rapid reduction in primary system inventory. An assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, Regulating Bank Insertion Limits; LCO 3.2.1, "Enthalpy Rise Hot Channel Factor (FH)," and LCO 3.2.2, AXIAL OFFSET (AO).
The flow resistance in the RCS directly affects the reactor coolant natural circulation flow rate established by THERMAL POWER, RCS pressure, and RCS temperature. The safety analyses assume flow rates that are based on a conservative value of flow resistance through the RCS.
Therefore the resistance must be verified to ensure that the assumptions in the safety analyses remain valid.
The pressurizer pressure operating limit and the RCS temperature limit specified in the COLR, as shown on the Analytical Design Operating Limits in FSAR Section 4.4 (Ref. 2), correspond to operating limits, with an allowance for steady state fluctuations and measurement errors.
These are the analytical initial conditions assumed in transient and LOCA analyses.
The RCS CHF parameters satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO This LCO specifies limits on the monitored process variables, pressurizer pressure and RCS cold temperature to ensure the core operates within the limits assumed in the safety analyses. It also specifies the limit on RCS flow resistance to ensure that the RCS flow is consistent with the flow assumed in the safety analyses. These variables are contained in the COLR to provide operating and analysis flexibility from fuel cycle to fuel cycle. Operating within these limits will result in meeting CHFR criterion in the event of a CHF-limited transient.
RCS P/T Limits B 3.4.3 NuScale US460 B 3.4.3-3 Draft Revision 1 BASES LCO (continued)
The LCO limits apply to all components of the RCS. These limits define allowable operating regions and permit a large number of fueloperating cycles while providing a wide margin to nonductile failure.
The limits for the rate of change of temperature control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, and cooldown P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves.
Violating the LCO limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCPB components. The consequences depend on several factors, as follows:
- a. The severity of the departure from the allowable operating P/T regime or the severity of the rate of change of temperature;
- b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
- c. The existences, sizes, and orientations of flaws in the vessel material.
APPLICABILITY The RCS P/T limits LCO provides a definition of acceptable operation for prevention of nonductile (brittle) failure in accordance with 10 CFR 50, Appendix G (Ref. 1). Although the P/T limits were developed to provide guidance for operation primarily during heatup or cooldown or required testing, they are applicable at all times in keeping with the concern for nonductile failure.
During MODE 1 other Technical Specifications provide limits for operation that can be more restrictive than, or can supplement these P/T limits.
LCO 3.4.1, RCS Pressure, Temperature, and Flow Resistance Critical Heat Flux (CHF) Limits, LCO 3.4.2, RCS Minimum Temperature for Criticality, and Safety Limit 2.1.2, Reactor Coolant System (RCS)
Pressure SL, also provide operational restrictions for pressure and temperature and maximum pressure. Furthermore, MODE 1 is above the temperature range of concern for nonductile failure, and stress analyses have been performed for normal maneuvering profiles, such as power ascension or descent.
Reporting Requirements 5.6 NuScale US460 5.6-2 Draft Revision 1 5.6 Reporting Requirements 5.6.3 Core Operating Limits Report (COLR)
- a.
Core operating limits shall be established prior to each reloadfuel cycle, or prior to any remaining portion of a reloadfuel cycle, and shall be documented in the COLR for the following:
3.1.1, SHUTDOWN MARGIN (SDM);
3.1.3, Moderator Temperature Coefficient (MTC);
3.1.4, Rod Group Alignment Limits; 3.1.5, Shutdown Bank Insertion Limits; 3.1.6, Regulating Bank Insertion Limits; 3.1.8, PHYSICS TESTS Exceptions; 3.1.9, Boron Dilution Control; 3.2.1, Enthalpy Rise Hot Channel Factor ( FH);
3.2.2, AXIAL OFFSET (AO);
3.4.1, RCS Pressure, Temperature, and Flow Resistance Critical Heat Flux (CHF) Limits; 3.5.3, Ultimate Heat Sink; 3.5.4, Emergency Core Cooling System Supplemental Boron (ESB), and 3.8.1, Nuclear Instrumentation.
- b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
[-----------------------------------REVIEWERS NOTE----------------------------------
The COL applicant shall confirm the validity of each listed document and the listed Specifications for the associated core operating limits, or state the valid NRC approved analytical method document and list of associated Specifications.
The COL applicant shall state the valid core reload analysis methodology document and list of associated Specifications.
]
Reporting Requirements 5.6 NuScale US460 5.6-4 Draft Revision 1 5.6 Reporting Requirements 5.6.3 Core Operating Limits Report (COLR) (continued)
- 4.
[NuScale SDA, NuScale FSAR, Section 4.4, Thermal and Hydraulic Design, Revision 0, December 2022; TR-0516-49416-P, Non-Loss-of-Coolant Accident Analysis Methodology, Revision 4, December 2022 (NuScale Proprietary);
TR-0516-49422-P, "Loss-of-Coolant Accident Evaluation Methodology," Revision 3, December 2022 (NuScale Proprietary);
TR-0915-17564-P-A, Subchannel Analysis Methodology, Revision 2, February 2019 (NuScale Proprietary); TR-108601-P, Statistical Subchannel Analysis Methodology, Supplement 1 to TR-0915-17564-P-A, Revision 2, Revision 2, December 2022; and TR-0716-50350-P, "Rod Ejection Accident Methodology,"
Revision 2, December 2022 (NuScale Proprietary).
(Methodology for Specification 3.4.1 - RCS Pressure, Temperature, and Flow Resistance CHF Limits.)]
- 5.
[NuScale SDA, NuScale FSAR, Section 4.3, Nuclear Design, Revision 0, December 2022; TR-0616-48793-P-A, Nuclear Analysis Codes and Methods Qualification, Revision 1, November 2018 (NuScale Proprietary); and TR-124587, Extended Passive Cooling and Reactivity Control Methodology Topical Report, Revision 0, December 2022, (NuScale Proprietary).
(Methodology for Specifications 3.5.3 - Ultimate Heat Sink, 3.5.4 - Emergency Core Cooling System Supplemental Boron (ESB), and 3.8.1 - Nuclear Instrumentation.)]
- c.
The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Passive Core Cooling Systems limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.
- d.
The COLR, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reloadfuel cycle to the NRC.