ML23067A186

Rev. 18 to Technical Specification, Bases Change, Chapter 3.2, Power Distribution Limits

ML23067A186
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Site:	Callaway
Issue date:	12/29/2022
From:	Union Electric Co, Ameren Missouri
To:	Office of Nuclear Reactor Regulation
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Text

CHAPTER TABLE OF CONTENTS CHAPTER B 3.2 POWER DISTRIBUTION LIMITS Section Page B 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

(FQ Methodology) .........................................................................................B 3.2.1-1 BACKGROUND .......................................................................................B 3.2.1-1 APPLICABLE SAFETY ANALYSES .......................................................B 3.2.1-2 LCO .........................................................................................................B 3.2.1-2 APPLICABILITY .....................................................................................B 3.2.1-4 ACTIONS ..................................................................................................B 3.2.1-4 SURVEILLANCE REQUIREMENTS .........................................................B 3.2.1-7 REFERENCES ......................................................................................B 3.2.1-11 B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor ..........................................................................................................B 3.2.2-1 BACKGROUND ......................................................................................B 3.2.2-1 APPLICABLE SAFETY ANALYSES ......................................................B 3.2.2-2 LCO ........................................................................................................B 3.2.2-3 APPLICABILITY .....................................................................................B 3.2.2-4 ACTIONS ..................................................................................................B 3.2.2-4 SURVEILLANCE REQUIREMENTS .........................................................B 3.2.2-6 REFERENCES........................................................................................ B 3.2.2-7 B 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology) ..........................................................B 3.2.3-1 BACKGROUND ......................................................................................B 3.2.3-1 APPLICABLE SAFETY ANALYSES ......................................................B 3.2.3-1 LCO ........................................................................................................B 3.2.3-2 APPLICABILITY .....................................................................................B 3.2.3-3 ACTIONS ..................................................................................................B 3.2.3-3 SURVEILLANCE REQUIREMENTS .........................................................B 3.2.3-3 REFERENCES .......................................................................................B 3.2.3-3 B 3.2.4 QUADRANT POWER TILT RATIO (QPTR) .................................................B 3.2.4-1 BACKGROUND ......................................................................................B 3.2.4-1 APPLICABLE SAFETY ANALYSES ......................................................B 3.2.4-1 CALLAWAY PLANT 3.2-i

CHAPTER TABLE OF CONTENTS (Continued)

Section Page LCO ........................................................................................................B 3.2.4-2 APPLICABILITY .....................................................................................B 3.2.4-2 ACTIONS ..................................................................................................B 3.2.4-2 SURVEILLANCE REQUIREMENTS .........................................................B 3.2.4-5 REFERENCES .......................................................................................B 3.2.4-7 3.2-ii

FQ(Z) (FQ Methodology)

B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (FQ(Z)) (FQ Methodology)

BASES BACKGROUND The purpose of the limits on the values of FQ Z is to limit the local (i.e., pellet) peak power density. The value of F Q Z varies along the axial height (Z) of the core.

F Q Z is defined as the maximum local fuel rod linear power density divided by the average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions. Therefore, F Q Z is a measure of the peak fuel pellet power within the reactor core.

During power operation, the global power distribution is limited by LCO 3.2.3, "Axial Flux Difference (AFD)," and LCO 3.2.4, "Quadrant Tilt Power Ratio (QPTR)," which are directly and continuously measured process variables. These LCOs, along with LCO 3.1.4, "Rod Group Alignment Limits," LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits," maintain the core limits on power distributions on a continuous basis.

F Q Z varies with fuel loading patterns, bank insertion, fuel burnup, and changes in axial power distribution.

F Q Z is not directly measurable but is inferred from a power distribution measurement obtained with either the movable incore detector system or from an OPERABLE power distribution monitoirng system (PDMS)

(Reference 5). The results of the three-dimensional power distribution measurement are analyzed to derive a measured value for F Q Z . These measurements are generally taken with the core at or near equilibrium conditions.

However, because this value represents an equilibrium condition, it does not include the variations in the value of FQ Z that are present during nonequilibrium situations, such as load following. To account for these possible variations, the steady state value of FQ Z is adjusted by an elevation dependent factor that accounts for the calculated worst case transient conditions.

Core monitoring and control under nonsteady state conditions are accomplished by operating the core within the limits of the appropriate (continued)

CALLAWAY PLANT B 3.2.1-1 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES BACKGROUND LCOs, including the limits on AFD, QPTR, and control and shutdown (continued) bank insertion.

APPLICABLE This LCO precludes core power distributions that violate the following fuel SAFETY design criteria:

ANALYSES

a. During a large break loss of coolant accident (LOCA), the peak cladding temperature must not exceed 2200F (Ref. 1);

b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a departure from nucleate boiling (DNB) condition;

c. During an ejected rod accident, the average fuel pellet enthalpy at the hot spot in irradiated fuel must not exceed 200 cal/gm (Ref. 2);

and

d. The control and shutdown rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth rod stuck fully withdrawn (Ref. 3).

Limits on F Q Z ensure that the value of the initial total peaking factor assumed in the accident analyses remains valid. Other criteria must also be met (e.g., maximum cladding oxidation, maximum hydrogen generation, coolable geometry, and long term cooling). However, the LOCA peak cladding temperature is typically most limiting.

FQ Z limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the FQ Z limit assumed in safety analyses for other postulated accidents. Therefore, this LCO provides conservative limits for other postulated accidents.

F Q Z satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The Heat Flux Hot Channel Factor, F Q Z , shall be limited by the following relationships:

CFQ FQ Z ------------- K Z for P 0.5 P

(continued)

CALLAWAY PLANT B 3.2.1-2 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES LCO CFQ (continued) FQ Z ------------- K Z for P 0.5 0.5 where:CFQ = FQRTP is the FQ Z limit at RTP provided in the COLR, K(Z) is the normalized FQ Z as a function of core height provided in the COLR, and P = THERMAL P OWER-RTP The actual values of CFQ and K(Z) are given in the COLR.

For Relaxed Axial Offset Control operation, FQ Z is approximated by FQC Z and FW Q Z . Thus, both FQ Z and FQ Z must meet the C W preceding limits on FQ Z .

An FQC Z evaluation requires obtaining a power distribution measurement in MODE 1. From the power distribution measurement results the measured value ( FM Q Z ) of FQ Z . is obtained. Then, FQC Z = FM Q Z UFQ where UFQ is a factor that accounts for fuel manufacturing tolerances and measurement uncertainty.

FQC Z is an excellent approximation for F Q Z when the reactor is at the steady state power at which the power distribution measurement was taken.

The expression for FW Q Z is:

FWQ Z = FQ Z W(Z)

C where W(Z) is a cycle dependent function that accounts for power distribution transients encountered during normal operation. W(Z) is included in the COLR.

The FQ Z limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200F during either a large or small break LOCA.

(continued)

CALLAWAY PLANT B 3.2.1-3 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES LCO This LCO requires operation within the bounds assumed in the safety (continued) analyses. Calculations are performed in the core design process to confirm that the core can be controlled in such a manner during operation that it can stay within the LOCA FQ Z limits. If FQ Z cannot be maintained within the LCO limits, reduction of the core power is required. If the power distribution measurements are performed at a power level less than 100%

RTP, then the FQC Z and FW Q Z values that would result from measurements if the core was at 100% RTP should be inferred from the available information. A comparison of these inferred values with FQRTP assures compliance with the LCO at all power levels.

Violating the LCO limits for FQ Z may produce unacceptable consequences if a design basis event occurs while FQ Z is outside its specified limits.

APPLICABILITY The FQ Z limits must be maintained in MODE 1 to prevent core power distributions from exceeding the limits assumed in the safety analyses.

Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require a limit on the distribution of core power.

ACTIONS A.1 Reducing THERMAL POWER by 1% RTP for each 1% by which FQC Z exceeds its limit, maintains an acceptable absolute power density. FQC Z is Q Z multiplied by a factor which accounts for manufacturing tolerances FM and measurement uncertainties. FM Q Z is the measured value of FQ Z .

The Completion Time of 15 minutes provides an acceptable time to reduce power in an orderly manner and without allowing the plant to remain in an unacceptable condition for an extended period of time.

The maximum allowable power level initially determined by Required Action A.1 may be affected by subsequent determinations of FQC Z and would require power reductions within 15 minutes of the FQC Z determination, if necessary to comply with the decreased maximum allowable power level. Decreases in FQC Z would allow increasing the maximum allowable power level and increasing power up to this revised limit.

(continued)

CALLAWAY PLANT B 3.2.1-4 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES ACTIONS A.1 (continued)

Calculate the percent FQC Z exceeds its limit by the following expression:

FQC Z

- 1 100 for P 0.5 max over z of --------------------------------

CFQ

- K Z P

FQC Z

- - 1 100 for P 0.5 max over z of -------------------------------

CFQ

- K Z 0.5 A.2 A reduction of the Power Range Neutron Flux - High trip setpoints by 1%

for each 1% by which FQC Z exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient considering the small likelihood of a severe transient in this time period and the preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1.

The maximum allowable Power Range Neutron Flux - High trip, setpoints initially determined by Required Action A.2 may be affected by subsequent determinations of FQC Z and would require Power Range Neutron Flux - High trip setpoint reductions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the FQC Z determination, if necessary to comply with the decreased maximum allowable Power Range Neutron Flux - High trip setpoints. Decreases in FQC Z would allow increasing the maximum allowable Power Range Neutron Flux - High trip setpoints.

A.3 Reduction in the Overpower T trip setpoints by 1% for each 1% by which FQC Z exceeds its limit, is a conservative action for protection against the consequences of severe transients with unanalyzed power distributions. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is sufficient considering the small likelihood of a severe transient in this time period, and the (continued)

CALLAWAY PLANT B 3.2.1-5 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES ACTIONS A.3 (continued) preceding prompt reduction in THERMAL POWER in accordance with Required Action A.1.

The maximum allowable Overpower T trip setpoints initially determined by Required Action A.3 may be affected by subsequent determinations of FQC Z and would require Overpower T trip setpoint reductions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the FQC Z determination, if necessary to comply with the decreased maximum allowable Overpower T trip setpoints. Decreases in FQC Z would allow increasing the maximum Overpower T trip setpoints.

A.4 Verification that FQC Z has been restored to within its limit, by performing SR 3.2.1.1 prior to increasing THERMAL POWER above the limit imposed by Required Action A.1, ensures that core conditions during operation at higher power levels are consistent with safety analyses assumptions.

Inherent in this action is identification of the cause of the out of limit condition, and the correction of the cause, to the extent necessary to allow safe operation at the higher power level. The allowable power level is determined by evaluating FQC Z for the higher power level. SR 3.2.1.1 must be satisfied prior to increasing power above the higher allowable power level or restoration of any reduced Reactor Trip System Setpoints.

B.1 If it is found that the maximum calculated value of FQ Z that can occur during normal maneuvers, FW Q Z , exceeds its specified limits, there exists a potential for FQC Z to become excessively high if a normal operational transient occurs. Reducing both the positive and negative AFD limits by 1% for each 1% by which FW Q Z exceeds its limit within the allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, restricts the axial flux distribution such that even if a transient occurred, core peaking factor limits are not exceeded.

(continued)

CALLAWAY PLANT B 3.2.1-6 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES ACTIONS B.1 (continued)

Calculate the percent FW Q Z exceeds its limit by the following expression:

FQC Z W Z

- - 1 100 for P max over z of -----------------------------------

CFQ

---

K Z P

FQC Z W Z max over z of ------------------------------------ - 1 100 for P < 0.5 CFQ

---

K Z 0.5 C.1 If Required Actions A.1 through A.4 or B.1 are not met within their associated Completion Times, the plant must be placed in a mode or condition in which the LCO requirements are not applicable. This is done by placing the plant in 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 />.

This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.1.1 and SR 3.2.1.2 are modified by a Note. The Note applies REQUIREMENTS during power ascensions following a plant shutdown (leaving MODE 1).

The Note allows for power ascensions if the surveillances are not current.

It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map can be obtained. This allowance is modified, however, by one of the Frequency conditions that requires verification that FQC Z and FW Q Z are within their specified limits after a power rise of more than 10% RTP over the THERMAL POWER at which they were last verified to be within specified limits. Because FQC Z and FW Q Z could not have previously been (continued)

CALLAWAY PLANT B 3.2.1-7 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES SURVEILLANCE measured in a reload core, there is a second Frequency condition, REQUIREMENTS applicable only for reload cores, that requires determination of these (continued) parameters before exceeding 75% RTP. This ensures that some determination of FQC Z and FW Q Z are made at a lower power level at which adequate margin is available before going to 100% RTP. Also, this Frequency condition, together with the Frequency condition requiring verification of FQC Z and FW Q Z following a power increase of more than 10%, ensures that they are verified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from when equilibrium conditions at RTP (or any other power level for extended operation) are achieved. Equilibrium conditions are achieved when the core is sufficiently stable such that the uncertainty allowances associated with the measurement are valid. In the absence of these Frequency conditions, it is possible to increase power to RTP and operate for a duration allowed by the Surveillance Frequency Control Program without verification of FQC Z and FW Q Z . The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ was last measured.

SR 3.2.1.1 Verification that FQC Z is within its specified limits involves increasing Q Z to allow for manufacturing tolerance and measurement FM uncertainties in order to obtain FQC Z . Specifically, FM Q Z is the measured value of FQ Z obtained from core power distribution measurement results and FQC Z = FM Q Z UFQ (Ref. 4). The value of UFQ is determined using the formulation provided in the COLR. FQC Z is then compared to its specified limits.

The limit with which FQC Z is compared varies inversely with power above 50% RTP and directly with a function called K(Z) provided in the COLR.

Performing this Surveillance in MODE 1 prior to exceeding 75% RTP, or at a reduced power level at any other time, and meeting the 100% RTP FQ Z limit, provides assurance that the FQC Z limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased.

(continued)

CALLAWAY PLANT B 3.2.1-8 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES SURVEILLANCE SR 3.2.1.1 (continued)

REQUIREMENTS If THERMAL POWER has been increased by 10% RTP since the last determination of FQC Z , another evaluation of this factor is required within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions (to ensure that FQC Z values are being reduced sufficiently with power increase to stay within the LCO limits).

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.2.1.2 The nuclear design process includes calculations performed to determine that the core can be operated within the FQ Z limits. Because power distribution measurements are taken either at or near equilibrium conditions, the variations in power distribution resulting from normal operational maneuvers are not typically present in the flux map data.

These variations are, however, conservatively calculated by considering a wide range of unit maneuvers in normal operation.

The maximum peaking factor increase over steady state values, calculated as a function of core elevation, Z, is called W(Z). Multiplying the measured total peaking factor, FQC Z , by W(Z) gives the maximum FQ Z calculated to occur in normal operation, FW Q Z.

The limit with which FW Q Z is compared varies inversely with power and directly with the function K(Z) provided in the COLR.

The W(Z) curve is provided in the COLR for discrete core elevations. Flux map data are typically taken for 30 to 75 core elevations. FW Q Z evaluations are normally not applicable for the following axial core regions, measured in percent of core height:

a. Lower core region, from 0 to 15% inclusive; and

b. Upper core region, from 85 to 100% inclusive.

The top and bottom 15% of the core are excluded from the evaluation because of the low probability that these regions would be more limiting in the safety analyses and because of the difficulty of making a precise (continued)

CALLAWAY PLANT B 3.2.1-9 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES SURVEILLANCE SR 3.2.1.2 (continued)

REQUIREMENTS measurement in these regions. However, it is permissable to exclude a smaller region from the evaluation. This is desirable if, for example, the limiting elevation is in the upper or lower 15% of the core based on cycle-specific supporting analyses.

This Surveillance has been modified by a Note that may require that more frequent surveillances be performed. When FQC Z is measured, an evaluation of the expression below is required to account for any increase to FQ Z that may occur and cause the FQ Z limit to be exceeded before the next required FQ Z evaluation.

If the two most recent FQ Z evaluations show an increase in the expression FQC Z maximum over z ---------------

KZ Q Z increased by the it is required to meet the FQ Z limit with the last FW appropriate factor specified in the COLR, or to evaluate FQ Z more frequently, each 7 EFPD. (The 25% extension allowed by SR 3.0.2 applies to this frequency.) These alternative requirements prevent FQ Z from exceeding its limit for any significant period of time without detection.

Performing the Surveillance in MODE 1 prior to exceeding 75% RTP, or at a reduced power at any other time, and verifying the inferred results for 100% RTP meet the 100% RTP FQ Z limit, provides assurance that the FQ Z limit will be met when RTP is achieved, because peaking factors are generally decreased as power level is increased.

FQ Z is verified at power levels 10% RTP above the THERMAL POWER of its last verification, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions to ensure that FQ Z is within its limit at higher power levels.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

(continued)

CALLAWAY PLANT B 3.2.1-10 Revision 12

FQ(Z) (FQ Methodology)

B 3.2.1 BASES (Continued)

REFERENCES 1. 10 CFR 50.46, 1974.

2. FSAR, Section 15.4.8.

3. 10 CFR 50, Appendix A, GDC 26.

4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.

5. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

6. WCAP-12472-P-A, Addendum 1-A CALLAWAY PLANT B 3.2.1-11 Revision 12

FNH B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor FNH BASES BACKGROUND The purpose of this LCO is to establish limits on the power density at any point in the core so that the fuel design criteria are not exceeded and the accident analysis assumptions remain valid. The design limits on local (pellet) and integrated fuel rod peak power density are expressed in terms of hot channel factors. Control of the core power distribution with respect to these factors ensures that local conditions in the fuel rods and coolant channels do not challenge core integrity at any location during either normal operation or a postulated accident analyzed in the safety analyses.

FNH is defined as the ratio of the integral of the linear power along the fuel rod with the highest integrated power to the average integrated fuel rod power. Therefore, FNH is a measure of the maximum total power produced in a fuel rod. FNH is sensitive to fuel loading patterns, bank insertion, and fuel burnup.

FNH is not directly measurable but is inferred from a power distribution measurement obtained with either the movable incore detector system or from an OPERABLE power distribution monitoring system (PDMS)

(Reference 4). Specifically, the results of the three dimensional power distribution measurement are analyzed to determine FNH . This factor is calculated at a frequency controlled by the Surveillance Frequency Control Program. However, during power operation, the global power distribution is monitored by LCO 3.2.3, "Axial Flux Difference (AFD)," and LCO 3.2.4, "Quadrant Power Tilt Ratio (QPTR)," which address directly and continuously measured process variables. Compliance with these LCOs, along with the LCOs governing shutdown and control rod insertion and alignment, maintains the core limits on power distribution on a continuous basis.

The COLR provides peaking factor limits that ensure that the design basis value of the departure from nucleate boiling (DNB) is met for normal operation, operational transients, and any transient condition arising from events of moderate frequency. All DNB limited transient events are assumed to begin with an FNH value that satisfies the LCO requirements.

(continued)

CALLAWAY PLANT B 3.2.2-1 Revision 12

FNH B 3.2.2 BASES BACKGROUND Operation outside the LCO limits may produce unacceptable (continued) consequences if a DNB limiting event occurs. The DNB design basis ensures that there is no overheating of the fuel that results in possible cladding perforation with the release of fission products to the reactor coolant.

APPLICABLE Limits on FNH preclude core power distributions that exceed the following SAFETY fuel design limits:

ANALYSES

a. There must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hottest fuel rod in the core does not experience a DNB condition;

b. During a large break loss of coolant accident (LOCA), peak cladding temperature (PCT) must not exceed 2200F;

c. During an ejected rod accident, the average fuel pellet enthalpy at the hot spot in irradiated fuel must not exceed 200 cal/gm (Ref. 1);

and

d. Fuel design limits required by GDC 26 (Ref. 2) for the condition when control and shutdown rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth rod stuck fully withdrawn.

For transients that may be DNB limited, the Reactor Coolant System flow and FNH are the core parameters of most importance. The limits on FNH ensure that the DNB design basis is met for normal operation, operational transients, and any transients arising from events of moderate frequency.

The DNB design basis is met by limiting the minimum DNBR to the 95/95 DNB criterion applicable to a specific DNBR correlation. This value provides a high degree of assurance that the hottest fuel rod in the core does not experience a DNB condition.

The allowable FNH limit increases with decreasing power level. This functionality in FNH is included in the analyses that provide the Reactor Core Safety Limits (SLs) of SL 2.1.1.

Therefore, any DNB events in which the calculation of the core limits is modeled implicitly use this variable value of FNH in the analyses.

Likewise, all transients that may be DNB limited are assumed to begin with an initial FNH as a function of power level defined by the COLR limit equation.

(continued)

CALLAWAY PLANT B 3.2.2-2 Revision 12

FNH B 3.2.2 BASES APPLICABLE The LOCA safety analysis indirectly models FNH as an input parameter.

SAFETY The Nuclear Heat Flux Hot Channel Factor ( F Q Z ) and the axial peaking ANALYSES factors are inserted directly into the LOCA safety analyses that verify the (continued) acceptability of the resulting peak cladding temperature (Ref. 3).

The fuel is protected in part by Technical Specifications, which ensure that the initial conditions assumed in the safety and accident analyses remain valid. The following LCOs ensure this: LCO 3.2.3, Axial Flux Difference (AFD), LCO 3.2.4, Quadrant Power Tilt Ratio (QPTR),

LCO 3.1.4 Rod Group Alignment Limits, LCO 3.1.5 Shutdown Bank Insertion Limits, LCO 3.1.6, "Control Bank Insertion Limits," LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor ( FNH )" and LCO 3.2.1, "Heat Flux Hot Channel Factor ( FQ Z )."

FNH and FQ Z are measured periodically using the movable incore detector system. Measurements are generally taken with the core at, or near, steady state conditions. Core monitoring and control under transient conditions (Condition I events) are accomplished by operating the core within the limits of the LCOs on AFD, QPTR, and Bank Insertion Limits.

FNH satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO FNH shall be maintained within the limits of the relationship provided in the COLR.

The FNH limit is representative of the coolant flow channel with the maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for a DNB.

The limiting value of FNH , described by the equation contained in the COLR, is the design radial peaking factor used in the unit safety analyses.

A power multiplication factor in this equation includes an additional allowance for higher radial peaking from reduced thermal feedback and greater control bank insertion at low power levels. The limiting value of FNH is allowed to increase by a cycle-dependent factor, PFH , specified in the COLR for each 1% RTP reduction in THERMAL POWER.

(continued)

CALLAWAY PLANT B 3.2.2-3 Revision 12

FNH B 3.2.2 BASES LCO If the power distribution measurements are performed at a power level (continued) less than 100% RTP, then the FNH values that would result from measurements if the core was at 100% RTP should be inferred from the available information. A comparison of these inferred values with FHRTP assures compliance with the LCO at all power levels.

APPLICABILITY The FNH limits must be maintained in MODE 1 to preclude core power distributions from exceeding the fuel design limits for DNBR and PCT.

Applicability in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the coolant to require a limit on the distribution of core power.

ACTIONS A.1.1 With FNH exceeding its limit, the unit is allowed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore FNH to within limits. This restoration may, for example, involve realigning any misaligned rods or reducing power enough to bring FNH within its power dependent limit. When the FNH limit is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could significantly perturb the FNH value (e.g., static control rod misalignment) are considered in the safety analyses. However, the DNBR limit may be violated if a DNB limiting event occurs. Thus, the allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides an acceptable time to restore FNH to within its limits without allowing the plant to remain in an unacceptable condition for an extended period of time. The restoration of the peaking factor to within its limits by power reduction or control rod movement does not restore compliance with the LCO. Thus, even though actions are taken to satisfy Required Action A.1.1, Condition A cannot be exited until a valid surveillance demonstrates compliance with the LCO.

Condition A is modified by a Note that requires that Required Actions A.2 and A.3 must be completed whenever Condition A is entered. Even if Required Action A.1.1 is completed within the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time period, Required Action A.2 requires another measurement and calculation of FNH within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in accordance with SR 3.2.2.1.

Required Action A.3 requires that another determination of FNH must be done prior to exceeding 50% RTP, prior to exceeding 75% RTP, and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching or exceeding 95% RTP, however THERMAL POWER does not have to be reduced to comply with these (continued)

CALLAWAY PLANT B 3.2.2-4 Revision 12

FNH B 3.2.2 BASES ACTIONS A.1.1 (continued) requirements. In addition, Required Action A.2 is performed if power ascension is delayed past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

A.1.2.1 and A.1.2.2 If the value of FNH is not restored to within its specified limit either by adjusting a misaligned rod or by reducing THERMAL POWER, the alternative option is to reduce THERMAL POWER to < 50% RTP in accordance with Required Action A.1.2.1 and reduce the Power Range Neutron Flux - High to 55% RTP in accordance with Required Action A.1.2.2. Reducing power to < 50% RTP increases the DNB margin and does not likely cause the DNBR limit to be violated in steady state operation. The reduction in trip setpoints ensures that continuing operation remains at an acceptable low power level with adequate DNBR margin. The allowed Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for Required Action A.1.2.1 is consistent with those allowed for in Required Action A.1.1 and provides an acceptable time to reach the required power level from full power operation without allowing the plant to remain in an unacceptable condition for an extended period of time. The Completion Times of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for Required Actions A.1.1 and A.1.2.1 are not additive.

The allowed Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to reset the trip setpoints per Required Action A.1.2.2 recognizes that, once power is reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints; however, for extended operations at the reduced power level, the reduced trip setpoints are required to protect against transients involving positive reactivity excursions. This is a sensitive operation that may inadvertently actuate the Reactor Trip System.

A.2 Once actions have been taken to restore FNH to within its limits per Required Action A.1.1, or the power level has been reduced to

< 50% RTP per Required Action A.1.2.1, a power distribution measurement (SR 3.2.2.1) must be obtained and the measured value of FNH verified not to exceed the allowed limit at the lower power level. The unit is provided 20 additional hours to perform this task over and above the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> allowed by either Action A.1.1 or Action A.1.2.1. The Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable because of the increase in the DNB margin, which is obtained at lower power levels, and the low probability of (continued)

CALLAWAY PLANT B 3.2.2-5 Revision 12

FNH B 3.2.2 BASES ACTIONS A.2 (continued) having a DNB limiting event within this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period. Additionally, operating experience has indicated that this Completion Time is sufficient to obtain an incore flux map, perform the required calculations, and evaluate FNH .

A.3 Verification that FNH is within its specified limits after an out of limit occurrence ensures that the cause that led to the FNH exceeding its limit is identified, to the extent necessary, and corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the FNH limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is 95% RTP. SR 3.2.2.1 must be satisfied prior to increasing power above the allowable power level or restoration of any reduced Reactor Trip System setpoints. When FNH is measured at reduced power levels, the allowable power level is determined by evaluating FNH for higher power levels.

This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action.

B.1 When Required Actions A.1.1 through A.3 cannot be completed within their required Completion Times, the plant must be placed in a mode in which the LCO requirements are not applicable. This is done by placing the plant in 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 />. The allowed Completion Time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS SR 3.2.2.1 is modified by a Note. The Note applies during power ascensions following a plant shutdown (leaving Mode 1). The note allows for power ascensions if the surveillances are not current. It states that THERMAL POWER may be increased until an equilibrium power level has been achieved at which a power distribution map can be obtained.

Equilibrium conditions are achieved when the core is sufficiently stable (continued)

CALLAWAY PLANT B 3.2.2-6 Revision 12

FNH B 3.2.2 BASES SURVEILLANCE SR 3.2.2.1 (continued)

REQUIREMENTS such that the uncertainty allowances associated with the measurement are valid.

The value of FNH is determined by either using the movable incore detector system to obtain a flux distribution map or from the power distribution information provided by an OPERABLE PDMS. A data reduction computer program then calculates the maximum value of FNH from the measured flux distribution map. The measured value of FNH must have the appropriate uncertainty included (4% for a flux distribution map and U H as defined in Reference 4 for a PDMS surveillance) before comparison to the limit. The value of U H is determined using the formula in the COLR.

After each refueling, FNH must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that FNH limits are met at the beginning of each fuel cycle. Performing this Surveillance in Mode 1 prior to exceeding 75% RTP, or at a reduced power level at any other time, and verifying the inferred results for 100% RTP meet the 100% RTP FNH limit, provides assurance that FNH limit will be met when RTP is achieved, because peaking factors generally decrease as power level is increased.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. FSAR, Section 15.4.8.

2. 10 CFR 50, Appendix A, GDC 26.

3. 10 CFR 50.46.

4. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

5. WCAP-12472-P-A, Addendum 1-A CALLAWAY PLANT B 3.2.2-7 Revision 12

AFD (RAOC Methodology)

B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Relaxed Axial Offset Control (RAOC) Methodology)

BASES BACKGROUND The purpose of this LCO is to establish limits on the values of the AFD in order to limit the amount of axial power distribution skewing to either the top or bottom of the core. By limiting the amount of power distribution skewing, core peaking factors are consistent with the assumptions used in the safety analyses. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in axial power distribution control.

RAOC is a calculational procedure that defines the allowed operational space of the AFD versus THERMAL POWER. The AFD limits are selected by considering a range of axial xenon distributions that may occur as a result of large variations of the AFD. Subsequently, power peaking factors and power distributions are examined to ensure that the loss of coolant accident (LOCA), loss of flow accident, and anticipated transient limits are met. Violation of the AFD limits invalidates the conclusions of the accident and transient analyses with regard to fuel cladding integrity.

The AFD is monitored on an automatic basis using the unit process computer, which has an AFD monitor alarm. The computer determines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message if the AFD for two or more OPERABLE excore channels is outside its specified limits.

RAOC calculations assume a wide range of xenon distributions and then confirm that the resulting power distributions satisfy the requirements of the accident analyses.

APPLICABLE The AFD is a measure of the axial power distribution skewing to either the SAFETY top or bottom half of the core. The AFD is sensitive to many core related ANALYSES parameters such as control bank positions, core power level, axial burnup, axial xenon distribution, and, to a lesser extent, reactor coolant temperature and boron concentration.

The allowed range of the AFD is used in the nuclear design process to confirm that operation within these limits produces core peaking factors and axial power distributions that meet safety analysis requirements.

(continued)

CALLAWAY PLANT B 3.2.3-1 Revision 10

AFD (RAOC Methodology)

B 3.2.3 BASES APPLICABLE The RAOC methodology (Ref. 2) establishes a xenon distribution library SAFETY with tentatively wide AFD limits. Axial power distribution calculations are ANALYSES then performed to demonstrate that normal operation power shapes are (continued) acceptable for the LOCA and loss of flow accident, and for initial conditions of anticipated transients. The tentative limits are adjusted as necessary to meet the safety analysis requirements.

The limits on the AFD ensure that the Heat Flux Hot Channel Factor

( FQ Z ) is not exceeded during either normal operation or in the event of xenon redistribution following power changes. The limits on the AFD also restrict the range of power distributions that are used as initial conditions in the analyses of Condition II, III, or IV events. This ensures that the fuel cladding integrity is maintained for these postulated accidents. The most important Condition IV event is the LOCA. The most important Condition III event is the loss of flow accident. The most important Condition II events are uncontrolled bank withdrawal and boration or dilution accidents. Condition II accidents simulated to begin from within the AFD limits are used to confirm the adequacy of the Overpower T and Overtemperature T trip setpoints.

The limits on the AFD satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The shape of the power profile in the axial (i.e., the vertical) direction is largely under the control of the operator through the manual operation of the control banks. Automatic rod control is available for insertion only.

Signals are available to the operator from the Nuclear Instrumentation System (NIS) excore neutron detectors (Ref. 3). Separate signals are taken from the top and bottom detectors. The AFD is defined as the difference in normalized flux signals between the top and bottom excore detectors in each detector well. For convenience, this flux difference is converted to provide flux difference units expressed as a percentage and labeled as % flux or %I.

The AFD limits are provided in the COLR. The AFD limits for RAOC do not depend on the target flux difference. However, the target flux difference may be used to minimize changes in the axial power distribution.

(continued)

CALLAWAY PLANT B 3.2.3-2 Revision 10

AFD (RAOC Methodology)

B 3.2.3 BASES LCO Violating this LCO on the AFD could produce unacceptable (continued) consequences if a Condition II, III, or IV event occurs while the AFD is outside its specified limits.

APPLICABILITY The AFD requirements are applicable in MODE 1 greater than or equal to 50% RTP when the combination of THERMAL POWER and core peaking factors are of primary importance in safety analysis.

For AFD limits developed using RAOC methodology, the value of the AFD does not affect the limiting accident consequences with THERMAL POWER < 50% RTP and for lower operating power MODES.

ACTIONS A.1 As an alternative to restoring the AFD to within its specified limits, Required Action A.1 requires a THERMAL POWER reduction to

< 50% RTP. This places the core in a condition for which the value of the AFD is not important in the applicable safety analyses.

A Completion Time of 30 minutes is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD, as indicated by the NIS excore channel, is within its specified limits. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. WCAP-8403 (nonproprietary), "Power Distribution Control and Load Following Procedures," Westinghouse Electric Corporation, September 1974.

2. WCAP-10216-P-A, Relaxation of Constant Axial Offset Control and FQ Surveillance Technical Specification, February 1994.

3. FSAR, Chapter 7.

CALLAWAY PLANT B 3.2.3-3 Revision 10

QPTR B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

BASES BACKGROUND The QPTR limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing, after refueling, and periodically during power operation.

The power density at any point in the core must be limited so that the fuel design criteria are maintained. Together, LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," LCO 3.2.4, LCO 3.1.4, "Rod Group Alignment Limits," LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits," provide limits on process variables that characterize and control the three dimensional power distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in the safety analyses.

APPLICABLE This LCO precludes core power distributions that violate the following fuel SAFETY design criteria:

ANALYSES

a. During a large break loss of coolant accident, the peak cladding temperature must not exceed 2200F (Ref. 1);

b. During a loss of forced reactor coolant flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 departure from nucleate boiling (DNB) criterion) that the hot fuel rod in the core does not experience a DNB condition;

c. During an ejected rod accident, the average fuel pellet enthalpy at the hot spot in irradiated fuel must not exceed 200 cal/gm (Ref. 2);

and

d. The control and shutdown rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth rod stuck fully withdrawn (Ref. 3).

The LCO limits on the AFD, the QPTR, the Heat Flux Hot Channel Factor

( F Q Z ), the Nuclear Enthalpy Rise Hot Channel Factor ( FNH ), and bank insertion and rod group alignment are established to preclude core power distributions that exceed the safety analyses limits.

(continued)

CALLAWAY PLANT B 3.2.4-1 Revision 12

QPTR B 3.2.4 BASES APPLICABLE The QPTR limits ensure that FNH and F Q Z remain below their limiting SAFETY values by preventing an undetected change in the gross radial power ANALYSES distribution.

(continued)

In MODE 1, the FNH and F Q Z limits must be maintained to preclude core power distributions from exceeding design limits assumed in the safety analyses.

The QPTR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

LCO The QPTR limit of 1.02, above which corrective action is required, provides a margin of protection for both the DNB ratio and linear heat generation rate contributing to excessive power peaks resulting from X-Y plane power tilts. A limiting QPTR of 1.02 can be tolerated before the margin for uncertainty in FQ Z and ( FNH ) is possibly challenged.

APPLICABILITY The QPTR limit must be maintained in MODE 1 with THERMAL POWER

> 50% RTP to prevent core power distributions from exceeding the design limits.

Applicability in MODE 1 50% RTP and in other MODES is not required because there is either insufficient stored energy in the fuel or insufficient energy being transferred to the reactor coolant to require the implementation of a QPTR limit on the distribution of core power. The QPTR limit in these conditions is, therefore, not important. Note that the FNH and F Q Z LCOs still apply, but allow progressively higher peaking factors at 50% RTP or lower.

ACTIONS A.1 With the QPTR exceeding its limit, a power level reduction of 3% RTP for each 1% by which the QPTR exceeds 1.00 is a conservative tradeoff of total core power with peak linear power. The Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> allows sufficient time to identify the cause and correct the tilt, or reduce power, as necessary. Note that a power reduction may cause a change in the tilted condition.

The maximum allowable THERMAL POWER level initially determined by Required Action A.1 may be affected by subsequent determinations of QPTR. Increases in QPTR would require a THERMAL POWER reduction within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of QPTR determination, if necessary to comply with the decreased maximum allowable THERMAL POWER level. Decreases in (continued)

CALLAWAY PLANT B 3.2.4-2 Revision 12

QPTR B 3.2.4 BASES ACTIONS A.1 (continued)

QPTR would allow raising the maximum allowable THERMAL POWER level and increasing THERMAL POWER up to this revised limit.

A.2 After completion of Required Action A.1, the QPTR may still exceed its limits. As such, any additional changes in the QPTR are detected by requiring a check of the QPTR once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter. If the QPTR continues to increase, THERMAL POWER has to be reduced accordingly.

A 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Completion Time is sufficient because any additional change in QPTR would be relatively slow.

A.3 The peaking factors FNH and F Q Z , as approximated by FQC Z and FWQ Z  ; are of primary importance in ensuring that the power distribution remains consistent with the initial conditions used in the safety analyses.

Performing SRs on FNH and FQ Z within the Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.1 ensures that these primary indicators of power distribution are within their respective limits. Equilibrium conditions are achieved when the core is sufficiently stable at the intended operating conditions to support obtaining a power distribution measurement. Power distribution information can be obtained either by using the movable incore detectors or from an OPERABLE power distribution monitoring system (PDMS) (Reference 5). A Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.1 takes into consideration the rate at which peaking factors are likely to change, and the time required to stabilize the plant and perform power distribution measurement. If these peaking factors are not within their limits, the Required Actions associated with these Surveillances provide an appropriate response for the abnormal condition. If the QPTR remains above its specified limit, the peaking factor surveillances are required each 7 days thereafter to evaluate FNH and FQ Z with changes in power distribution. Relatively small changes are expected due to either burnup and xenon redistribution or correction of the cause for exceeding the QPTR limit.

(continued)

CALLAWAY PLANT B 3.2.4-3 Revision 12

QPTR B 3.2.4 BASES ACTIONS A.4 (continued)

Although FNH and FQ Z are of primary importance as initial conditions in the safety analyses, other changes in the power distribution may occur as the QPTR limit is exceeded and may have an impact on the validity of the safety analysis. A change in the power distribution can affect such reactor parameters as bank worths and peaking factors for rod malfunction accidents. When the QPTR exceeds its limit, it does not necessarily mean a safety concern exists. It does mean that there is an indication of a change in the gross radial power distribution that requires an investigation and evaluation that is accomplished by examining the incore power distribution. Specifically, the core peaking factors and the quadrant tilt must be evaluated because they are the factors that best characterize the core power distribution. This re-evaluation is required to ensure that, before increasing THERMAL POWER above the limit of Required Action A.1, the reactor core conditions are consistent with the assumptions in the safety analyses.

A.5 If the QPTR remains above the 1.02 limit and a re-evaluation of the safety analysis is completed and shows that safety requirements are met, the excore detectors are normalized to restore QPTR to within limit prior to increasing THERMAL POWER to above the limit of Required Action A.1.

This is done to detect any subsequent significant changes in QPTR.

Normalization is accomplished in such a manner that the indicated QPTR following normalization is near 1.00.

Required Action A.5 is modified by two Notes. Note 1 states that excore detectors are not normalized to restore QPTR to within limit until after the re-evaluation of the safety analysis has determined that core conditions at RTP are within the safety analysis assumptions (i.e., Required Action A.4).

Note 2 states that if Required Action A.5 is performed, then Required Action A.6 shall be performed. Required Action A.5 normalizes the excore detectors to restore QPTR to within limit, which restores compliance with LCO 3.2.4. Thus, Note 2 prevents exiting the Actions prior to completing a power distribution measurement to verify peaking factors per Required Action A.6. These Notes are intended to prevent any ambiguity about the required sequence of actions.

(continued)

CALLAWAY PLANT B 3.2.4-4 Revision 12

QPTR B 3.2.4 BASES ACTIONS A.6 (continued)

Once the excore detectors are normalized to restore QPTR to within limit (i.e., Required Action A.5 is performed), it is acceptable to return to full power operation. However, as an added check that the core power distribution is consistent with the safety analysis assumptions, Required Action A.6 requires verification that FQ Z , as approximated by FQC Z and FWQ Z , and F H are within their specified limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of N

achieving equilibrium conditions at RTP. Equilibrium conditions at RTP are achieved when the core is sufficiently stable at the intended operating conditions to support obtaining a power distribution measurement. As an added precaution, if the core does not reach equilibrium conditions at RTP within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but power is increased slowly, then the peaking factor surveillances must be performed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after increasing THERMAL POWER above the limit of Required Action A.1. These Completion Times are intended to allow adequate time to increase THERMAL POWER to above the limit of Required Action A.1, while not permitting the core to remain with unconfirmed power distributions for extended periods of time.

Required Action A.6 is modified by a Note that states that the peaking factor surveillances must be completed when the excore detectors have been normalized to restore QPTR to within limit (i.e., Required Action A.5).

The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limit.

B.1 If Required Actions A.1 through A.6 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% 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 of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems.

SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is 75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1.

(continued)

CALLAWAY PLANT B 3.2.4-5 Revision 12

QPTR B 3.2.4 BASES SURVEILLANCE SR 3.2.4.1 (continued)

REQUIREMENTS This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

For those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is not required until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the inputs from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is > 75% RTP.

With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

For purposes of monitoring the QPTR when one power range channel is inoperable, the moveable incore detectors may be used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is performed with a full incore flux map or two sets of four thimble locations with quarter core symmetry. The two sets of four symmetric thimbles is a set of eight unique detector locations. These locations are C-8, E-5, E-11, H-3, H-13, L-5, L-11, and N-8.

The symmetric thimble flux map can be used to generate symmetric thimble "tilt." This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore QPTR. If one of the symmetric thimbles is not available, then other pairs (triples) of symmetric thimbles can be monitored to gain information about the quadrant with the out-of-service thimble, provided the reference case is set up with the same thimble groupings (Ref. 4). Therefore, incore monitoring of QPTR can be used to confirm that QPTR is within limits.

(continued)

CALLAWAY PLANT B 3.2.4-6 Revision 12

QPTR B 3.2.4 BASES SURVEILLANCE SR 3.2.4.2 (continued)

REQUIREMENTS With one NIS channel inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occurred, which might cause the QPTR limit to be exceeded, the incore tilt result may be compared against previous tilt values either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent power distribution measurement data.

REFERENCES 1. 10 CFR 50.46.

2. FSAR, Section 15.4.8.

3. 10 CFR 50, Appendix A, GDC 26.

4. Westinghouse Recommendations on Monitoring QPTR with One Power Range Channel Out of Service, (Proprietary).

5. WCAP-12472-P-A, BEACON Core Monitoring and Operations Support System, August 1994.

6. WCAP-12472-P-A, Addendum 1-A CALLAWAY PLANT B 3.2.4-7 Revision 12