ML100550497

Watts Bar Nuclear Plant, Unit 2, Developmental Revision B - Technical Specifications Bases B 3.2 - Power Distribution Limits

ML100550497
Person / Time
Site:	Watts Bar
Issue date:	02/02/2010
From:	Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation
References
Download: ML100550497 (30)

Text

F Q (Z)B 3.2.1 (continued)

Watts Bar - Unit 2 B 3.2-1 (developmental)

B B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Heat Flux Hot Channel Factor (F Q (Z))

BASES BACKGROUND The purpose of the limits on the values of F Q (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 adjusted for uncertainty. 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 POWER TILT RATIO (QPTR)," which are directly and

continuously measured process variables. These LCOs, along with

LCO 3.1.7, "Control Bank Insertion Limits," maintain the core limits on

power distributions on a continuous basis.

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

F Q (Z) is measured periodically using the Power Distribution Monitoring System (PDMS). These measurements are generally taken with the core at or near steady state conditions.

Using the measured three dimensional power distributions, it is possible

to derive a measured value for F Q (Z). However, because this value represents a steady state condition, it does not include the variations in

the value of F Q (Z) that are present during nonequilibrium situations, such as load following.

To account for these possible variations, the steady state value of F Q (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

LCOs, including the limits on AFD, QPTR, and control rod insertion.

F Q (Z)B 3.2.1 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.2-2 (developmental)

A APPLICABLE SAFETY ANALYSES This LCO precludes core power distributions that violate the following fuel

design criteria:

a. During a loss of coolant accident (LOCA), the peak cladding temperature must not exceed 2200°F for small breaks, and there

must be a high level of probability that the peak cladding temperature

does not exceed 2200 F for large breaks (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 energy deposition to the fuel must not exceed 280 cal/gm (Ref. 2); and d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control 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

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 F Q (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 the NRC Policy Statement.

F Q (Z)B 3.2.1 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.2-3 (developmental)

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

5.0PforZK P CFQZF Q FZ CFQKZforP Q 05 05.. where: CFQ is the F Q (Z) limit at RTP provided in the COLR, K(Z) is the normalized F Q (Z) as a function of core height provided in the COLR, and P = THERMAL POWER R TP The actual values of CFQ and K(Z) are given in the COLR; however, CFQ

is normally a number on the order of 2.4, and K(Z) is a function that looks

like the one provided in Figure B 3.2.1-1.

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

An F Q C (Z) evaluation requires obtaining an incore power distribution measurement in MODE 1.

The measured value, F M Q (Z), of F Q (Z) is obtained from the incore power distribution measurement and then corrected for fuel manufacturing tolerances and measurement uncertainty.

F Q (Z)B 3.2.1 BASES (continued)

Watts Bar - Unit 2 B 3.2-4 (developmental)

B LCO (continued)

Using the PDMS to obtain the incore power distribution measurement:

F C Q (Z) = 1.03 F M Q (Z) (1+U Q/100) where 1.03 is the factor that accounts for the fuel manufacturing tolerances and the factor (1+U Q/100), which accounts for measurement uncertainty, is calculated and applied automatically by the BEACON software (Ref. 4).

F Q C (Z) is an approximation of the steady state F Q (Z).

The expression for F Q W (Z) is:

F Q W (Z) = F Q C (Z) W(Z)/P for P > 0.5 F Q W (Z) = F Q C (Z) W(Z)/0.5 for P <

0.5 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 F Q (Z) limits define limiting values for core power peaking that precludes peak cladding temperatures above 2200°F during a small

break LOCA, and assures with a high level of probability that the peak

cladding temperature does not exceed 2200°F for large breaks (Ref. 1).

This LCO requires operation within the bounds assumed in the safety

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 F Q (Z) limits. If F Q (Z) cannot be maintained within the LCO limits, reduction of the core power is required.

Violating the LCO limits for F Q (Z) produces unacceptable consequences if a design basis event occurs while F Q (Z) is outside its specified limits.

APPLICABILITY The F Q (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.

F Q (Z)B 3.2.1 BASES (continued)

Watts Bar - Unit 2 B 3.2-5 (developmental)

A ACTIONS A.1 Reducing THERMAL POWER by 1% RTP for each 1% by which F Q C (Z) exceeds its limit, maintains an acceptable absolute power density.

F Q C (Z) is F Q M (Z) multiplied by a factor accounting for manufacturing tolerances and measurement uncertainties. F Q M (Z) is the measured value of F Q (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.

A.2 A reduction of the Power Range Neutron Flux - High trip setpoints by

> 1% for each 1% by which F Q C (Z) exceeds its limit, is a conservative action for protection against the consequences of severe transients with

unanalyzed power distributions. The Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 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.

A.3 Reduction in the Overpower T trip setpoints by >

1% for each 1% by which F Q C (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.

A.4 Verification that F Q C (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.

F Q (Z)B 3.2.1 BASES (continued)

Watts Bar - Unit 2 B 3.2-6 (developmental)

A ACTIONS (continued)

B.1 If it is found that the maximum calculated value of F Q (Z) that can occur during normal maneuvers, F Q W (Z), exceeds its specified limits, there exists a potential for F Q C (Z) to become excessively high if a normal operational transient occurs. Reducing the AFD limits by 1% for each 1% by which F Q W (Z) exceeds its limit within the allowed Completion Time of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, restricts the axial flux distribution such that even if a transient

occurred, core peaking factors are not exceeded.

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

REQUIREMENTS SR 3.2.1.1 and SR 3.2.1.2 are modified by a Note. The Note applies

during the first power ascension after initial fuel loading and a refueling. 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 F Q C (Z) and F Q W (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 F Q C (Z) and F Q W (Z) could not have previously been measured in this core, there is a second Frequency condition that

requires determination of these parameters before exceeding 75% RTP.

This ensures that some determination of F Q C (Z) and F Q W (Z) is 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 F Q C (Z) and F Q W (Z) following a power increase of more than 10%, ensures that they are verified as soon as

RTP (or any other level for extended operation) is achieved.

F Q (Z)B 3.2.1 BASES (continued)

Watts Bar - Unit 2 B 3.2-7 (developmental)

B SURVEILLANCE REQUIREMENTS (continued)

In the absence of these Frequency conditions, it is possible to increase

power to RTP and operate for 31 days without verification of F Q C (Z) and F Q W (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 F Q was last measured.

SR 3.2.1.1

Verification that F Q C (Z) is within its specified limits involves increasing F Q M (Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain F Q C (Z). Specifically, F Q M (Z) is the measured value of F Q (Z) obtained from the incore power distribution measurement.

Using the PDMS to obtain the incore power distribution measurement:

F C Q (Z) = 1.03 F M Q (Z) (1+U Q/100) where 1.03 is the factor that accounts for the fuel manufacturing tolerances and the factor (1+U Q/100), which accounts for measurement uncertainty, is calculated and applied automatically by the BEACON software (Ref. 4).

The limit with which F Q C (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

ensures that the F Q C (Z) limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased.

If THERMAL POWER has been increased by >

10% RTP since the last determination of F Q C (Z), another evaluation of this factor is required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions at this higher power level (to ensure that F Q C (Z) values are being reduced sufficiently with power increase to stay within the LCO limits).

The Frequency of 31 EFPD is adequate to monitor the change of power

distribution with core burnup because such changes are slow and well

controlled when the plant is operated in accordance with the Technical

Specifications (TS).

F Q (Z)B 3.2.1 BASES Watts Bar - Unit 2 B 3.2-8 (developmental)

B SURVEILLANCE REQUIREMENTS (continued)

SR 3.2.1.2

The nuclear design process includes calculations performed to determine

that the core can be operated within the F Q (Z) limits. Because incore power distribution measurements are taken at or near steady state conditions, the variations in power distribution resulting from normal

operational maneuvers are not present in the incore power distribution measurement 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, F Q C (Z), by W(Z) and by dividing by P gives the maximum F Q (Z) calculated to occur in normal operation, F Q W (Z). Scaling the W(Z) factors by "1/P" accounts for the possibility that reactor power may be increased prior to the next F Q surveillance (Ref. 5).

The limit with which F Q W (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.

Incore power distribution measurement results are typically calculated at 30 to 75 core elevations. F Q W (Z) evaluations are not applicable for the following axial core regions, measured in percent of core height:

a. Lower core region, from 0 to 10% inclusive; and b. Upper core region, from 90 to 100% inclusive.

The top and bottom 10% of the core are excluded from the evaluation

because of the difficulty of making a precise measurement in these

regions.

This Surveillance has been modified by a Note that may require that more

frequent surveillances be performed. If F Q W (Z) is evaluated and found to be within its limit, an evaluation of the expression below is required to

account for any increase to F Q M (Z) that may occur and cause the F Q (Z) limit to be exceeded before the next required F Q (Z) evaluation.

F Q (Z)B 3.2.1 BASES Watts Bar - Unit 2 B 3.2-9 (developmental)

A SURVEILLANCE REQUIREMENTS

SR 3.2.1.2 (continued)

If the two most recent F Q (Z) evaluations show an increase in the expression

maximum over z FZ KZ Q C, it is required to meet the F Q (Z) limit with the last F Q W (Z) increased by the appropriate factor specified in the COLR, or to evaluate F Q (Z) more frequently, each 7 EFPD. These alternative requirements prevent F Q (Z) from exceeding its limit for any significant period of time without detection.

Performing the Surveillance in MODE 1 prior to exceeding 75% RTP

ensures that the F Q (Z) limit is met when RTP is achieved, because peaking factors are generally decreased as power level is increased.

F Q (Z) is verified at power levels >

10% RTP above the THERMAL POWER of its last verification, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium

conditions to ensure that F Q (Z) is within its limit at higher power levels.

The Surveillance Frequency of 31 EFPD is adequate to monitor the change of power distribution with core burnup. The Surveillance may be

done more frequently if required by the results of F Q (Z) evaluations.

The Frequency of 31 EFPD is adequate to monitor the change of power

distribution because such a change is sufficiently slow, when the plant is

operated in accordance with the TS, to preclude adverse peaking factors

between 31 day surveillances.

F Q (Z)B 3.2.1 BASES Watts Bar - Unit 2 B 3.2-10 (developmental)

B REFERENCES 1. Title 10, Code of Federal Regulations, Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water

Nuclear Power Reactors." 2. Regulatory Guide 1.77, Rev. 0, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized water Reactors,"

May 1974.

3. Title 10, Code of Federal Regulations, Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 26, "Reactivity Control System Redundancy and Capability." 4. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994, (Addendum 2, April 2002).

5. Westinghouse Technical Bulletin (TB) 08-4, "F Q Surveillance at Part Powers," July 15, 2008.

F Q (Z)B 3.2.1 BASES Watts Bar - Unit 2 B 3.2-11 (developmental)

A CORE HEIGHT

• For core height of 12 feet

Figure B 3.2.1-1 (page 1 of 1)

K(Z) - Normalized F Q (Z) as a Function of Core Height

F NH B 3.2.2 (continued)

Watts Bar - Unit 2 B 3.2-12 (developmental)

B B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor F H N 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.

F NH 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, F NH is a measure of the maximum total power produced in a fuel rod.

F NH is sensitive to fuel loading patterns, bank insertion, and fuel burnup.

F NH typically increases with control bank insertion and typically decreases with fuel burnup.

F NH is not directly measurable but is inferred from an incore power distribution measurement obtained with the Power Distribution Monitoring System (PDMS). Specifically, the resu lts of the three dimensional incore power distribution measurement are analyzed by a computer to determine F NH. This factor is calculated at least every 31 EFPD. 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.

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. The DNB design basis precludes DNB for

the hottest fuel rod in the core. All DNB limited transient events are

assumed to begin with an F NH value that satisfies the LCO requirements.

F NH B 3.2.2 BASES (continued)

Watts Bar - Unit 2 B 3.2-13 (developmental)

B BACKGROUND (continued)

Operation outside the LCO limits may produce unacceptable 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

SAFETY ANALYSES Limits on F NH preclude core power distributions that exceed the following fuel design limits: 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 loss of coolant accident (LOCA), the peak cladding temperature (PCT) must not exceed 2200 F for small breaks, and there must be a high level of probability that the peak cladding

temperature does not exceed 2200 F for large breaks (Ref. 3); c. During an ejected rod accident, the energy deposition to the fuel must not exceed 280 cal/gm (Ref. 1); and d. Fuel design limits required by GDC 26 (Ref. 2) for the condition when control rods must be capable of shutting down the reactor with a

minimum required SDM with the highest worth control rod stuck fully

withdrawn.

For transients that may be DNB limited, F NH is a significant core parameter. The limits on F NH 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 local DNB heat flux ratio to a value which satisfies

the 95/95 criterion for the DNB correlation used. Refer to the Bases for

the Reactor Core Safety Limits, B 2.1.1, for a discussion of the applicable

DNBR limits. The W-3 Correlation with a DNBR limit of 1.3 is applied in

the heated region below the first mixing vane grid. In addition, the

W-3 DNB correlation is applied in the analysis of accident conditions

where the system pressure is below the range of the WRB-2M correlation for RFA-2 fuel with IFMs. For system pressures in the range of 500 to

1000 psia, the W-3 correlation DNBR limit is 1.45 instead of 1.3.

Application of these criteria provides assurance that the hottest fuel rod in

the core does not experience a DNB.

F NH B 3.2.2 BASES (continued)

Watts Bar - Unit 2 B 3.2-14 (developmental)

B APPLICABLE SAFETY ANALYSES (continued)

The allowable F NH limit increases with decreasing power level. This functionality in F NH 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 F NH in the analyses. Likewise, all transients that may be DNB limited are assumed to begin with an initial F NH as a function of power level defined by the COLR limit equation.

The LOCA safety analyses that verify the acceptability of the resulting

peak cladding temperature (Ref. 3) model F NH as well as the Nuclear Heat Flux Hot Channel Factor (F Q (Z)).

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 3.1.7, "Control Bank Insertion Limits," LCO 3.2.2, "Nuclear

Enthalpy Rise Hot Channel Factor (F NH)," and LCO 3.2.1, "Heat Flux Hot Channel Factor (F Q (Z))."

F NH and F Q (Z) are measured periodically using the PDMS (Ref. 4).

Measurements are generally taken with the core at, or near, steady state

conditions. Core monitoring and control under transient conditions (Condition 1 events) are accomplished by operating the core within the

limits of the LCOs on AFD, QPTR, and Bank Insertion Limits.

F NH satisfies Criterion 2 of the NRC Policy Statement.

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

The F NH limit identifies 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 F NH , described by the equation contained in the COLR, is the design radial peaking factor used in the unit safety

analyses.

F NH B 3.2.2 BASES (continued)

Watts Bar - Unit 2 B 3.2-15 (developmental)

A LCO (continued)

A power multiplication factor in this equation includes an additional margin for higher radial peaking from reduced thermal feedback and

greater control rod insertion at low power levels. The limiting value of

F NH is allowed to increase 0.3% for every 1% RTP reduction in THERMAL POWER.

APPLICABILITY The F NH 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. Specifically, the design bases events that are sensitive to F NH in other MODES (MODES 2 through 5) have significant margin to DNB, and

therefore, there is no need to restrict F NH in these MODES.

ACTIONS A.1.1

With F NH 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 F NH to within its limits. This restoration may, for example, involve realigning any

misaligned rods or reducing power enough to bring F NH within its power dependent limit. When the F NH limit is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could

significantly perturb the F NH 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 F NH to within its limits without allowing the plant to remain in an unacceptable condition for

an extended period of time.

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. Thus, if

power is not reduced because this Required Action 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 nevertheless requires another

measurement and calculation of F NH 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.

However, if power is reduced below 50% RTP, Required Action A.3

requires that another determination of F NH 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. 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 />.

F NH B 3.2.2 BASES (continued)

Watts Bar - Unit 2 B 3.2-16 (developmental)

B ACTIONS (continued)

A.1.2.1 and A.1.2.2 If the value of F NH 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 RTP 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 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 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. This is a sensitive operation that may

inadvertently trip the Reactor Protection System.

A.2 Once the power level has been reduced to < 50% RTP per Required

Action A.1.2.1, an incore power distribution measurement (SR 3.2.2.1) must be obtained and the measured value of F NH 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 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

the incore power distribution measurement, perform the required calculations, and evaluate F NH.

F NH B 3.2.2 BASES (continued)

Watts Bar - Unit 2 B 3.2-17 (developmental)

B ACTIONS (continued)

A.3 Verification that F NH is within its specified limits after an out of limit occurrence ensures that the cause that led to the F NH exceeding its limit is corrected, and that subsequent operation proceeds within the LCO

limit. This Action demonstrates that the F NH 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.

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

REQUIREMENTS SR 3.2.2.1

The value of F NH is determined by using the PDMS to obtain an incore power distribution measurement. A data reduction computer program then calculates the maximum value of F NH from the measured flux distributions. The measured value of F NH must be multiplied by a factor to account for measurement uncertainty before making comparisons to

the F NH limit. When the PDMS is used to obtain the incore power distribution measurement, the factor (1+UH/100) is calculated and applied automatically by the BEACON software (References 4 and 5).

After the initial fuel loading and each refueling, F NH must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that

F NH limits are met at the beginning of each fuel cycle.

The 31 EFPD Frequency is acceptable because the power distribution

changes relatively slowly over this amount of fuel burnup. Accordingly, this Frequency is short enough that the F NH limit cannot be exceeded for any significant period of operation.

F NH B 3.2.2 BASES (continued)

Watts Bar - Unit 2 B 3.2-18 (developmental)

B REFERENCES 1. Regulatory Guide 1.77, Rev. 0, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors,"

May 1974.

2. Title 10, Code of Federal Regulations, Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 26, "Reactivity Control System Redundancy and Capability." 3. Title 10, Code of Federal Regulations, Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water

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

5. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," Addendum 2, April 2002.

AFD B 3.2.3 (continued)

Watts Bar - Unit 2 B 3.2-19 (developmental)

A B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

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.

Relaxed Axial Offset Control (RAOC) methodology 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 invalidate the conclusions of the accident and transient

analyses with regard to fuel cladding integrity.

Although the RAOC defines limits that must be met to satisfy safety

analyses, typically an operating scheme, Constant Axial Offset Control (CAOC), is used to control axial power distribution in day to day operation (Ref. 1). CAOC requires that the AFD be controlled within a narrow

tolerance band around a burnup dependent target to minimize the

variation of axial peaking factors and axial xenon distribution during unit

maneuvers.

The CAOC operating space is typically smaller and lies within the RAOC

operating space. Control within the CAOC operating space constrains

the variation of axial xenon distributions and axial power distributions.

RAOC calculations assume a wide range of xenon distributions and then

confirm that the resulting power distributions satisfy the requirements of

the accident analyses.

AFD B 3.2.3 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.2-20 (developmental)

A APPLICABLE SAFETY ANALYSES The AFD is a measure of the axial power distribution skewing to either the

top or bottom half of the core. The AFD is sensitive to many core related

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.

The RAOC methodology (Ref. 2) establishes a xenon distribution library

with tentatively wide AFD limits. One dimensional axial power distribution

calculations are then performed to demonstrate that normal operation

power shapes are 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 (F Q (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 2, 3, or 4 events. This ensures that the fuel

cladding integrity is maintained for these postulated accidents. The most

important Condition 4 event is the LOCA. The most important Condition 3

event is the loss of flow accident. The most important Condition 2 events

are uncontrolled bank withdrawal and boration or dilution accidents.

Condition 2 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 the NRC Policy Statement.

AFD B 3.2.3 BASES (continued)

(continued)

Watts Bar - Unit 2 B 3.2-21 (developmental)

A 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 or automatic motion of control banks. The automatic motion of the control banks is in response to temperature deviations resulting from manual operation of the Chemical and Volume Control

System to change boron concentration or from power level changes.

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.

The AFD limits are provided in the COLR. Figure B 3.2.3-1 shows typical

RAOC AFD limits. 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.

Violating this LCO on the AFD could produce unacceptable

consequences if a Condition 2, 3, or 4 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.

AFD B 3.2.3 BASES (continued)

Watts Bar - Unit 2 B 3.2-22 (developmental)

A SURVEILLANCE REQUIREMENTS SR 3.2.3.1

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 immediately if the AFD for two or more

OPERABLE excore channels is outside its specified limits.

This Surveillance verifies that the AFD, as indicated by the NIS excore

channel, is within its specified limits and is consistent with the status of

the AFD monitor alarm. With the AFD monitor alarm inoperable, the AFD is monitored every hour to detect operation outside its limit. The Frequency of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is based on operating experience regarding the

amount of time required to vary the AFD, and the fact that the AFD is

closely monitored. With the AFD monitor alarm OPERABLE, the

Surveillance Frequency of 7 days is adequate considering that the AFD is

monitored by a computer and any deviation from requirements is alarmed.

REFERENCES

1. WCAP-8385 (Proprietary), "Power Distribution Control and Load Following Procedures," Westinghouse Electric Corporation, September 1974.

2. R. W. Miller et al., "Relaxation of Constant Axial Offset Control:

F Q Surveillance Technical Specification," WCAP-10216-P-A, June 1983.

3. Watts Bar FSAR, Section 7.7, "Control Systems."

AFD B 3.2.3 BASES Figure B 3.2.3-1 TYPICAL AXIAL FLUX DIFFERENCE Acceptable Operation Limits as a Function of RATED THERMAL POWER Watts Bar - Unit 2 B 3.2-23 (developmental)

A QPTR B 3.2.4 (continued)

Watts Bar - Unit 2 B 3.2-24 (developmental)

B 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, and LCO 3.1.7, "Control Rod 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

SAFETY ANALYSES This LCO precludes core power distributions that violate the following fuel

design criteria:

a. During a large break loss of coolant accident, the peak cladding temperature must not exceed 2200 F (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 energy deposition to the fuel must not exceed 280 cal/gm (Ref. 2); and d. The control rods must be capable of shutting down the reactor with a minimum required SDM with the highest worth control 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 (F NH), rod group alignment, sequence, overlap, and control bank insertion are established to preclude core power distributions that exceed the safety analyses

limits.

QPTR B 3.2.4 BASES (continued)

Watts Bar - Unit 2 B 3.2-25 (developmental)

A APPLICABLE SAFETY ANALYSES (continued)

The QPTR limits ensure that F NH and F Q (Z) remain below their limiting values by preventing an undetected change in the gross radial power

distribution.

In MODE 1, the F NH 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 the NRC Policy Statement.

LCO The QPTR limit of 1.02, at 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 F Q (Z) and (F NH) 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

F NH 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. Note that

the power reduction itself may cause a change in the tilted condition.

QPTR B 3.2.4 BASES (continued)

Watts Bar - Unit 2 B 3.2-26 (developmental)

B ACTIONS (continued)

A.2 After completion of Required Action A.1, the QPTR Alarm may still be in

its alarmed state. 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-hour Completion Time is sufficient because any

additional change in QPTR would be relatively slow.

A.3 The peaking factors F NH and F Q (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 F NH and F Q (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 /> ensures that these primary indicators of power distribution are within their respective limits.

A Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> takes into consideration the rate at which

peaking factors are likely to change, and the time required to stabilize the

plant and perform an incore power distribution measurement. If these peaking factors are not within their limits, the Required Actions of 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 F NH and F Q (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.

A.4 Although F NH and F Q (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 ex ists. 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 to above the limit of

Required Action A.1, the reactor core conditions are consistent with the

assumptions in the safety analyses.

QPTR B 3.2.4 BASES (continued)

Watts Bar - Unit 2 B 3.2-27 (developmental)

A ACTIONS (continued)

A.5 If the QPTR has exceeded 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 recalibrated to show a QPTR of 1.0 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.

Required Action A.5 is modified by a Note that states that the QPTR is

not zeroed out 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). This Note is intended to prevent

any ambiguity about the required sequence of actions.

A.6 Once the flux tilt is zeroed out (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 at RTP is consistent with the safety

analysis assumptions, Required Action A.6 requires verification that F Q (Z) and F NH 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 reaching RTP.

As an added precaution, if the core power does not reach RTP within

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, but 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 /> of the time when the ascent to power

was begun. 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 may only be done after the excore detectors have

been calibrated to show zero tilt (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 calibrated to show zero tilt and the core returned to power.

QPTR B 3.2.4 BASES (continued)

Watts Bar - Unit 2 B 3.2-28 (developmental)

A ACTIONS (continued)

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

REQUIREMENTS SR 3.2.4.1

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

if more than one input from power range neutron flux channels are

inoperable.

This Surveillance verifies that the QPTR, as indicated by the Nuclear

Instrumentation System (NIS) excore channels, is within its limits. The Frequency of 7 days when the QPTR alarm is OPERABLE is acceptable

because of the low probability that this alarm can remain inoperable

without detection.

When the QPTR alarm is inoperable, the Frequency is increased to

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This Frequency is adequate to detect any relatively slow

changes in QPTR, because for those changes of QPTR 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 required

only when the input from one or more power range neutron flux channels

are inoperable and the THERMAL POWER is 75% RTP.

QPTR B 3.2.4 BASES Watts Bar - Unit 2 B 3.2-29 (developmental)

B SURVEILLANCE REQUIREMENTS SR 3.2.4.2 (continued)

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. Performing SR 3.2.4.2 at a Frequency

of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> provides an accurate alternative means for ensuring that any

tilt remains within its limits.

For the purpose of monitoring the QPTR when the input from one or more power range neutron flux channels is inoperable, incore power distribution measurement information is used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and the reference normalized symmetric power distribution. The incore power distribution measurement information can be used to generate an incore "tilt." This tilt can be compared to the reference incore tilt to generate an incore QPTR. Therefore, incore QPTR can be used to confirm that excore QPTR is within limits.

The incore power distribution measurement information can be obtained from an OPERABLE Power Distribution Monitoring System (PDMS) (Ref.

4). The reference normalized symmetric power distribution is available from the last incore power distribution measurement information used to calibrate the excore axial offset. The reference incore power distribution measurement information is obtained from an OPERABLE PDMS.

With the input from one or more power range neutron flux channels inoperable, the indicated QPTR 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 normalized quadrant tilt is compared against the reference normalized quadrant tilt. Nominally, quadrant tilt from the surveillance should be within 2% of the tilt shown by the reference incore power distribution measurement information.

REFERENCES 1. Title 10, Code of Federal Regulations, Part 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water

Nuclear Power Reactors."

QPTR B 3.2.4 BASES Watts Bar - Unit 2 B 3.2-30 (developmental)

B 2. Regulatory Guide 1.77, Rev. 0, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors," May 1974.

3. Title 10, Code of Federal Regulations, Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," GDC 26, "Reactivity Control System Redundancy and Capability." 4. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994 (Addendum 2, April 2002).