ML20134H887

Technical Specification Bases 3.2

ML20134H887
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Site:	Grand Gulf
Issue date:	05/12/2020
From:	Siva Lingam NRC/NRR/DORL/LPL4
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Lingam S
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Download: ML20134H887 (19)

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APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Average Planar Linear Heat Generation Rate (APLHGR)

BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that the fuel design limits identified in Reference 1 are not exceeded during anticipated operational occurrences (AOOs) and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel design limits are presented in the UFSAR, Chapters 4, 6, and 15, and in References 2 and 6. The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs), anticipated operational transients, and normal operations that determine APLHGR limits are presented in UFSAR, Chapters 4, 6, and 15, and in References 2, 3, 4 and 6.

Fuel design evaluations are performed to demonstrate that the 1% limit on the fuel cladding strain and other fuel design limits described in References 1 and 6 are not exceeded during AOOs for operation with LHGR up to the operating limit LHGR. APLHGR limits are developed as a function of exposure and, along with the LHGR limits, ensure adherence to fuel design limits during the limiting AOOs (Refs. 2 and 3).

LOCA analyses are performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46. The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A discussion of the fuel vendor analysis codes is provided in References 5 and 6 for Siemens and GE, respectively. The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly. The Siemens APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA (continued)

GRAND GULF B 3.2-1 LBDCR 20026

APLHGR B 3.2.1 BASES (continued)

SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is t 21.8% RTP and periodically thereafter. They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER t 21.8% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels. The time interval based Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. UFSAR, Chapter 4.

2. UFSAR, Chapter 15, Appendix 15C.

3. UFSAR, Chapter 15, Appendix 15D.

4. XN-NF-80-19(P)(A), "Exxon Nuclear Methodology for Boiling Water Reactors, Neutronics Methods for Design and Analysis," Volume 1 (as supplemented).

5. XN-NF-80-19(A), "Exxon Nuclear Methodology for Boiling Water Reactors, ECCS Evaluation Model," Volume 2 (as supplemented).

6. NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR)."

GRAND GULF B 3.2-4 LBDCR 18127

MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Minimum Critical Power Ratio (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power. The operating limit MCPR is established to ensure that no fuel damage results during anticipated operational occurrences (AOOs) and that 99.9% of the fuel rods are not susceptible to boiling transition if the limit is not violated.

Although fuel damage does not necessarily occur if a fuel rod actually experiences boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs. Based on these experimental data, correlations have been developed to predict critical bundle power (i.e., the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling). Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the AOOs to establish the operating limit MCPR are presented in the UFSAR, Chapters 4, 6, and 15, and References 2, 3, 4, and 5. To ensure that the MCPR Safety Limit (SL) is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR).

The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest change in CPR (CPR). When the largest CPR is combined with the SLMCPR99.9%, the required operating limit MCPR is obtained.

MCPR99.9% is determined to ensure more than 99.9% of the fuel rods in the core are not susceptible to boiling transition using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the approved Critical Power correlations. Details of the MCPR99.9% calculation are given in Reference 2. Reference 2 also includes a tabulation of the uncertainties and the nominal values of the parameters used in the MCPR99.9%

statistical analysis.

(continued)

GRAND GULF B 3.2-5 LBDCR 18129

MCPR B 3.2.2 BASES APPLICABLE The MCPR operating limits are derived from the MCPR99.9% value and SAFETY ANALYSES the transient analysis, and are dependent on the operating core flow (continued) and power state (MCPRf and MCPRp, respectively) to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 3, 4, and 5). Flow dependent MCPR limits are determined by steady state thermal hydraulic methods using the three dimensional BWR simulator code (Ref. 6) and the steady state thermal hydraulic code (Ref. 2). MCPRf curves are provided based on the maximum credible flow runout transient for Loop Manual operation. The result of a single failure or single operator error during Loop Manual operation is the runout of only one loop because both recirculation loops are under independent control.

Power dependent MCPR limits (MCPRp) are determined by approved transient analysis models. (Ref. 7). The MCPRp limits are established for a set of exposure intervals. The limiting transients are analyzed at the limiting exposure for each interval. Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, high and low flow MCPRp operating limits are provided for operating between 21.8% RTP and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of the NRC Policy Statement.

LCO The MCPR operating limits specified in the COLR MCPR99.9% value, MCPRf values, and MCPRp values are the result of the Design Basis Accident (DBA) and transient analysis. The MCPR operating limits are determined by the larger of the MCPRf and MCPRp limits, which are based on the MCPR99.9% limit specified in the COLR.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 21.8% RTP, the reactor is operating at a slow recirculation pump speed and the moderator void ratio is small. Surveillance of thermal limits below 21.8% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

(continued)

GRAND GULF B 3.2-6 LBDCR 18129

MCPR B 3.2.2 BASES (continued)

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is t 21.8% RTP and periodically thereafter. It is compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER reaches t 21.8% RTP is acceptable given the large inherent margin to operating limits at low power levels.

The time interval based Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

SR 3.2.2.2 Because the transient analyses may take credit for conservatism in the control rod scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analyses. SR 3.2.2.2 determines the actual scram speed distribution and compares it with the assumed distribution. The MCPR operating limit is then determined based either on the applicable limit associated with scram times of LCO 3.1.4, "Control Rod Scram Times," or the realistic scram times. The scram time dependent MCPR limits are contained in the COLR. This determination must be performed and any necessary changes must be implemented within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each set of control rod scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle or after maintenance that could affect scram times. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is acceptable due to the relatively minor changes in the actual control rod scram speed distribution expected during the fuel cycle.

REFERENCES 1. NUREG-0562, "Fuel Failures As A Consequence of Nucleate Boiling or Dry Out," June 1979.

2. NEDE-24011-P-A General Electric Standard Application for Reactor Fuel (GESTAR II).

3. UFSAR, Chapter 15, Appendix 15B.

4. UFSAR, Chapter 15, Appendix 15C.

(continued)

GRAND GULF B 3.2-8 LBDCR 18127

LHGR B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 Linear Heat Generation Rate (LHGR)

BASES BACKGROUND The LHGR is a measure of the heat generation rate of a fuel rod in a fuel assembly at any axial location. Limits on the LHGR are specified to ensure that fuel design limits are not exceeded anywhere in the core during normal operation, including anticipated operational occurrences (AOOs). Exceeding the LHGR limit could potentially result in fuel damage and subsequent release of radioactive materials. Fuel design limits are specified to ensure that fuel system damage, fuel rod failure or inability to cool the fuel does not occur during the anticipated operating conditions identified in UFSAR Chapters 6 and 15.

APPLICABLE The analytical methods and assumptions used in evaluating SAFETY ANALYSES the fuel system design are presented in the UFSAR, Chapters 4, 6, and 15, and in References 1 and 2. The fuel assembly is designed to ensure (in conjunction with the core nuclear and thermal hydraulic design, plant equipment, instrumentation, and protection system) that fuel damage will not result in the release of radioactive materials in excess of the guidelines of 10 CFR, Parts 20 and 50. The mechanisms that could cause fuel damage during operational transients and that are considered in fuel evaluations are:

a. Rupture of the fuel rod cladding caused by strain from the relative expansion of the UO2 pellet; and

b. Severe overheating of the fuel rod cladding caused by inadequate cooling.

A value of 1% strain of the fuel cladding has been defined as the limit below which fuel damage caused by overstraining of the fuel cladding is not expected to occur (Ref. 3).

Fuel design evaluations have been performed and demonstrate that the 1% fuel cladding strain design limit is not exceeded during continuous operation with LHGRs up to the (continued)

GRAND GULF B 3.2-9 LBDCR 20026

LHGR B 3.2.3 BASES ACTIONS A.1 (continued) restore the LHGR(s) to within its required limit(s) such that the plant is operating within analyzed conditions and within the design limits of the fuel rods. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is normally sufficient to restore the LHGR(s) to within its limit and is acceptable based on the low probability of a transient or Design Basis Accident occurring simultaneously with the LHGR out of specification.

B.1 If the LHGR cannot be restored to within its required limit within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 21.8% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to < 21.8% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The LHGRs are required to be initially calculated within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is t 21.8% RTP and periodically thereafter. They are compared with the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER t 21.8% RTP is achieved is acceptable given the large inherent margin to operating limits at lower power levels. The time interval based Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. UFSAR, Chapter 15.

2. UFSAR, Chapter 4.

3. NUREG-0800, "Standard Review Plan," Section 4.2, II.A.2(g),

Revision 2, July 1981.

GRAND GULF B 3.2-11 LBDCR 18127