ML23067A186
| ML23067A186 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 12/29/2022 |
| From: | Union Electric Co, Ameren Missouri |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML23067A139 | List:
|
| References | |
| ULNRC-06782 | |
| Download: ML23067A186 (1) | |
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 is to limit the local F Q Z (i.e., pellet) peak power density. The value of varies along the axial F Q Z 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, is a measure of the F Q Z 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 Li mits," 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. These F Q Z 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 that are present during F Q Z nonequilibrium situations, such as l oad following. To account for these possible variations, the steady state value of is adjusted by an F Q Z 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
(conine
CALLAWAY PLANT B 3.2.1-1 Revision 12 FQ(Z) (FQ Methodology)
B 3.2.1
BASES
LCOs, including the limits on AFD, QPTR, and control and shutdown (conine) 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 ensure that the value of the initial total peaking factor F Q Z 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.
F Q Z limits assumed in the LOCA analysis are typically limiting relative to (i.e., lower than) the limit assumed in safety analyses for other F Q Z 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,, shall be limited by the following F Q Z relationships:
FQ Z CFQ for P 0.5------------P-KZ
ine
CALLAWAY PLANT B 3.2.1-2 Revision 12 FQ(Z) (FQ Methodology)
B 3.2.1
BASES
LCO F CFQ for P 0.5------------- KZ (conine) Q Z 0.5
where:CFQ = is the limit at RTP provided in the COLR, F PQFQ Z K(Z) is the normalized as a function of core height FQ Z provided in the COLR, and
P THERMAL P =-----------------------------------------------------------RTP-
The actual values of CFQ and K(Z) are given in the COLR.
For Relaxed Axial Offset Control operation, is approximated by F Q Z F CQ Z and. Thus, both and must meet the F WQ ZF CQ ZF WQ Z preceding limits on.FQ Z
An evaluation requires obtaining a power distribution measurement F CQ Z in MODE 1. From the power distribution measurement results the measured value ( ) of. is obtained. Then,F MQ ZFQ Z
F CQ Z = F MQ ZU FQ
where UFQ is a factor that accounts for fuel manufacturing tolerances and measurement uncertainty.
F CQ Z is an excellent approximation for when the reactor is at the F Q Z steady state power at which the power distribution measurement was taken.
The expression for is:F WQ Z
F WQ Z=WF CQ Z(Z)
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 limits define limiting values for core power peaking that F Q Z precludes peak cladding temperatures above 2200F during either a large or small break LOCA.
ine
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 (conine) 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 limits. If cannot be maintained FQ ZFQ Z 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 and values that would result from F CQ ZF WQ Z measurements if the core was at 100% RTP should be inferred from the available information. A comparison of these inferred values with F PQ assures compliance with the LCO at all power levels.
Violating the LCO limits for may produce unacceptable FQ Z consequences if a design basis event occurs while is outside its FQ Z specified limits.
APPLICABILITY The limits must be maintained in MODE 1 to prevent core power FQ Z distributions from exceeding the limits assumed in the safety analyses.
Applicability in other MODES is no t 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 F CQ Z exceeds its limit, maintains an acceptable absolute power density. is F CQ Z F MQ Z multiplied by a factor which accounts for manufacturing tolerances and measurement uncertainties. is the measured value of. F MQ ZFQ 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 and F CQ Z would require power reductions within 15 minutes of the F CQ Z determination, if necessary to comply with the decreased maximum allowable power level. Decreases in would allow increasing the F CQ Z maximum allowable power level and increasing power up to this revised limit.
ine
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 exceeds its limit by the following expression:FCQ Z
F C x over z of-------------------------------Q Z-- 100 f o r P 0.51 CFQ
KZ P
F C max over z -------------------------------Q Z-100 f o r P 0.51-CFQ
KZ 0.5
A.2
A reduction of the Power Range Neutron Flux - High trip setpoints by 1%
for each 1% by which exceeds its limit, is a conservative action for F CQ Z 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 and would require Power Range F CQ Z 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 F CQ Z determination, if necessary to comply with the decreased maximum allowable Power Range Neutron Flux - High trip setpoints. Decreases in F CQ 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 exceeds its limit, is a conservative action for protection F CQ Z 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
ine
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 FCQ 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 determination, if necessary to comply with the F CQ Z decreased maximum allowable Overpower T trip setpoints. Decreases in F CQ Z would allow increasing the maximum Overpower T trip setpoints.
A.4
Verification that has been restored to within its limit, by performing F CQ Z SR 3.2.1.1 prior to increasing THER MAL 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 for the higher power level. SR 3.2.1.1 F CQ Z 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 that can occur FQ Z during normal maneuvers,, exceeds its specified limits, there F WQ Z exists a potential for to become excessively high if a normal F CQ Z operational transient occurs. Reducing both the positive and negative AFD limits by 1% for each 1% by which exceeds its limit within F WQ Z 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.
ine
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 exceeds its limit by the following expression:FWQ Z
WZ max over z F C------------------Q Z------------------- 100 f o r P 1 CFQ
KZ P
WZ max over z F C------------------Q Z------------------- 100 f o r P 0.51<
CFQ
KZ 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 and are within their F CQ ZF WQ Z 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 and could not have previously been F CQ ZF WQ Z
ine
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, applicable only for reload cores, th at requires determination of these (conine) parameters before exceeding 75% RTP. This ensures that some determination of and are made at a lower power level at FCQ ZF WQ Z which adequate margin is available before going to 100% RTP. Also, this Frequency condition, together with the Frequency condition requiring verification of and following a power increase of more than F CQ ZF WQ Z 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 Pr ogram without verification of F CQ Z and. The Frequency condition is not intended to require verification F WQ Z of these parameters after every 10% increase in power level above the last verification. It only requires verifi cation after a power level is achieved for extended operation that is 10% higher than that power at which was FQ last measured.
Verification that is within its specified limits involves increasing F CQ Z F MQ Z to allow for manufacturing tolerance and measurement uncertainties in order to obtain. Specifically, is the measured F CQ ZF MQ Z value of obtained from core power distribution measurement results FQ Z and = UFQ (Ref. 4). The value of UFQ is determined using F CQ ZF MQ Z the formulation provided in the COLR. is then compared to its F CQ Z specified limits.
The limit with which is compared varies inversely with power above F CQ Z 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 limit is met when RTP is F CQ Z achieved, because peaking factors generally decrease as power level is increased.
ine
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)
If THERMAL POWER has been increased by 10% RTP since the last determination of, another evaluation of this factor is required within FCQ Z 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 F CQ 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.
The nuclear design process includes calculations performed to determine that the core can be operated within the limits. Because power FQ Z 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, conser vatively 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,, by W(Z) gives the maximum F CQ ZFQ Z calculated to occur in normal operation,.F WQ Z
The limit with which is compared varies inversely with power and F WQ Z 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. F WQ 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 thes e regions would be more limiting in the safety analyses and because of the difficulty of making a precise
ine
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)
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 is measured, an FCQ Z evaluation of the expression below is required to account for any increase to that may occur and cause the limit to be exceeded before FQ ZFQ Z the next required evaluation.FQ Z
If the two most recent evaluations show an increase in the FQ Z expression
Q Z m or z F C---------------KZ
it is required to meet the limit with the last increased by the F Q ZF WQ Z appropriate factor specified in the COLR, or to evaluate more FQ Z 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 limit, provides assurance that the FQ Z 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 is within its limit at higher power levels.FQ Z
The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.
ine
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.
- 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 FN H B 3.2.2
B 3.2 POWER DISTRIBUTION LIMITS
B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor FN H
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.
FN H 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, is a measure of the maximum total power FN H produced in a fuel rod. is sensitive to fuel loading patterns, bank FN H insertion, and fuel burnup.
FN H 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. This factor is FN H 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 value that satisfies the LCO requirements.FN H
(continued)
CALLAWAY PLANT B 3.2.2-1 Revision 12 FN H 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 preclude core power distributions that exceed the following FN H 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 are the core parameters of most importance. The limits on FN HFN H 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 limit increases with decreasing power level. This FN H functionality in is included in the analyses that provide the Reactor FN H 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 in the analyses. FN H Likewise, all transients that may be DNB limited are assumed to begin with an initial as a function of power level defined by the COLR limit FN H equation.
(continued)
CALLAWAY PLANT B 3.2.2-2 Revision 12 FN H B 3.2.2
BASES
APPLICABLE The LOCA safety analysis indirectly models as an input parameter. FN H SAFETY The Nuclear Heat Flux Hot Channel Factor ( ) and the axial peaking F Q Z 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 ( )" and LCO 3.2.1, "Heat F N H Flux Hot Channel Factor ( )."F Q Z
F N H and are measured periodically using the movable incore F Q Z 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.
F N H satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO shall be maintained within the limits of the relationship provided in the FN H COLR.
The limit is representative of the coolant flow channel with the FN H maximum enthalpy rise. This channel has the least heat removal capability and thus the highest probability for a DNB.
The limiting value of, described by the equation contained in the FN H 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 FN H is allowed to increase by a cycle-dependent factor,, specified in PF H the COLR for each 1% RTP reduction in THERMAL POWER.
(continued)
CALLAWAY PLANT B 3.2.2-3 Revision 12 FN H B 3.2.2
BASES
LCO If the power distribution measurements are performed at a power level (continued) less than 100% RTP, then the values that would result from FN H measurements if the core was at 100% RTP should be inferred from the available information. A comparison of these inferred values with F RTP H assures compliance with the LCO at all power levels.
APPLICABILITY The limits must be maintained in MODE 1 to preclude core power FN H distributions from exceeding the fuel design limits for DNBR and PCT.
Applicability in other MODES is no t 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 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 to FN HFN H within limits. This restoration may, for example, involve realigning any misaligned rods or reducing power enough to bring within its power FN H dependent limit. When the limit is exceeded, the DNBR limit is not FN H likely violated in steady state operation, because events that could significantly perturb the value (e.g., static control rod misalignment) FN H 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 to within its FN H 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 th at 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 FN H 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 must be FN H 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 FN H 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 is not restored to within its specified limit either by FN H 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 to within its limits per FN H 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 FN H 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 FN H 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.FN H
A.3
Verification that is within its specified limits after an out of limit FN H occurrence ensures that the cause that led to the exceeding its limit is FN H identified, to the extent necessary, and corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the limit is within the LCO limits prior to exceeding 50% RTP, again FN H 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 is measured at reduced power FN H levels, the allowable power level is determined by evaluating for FN H 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 FN H 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 is determined by either using the movable incore FN H 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 FN H from the measured flux distribution map. The measured value of FN H must have the appropriate uncertainty included (4% for a flux distribution map and as defined in Reference 4 for a PDMS surveillance) before U H comparison to the limit. The value of is determined using the formula U H in the COLR.
After each refueling, must be determined in MODE 1 prior to FN H exceeding 75% RTP. This requirement ensures that limits are met at FN H 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 FN H limit, provides assurance that limit will be met when RTP is FN H 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.
- 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 then performed to demonstrate that normal operation power shapes are (conine) 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
( ) is not exceeded during either normal operation or in the event of F Q Z 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 co nvenience, 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.
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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 di mensional 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
( ), the Nuclear Enthalpy Rise Hot Channel Factor ( ), and bank F Q ZF N H 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 and remain below their limiting FN HF Q Z SAFETY values by preventing an undetected change in the gross radial power distribution.
(conine)
In MODE 1, the and limits must be maintained to preclude core F N HF Q Z 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 and ( ) is possibly challenged.F Q ZF N H
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 FN H and LCOs still apply, but allow progressively higher peaking F Q Z 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 ine
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 and, as approximated by and FN HF Q ZF CQ Z F WQ 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 and 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 /> F N HF Q Z after achieving equilibrium conditions from a THERMAL POWER reduction per Required Action A.1 ensure s that these primary indicators of power distribution are within their res pective 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 and with changes F N HF Q Z 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.
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CALLAWAY PLANT B 3.2.4-3 Revision 12 QPTR B 3.2.4
BASES
ACTIONS A.4 (conine)
Although and are of primary importance as initial conditions in FN HF Q Z 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.
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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, as approximated by and F Q ZF CQ Z F WQ Z, and 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 F N H 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.
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 c onfirm 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, quadran t 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.
- 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