ML20209A120

Technical Specifications Base, Chapter B 3.1, Reactivity Controls Systems

ML20209A120
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Site:	Callaway
Issue date:	11/13/2019
From:	Ameren Missouri, Union Electric Co
To:	Office of Nuclear Reactor Regulation
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CHAPTER TABLE OF CONTENTS CHAPTER B 3.1 REACTIVITY CONTROL SYSTEMS Section Page B 3.1.1 SHUTDOWN MARGIN (SDM) ......................................................................B 3.1.1-1 BACKGROUND ........................................................................................B 3.1.1-1 APPLICABLE SAFETY ANALYSES .......................................................B 3.1.1-2 LCO ..........................................................................................................B 3.1.1-4 APPLICABILITY .......................................................................................B 3.1.1-4 ACTIONS ..................................................................................................B 3.1.1-5 SURVEILLANCE REQUIREMENTS .........................................................B 3.1.1-5 REFERENCES..........................................................................................B 3.1.1-6 B 3.1.2 Core Reactivity .............................................................................................B 3.1.2-1 BACKGROUND .......................................................................................B 3.1.2-1 APPLICABLE SAFETY ANALYSES .............................................................................................B 3.1.2-2 LCO ........................................................................................................B 3.1.2-3 APPLICABILITY ......................................................................................B 3.1.2-3 ACTIONS ..................................................................................................B 3.1.2-4 SURVEILLANCE REQUIREMENTS .....................................................................................B 3.1.2-5 REFERENCES..........................................................................................B 3.1.2-5 B 3.1.3 Moderator Temperature Coefficient (MTC) ..................................................B 3.1.3-1 BACKGROUND .......................................................................................B 3.1.3-1 APPLICABLE SAFETY ANALYSES .......................................................B 3.1.3-2 LCO .........................................................................................................B 3.1.3-3 APPLICABILITY ......................................................................................B 3.1.3-4 ACTIONS ..................................................................................................B 3.1.3-4 SURVEILLANCE REQUIREMENTS .........................................................B 3.1.3-5 REFERENCES..........................................................................................B 3.1.3-7 B 3.1.4 Rod Group Alignment Limits .........................................................................B 3.1.4-1 BACKGROUND .......................................................................................B 3.1.4-1 APPLICABLE SAFETY ANALYSES .......................................................B 3.1.4-2 CALLAWAY PLANT 3.1-i

CHAPTER TABLE OF CONTENTS (Continued)

Section Page LCO .........................................................................................................B 3.1.4-4 APPLICABILITY ......................................................................................B 3.1.4-4 ACTIONS ..................................................................................................B 3.1.4-5 SURVEILLANCE REQUIREMENTS .........................................................B 3.1.4-8 REFERENCES..........................................................................................B 3.1.4-9 B 3.1.5 Shutdown Bank Insertion Limits ....................................................................B 3.1.5-1 BACKGROUND .......................................................................................B 3.1.5-1 APPLICABLE SAFETY ANALYSES .......................................................B 3.1.5-2 LCO .........................................................................................................B 3.1.5-3 APPLICABILITY ......................................................................................B 3.1.5-3 ACTIONS ..................................................................................................B 3.1.5-3 SURVEILLANCE REQUIREMENTS .........................................................B 3.1.5-4 REFERENCES..........................................................................................B 3.1.5-4 B 3.1.6 Control Bank Insertion Limits ........................................................................B 3.1.6-1 BACKGROUND .......................................................................................B 3.1.6-1 APPLICABLE SAFETY ANALYSES .......................................................B 3.1.6-2 LCO .........................................................................................................B 3.1.6-3 APPLICABILITY ......................................................................................B 3.1.6-3 ACTIONS ..................................................................................................B 3.1.6-4 SURVEILLANCE REQUIREMENTS .........................................................B 3.1.6-5 REFERENCES..........................................................................................B 3.1.6-5 B 3.1.7 Rod Position Indication .................................................................................B 3.1.7-1 BACKGROUND .......................................................................................B 3.1.7-1 APPLICABLE SAFETY ANALYSES .......................................................B 3.1.7-2 LCO .........................................................................................................B 3.1.7-2 APPLICABILITY ......................................................................................B 3.1.7-3 ACTIONS ..................................................................................................B 3.1.7-3 SURVEILLANCE REQUIREMENTS .........................................................B 3.1.7-7 REFERENCES..........................................................................................B 3.1.7-7 B 3.1.8 PHYSICS TESTS Exceptions - MODE 2 .....................................................B 3.1.8-1 BACKGROUND .......................................................................................B 3.1.8-1 APPLICABLE SAFETY ANALYSES .......................................................B 3.1.8-3 LCO .........................................................................................................B 3.1.8-4 APPLICABILITY ......................................................................................B 3.1.8-5 ACTIONS ..................................................................................................B 3.1.8-5 SURVEILLANCE REQUIREMENTS .........................................................B 3.1.8-6 REFERENCES..........................................................................................B 3.1.8-7 3.1-ii

CHAPTER TABLE OF CONTENTS (Continued)

Section Page B 3.1.9 RCS Boron Limitations < 500°F ....................................................................B 3.1.9-1 BACKGROUND ........................................................................................B 3.1.9-1 APPLICABLE SAFETY ANALYSES .........................................................B 3.1.9-2 LCO ..........................................................................................................B 3.1.9-3 APPLICABILITY ........................................................................................B 3.1.9-3 ACTIONS ..................................................................................................B 3.1.9-4 SURVEILLANCE REQUIREMENTS .........................................................B 3.1.9-5 REFERENCES .........................................................................................B 3.1.9-5 3.1-iii

SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)

BASES BACKGROUND According to GDC 26 (Ref. 1), the reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under cold conditions. Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel.

SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtained immediately following the insertion of all shutdown and control rods, assuming that the single rod cluster assembly of highest reactivity worth is fully withdrawn.

The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable control assemblies and soluble boric acid in the Reactor Coolant System (RCS). The Rod Control System can compensate for the reactivity effects of the fuel and water temperature changes accompanying power level changes over the range from full load to no load. In addition, the Rod Control System, together with the boration system, provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel damage limits, assuming that the rod of highest reactivity worth remains fully withdrawn. The Chemical and Volume Control System can control the soluble boron concentration to compensate for fuel depletion during operation and all xenon burnout reactivity changes and can maintain the reactor subcritical under cold conditions.

During power operation, SDM control is ensured by operating with the shutdown banks fully withdrawn and the control banks within the limits of LCO 3.1.6, "Control Bank Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration.

(continued)

CALLAWAY PLANT B 3.1.1-1 Revision 10

SDM B 3.1.1 BASES (Continued)

APPLICABLE The minimum required SDM is assumed as an initial condition in safety SAFETY analyses. The safety analysis establishes an SDM that ensures specified ANALYSES acceptable fuel design limits are not exceeded for normal operation and AOOs, with the assumption of the highest worth rod stuck out on scram.

For MODE 5, the primary safety analysis that relies on the SDM limits is the boron dilution analysis.

The acceptance criteria for the SDM requirements are that specified acceptable fuel design limits are not exceeded. This is done by ensuring that:

a. The reactor can be made subcritical from all operating conditions, transients, and Design Basis Events;

b. The reactivity transients associated with postulated accident conditions are controllable within acceptable limits (departure from nucleate boiling ratio (DNBR), fuel centerline temperature limits for AOOs, and 200 cal/gm average fuel pellet enthalpy at the hot spot in irradiated fuel for the rod ejection accident); and

c. The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

The most limiting accidents for the SDM requirements are the main steam line break (MSLB) and inadvertent boron dilution accidents, as described in the FSAR (Refs. 2 and 3). In addition to the limiting MSLB transient, the SDM requirement is also used in the analyses of the following events:

a. Inadvertent boron dilution;

b. An uncontrolled rod withdrawal from subcritical or low power condition (automatic rod withdrawal is no longer available);

c. Startup of an inactive reactor coolant pump (RCP); and

d. Rod ejection.

The increased steam flow resulting from a pipe break in the main steam system causes an increased energy removal from the affected steam generator (SG), and consequently the RCS. This results in a reduction of the reactor coolant temperature. The resultant coolant shrinkage causes a reduction in pressure. In the presence of a negative moderator temperature coefficient, this cooldown causes an increase in core (continued)

CALLAWAY PLANT B 3.1.1-2 Revision 10

SDM B 3.1.1 BASES APPLICABLE reactivity. As RCS temperature decreases, the severity of an MSLB SAFETY decreases until the MODE 5 value is reached. The most limiting MSLB, ANALYSES with respect to potential fuel damage before a reactor trip occurs, is a (continued) guillotine break of a main steam line inside containment initiated at the end of core life with RCS Tavg equal to 557F. The positive reactivity addition from the moderator temperature decrease will terminate when the affected SG boils dry, thus terminating RCS heat removal and cooldown. Following the MSLB, a post trip return to power may occur; however, no fuel damage occurs as a result of the post trip return to power, and THERMAL POWER does not violate the Safety Limit (SL) requirement of SL 2.1.1. Although the severity of an MSLB is reduced at lower RCS temperatures, e.g., in MODES 3 and 4, SDM requirements still apply to ensure that the limiting MSLB analyzed at the end of core life with RCS Tavg equal to 557°F remains bounding. In MODES 3 and 4 the required SDM with automatic safety injection (SI) capability blocked below P-11 is greater than the SDM required below P-11 with SI capability unblocked, to ensure that an MSLB occurring at the analyzed conditions remains bounding, as described in Westinghouse NSAL-02-14 (Ref. 5).

In the boron dilution analysis, the required SDM defines the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. These values, in conjunction with the configuration of the RCS and the assumed dilution flow rate, directly affect the results of the analysis. This event is most limiting at the beginning of core life, when critical boron concentrations are highest. The SDM must be adequate to allow sufficient time for the BDMS to detect a flux multiplication greater than its setpoint and initiate valve swapover to prevent inadvertent criticality.

Depending on the system initial conditions and reactivity insertion rate, the uncontrolled rod withdrawal transient is terminated by either a high power level trip or a high pressurizer pressure trip. In all cases, power level, RCS pressure, linear heat rate, and the DNBR do not exceed allowable limits.

The startup of an inactive RCP is administratively precluded in MODES 1 and 2. In MODE 3, the startup of an inactive RCP can not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. The maximum positive reactivity addition that can occur due to an inadvertent RCP start is less than half the minimum required SDM. Startup of an idle RCP cannot, therefore, produce a return to power from the hot standby condition.

The ejection of a control rod rapidly adds reactivity to the reactor core, causing both the core power level and heat flux to increase with corresponding increases in reactor coolant temperatures and pressure.

(continued)

CALLAWAY PLANT B 3.1.1-3 Revision 10

SDM B 3.1.1 BASES APPLICABLE The ejection of a rod also produces a time dependent redistribution of SAFETY core power. Depending on initial power level, this accident is terminated ANALYSES by the power range neutron flux - high or low reactor trip setpoint in the (continued) FSAR analyses.

SDM satisfies Criterion 2 of 10CFR50.36(c)(2)(ii). Even though it is not directly observed from the control room, SDM is considered an initial condition process variable because it is periodically monitored to ensure that the unit is operating within the bounds of accident analysis assumptions.

LCO SDM is a core design condition that can be ensured during operation through control rod positioning (control and shutdown banks) and through the soluble boron concentration.

The MSLB (Ref. 2) and the boron dilution (Ref. 3) accidents are the most limiting analyses that establish the SDM value of the LCO. For MSLB accidents, if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 100, "Reactor Site Criteria," limits (Ref. 4).

For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be sufficient. The required SDM limits are specified in the COLR.

APPLICABILITY In MODE 2 with keff < 1.0 and in MODES 3, 4, and 5 the SDM requirements are applicable to provide sufficient negative reactivity to meet the assumptions of the safety analyses discussed above. In MODE 6, the shutdown reactivity requirements are given in LCO 3.9.1, "Boron Concentration." In MODES 1 and 2, SDM is ensured by complying with LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, Control Bank Insertion Limits.

Since this Specification has no LCO 3.0.4.c allowance, MODE 5 can not be entered from MODE 6 while not meeting the SDM limits. This assures that the initial condition assumptions of an inadvertent boron dilution event in MODE 5 are met. The risk assessments of LCO 3.0.4.b may only be utilized for systems and components, not Criterion 2 values or parameters such as SDM. Therefore, a risk assessment per LCO 3.0.4.b to allow MODE changes with single or multiple system/equipment inoperabilities may not be used to allow a MODE change into or ascending within this LCO while not meeting the SDM limits, even if the risk assessment specifically includes consideration of SDM.

(continued)

CALLAWAY PLANT B 3.1.1-4 Revision 10

SDM B 3.1.1 BASES (Continued)

ACTIONS A.1 If the SDM requirements are not met, boration must be initiated promptly.

A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. It is assumed that boration will be continued until the SDM requirements are met.

In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the borated water source should be a highly concentrated solution, such as that normally found in the boric acid storage tanks, or the refueling water storage tank. The operator should borate with the best source available for the plant conditions.

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS In MODES 1 and 2, SDM is verified by observing that the requirements of LCO 3.1.5 and LCO 3.1.6 are met. In the event that a rod is known to be untrippable, however, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum worth.

In MODES 2 (with keff < 1.0), 3, 4, and 5, the SDM is verified by performing a reactivity balance calculation, considering the listed reactivity effects:

a. RCS boron concentration (may include allowances for boron-10 depletion);

b. Control and shutdown rod position;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation;

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

(continued)

CALLAWAY PLANT B 3.1.1-5 Revision 10

SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS In the event that a rod is known to be untrippable and not fully inserted, SDM verification must account for the worth of the untrippable rod as well as another rod of maximum worth.

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

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Chapter 15, Section 15.1.5.

3. FSAR, Chapter 15, Section 15.4.6.

4. 10 CFR 100.

5. Westinghouse NSAL-02-14.

CALLAWAY PLANT B 3.1.1-6 Revision 10

Core Reactivity B 3.1.2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Core Reactivity BASES BACKGROUND According to GDC 26, GDC 28, and GDC 29 (Ref. 1), reactivity shall be controllable, such that subcriticality is maintained under cold conditions, and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Therefore, reactivity balance is used as a measure of the predicted versus measured core reactivity during power operation. The periodic confirmation of core reactivity is necessary to ensure that Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity difference could be the result of unanticipated changes in fuel, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3.1.1, SHUTDOWN MARGIN (SDM)) in ensuring the reactor can be brought safely to cold, subcritical conditions.

When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers producing zero net reactivity.

Excess reactivity can be inferred from the boron letdown curve (or critical boron curve), which provides an indication of the soluble boron concentration in the Reactor Coolant System (RCS) versus cycle burnup.

Periodic measurement of the RCS boron concentration for comparison with the predicted value with other variables fixed (such as rod height, temperature, pressure, and power), provides a convenient method of ensuring that core reactivity is within design expectations and that the calculational models used to generate the safety analysis are adequate.

In order to achieve the required fuel cycle energy output, the uranium enrichment, in the new fuel loading and in the fuel remaining from the previous cycle, provides excess positive reactivity beyond that required to sustain steady state operation throughout the cycle. When the reactor is (continued)

CALLAWAY PLANT B 3.1.2-1 Revision 10

Core Reactivity B 3.1.2 BASES BACKGROUND critical at RTP and moderator temperature, the excess positive reactivity (continued) is compensated by burnable absorbers (if any), control rods, whatever neutron poisons (mainly xenon and samarium) are present in the fuel, and the RCS boron concentration.

When the core is producing THERMAL POWER, the fuel is being depleted and excess reactivity is decreasing. As the fuel depletes, the RCS boron concentration is reduced to decrease negative reactivity and maintain constant THERMAL POWER. The boron letdown curve is based on steady state operation at RTP. Therefore, deviations from the predicted boron letdown curve may indicate deficiencies in the design analysis, deficiencies in the calculational models, or abnormal core conditions, and must be evaluated.

APPLICABLE The acceptance criteria for core reactivity are that the reactivity balance SAFETY limit ensures plant operation is maintained within the assumptions of the ANALYSES safety analyses.

Accurate prediction of core reactivity is either an explicit or implicit assumption in the accident analysis evaluations. Every accident evaluation (Ref. 2) is, therefore, dependent upon accurate evaluation of core reactivity. In particular, SDM and reactivity transients, such as control rod withdrawal accidents (automatic rod withdrawal is no longer available) or rod ejection accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity balance additionally ensures that the nuclear methods provide an accurate representation of the core reactivity.

Design calculations and safety analyses are performed for each fuel cycle for the purpose of predetermining reactivity behavior and the RCS boron concentration requirements for reactivity control during fuel depletion.

The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted RCS boron concentrations for identical core conditions at beginning of cycle (BOC) do not agree, then the assumptions used in the reload cycle design analysis or the calculational models used to predict soluble boron requirements may not be accurate. If reasonable agreement between measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured boron concentration. Thereafter, any significant deviations (continued)

CALLAWAY PLANT B 3.1.2-2 Revision 10

Core Reactivity B 3.1.2 BASES APPLICABLE in the measured boron concentration from the predicted boron letdown SAFETY curve that develop during fuel depletion may be an indication that the ANALYSES calculational model is not adequate for core burnups beyond BOC, or that (continued) an unexpected change in core conditions has occurred.

The normalization of predicted RCS boron concentration to the measured value may be performed after reaching RTP following startup from a refueling outage, with the control rods in their normal positions for power operation. The normalization is performed at BOC conditions, so that core reactivity relative to predicted values can be continually monitored and evaluated as core conditions change during the cycle.

Core reactivity satisfies Criterion 2 of 10CFR50.36(c)(2)(ii).

LCO Long term core reactivity behavior is a result of the core physics design and cannot be easily altered once the core design is fixed. During operation, therefore, the LCO can only be ensured through measurement and tracking, and appropriate actions taken as necessary. Large differences between actual and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the Nuclear Design Methodology are larger than expected. A limit on the reactivity balance of 1% k/k has been established based on engineering judgment. A 1% deviation in reactivity from that predicted is larger than expected for normal operation and should therefore be evaluated.

When measured core reactivity is within 1% k/k of the predicted value at steady state thermal conditions, the core is considered to be operating within acceptable design limits. Since deviations from the limit are normally detected by comparing predicted and measured steady state RCS critical boron concentrations, the difference between measured and predicted values would be approximately 100 ppm (depending on the boron worth) before the limit is reached. These values are well within the uncertainty limits for analysis of boron concentration samples, so that spurious violations of the limit due to uncertainty in measuring the RCS boron concentration are unlikely.

APPLICABILITY The limits on core reactivity must be maintained during MODES 1 and 2 because a reactivity balance must exist when the reactor is critical or producing THERMAL POWER. As the fuel depletes, core conditions are changing, and confirmation of the reactivity balance ensures the core is (continued)

CALLAWAY PLANT B 3.1.2-3 Revision 10

Core Reactivity B 3.1.2 BASES APPLICABILITY operating as designed. This Specification does not apply in MODES 3, 4, (continued) and 5 because the reactor is shut down and the reactivity balance is not changing.

In MODE 6, fuel loading results in a continually changing core reactivity.

Boron concentration requirements (LCO 3.9.1, "Boron Concentration")

ensure that fuel movements are performed within the bounds of the safety analysis. An SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling).

ACTIONS A.1 and A.2 Should an anomaly develop between measured and predicted core reactivity, an evaluation of the core design and safety analysis must be performed. Core conditions are evaluated to determine their consistency with input to design calculations. Measured core and process parameters are evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models are reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 7 days is based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis.

Following evaluations of the core design and safety analysis, the cause of the reactivity anomaly may be resolved. If the cause of the reactivity anomaly is a mismatch in core conditions at the time of RCS boron concentration sampling, then a recalculation of the RCS boron concentration requirements may be performed to demonstrate that core reactivity is behaving as expected. If an unexpected physical change in the condition of the core has occurred, it must be evaluated and corrected, if possible. If the cause of the reactivity anomaly is in the calculation technique, then the calculational models must be revised to provide more accurate predictions. If any of these results are demonstrated, and it is concluded that the reactor core is acceptable for continued operation, then the boron letdown curve may be renormalized and power operation may continue. If operational restrictions or additional SRs are necessary to ensure the reactor core is acceptable for continued operation, then they must be defined.

(continued)

CALLAWAY PLANT B 3.1.2-4 Revision 10

Core Reactivity B 3.1.2 BASES ACTIONS A.1 and A.2 (continued)

The required Completion Time of 7 days is adequate for preparing whatever operating restrictions or surveillances that may be required to allow continued reactor operation.

B.1 If the core reactivity cannot be restored to within the 1% k/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If the SDM for MODE 3 is not met, then the boration required by LCO 3.1.1 Required Action A.1 would occur. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations (may include allowances for boron-10 depletion). The comparison is made, considering that other core conditions are fixed or stable, including control and shutdown rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization (if necessary) of predicted core reactivity to the measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.

2. FSAR, Chapter 15.

CALLAWAY PLANT B 3.1.2-5 Revision 10

MTC B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Moderator Temperature Coefficient (MTC)

BASES BACKGROUND According to GDC 11 (Ref. 1), the reactor core and its interaction with the Reactor Coolant System (RCS) must be designed for inherently stable power operation, even in the possible event of an accident. In particular, the net reactivity feedback in the system must compensate for any unintended reactivity increases.

The MTC relates a change in core reactivity to a change in reactor coolant temperature (a positive MTC means that reactivity increases with increasing moderator temperature; conversely, a negative MTC means that reactivity decreases with increasing moderator temperature). The reactor is designed to operate with a negative MTC over the largest possible range of fuel cycle operation. Therefore, a coolant temperature increase will cause a reactivity decrease, so that the coolant temperature tends to return toward its initial value. Reactivity increases that cause a coolant temperature increase will thus be self limiting, and stable power operation will result.

MTC values are predicted at selected burnups during the safety evaluation analysis and are confirmed to be acceptable by measurements. Reload cores are designed so that the beginning of cycle (BOC) MTC is less than zero when THERMAL POWER is at RTP. The actual value of the MTC is dependent on core characteristics, such as fuel loading and reactor coolant soluble boron concentration. The core design may require additional fixed distributed poisons to yield an MTC at BOC within the range analyzed in the plant accident analysis. The end of cycle (EOC) MTC is also limited by the requirements of the accident analysis. Fuel cycles that are designed to achieve high burnups or that have changes to other characteristics are evaluated to ensure that the MTC does not exceed the EOC limit.

The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting conditions assumed in the FSAR accident and transient analyses.

If the LCO limits are not met, the unit response during transients may not be as predicted. The core could violate criteria that prohibit a return to criticality, or the departure from nucleate boiling ratio criteria of the (continued)

CALLAWAY PLANT B 3.1.3-1 Revision 3

MTC B 3.1.3 BASES BACKGROUND approved correlation may be violated, which could lead to a loss of the (continued) fuel cladding integrity.

The SRs for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits, since this coefficient changes slowly, due principally to the reduction in RCS boron concentration associated with fuel burnup.

APPLICABLE The acceptance criteria for the specified MTC are:

SAFETY ANALYSES a. The MTC values must remain within the bounds of those used in the accident analysis (Ref. 2); and

b. The MTC must be such that inherently stable power operations result during normal operation and accidents, such as overheating and overcooling events.

The FSAR, Chapter 15 (Ref. 2), contains analyses of accidents that result in both overheating and overcooling of the reactor core. MTC is one of the controlling parameters for core reactivity in these accidents. Both the most positive value and most negative value of the MTC are important to safety, and both values must be bounded. Values used in the analyses consider worst case conditions to ensure that the accident results are bounding (Ref. 3).

The consequences of accidents that cause core overheating must be evaluated when the MTC is positive. Such accidents include the rod withdrawal transient from either zero (Ref. 2) or RTP (automatic rod withdrawal is no longer available), loss of main feedwater flow, and loss of forced reactor coolant flow. The consequences of accidents that cause core overcooling must be evaluated when the MTC is negative. Such accidents include sudden feedwater flow increase and sudden decrease in feedwater temperature.

In order to ensure a bounding accident analysis, the MTC is assumed to be its most limiting value for the analysis conditions appropriate to each accident. The bounding value is determined by considering rodded and unrodded conditions, whether the reactor is at full or zero power, and whether it is the BOC or EOC life. The most conservative combination appropriate to the accident is then used for the analysis (Ref. 2).

MTC values are bounded in reload safety evaluations assuming steady state conditions at BOC and EOC. An EOC measurement is conducted (continued)

CALLAWAY PLANT B 3.1.3-2 Revision 3

MTC B 3.1.3 BASES APPLICABLE at conditions when the RCS boron concentration reaches a boron SAFETY concentration equivalent to 300 ppm at an equilibrium, all rods out, RTP ANALYSES condition. The measured value may be extrapolated to project the EOC (continued) value, in order to confirm reload design predictions.

The most negative MTC value, equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved: (1) a conversion of the MDC used in the FSAR accident analyses to its equivalent MTC, based on the rate of change of moderator density with temperature at RATED THERMAL POWER conditions, and (2) adding margin to this value to account for the largest difference in MTC observed between an EOC, all rods withdrawn, RATED THERMAL POWER condition and an envelope of those most adverse conditions of moderator temperature and pressure, rods inserted to their insertion limits, axial power skewing, and xenon concentration that can occur in normal operation within Technical Specification limits and lead to a significantly more negative EOC MTC at RATED THERMAL POWER.

These corrections transformed the MDC value used in the FSAR accident analyses into the limiting End of Cycle (EOC) Life MTC value. The 300 ppm surveillance limit MTC value represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by adding an allowance for burnup and soluble boron concentration changes to the limiting EOC MTC value.

MTC satisfies Criterion 2 of 10CFR50.36(c)(2)(ii). Even though it is not directly observed and controlled from the control room, MTC is considered an initial condition process variable because of its dependence on boron concentration.

LCO LCO 3.1.3 requires the MTC to be within specified limits of the COLR to ensure that the core operates within the assumptions of the accident analysis. During the reload core safety evaluation, the MTC is analyzed to determine that its values remain within the bounds of the original accident analysis during operation.

Assumptions made in safety analyses require that the MTC be less positive than a given upper bound and less negative than a given lower bound. The MTC is most positive near BOC; this upper bound must not be exceeded. This maximum upper limit occurs near BOC, all rods out (ARO), hot zero power conditions. At EOC the MTC takes on its most (continued)

CALLAWAY PLANT B 3.1.3-3 Revision 3

MTC B 3.1.3 BASES LCO negative value, when the lower bound becomes important. This LCO (continued) exists to ensure that both the upper and lower bounds are not exceeded.

During operation, therefore, the conditions of the LCO can only be ensured through measurement. The Surveillance checks at BOC and EOC on MTC provide confirmation that the MTC is behaving as anticipated so that the acceptance criteria are met.

The LCO establishes a maximum positive value that cannot be exceeded.

The BOC positive limit and the EOC negative limit are established in the COLR to allow specifying limits for each particular cycle. This permits the unit to take advantage of improved fuel management and changes in unit operating schedule.

APPLICABILITY Technical Specifications place both LCO and SR values on MTC, based on the safety analysis assumptions described above.

In MODE 1, the limits on MTC must be maintained to ensure that any accident initiated from THERMAL POWER operation will not violate the design assumptions of the accident analysis. In MODE 2 with the reactor critical, the upper limit must also be maintained to ensure that startup and subcritical accidents (such as the uncontrolled control rod assembly or group withdrawal) will not violate the assumptions of the accident analysis. The lower MTC limit must be maintained in MODES 2 and 3, in addition to MODE 1, to ensure that cooldown accidents will not violate the assumptions of the accident analysis. In MODES 4, 5, and 6, this LCO is not applicable, since no Design Basis Accidents using the MTC as an analysis assumption are initiated from these MODES.

ACTIONS A.1 If the BOC MTC limit is violated, administrative withdrawal limits for control banks must be established to maintain the MTC within its limits.

The MTC becomes more negative with control bank insertion and decreased boron concentration. A Completion Time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> provides enough time for evaluating the MTC measurement and computing the required bank withdrawal limits.

As cycle burnup is increased, the RCS boron concentration will initially be increased due to burnable absorber depletion (between roughly 150-3000 MWd/MTU, depending on cycle energy requirements, burnable absorber (continued)

CALLAWAY PLANT B 3.1.3-4 Revision 3

MTC B 3.1.3 BASES ACTIONS A.1 (continued) loading, etc.) and then be reduced. The reduced boron concentration causes the MTC to become more negative. Using physics calculations, the time in cycle life at which the calculated MTC will meet the LCO requirement can be determined. At this point in core life Condition A no longer exists. The unit is no longer in the Required Action, so the administrative withdrawal limits are no longer in effect.

B.1 If the required administrative withdrawal limits at BOC are not established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the unit must be brought to MODE 2 with keff < 1.0 to prevent operation with an MTC that is more positive than that assumed in safety analyses.

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, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

C.1 Exceeding the EOC MTC limit means that the safety analysis assumptions for the EOC accidents that use a bounding negative MTC value may be invalid. If the EOC MTC limit is exceeded, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The allowed Completion Time is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.3.1 REQUIREMENTS This SR requires measurement of the MTC at BOC prior to entering MODE 1 in order to demonstrate compliance with the most positive MTC LCO. Meeting the BOC limit prior to entering MODE 1 ensures that the limit will also be met at higher power levels.

(continued)

CALLAWAY PLANT B 3.1.3-5 Revision 3

MTC B 3.1.3 BASES SURVEILLANCE SR 3.1.3.1 (continued)

REQUIREMENTS The BOC MTC value for ARO will be inferred from isothermal temperature coefficient measurements obtained during the physics tests after refueling. The ARO value can be directly compared to the BOC MTC limit of the LCO. If required, measurement results and predicted design values can be used to establish administrative withdrawal limits for control banks.

SR 3.1.3.2 In similar fashion, the LCO demands that the MTC be less negative than the specified value for EOC full power conditions. This measurement may be performed at any THERMAL POWER, but its results must be extrapolated to the conditions of RTP and all banks withdrawn in order to make a proper comparison with the LCO value. Because the RTP MTC value will gradually become more negative with further core depletion and boron concentration reduction, a 300 ppm SR value of MTC should necessarily be less negative than the EOC LCO limit. The 300 ppm SR value is sufficiently less negative than the EOC LCO limit value to ensure that the LCO limit will be met when the 300 ppm Surveillance criterion is met.

SR 3.1.3.2 is modified by three Notes that include the following requirements:

1. The SR is required to be performed once each cycle within 7 effective full power days (EFPDs) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm.

2. If the 300 ppm Surveillance limit is exceeded, it is possible that the EOC limit on MTC could be reached before the planned EOC.

Because the MTC changes slowly with core depletion, the Frequency of 14 effective full power days is sufficient to avoid exceeding the EOC limit. (The 25% extension allowed by SR 3.0.2 applies to this frequency.)

3. The Surveillance limit for RTP boron concentration of 60 ppm is conservative. If the measured MTC at 60 ppm is less negative than the 60 ppm Surveillance limit, the EOC limit will not be exceeded because of the gradual manner in which MTC changes with core burnup.

(continued)

CALLAWAY PLANT B 3.1.3-6 Revision 3

MTC B 3.1.3 BASES (Continued)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 11.

2. FSAR, Chapter 15.

3. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

CALLAWAY PLANT B 3.1.3-7 Revision 3

Rod Group Alignment Limits B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Rod Group Alignment Limits BASES BACKGROUND The OPERABILITY (i.e., trippability) of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design,"

GDC 26, "Reactivity Control System Redundancy and Capability" (Ref. 1),

and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants" (Ref. 2).

Mechanical or electrical failures may cause a control or shutdown rod to become inoperable or to become misaligned from its group. Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown. Therefore, rod alignment and OPERABILITY are related to core operation in design power peaking limits and the core design requirement of a minimum SDM.

Limits on rod alignment have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Rod cluster control assemblies (RCCAs), or rods, are moved by their control rod drive mechanisms (CRDMs). Each CRDM moves its RCCA one step (approximately 5/8 inch) at a time, but at varying rates (steps per minute) depending on the signal output from the Rod Control System.

The RCCAs are divided among four control banks and five shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. If a bank of RCCAs consists of two groups, the groups are moved in a staggered fashion, but always within one step of each other. All control banks contain two rod groups. Two shutdown banks (A and B) contain two rod groups and the remaining three shutdown banks (C, D, and E) contain one rod group.

(continued)

CALLAWAY PLANT B 3.1.4-1 Revision 12

Rod Group Alignment Limits B 3.1.4 BASES BACKGROUND The shutdown banks are maintained either in the fully inserted or fully (continued) withdrawn position. The control banks are moved in an overlap pattern, using the following withdrawal sequence: When control bank A reaches a predetermined height in the core, control bank B begins to move out with control bank A. Control bank A stops at the position of maximum withdrawal, and control bank B continues to move out. When control bank B reaches a predetermined height, control bank C begins to move out with control bank B. This sequence continues until control banks A, B, and C are at the fully withdrawn position, and control bank D is approximately halfway withdrawn. The insertion sequence is the opposite of the withdrawal sequence. The control rods are arranged in a radially symmetric pattern, so that control bank motion does not introduce radial asymmetries in the core power distributions.

The axial position of shutdown rods and control rods is indicated by two separate and independent systems, which are the Bank Demand Position Indication System (commonly called group step counters) and the Digital Rod Position Indication (DRPI) System.

The Bank Demand Position Indication System counts the pulses from the rod control system that moves the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by the group step counter for that group. The Bank Demand Position Indication System is considered highly precise ( 1 step or 5/8 inch).

However, if rod movement does not occur upon demand, the demand step counter will still count the pulse and incorrectly reflect the position of the rod.

The DRPI System provides a highly accurate indication of actual rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one data system fails, the DRPI will go on half accuracy. The DRPI system is capable of monitoring rod position within at least 12 steps with either full accuracy or half accuracy.

APPLICABLE Rod misalignment accidents are analyzed in the safety analysis (Ref. 3).

SAFETY The acceptance criteria for addressing rod inoperability or misalignment ANALYSES are that:

a. There be no violations of:

(continued)

CALLAWAY PLANT B 3.1.4-2 Revision 12

Rod Group Alignment Limits B 3.1.4 BASES APPLICABLE 1. specified acceptable fuel design limits, or SAFETY ANALYSES 2. Reactor Coolant System (RCS) pressure boundary (continued) integrity; and

b. The core remains subcritical after accident transients.

Two types of misalignment are distinguished. During movement of a rod group, one rod may stop moving, while the other rods in the group continue. This condition may cause excessive power peaking. The second type of misalignment occurs if one rod fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the control and shutdown rods to meet the SDM requirement, with the maximum worth rod stuck fully withdrawn.

Two types of analysis are performed in regard to static rod misalignment (Ref. 3). With control and shutdown banks at their insertion limits, one type of analysis considers the case when any one rod is completely inserted into the core. The second type of analysis considers the case of a completely withdrawn single rod from control bank D, inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio in both of these cases bounds the situation when a rod is misaligned from its group by 12 steps.

Another type of misalignment occurs if one RCCA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition is assumed in the evaluation to determine that the required SDM is met with the maximum worth RCCA also fully withdrawn (Ref. 4).

The Required Actions in this LCO ensure that either deviations from the alignment limits will be corrected or that THERMAL POWER will be adjusted so that excessive local linear heat rates (LHRs) will not occur, and that the requirements on SDM and ejected rod worth are preserved.

Continued operation of the reactor with a misaligned rod is allowed if the heat flux hot channel factor ( FQ Z ) and the nuclear enthalpy rise hot N

channel factor ( F H ) are verified to be within their limits in the COLR and the safety analysis is verified to remain valid. When a rod is misaligned, the assumptions that are used to determine the rod insertion limits, AFD limits, and quadrant power tilt limits are not preserved. Therefore, the N

limits may not preserve the design peaking factors, and FQ Z and F H must be verified directly by core power distribution measurement information. Bases Section 3.2 (Power Distribution Limits) contains (continued)

CALLAWAY PLANT B 3.1.4-3 Revision 12

Rod Group Alignment Limits B 3.1.4 BASES APPLICABLE more complete discussions of the relation of FQ Z and F H N to the SAFETY operating limits.

ANALYSES (continued) Shutdown and control rod OPERABILITY and alignment are directly related to power distributions and SDM, which are initial conditions assumed in safety analyses. Therefore they satisfy Criterion 2 of 10CFR50.36(c)(2)(ii).

LCO The limits on shutdown or control rod alignments ensure that the assumptions in the safety analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the assumed reactivity will be available and will be inserted. The OPERABILITY requirements (i.e.,

trippability) are separate from the alignment requirements, which ensure that the RCCAs and banks maintain the correct power distribution and rod alignment. The rod OPERABILITY requirement is satisfied provided the rod will fully insert in the required rod drop time assumed in the safety analyses. Rod control malfunctions that result in the inability to move a rod (e.g., rod urgent failures), but do not impact trippability or required rod drop time, do not result in rod inoperability.

The requirement to maintain the rod alignment to within plus or minus 12 steps of their group step counter demand position is conservative.

The minimum misalignment assumed in safety analysis is 24 steps (15 inches), and in some cases a total misalignment from fully withdrawn to fully inserted is assumed.

Failure to meet the requirements of this LCO may produce unacceptable power peaking factors and LHRs, or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis.

APPLICABILITY The requirements on RCCA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (i.e., trippability) and alignment of rods have the potential to affect the safety of the plant.

In MODES 3, 4, 5, and 6, the alignment limits do not apply because the rods are typically fully inserted and the reactor is shut down and not producing fission power. In the shutdown MODES, the OPERABILITY of the shutdown and control rods has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.1, SHUTDOWN MARGIN (SDM), for SDM in MODES 2 with keff < 1.0, 3, 4, and 5 and LCO 3.9.1, (continued)

CALLAWAY PLANT B 3.1.4-4 Revision 12

Rod Group Alignment Limits B 3.1.4 BASES APPLICABILITY "Boron Concentration," for boron concentration requirements during (continued) refueling.

ACTIONS A.1.1 and A.1.2 When one or more rods are inoperable (i.e., untrippable), there is a possibility that the required SDM may be adversely affected. Under these conditions, it is important to determine the SDM, and if it is less than the required value, initiate boration until the required SDM is recovered. The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is adequate for determining SDM and, if necessary, for initiating emergency boration and restoring SDM.

A rod is considered trippable if it was demonstrated OPERABLE during the last performance of SR 3.1.4.2 and met the rod drop time criteria during the last performance of SR 3.1.4.3.

In this situation, SDM verification must account for the absence of the negative reactivity of the untrippable rod(s), as well as the rod of maximum worth.

A.2 If the inoperable rod(s) cannot be restored to OPERABLE status, the plant must be brought to a MODE or condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 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 is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

B.1 When a rod becomes misaligned, it can usually be moved and is still trippable (i.e., OPERABLE). If the rod can be realigned within the Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, local xenon redistribution during this short interval will not be significant, and operation may proceed without further restriction.

An alternative to realigning a single misaligned RCCA to the group demand position is to align the remainder of the group to the position of (continued)

CALLAWAY PLANT B 3.1.4-5 Revision 12

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS B.1 (continued) the misaligned RCCA. However, this must be done without violating the bank sequence, overlap, and insertion limits specified in LCO 3.1.5, Shutdown Bank Insertion Limits, and LCO 3.1.6, "Control Bank Insertion Limits." The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> gives the operator sufficient time to adjust the rod positions in an orderly manner.

B.2.1.1 and B.2.1.2 With a misaligned rod, SDM must be verified to be within limit or boration must be initiated to restore SDM to within limit.

In many cases, realigning the remainder of the group to the misaligned rod may not be desirable. For example, realigning control bank B to a rod that is misaligned 15 steps from the top of the core would require a significant power reduction, since control bank D must be fully inserted and control bank C must be inserted to approximately 100 steps.

Power operation may continue with one RCCA OPERABLE (i.e. trippable) but misaligned, provided that SDM is verified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> represents the time necessary for determining the actual unit SDM and, if necessary, aligning and starting the necessary systems and components to initiate boration.

B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 For continued operation with a misaligned rod, reactor power must be reduced, SDM must periodically be verified within limits, hot channel N

factors ( FQ Z and F H ) must be verified within limits, and the safety analyses must be re-evaluated to confirm continued operation is permissible.

Reduction of power to 75% RTP ensures that local LHR increases due to a misaligned RCCA will not cause the core design criteria to be exceeded (Ref. 3). 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 /> gives the operator sufficient time to accomplish an orderly power reduction without challenging the Reactor Trip System.

When a rod is known to be misaligned, there is a potential to impact the SDM. Since the core conditions can change with time, periodic (continued)

CALLAWAY PLANT B 3.1.4-6 Revision 12

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS B.2.2, B.2.3, B.2.4, B.2.5, and B.2.6 (continued) verification of SDM is required. A Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient to ensure this requirement continues to be met.

N Verifying that FQ Z , as approximated by F QC Z and F W Q Z , and F H are within the required limits ensures that current operation at 75% RTP with a rod misaligned is not resulting in power distributions that may invalidate safety analysis assumptions at full power. The Completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> allows sufficient time to obtain flux maps of the core power distribution using core power distribution measurement information and to N

calculate FQ Z and F H .

Once current conditions have been verified acceptable, time is available to perform evaluations of accident analysis to determine that core limits will not be exceeded during a Design Basis Event for the duration of operation under these conditions. The accident analyses presented in FSAR Chapter 15 (Ref. 5) that may be adversely affected will be evaluated to ensure that the analyses results remain valid for the duration of continued operation under these conditions. A Completion Time of 5 days is sufficient time to obtain the required input data and to perform the analysis.

C.1 When Required Actions cannot be completed within their Completion Time, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, which obviates concerns about the development of undesirable xenon or power distributions. 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, for reaching MODE 3 from full power conditions in an orderly manner and without challenging the plant systems.

D.1.1 and D.1.2 More than one control rod becoming misaligned from its group demand position is not expected, and has the potential to reduce SDM. Therefore, SDM must be evaluated. One hour allows the operator adequate time to determine SDM. Restoration of the required SDM, if necessary, requires increasing the RCS boron concentration to provide negative reactivity, as described in the Bases or LCO 3.1.1. The required Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for initiating boration is reasonable, based on the time required for (continued)

CALLAWAY PLANT B 3.1.4-7 Revision 12

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS D.1.1 and D.1.2 (continued) potential xenon redistribution, the low probability of an accident occurring, and the steps required to complete the action. This allows the operator sufficient time to align the required valves and start the boric acid pumps.

Boration will continue until the required SDM is restored.

D.2 If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable. To achieve this status, the unit must be brought to at least MODE 3 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 is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Periodic verification that individual rod positions are within alignment limits provides a history that allows the operator to detect a rod that is beginning to deviate from its expected position. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.1.4.2 Verifying each rod is OPERABLE would require that each rod be tripped.

However, in MODES 1 and 2, tripping each rod would result in radial or axial power tilts, or oscillations. Exercising each individual rod provides confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each rod by 10 steps will not cause radial or axial power tilts, or oscillations, to occur.

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. Between or during required performances of SR 3.1.4.2 (determination of rod OPERABILITY by movement), if a rod(s) is discovered to be immovable, but remains trippable, the rod(s) is considered to be OPERABLE until the surveillance interval expires. At any (continued)

CALLAWAY PLANT B 3.1.4-8 Revision 12

Rod Group Alignment Limits B 3.1.4 BASES SURVEILLANCE SR 3.1.4.2 (continued)

REQUIREMENTS (continued) time, if a rod(s) is immovable, a determination of the trippability (OPERABILITY) of the rod(s) must be made, and appropriate action taken.

SR 3.1.4.3 Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior to reactor criticality, after reactor vessel head removal, ensures that the reactor internals and rod drive mechanism will not interfere with rod motion or rod drop time, and that no degradation in these systems has occurred that would adversely affect rod motion or drop time. This testing is performed with all RCPs operating and the average moderator temperature 500F to simulate a reactor trip under actual conditions.

This Surveillance is performed during a plant outage, due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26.

2. 10 CFR 50.46.

3. FSAR, Chapter 15, Section 15.4.3.

4. FSAR, Section 4.3.1.

5. FSAR, Chapter 15.

CALLAWAY PLANT B 3.1.4-9 Revision 12

Shutdown Bank Insertion Limits B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Shutdown Bank Insertion Limits BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available ejected rod worth, SDM and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design,"

GDC 26, "Reactivity Control System Redundancy and Capability, " GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

The rod cluster control assemblies (RCCAs) are divided among four control banks and five shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. If a bank of RCCAs consists of two groups, the groups are moved in a staggered fashion, but always within one step of each other. Three shutdown banks (C, D, and E) consist of a single group. See LCO 3.1.4, "Rod Group Alignment Limits," for control and shutdown rod OPERABILITY and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements.

The control banks are used for precise reactivity control of the reactor.

The positions of the control banks are controlled automatically or manually using the Rod Control System. Automatic rod control is available for insertion only. They are capable of adding negative reactivity very quickly (compared to borating). The control banks must be maintained above designed insertion limits and are typically near the fully withdrawn position during normal full power operations.

Hence, they are not capable of adding a large amount of positive reactivity. Boration or dilution of the Reactor Coolant System (RCS) compensates for the reactivity changes associated with large changes in RCS temperature. The design calculations are performed with the assumption that the shutdown banks are withdrawn first. The shutdown banks can be fully withdrawn without the core going critical. This (continued)

CALLAWAY PLANT B 3.1.5-1 Revision 10

Shutdown Bank Insertion Limits B 3.1.5 BASES BACKGROUND provides available negative reactivity in the event of boration errors. The (continued) shutdown banks are controlled manually by the control room operator.

During normal unit operation, the shutdown banks are either fully withdrawn or fully inserted. The shutdown banks must be completely withdrawn from the core, prior to withdrawing any control banks during an approach to criticality. The shutdown banks are then left in this position until the reactor is shut down. They affect core power and burnup distribution, and add negative reactivity to shut down the reactor upon receipt of a reactor trip signal.

APPLICABLE On a reactor trip, all RCCAs (shutdown banks and control banks), except SAFETY the most reactive RCCA, are assumed to insert into the core. The ANALYSES shutdown banks shall be at or above their insertion limits and available to insert the maximum amount of negative reactivity on a reactor trip signal.

The control banks may be partially inserted in the core, as allowed by LCO 3.1.6, "Control Bank Insertion Limits." The shutdown bank and control bank insertion limits are established to ensure that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") following a reactor trip from full power. The combination of control banks and shutdown banks (less the most reactive RCCA, which is assumed to be fully withdrawn) is sufficient to take the reactor from full power conditions at rated temperature to zero power, and to maintain the required SDM at rated no load temperature (Ref. 3). The shutdown bank insertion limit also limits the reactivity worth of an ejected shutdown rod.

The acceptance criteria for addressing shutdown and control rod bank insertion limits and inoperability or misalignment is that:

a. There be no violations of:

1. specified acceptable fuel design limits, or

2. RCS pressure boundary integrity; and

b. The core remains subcritical after accident transients.

As such, the shutdown bank insertion limits affect safety analysis involving core reactivity and SDM (Ref. 3).

The shutdown bank insertion limits preserve an initial condition assumed in the safety analyses and, as such, satisfy Criterion 2 of 10CFR50.36(c)(2)(ii).

(continued)

CALLAWAY PLANT B 3.1.5-2 Revision 10

Shutdown Bank Insertion Limits B 3.1.5 BASES (Continued)

LCO The shutdown banks must be within their insertion limits any time the reactor is critical or approaching criticality. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip.

The shutdown bank insertion limits are defined in the COLR.

APPLICABILITY The shutdown banks must be within their insertion limits, with the reactor in MODE 1 and in MODE 2 with any control bank not fully inserted. The applicability in MODE 2 begins at initial control bank withdrawal, during an approach to criticality, and continues throughout MODE 2, until all control bank rods are again fully inserted by reactor trip or by shutdown. This ensures that a sufficient amount of negative reactivity is available to shut down the reactor and maintain the required SDM following a reactor trip.

The shutdown banks do not have to be within their insertion limits in MODE 3, unless an approach to criticality is being made. In MODE 3, 4, 5, or 6, the shutdown banks are typically fully inserted in the core and contribute to the SDM. Refer to LCO 3.1.1 for SDM requirements in MODES 2 with keff < 1.0, 3, 4, and 5. LCO 3.9.1, "Boron Concentration,"

ensures adequate SDM in MODE 6.

The Applicability requirements have been modified by a Note indicating the LCO requirement is suspended during SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the shutdown bank to move below the LCO limits, which would normally violate the LCO.

ACTIONS A.1.1, A.1.2 and A.2 When one or more shutdown banks is not within insertion limits, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is allowed to restore the shutdown banks to within the insertion limits.

This is necessary because the available SDM may be significantly reduced, with one or more of the shutdown banks not within their insertion limits. Also, verification of SDM or initiation of boration within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is required, since the SDM in MODES 1 and 2 is ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1). If shutdown banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1.

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 /> provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

(continued)

CALLAWAY PLANT B 3.1.5-3 Revision 10

Shutdown Bank Insertion Limits B 3.1.5 BASES ACTIONS B.1 (continued)

If the shutdown banks cannot be restored to within their insertion limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the unit must be brought to MODE 3 where the LCO is not applicable. 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, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip.

This SR and Frequency ensure that the shutdown banks are withdrawn before the control banks are withdrawn during a unit startup.

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

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28.

2. 10 CFR 50.46.

3. FSAR, Chapter 15, Section 15.1.5.

CALLAWAY PLANT B 3.1.5-4 Revision 10

Control Bank Insertion Limits B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Control Bank Insertion Limits BASES BACKGROUND The insertion limits of the shutdown and control rods are initial assumptions in all safety analyses that assume rod insertion upon reactor trip. The insertion limits directly affect core power and fuel burnup distributions and assumptions of available SDM, and initial reactivity insertion rate.

The applicable criteria for these reactivity and power distribution design requirements are 10 CFR 50, Appendix A, GDC 10, "Reactor Design,"

GDC 26, "Reactivity Control System Redundancy and Capability, "

GDC 28, "Reactivity Limits" (Ref. 1), and 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors" (Ref. 2). Limits on rod insertion have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

The rod cluster control assemblies (RCCAs) are divided among four control banks and five shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control. A group consists of two or more RCCAs that are electrically paralleled to step simultaneously. If a bank of RCCAs consists of two groups, the groups are moved in a staggered fashion, but always within one step of each other. Three shutdown banks (C, D, and E) consist of a single group. See LCO 3.1.4, "Rod Group Alignment Limits," for control and shutdown rod OPERABILITY and alignment requirements, and LCO 3.1.7, "Rod Position Indication," for position indication requirements.

The control bank insertion limits are specified in the COLR. The control banks are required to be at or above the insertion limit lines.

The COLR also indicates how the control banks are moved in an overlap pattern. Overlap is the distance traveled together by two control banks.

The control banks are used for precise reactivity control of the reactor.

The positions of the control banks are controlled automatically or manually using the Rod Control System. Automatic rod control is available for insertion only. They are capable of adding reactivity very quickly (compared to borating or diluting).

(continued)

CALLAWAY PLANT B 3.1.6-1 Revision 12

Control Bank Insertion Limits B 3.1.6 BASES BACKGROUND The power density at any point in the core must be limited, so that the fuel (continued) design criteria are maintained. Together, 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 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)," and LCO 3.2.4, "QUADRANT POWER TILT RATIO (QPTR)," provide limits on control component operation and on monitored process variables, which ensure that the core operates within the fuel design criteria.

The shutdown and control bank insertion and alignment limits, AFD, and QPTR are process variables that together characterize and control the three dimensional power distribution of the reactor core. Additionally, the control bank insertion limits control the reactivity that could be added in the event of a rod ejection accident, and the shutdown and control bank insertion limits ensure the required SDM is maintained.

Operation within the subject LCO limits will prevent fuel cladding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), loss of flow, ejected rod, or other accident requiring termination by a Reactor Trip System (RTS) trip function.

APPLICABLE The shutdown and control bank insertion limits, AFD, and QPTR LCOs SAFETY are required to prevent power distributions that could result in fuel ANALYSES cladding failures in the event of a LOCA, loss of flow, ejected rod, or other accident requiring termination by an RTS trip function.

The acceptance criteria for addressing shutdown and control bank insertion limits and inoperability or misalignment are that:

a. There be no violations of:

1. specified acceptable fuel design limits, or

2. Reactor Coolant System pressure boundary integrity; and

b. The core remains subcritical after accident transients.

As such, the shutdown and control bank insertion limits affect safety analysis involving core reactivity and power distributions (Ref. 3).

(continued)

CALLAWAY PLANT B 3.1.6-2 Revision 12

Control Bank Insertion Limits B 3.1.6 BASES APPLICABLE The SDM requirement is ensured by limiting the control and shutdown SAFETY bank insertion limits so that allowable inserted worth of the RCCAs is ANALYSES such that sufficient reactivity is available in the rods to shut down the (continued) reactor to hot zero power with a reactivity margin that assumes the maximum worth RCCA remains fully withdrawn upon trip (Ref. 4).

Operation at the insertion limits or AFD limits may approach the maximum allowable linear heat generation rate or peaking factor with the allowed QPTR present. Operation at the insertion limit may also indicate the maximum ejected RCCA worth could be equal to the limiting value in fuel cycles that have sufficiently high ejected RCCA worths.

The control and shutdown bank insertion limits ensure that safety analyses assumptions for SDM, ejected rod worth, and power distribution peaking factors are preserved (Ref. 5).

The insertion limits satisfy Criterion 2 of 10CFR50.36(c)(2)(ii), in that they are initial conditions assumed in the safety analysis.

LCO The limits on control banks sequence, overlap, and physical insertion, as defined in the COLR, must be maintained because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring that ejected rod worth is maintained, and ensuring adequate negative reactivity insertion is available on trip. The overlap between control banks provides more uniform rates of reactivity insertion and withdrawal and is imposed to maintain acceptable power peaking during control bank motion.

APPLICABILITY The control bank sequence, overlap, and physical insertion limits shall be maintained with the reactor in MODES 1 and 2 with keff 1.0. These limits must be maintained, since they preserve the assumed power distribution, ejected rod worth, SDM, and reactivity rate insertion assumptions. Applicability in MODES 3, 4, and 5 is not required, since neither the power distribution nor ejected rod worth assumptions would be exceeded in these MODES.

The applicability requirements have been modified by a Note indicating the LCO requirements are suspended during the performance of SR 3.1.4.2. This SR verifies the freedom of the rods to move, and requires the control bank to move below the LCO limits, which would violate the LCO.

(continued)

CALLAWAY PLANT B 3.1.6-3 Revision 12

Control Bank Insertion Limits B 3.1.6 BASES (Continued)

ACTIONS A.1.1, A.1.2, A.2, B.1.1, B.1.2, and B.2 When the control banks are outside the acceptable insertion limits, they must be restored to within those limits. This restoration can occur in two ways:

a. Reducing power to be consistent with rod position; or

b. Moving rods to be consistent with power.

Also, verification of SDM or initiation of boration to regain SDM is required within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, since the SDM in MODES 1 and 2 normally ensured by adhering to the control and shutdown bank insertion limits (see LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") has been upset. If control banks are not within their insertion limits, then SDM will be verified by performing a reactivity balance calculation, considering the effects listed in the BASES for SR 3.1.1.1.

Similarly, if the control banks are found to be out of sequence or in the wrong overlap configuration, they must be restored to meet the limits.

Operation beyond the LCO limits is allowed for a short time period in order to take conservative action because the simultaneous occurrence of either a LOCA, loss of flow accident, ejected rod accident, or other accident during this short time period, together with an inadequate power distribution or reactivity capability, has an acceptably low probability.

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 /> for restoring the banks to within the insertion, sequence, and overlap limits provides an acceptable time for evaluating and repairing minor problems without allowing the plant to remain in an unacceptable condition for an extended period of time.

Failure of overlap or sequence support equipment does not require entering the ACTIONS as long as sequence and overlap limits are maintained.

C.1 If Required Actions A.1 and A.2, or B.1 and B.2 cannot be completed within the associated Completion Times, the plant must be brought to MODE 3, where the LCO is not applicable. 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, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

(continued)

CALLAWAY PLANT B 3.1.6-4 Revision 12

Control Bank Insertion Limits B 3.1.6 BASES (Continued)

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.

The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. If the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Performing the ECP calculation within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to criticality avoids a large error from changes in xenon concentration, but allows the operator some flexibility to schedule the ECP calculation with other startup activities.

SR 3.1.6.2 Periodic verification of control bank insertion is required to ensure that the control banks do not exceed the insertion limits specified in the COLR.

This surveillance supports safety analysis assumptions. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.1.6.3 When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. The verification of compliance with the sequence and overlap limits specified in the COLR consists of an observation that the static rod positions of those control banks not fully withdrawn from the core are within the limits specified in the COLR. Bank sequence and overlap must also be maintained during rod movement, implicit within the LCO. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, GDC 28.

2. 10 CFR 50.46.

3. FSAR, Chapter 15.

4. FSAR, Chapter 15, Section 15.1.5.

(continued)

CALLAWAY PLANT B 3.1.6-5 Revision 12

Control Bank Insertion Limits B 3.1.6 BASES REFERENCES 5. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation (continued) Methodology," July 1985.

CALLAWAY PLANT B 3.1.6-6 Revision 12

Rod Position Indication B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEM B 3.1.7 Rod Position Indication BASES BACKGROUND According to GDC 13 (Ref. 1), instrumentation to monitor variables and systems over their operating ranges during normal operation, anticipated operational occurrences, and accident conditions must be OPERABLE.

LCO 3.1.7 is required to ensure OPERABILITY of the control and shutdown rod position indicators to determine rod positions and thereby ensure compliance with the rod alignment and insertion limits.

The OPERABILITY, including position indication, of the shutdown and control rods is an initial assumption in all safety analyses that assume rod insertion upon reactor trip. Maximum rod misalignment is an initial assumption in the safety analysis that directly affects core power distributions and assumptions of available SDM. Rod position indication is required to assess OPERABILITY and misalignment.

Mechanical or electrical failures may cause a control or shutdown rod to become inoperable or to become misaligned from its group. Rod inoperability or misalignment may cause increased power peaking, due to the asymmetric reactivity distribution and a reduction in the total available rod worth for reactor shutdown.

Limits on rod alignment and OPERABILITY have been established, and all rod positions are monitored and controlled during power operation to ensure that the power distribution and reactivity limits defined by the design power peaking and SDM limits are preserved.

Rod cluster control assemblies (RCCAs), or rods, are moved out of the core (up or withdrawn) or into the core (down or inserted) by their rod drive mechanisms. The RCCAs are divided among four control banks and five shutdown banks. Each bank may be further subdivided into two groups to provide for precise reactivity control.

The axial position of shutdown rods and control rods are determined by two separate and independent systems: the Bank Demand Position Indication System (commonly called group step counters) and the Digital Rod Position Indication (DRPI) System.

The Bank Demand Position Indication System counts the pulses from the Rod Control System that move the rods. There is one step counter for each group of rods. Individual rods in a group all receive the same signal to move and should, therefore, all be at the same position indicated by (continued)

CALLAWAY PLANT B 3.1.7-1 Revision 12

Rod Position Indication B 3.1.7 BASES BACKGROUND the group step counter for that group. The Bank Demand Position (continued) Indication System is considered highly precise ( 1 step or 5/8 inch).

However, if rod movement does not occur upon demand, the demand step counter will still count the pulse and incorrectly reflect the position of the rod.

The DRPI System provides a highly accurate indication of actual rod position, but at a lower precision than the step counters. This system is based on inductive analog signals from a series of coils spaced along a hollow tube. To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one data system fails, the DRPI will go on half accuracy. The DRPI system is capable of monitoring rod position within at least 12 steps with either full accuracy or half accuracy.

APPLICABLE Control and shutdown rod position accuracy is essential during power SAFETY operation. Power peaking, ejected rod worth, or SDM limits may be ANALYSES violated in the event of a Design Basis Accident (Ref. 2), with control or shutdown rods operating outside their limits undetected. Therefore, the acceptance criteria for rod position indication is that rod positions must be known with sufficient accuracy in order to verify the core is operating within the bank sequence, overlap, design peaking limits, ejected rod worth, and with minimum SDM (LCO 3.1.5, "Shutdown Bank Insertion Limits," and LCO 3.1.6, "Control Bank Insertion Limits"). The rod positions must also be known in order to verify the alignment limits are preserved (LCO 3.1.4, "Rod Group Alignment Limits"). Rod positions are continuously monitored to provide operators with information that ensures the plant is operating within the bounds of the accident analysis assumptions.

The rod position indicator channels satisfy Criterion 2 of 10CFR50.36(c)(2)(ii). The rod position indicators monitor rod position, which is an initial condition of the accident.

LCO LCO 3.1.7 specifies that the DRPI System and the Bank Demand Position Indication System be OPERABLE for each control and shutdown rod. For the rod position indicators to be OPERABLE requires meeting the SR of the LCO and the following:

a. The DRPI System, on either full accuracy or half accuracy, indicates within 12 steps of the group step counter demand position as required by LCO 3.1.4, "Rod Group Alignment Limits";

and (continued)

CALLAWAY PLANT B 3.1.7-2 Revision 12

Rod Position Indication B 3.1.7 BASES LCO b. The Bank Demand Indication System has been calibrated either in (continued) the fully inserted position or to the DRPI System.

The 12 step agreement limit between the Bank Demand Position Indication System and the DRPI System indicates that the Bank Demand Position Indication System is adequately calibrated, and can be used for indication of the measurement of rod bank position.

A deviation of less than the allowable limit, given in LCO 3.1.4, in position indication for a single rod, ensures high confidence that the position uncertainty of the corresponding rod group is within the assumed values used in the analysis (that specified rod group insertion limits).

These requirements ensure that rod position indication during power operation and PHYSICS TESTS is accurate, and that design assumptions are not challenged. OPERABILITY of the position indicator channels ensures that inoperable, misaligned, or mispositioned rods can be detected. Therefore, power peaking, ejected rod worth, and SDM can be controlled within acceptable limits.

APPLICABILITY The requirements on the DRPI and step counters are only applicable in MODES 1 and 2 (consistent with LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6),

because these are the only MODES in which power is generated, and the OPERABILITY and alignment of rods have the potential to affect the safety of the plant. In the shutdown MODES, the OPERABILITY of the shutdown and control banks has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the Reactor Coolant System.

ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator. This is acceptable because the Required Actions for each Condition provide appropriate compensatory actions for each inoperable position indicator.

A.1 When one DRPI per group fails, the position of the rod may still be determined indirectly by use of core power distribution measurement information. Core power distribution measurement information can be obtained from flux maps using the movable incore detectors, or from an OPERABLE power distribution monitoring system (PDMS) (Reference 5).

(continued)

CALLAWAY PLANT B 3.1.7-3 Revision 12

Rod Position Indication B 3.1.7 BASES ACTIONS A.1 (continued)

The Required Action may also be satisfied by ensuring at least once per N

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that FQ Z satisfies LCO 3.2.1, F H satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. The alternate use of peaking factor and SDM verification is limited to those rodded core locations where rod position can not be determined by power distribution measurement infomation. These locations are either not instrumented (or has an out of service incore thimble) or not face-adjacent to instrumented assembly locations. Based on experience, normal power operation does not require excessive movement of banks. If a bank has been significantly moved, the Required Action of C.1 or C.2 below is required. Therefore, verification of RCCA position within 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 adequate for allowing continued full power operation, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. While in Condition A, the performance of Required Action A.1 can be used with the Bank Demand Position Indication System (group step counters) to verify alignment limits are met for SR 3.1.4.1.

A.2 Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 3).

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 /> is reasonable, based on operating experience, for reducing power to 50% RTP from full power conditions without challenging plant systems and allowing for rod position determination by Required Action A.1 above.

B.1, B.2, B.3, and B.4 When more than one DRPI per group fail, additional actions are necessary to ensure that acceptable power distribution limits are maintained, minimum SDM is maintained, and the potential effects of rod misalignment on associated accident analyses are limited. Placing the Rod Control System in manual assures unplanned rod motion will not occur. Together with the indirect position determination available via power distribution measurement information, this will minimize the potential for rod misalignment.

(continued)

CALLAWAY PLANT B 3.1.7-4 Revision 12

Rod Position Indication B 3.1.7 BASES ACTIONS B.1, B.2, B.3, and B.4 (continued)

The Immediate Completion Time for placing the Rod Control System in manual reflects the urgency with which unplanned rod motion must be prevented while in this Condition. Monitoring and recording reactor coolant system Tavg help to assure that significant changes in power distribution and SDM are avoided. The once per hour Completion Time is acceptable because only minor fluctuations in RCS temperature are expected at steady state plant operating conditions.

The position of the rods may be determined indirectly by use of power distribution measurement information. The Required Action may also be satisfied by ensuring at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that FQ Z satisfies N

LCO 3.2.1, F H satisfies LCO 3.2.2, and SHUTDOWN MARGIN is within the limits provided in the COLR, provided the nonindicating rods have not been moved. The alternate use of peaking factor and SDM verification is limited to those rodded core locations where rod position can not be determined by power distribution measurement information. These locations are either not instrumented (or has an out of service incore thimble) or not face-adjacent to instrumented assembly locations.

Verification of RCCA position once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is adequate for allowing continued full power operation for a limited, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, since the probability of simultaneously having a rod significantly out of position and an event sensitive to that rod position is small. While in Condition B, the performance of Required Action B.3 can be used with the Bank Demand Position Indication System (group step counters) to verify alignment limits are met for SR 3.1.4.1. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Completion Time provides sufficient time to troubleshoot and restore the DRPI system to operation while avoiding the plant challenges associated with a shutdown without full rod position indication (Ref. 4).

Based on operating experience, normal power operation does not require excessive rod movement. If one or more rods has been significantly moved, the Required Action of C.1 or C.2 below is required.

C.1 and C.2 These Required Actions clarify that when one or more rods with inoperable DRPIs have been moved in excess of 24 steps in one direction, since the position was last determined, the Required Actions of A.1 and A.2 or B.1 are still appropriate but must be initiated promptly under Required Action C.1 to begin indirectly verifying that these rods are still properly positioned, relative to their group positions.

If, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the rod positions have not been determined, THERMAL POWER must be reduced to 50% RTP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to avoid (continued)

CALLAWAY PLANT B 3.1.7-5 Revision 12

Rod Position Indication B 3.1.7 BASES ACTIONS C.1 and C.2 (continued) undesirable power distributions that could result from continued operation at > 50% RTP, if one or more rods are misaligned by more than 24 steps.

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 period of time to verify the rod positions using the movable incore detectors or other power distribution measurement methods.

D.1.1 and D.1.2 With one demand position indicator per bank inoperable, the rod positions can be determined by the DRPI System. Since normal power operation does not require excessive movement of rods, verification by administrative means that the digital rod position indicators are OPERABLE and the most withdrawn rod and the least withdrawn rod are 12 steps apart within the allowed Completion Time of once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is adequate. While in Condition D, the performance of Required Actions D.1.1 and D.1.2 can be used with the plant computer demand position (which receives the same pulses as the group step counters) to verify alignment limits are met for SR 3.1.4.1.

D.2 Reduction of THERMAL POWER to 50% RTP puts the core into a condition where rod position is not significantly affecting core peaking factors (Ref. 3). 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 /> provides an acceptable period of time to verify the rod positions per Required Actions D.1.1 and D.1.2 or reduce power to 50% RTP.

E.1 If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 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 is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

(continued)

CALLAWAY PLANT B 3.1.7-6 Revision 12

Rod Position Indication B 3.1.7 BASES (Continued)

SURVEILLANCE SR 3.1.7.1 REQUIREMENTS Verification that the DRPI agrees with the demand position within 12 steps ensures that the DRPI is operating correctly. Verification at 24, 48, 120, and 228 steps withdrawn for the control banks and at 18, 210, and 228 steps withdrawn for the shutdown banks provides assurance that the DRPI is operating correctly over the full range of indication. Since the DRPI does not display the actual shutdown rod positions between 18 and 210 steps, only points within the indicated ranges are required in comparison.

This surveillance is performed prior to reactor criticality after each removal of the reactor head, since there is potential for unnecessary plant transients if the SR were performed with the reactor at power.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 13.

2. FSAR, Chapter 15.

3. WCAP-10216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control and FQ Surveillance Technical Specification," February 1994.

4. Amendment 61 to Callaway Plant Facility Operating License NPF-30, February 1, 1991.

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 3.1.7-7 Revision 12

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 PHYSICS TESTS Exceptions - MODE 2 BASES BACKGROUND The primary purpose of the MODE 2 PHYSICS TESTS exceptions is to permit relaxations of existing LCOs to allow certain PHYSICS TESTS to be performed.

Section XI of 10 CFR 50, Appendix B (Ref. 1), requires that a test program be established to ensure that structures, systems, and components will perform satisfactorily in service. All functions necessary to ensure that the specified design conditions are not exceeded during normal operation and anticipated operational occurrences must be tested.

This testing is an integral part of the design, construction, and operation of the plant. Requirements for notification of the NRC, for the purpose of conducting tests and experiments, are specified in 10 CFR 50.59 (Ref. 2).

The key objectives of a test program are to (Ref. 3):

a. Ensure that the facility has been adequately designed;

b. Validate the analytical models used in the design and analysis;

c. Verify the assumptions used to predict unit response;

d. Ensure that installation of equipment in the facility has been accomplished in accordance with the design; and

e. Verify that the operating and emergency procedures are adequate.

To accomplish these objectives, testing is performed prior to initial criticality, during startup, during low power operations, during power ascension, at high power, and after each refueling. The PHYSICS TESTS requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions and that the core can be operated as designed.

PHYSICS TESTS procedures are written and approved in accordance with established formats. The procedures include all information necessary to permit a detailed execution of the testing required to ensure that the design intent is met. PHYSICS TESTS are performed in accordance with these procedures and test results are approved prior to continued power escalation and long term power operation.

(continued)

CALLAWAY PLANT B 3.1.8-1 Revision 11

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES BACKGROUND The PHYSICS TESTS typically required for reload fuel cycles in (continued) MODE 2 include:

a. Critical Boron Concentration - Control Rods Withdrawn;

b. Critical Boron Concentration - Control Rods Inserted;

c. Control Rod Worth; and

d. Isothermal Temperature Coefficient (ITC).

These and other supplementary tests may be required to calibrate the nuclear instrumentation or to diagnose operational problems. These tests may cause the operating controls and process variables to deviate from their LCO requirements during their performance.

a. The Critical Boron Concentration - Control Rods Withdrawn Test measures the critical boron concentration at hot zero power (HZP). With all rods out, the lead control bank is at or near its fully withdrawn position. HZP is where the core is critical (keff = 1.0),

and the Reactor Coolant System (RCS) is at design temperature and pressure for zero power. Performance of this test should not violate any of the referenced LCOs.

b. The Critical Boron Concentration - Control Rods Inserted Test measures the critical boron concentration at HZP, with a bank having a worth of at least 1% k/k at or near its fully inserted position in the core. With the core at HZP and all banks fully withdrawn, the boron concentration of the reactor coolant is gradually lowered in a continuous manner. The selected bank is then inserted to make up for the decreasing boron concentration.

The boron concentration is then measured with the selected bank at or near its fully inserted position. This test may be performed concurrently with the Control Rod Worth Test described below.

Performance of this test could violate LCO 3.1.4, "Rod Group Alignment Limits"; LCO 3.1.5, "Shutdown Bank Insertion Limits";

or LCO 3.1.6, "Control Bank Insertion Limits."

c. The Control Rod Worth Test is used to measure the reactivity worth of selected banks. This test is performed at HZP and has four alternative methods of performance. The first method, the Boron Exchange Method, varies the reactor coolant boron concentration and moves the selected bank in response to the changing boron concentration. The reactivity changes are measured with a reactivity computer. This sequence is continued (continued)

CALLAWAY PLANT B 3.1.8-2 Revision 11

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES BACKGROUND for the remaining selected banks. The second method, the Rod (continued) Swap Method, measures the worth of a predetermined reference bank using the Boron Exchange Method above. The reference bank is then nearly fully inserted into the core. The selected bank is then inserted into the core as the reference bank is withdrawn.

The HZP critical conditions are then determined with the selected bank fully inserted into the core. The worth of the selected bank is inferred, based on the position of the reference bank with respect to the selected bank. This sequence is repeated as necessary for the remaining selected banks. The third method, the Boron Endpoint Method, moves the selected bank over its entire length of travel and then varies the reactor coolant boron concentration to achieve HZP criticality again. The difference in boron concentration is the worth of the selected bank. This sequence is repeated for the remaining selected banks. The fourth method, the Dynamic Rod Worth Measurement Method, fully inserts and withdraws the selected bank into the core while measuring reactivity changes with a reactivity computer. Since the reactor is not maintained critical while the selected bank is inserted, the measured reactivity is corrected based on design predictions to obtain the actual measured bank worth. The insertion and withdrawal sequence is repeated for each selected bank to obtain their worths. Performance of this test by any of the four methods could violate LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6.

d. The ITC Test measures the ITC of the reactor. This test is performed at HZP and has two methods of performance. The first method, the Slope Method, varies RCS temperature in a slow and continuous manner. The reactivity change is measured with a reactivity computer as a function of the temperature change. The ITC is the slope of the reactivity versus the temperature plot. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs. The second method, the Endpoint Method, changes the RCS temperature and measures the reactivity at the beginning and end of the temperature change. The ITC is the total reactivity change divided by the total temperature change. The test is repeated by reversing the direction of the temperature change, and the final ITC is the average of the two calculated ITCs.

Performance of this test could violate LCO 3.4.2, "RCS Minimum Temperature for Criticality."

APPLICABLE The fuel is protected by LCOs that preserve the initial conditions of the SAFETY core assumed during the safety analyses. The methods for development ANALYSES of the LCOs that are excepted by this LCO are described in the (continued)

CALLAWAY PLANT B 3.1.8-3 Revision 11

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES APPLICABLE Westinghouse Reload Safety Evaluation Methodology Report (Ref. 5).

SAFETY The above mentioned PHYSICS TESTS, and other tests that may be ANALYSES required to calibrate nuclear instrumentation or to diagnose operational (continued) problems, may require the operating control or process variables to deviate from their LCO limitations.

The FSAR defines requirements for initial testing of the facility, including PHYSICS TESTS. Section 14.2 summarizes the zero, low power, and power tests. Reload fuel cycle PHYSICS TESTS are performed in accordance with Technical Specification requirements, fuel vendor guidelines, and established industry practices. Although these PHYSICS TESTS are generally accomplished within the limits for all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. When one or more of the requirements specified in LCO 3.1.3, "Moderator Temperature Coefficient (MTC)," LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 are suspended for PHYSICS TESTS, the fuel design criteria are preserved as long as the power level is limited to 5% RTP, the reactor coolant lowest operating loop temperature is kept 541F, and SDM is within the limits specified in the COLR.

The PHYSICS TESTS include measurement of core nuclear parameters or the exercise of control components that affect process variables.

Among the process variables involved are MTC and RCS Average Temperature, which represent initial conditions of the unit safety analyses.

Also involved are the movable control components (control and shutdown rods), which are required to shut down the reactor. The limits for these variables are specified for each fuel cycle in the COLR. PHYSICS TESTS meet the criteria for inclusion in the Technical Specifications, since the components and process variable LCOs suspended during PHYSICS TESTS meet Criteria 1, 2, and 3 of 10CFR50.36(c)(2)(ii).

Reference 6 allows special test exceptions (STEs) to be included as part of the LCO that they affect. It was decided, however, to retain this STE as a separate LCO because it was less cumbersome and provided additional clarity.

LCO This LCO allows the reactor parameters of MTC and minimum temperature for criticality to be outside their specified limits. In addition, it allows selected control and shutdown rods to be positioned outside of their specified alignment and insertion limits. One Power Range Neutron Flux channel may be bypassed, reducing the number of required channels from 4 to 3. Operation beyond specified limits is permitted for (continued)

CALLAWAY PLANT B 3.1.8-4 Revision 11

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES LCO the purpose of performing PHYSICS TESTS and poses no threat to fuel (continued) integrity, provided the SRs are met.

The requirements of LCO 3.1.3, LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, and LCO 3.4.2 may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, 6, and 18.e, may be reduced to 3 required channels during the performance of PHYSICS TESTS provided:

a. RCS lowest operating loop average temperature is 541F;

b. SDM is within the limits specified in the COLR; and

c. THERMAL POWER is 5% RTP.

APPLICABILITY This LCO is applicable in MODE 2 when performing low power PHYSICS TESTS. The applicable PHYSICS TESTS are performed in MODE 2 at HZP.

ACTIONS A.1 and A.2 If the SDM requirement is not met, boration must be initiated promptly. A Completion Time of 15 minutes is adequate for an operator to correctly align and start the required systems and components. The operator should begin boration with the best source available for the plant conditions. Boration will be continued until SDM is within limit.

Suspension of PHYSICS TESTS exceptions requires restoration of each of the applicable LCOs to within specification.

B.1 When THERMAL POWER is > 5% RTP, the only acceptable action is to open the reactor trip breakers (RTBs) to prevent operation of the reactor beyond its design limits. Immediately opening the RTBs will shut down the reactor and prevent operation of the reactor outside of its design limits.

C.1 When the RCS lowest operating loop Tavg is < 541F, the appropriate action is to restore Tavg to within its specified limit. The allowed (continued)

CALLAWAY PLANT B 3.1.8-5 Revision 11

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES ACTIONS C.1 (continued)

Completion Time of 15 minutes provides time for restoring Tavg to within limits without allowing the plant to remain in an unacceptable condition for an extended period of time. Operation with the reactor critical and with an operating loops temperature below 541F could violate the assumptions for accidents analyzed in the safety analyses.

D.1 If the Required Actions cannot be completed within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, for reaching MODE 3 in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.8.1 REQUIREMENTS The required power range and intermediate range channels must be OPERABLE in MODE 2 by LCO 3.3.1, "Reactor Trip System (RTS)

Instrumentation." A CHANNEL OPERATIONAL TEST is performed on each OPERABLE power range and intermediate range channel prior to initiation of the PHYSICS TESTS. This will ensure that the RTS is properly aligned to provide the required degree of core protection during the performance of the PHYSICS TESTS.

SR 3.1.8.2 Verification that the RCS lowest operating loop Tavg is 541F will ensure that the unit is not operating in a condition that could invalidate the safety analyses. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

SR 3.1.8.3 Verification that the THERMAL POWER is 5% RTP will ensure that the plant is not operating in a condition that could invalidate the safety analyses. The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

(continued)

CALLAWAY PLANT B 3.1.8-6 Revision 11

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES SURVEILLANCE SR 3.1.8.4 REQUIREMENTS (continued) Verification that the SDM is within limits specified in the COLR ensures that, for the specific RCCA and RCS temperature manipulations performed during PHYSICS TESTS, the plant is not operating in a manner that could invalidate the safety analysis assumptions.

During PHYSICS TESTS in which the requirements of LCOs 3.1.4, 3.1.5, and 3.1.6 are satisfied, the SDM surveillance consists of a verification that the rod insertion limits of LCOs 3.1.5 and 3.1.6 are met.

During PHYSICS TESTS in which the requirements of LCO 3.1.4, LCO 3.1.5, or LCO 3.1.6 are not met, the SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects:

a. RCS boron concentration (may include allowances for boron-10 depletion);

b. Control and shutdown rod position;

c. RCS average temperature;

d. Fuel burnup based on gross thermal energy generation;

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation when the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

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

REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50.59.

3. Regulatory Guide 1.68, Revision 2, August, 1978.

4. Not Used.

(continued)

CALLAWAY PLANT B 3.1.8-7 Revision 11

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES REFERENCES 5. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation (continued) Methodology Report," July 1985.

6. WCAP-11618, including Addendum 1, April 1989.

CALLAWAY PLANT B 3.1.8-8 Revision 11

RCS Boron Limitations < 500°F B 3.1.9 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.9 RCS Boron Limitations < 500°F BASES BACKGROUND The control rod drive mechanisms (CRDMs) are wired into pre-selected RCCA banks, such that the RCCA banks can only be withdrawn in their proper withdrawal sequence during the normal mode of operation (i.e., not in the bank select mode). The control of the power supplied to the RCCA banks is such that no more than two RCCA banks can be withdrawn at any time.

When the RCCA banks are capable of being withdrawn from the core, i.e.,

power supplied to the CRDMs during an approach to criticality for reactor startup, or during maintenance and surveillance testing, there is the potential for an inadvertent RCCA bank withdrawal due to a malfunction of the control rod drive system.

Westinghouse NSAL-00-016 (Ref. 1) discussed the reactor trip functions assumed in the analysis of an Uncontrolled RCCA Bank Withdrawal from a Low Power or Subcritical Condition event (RWFS) (Ref. 2). The primary protection for an RWFS event is provided by the Power Range Neutron Flux - Low trip Function. The Source Range Neutron Flux trip Function is implicitly credited as the primary reactor trip function for an RWFS event in MODES 3, 4, or 5 since the Power Range Neutron Flux - Low trip Function is not required to be OPERABLE throughout these MODES. However, the Source Range Neutron Flux trip Function response time is listed as not applicable (Ref. 3) and that trip function is not response time tested per SR 3.3.1.16. Therefore, the Source Range Neutron Flux trip Function can not be credited to provide protection for an RWFS event in MODES 3, 4, and 5.

NSAL-00-016 also identified that the Power Range Neutron Flux - Low trip Function may not be OPERABLE at RCS temperatures significantly below the hot zero power T-avg due to calibration issues associated with shielding caused by the cold water in the downcomer region of the reactor vessel. The low RCS temperature limit for OPERABILITY of the Power Range Neutron Flux - Low trip Function is 500°F. Therefore, the Power Range Neutron Flux - Low trip Function may not provide the required protection in MODE 3 when the RCS temperature is < 500°F, nor in MODES 4 and 5, due to the calibration issues discussed above.

Borating the RCS to greater than the all rods out (ARO) critical boron concentration when the RCCA banks are capable of being withdrawn (continued)

CALLAWAY PLANT B 3.1.9-1 Revision 11

RCS Boron Limitations < 500°F B 3.1.9 BASES BACKGROUND provides sufficient SHUTDOWN MARGIN in the event of an RWFS (continued) transient when the RCS temperature is < 500°F.

APPLICABLE The RCCA bank withdrawal transient addressed by this LCO is the RWFS SAFETY event. An RCCA bank withdrawal event at power is also analyzed, but ANALYSES that event is mitigated by equipment covered by the requirements of other Technical Specifications that are applicable in MODE 1, such as the Power Range Neutron Flux - High, Power Range Neutron Flux Rate - High Positive Rate, and Overtemperature T trip Functions. The RWFS event assumes a positive reactivity insertion rate that is equal to the worth obtained from the simultaneous withdrawal of the combination of the two sequential control banks with the highest combined worth moving together with 100% overlap at the maximum withdrawal speed. The RWFS event is assumed to be terminated by the Power Range Neutron Flux - Low trip Function. The Source Range Neutron Flux and Intermediate Range Neutron Flux trip Functions are also available to terminate an RWFS event, but are not explicitly credited in the safety analyses to terminate the event.

The Power Range Neutron Flux - Low trip Function is available to provide the required protection for an RWFS event when the RCS temperature is

> 500°F. This temperature limitation is due to calibration issues associated with shielding caused by cold water in the downcomer region of the reactor vessel. Additionally, although not explicitly analyzed in MODES 3, 4, and 5 below 500°F, the Source Range Neutron Flux trip Function is implicitly credited to provide protection for an RWFS event in these MODES.

Since there is no explicit RCCA bank withdrawal analysis that is performed in MODE 3 when the RCS temperature is below 500°F, nor in MODES 4 and 5, and the Power Range Neutron Flux - Low trip Function can not be credited to mitigate an RWFS event with the RCS temperature below 500°F, LCO 3.1.9 requires that the RCS boron concentration be greater than the ARO critical boron concentration when the Rod Control System is capable of rod withdrawal in these MODES. This requirement provides sufficient SHUTDOWN MARGIN to prevent the undesirable consequences (i.e., inadvertent criticality) that could result from an RWFS event.

RCS Boron Limitations < 500°F satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

(continued)

CALLAWAY PLANT B 3.1.9-2 Revision 11

RCS Boron Limitations < 500°F B 3.1.9 BASES (Continued)

LCO This LCO requires that the boron concentration of the RCS be greater than the ARO critical boron concentration to provide adequate SHUTDOWN MARGIN in the event of an RWFS transient.

APPLICABILITY In the event of an RWFS transient, this LCO must be applicable to provide adequate SHUTDOWN MARGIN in the following MODES and specified conditions:

• In MODE 2 with keff < 1.0 with any RCS cold leg temperature

< 500°F and with the Rod Control System capable of rod withdrawal;

• In MODE 3 with any RCS cold leg temperature < 500°F and with the Rod Control System capable of rod withdrawal; and

• In MODES 4 and 5 with the Rod Control System capable of rod withdrawal.

In MODE 6 the requirements of LCO 3.1.9 are not applicable because the Rod Control System is not capable of rod withdrawal.

When protection is required to mitigate an RWFS event while operating under specified conditions other than those above in MODES 2 and 3, LCO 3.3.1, "Reactor Trip System Instrumentation," assures that the Power Range Neutron Flux - Low trip Function is OPERABLE to mitigate the event.

In MODE 1 the requirements of LCO 3.1.9 are not applicable since an uncontrolled RCCA bank withdrawal event at power would be mitigated by the Power Range Neutron Flux - High trip Function, or the Power Range Neutron Flux Rate - High Positive Rate trip Function, or the Overtemperature T trip Function, all of which are required to be OPERABLE by LCO 3.3.1, "Reactor Trip System Instrumentation."

Since this Specification has no LCO 3.0.4.c allowance, MODE 5 can not be entered from MODE 6 while not meeting the RCS boron concentration limits. The risk assessments of LCO 3.0.4.b may only be utilized for systems and components, not Criterion 2 values or parameters such as RCS boron concentration. Therefore, a risk assessment per LCO 3.0.4.b to allow MODE changes with single or multiple system/equipment inoperabilities can not be used to allow a MODE change into, or ascending within, this LCO while not meeting the RCS boron concentration limits, (continued)

CALLAWAY PLANT B 3.1.9-3 Revision 11

RCS Boron Limitations < 500°F B 3.1.9 BASES APPLICABILITY even if the risk assessment specifically includes consideration of RCS (continued) boron concentration.

ACTIONS A.1 If the RCS boron concentration is not within limit, action must be taken immediately to restore the boron concentration to within limit. Borating the RCS to a boron concentration greater than the ARO critical boron concentration provides sufficient SHUTDOWN MARGIN if an RWFS event should occur. Initiating action immediately to restore the boron concentration to within limit provides assurance that the LCO requirement will be restored in a timely manner. The Completion Time is reasonable, considering the low probability of an RWFS event occurring while restoring the boron concentration to within limit. Additionally, although not explicitly credited as a primary trip function, the Source Range Neutron Flux trip Function would provide protection for an RWFS event during this period of time.

A.2 If the RCS boron concentration is not within limit, an alternate action is to make the Rod Control System incapable of rod withdrawal (e.g., by de-energizing all CRDMs, by opening the RTBs, or de-energizing the motor generator (MG) sets). This action precludes an RWFS event from occurring with an inadequate SHUTDOWN MARGIN. Initiating action immediately to make the Rod Control System incapable of rod withdrawal provides adequate assurance that the plant is promptly placed in a condition in which the boron concentration requirements of the LCO are no longer required to mitigate the consequences of an RWFS event.

A.3 If the RCS boron concentration is not within limit, another alternate action is to restore all RCS cold leg temperatures to 500°F. At this RCS temperature the Power Range Neutron Flux - Low trip Function would be available to provide the necessary protection should an RWFS event occur. Initiating action immediately to restore all RCS cold leg temperatures to 500°F provides adequate assurance that the plant is promptly placed in a condition in which the boron concentration requirements of the LCO are no longer necessary.

(continued)

CALLAWAY PLANT B 3.1.9-4 Revision 11

RCS Boron Limitations < 500°F B 3.1.9 BASES ACTIONS A.3 (continued)

Additionally, although not explicitly credited as a primary trip function, the Source Range Neutron Flux trip Function would provide protection for an RWFS event while RCS temperature is being increased.

Required Action A.3 is modified by a Note that states that it is not applicable in MODES 4 and 5. The Note provides assurance that this Required Action would only be taken in MODES 2 and 3 (i.e., during a plant startup) when the RCS temperature can readily be increased to 500°F. After the RCS cold leg temperatures are increased to 500°F, the requirements of LCO 3.1.9 are no longer applicable and protection for an RWFS event would be provided by the Power Range Neutron Flux -

Low trip Function, which is required to be OPERABLE by LCO 3.3.1, "Reactor Trip System Instrumentation."

SURVEILLANCE SR 3.1.9.1 REQUIREMENTS This SR ensures that the RCS boron concentration is within limit. The boron concentration is determined periodically by chemical analysis.

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

REFERENCES 1. Westinghouse Nuclear Safety Advisory Letter NSAL-00-016, Rod Withdrawal from Subcritical Protection in Lower Modes, December 4, 2000.

2. FSAR Section 15.4.1.

3. FSAR Table 16.3-1.

CALLAWAY PLANT B 3.1.9-5 Revision 11