Text

Technical Specifications, for Inserting and Remove of Amendment 148, 149, and 133
	ML021980536
Person / Time
Site:	Callaway
Issue date:	07/09/2002
From:	; AmerenUE
To:	; Office of Nuclear Reactor Regulation
References
	489080
	Download: ML021980536 (41)
	v • d • e

DATE: 07/09/02 AMEREN/UE PAGE: 63

,-TIME: 07:43:00 DOCUMENT CONTROL SYSTEM ARDC8801 DOCUMENT TRANSMITTAL TRANSMITTAL NUMBER: 489080 RETURN ACKNOWLEDGED TRANSMITTAL AND TO CONTROL NUMBER: 423U SUPERSEDED DOCUMENTS (IF APPLICABLE) TO:

TITLE: OTHER ADMINISTRATION RECORDS DEPT: NUCLEAR REGULATORY COMM. AMEREN/UE LOCATION: USNRC-WASH DC CALLAWAY PLANT TRANSMITTAL DATE: 20020709 P.O. BOX 620 FULTON, MO 65251 TRAN DOC RET ALT ALT CODE TYPE DOCUMENT NUMBER REV REV MED COPY MED COPY AFFECTED DOCUMENT A CNOT AMENDMENT 148 C 1 TECH SPEC A CNOT 02-008 TSB C 1 TECH SPEC BASES ACKNOWLEDGED BY: DATE:

July 9, 2002 DIRECTIONS FOR INSERTING AMENDMENT 148 TO TECH. SPEC.

Remove and insert pages as listed below. Dashes (---) in the remove or insert column of the directions indicate no actions required.

REMOVE INSERT LIST OF EFFECTIVE PAGES Amendment 149 Amendment 148 (total pages 5) .(total pages 5)

SPECIFICATION 3.3-12, Amendment 133 .3.3-12, Amendment 148

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Table of Contents Chapter 3.0 1 133 3.0-1 133 2 135 3.0-2 133 3 135 3.0-3 133 4 133 3.0-4 147 3.0-5 133 Chapter 1.0 Chapter 3.1 1.1-1 135 1.1-2 133 3.1-1 133 1.1-3 133 3. 1-2 133 1.1-4 133 3. 1-3 133 1.1-5 133 3. 1-4 133 1.1-6 135 3. 1-5 133 1.1-7 133 3. 1-6 133 1.2-1 133 3. 1-7 133 1.2-2 133 3. 1-8 133 1.2-3 133 3. 1-9 133 1.3-1 133 3. 1-10 133 1.3-2 133 3. 1-11 133 1.3-3 133 3. 1-12 133 1.3-4 133 3. 1-13 133 1.3-5 133 3. 1-14 133 1.3-6 133 3. 1-15 133 1.3-7 133 3. 1-16 133 1.3-8 133 3. 1-17 133 1.3-9 133 3. 1-18 133 1.3-10 135 3. 1-19 133 1.3-11 133 3. 1-20 133 1.3-12 133 1.3-13 133 1.4-1 133 Chapter 3.2 1.4-2 133 1.4-3 133 3.2-1 133 1.4-4 133 3.2-2 133 3.2-3 133 3.2-4 133 Chapter 2.0 3.2-5 133 3.2-6 133 2.0-1 133 3.2-7 133 2.0-2 133 3.2-8 133 1 Amendment No. 148

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Chapter 3.2 Chapter 3.3 (continued) (continued) 3.2-9 133 3.3-33 133 3.2-10 133 3.3-34 133 3.2-11 133 3.3-35 133 3.2-12 133 3.3-36 137 3.2-13 133 3.3-37 133 3.3-38 133 3.3-39 133 Chapter 3.3 3.3-40 133 3.3-41 133 3.3-1 133 3.3-42 133 3.3-2 133 3.3-43 141 3.3-3 133 3.3-44 141 3.3-4 133 3.3-45 133 3.3-5 149 3.3-46 133 3.3-6 133 3.3-47 133 3.3-7 133 3.3-48 133 3.3-8 133 3.3-49 133 3.3-9 133 3.3-50 133 3.3-10 133 3.3-51 133 3.3-11 133 3.3-52 133 3.3-12 148 3.3-53 133 3.3-13 133 3.3-54 133 3.3-14 133 3.3-55 133 3.3-15 133 3.3-56 133 3.3-16 133 3.3-57 133 3.3-17 133 3.3-58 133 3.3-18 133 3.3-59 133 3.3-19 133 3.3-60 133 3.3-20 133 3.3-61 133 3.3-21 133 3.3-62 133 3.3-22 133 3.3-63 133 3.3-23 133 3.3-64 133 3.3-24 133 3.3-65 133 3.3-25 133 3.3-66 133 3.3-26 133 3.3-67 133 3.3-27 133 3.3-68 133 3.3-28 137 3.3-69 133 3.3-29 133 3.3-70 149 3.3-30 133 3.3-71 133 3.3-31 133 3.3-72 133 3.3-32 133 2 Amendment No. 148

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Chapter 3.4 Chapter 3.4 (continued) 3.4-1 133 3.4-42 133 3.4-2 133 3.4-43 133 3.4-3 133 3.4-4 133 3.4-5 133 Chapter 3.5 3.4-6 133 3.4-7 149 3.5-1 133 3.4-8 149 3.5-2 133 3.4-9 133 3.5-3 133 3.4-10 149 3.5-4 133 3.4-11 149 3.5-5 140 3.4-12 149 3.5-6 133 3.4-13 149 3.5-7 133 3.4-14 133 3.5-8 133 3.4-15 149 3.5-9 133 3.4-16 149 3.5-10 146 3.4-17 133 3.5-11 146 3.4-18 133 3.4-19 137 3.4-20 137 Chapter 3.6 3.4-21 137 3.4-22 137 3.6-1 133 3.4-23 133 3.6-2 133 3.4-24 133 3.6-3 133 3.4-25 133 3.6-4 133 3.4-26 133 3.6-5 133 3.4-27 133 3.6-6 133 3.4-28 133 3.6-7 133 3.4-29 133 3.6-8 133 3.4-30 133 3.6-9 133 3.4-31 133 3.6-10 133 3.4-32 133 3.6-11 133 3.4-33 133 3.6-12 133 3.4-34 133 3.6-13 133 3.4-35 133 3.6-14 133 3.4-36 133 3.6-15 133 3.4-37 135 3.6-16 133 3.4-38 133 3.6-17 133 3.4-39 133 3.6-18 133 3.4-40 133 3.6-19 133 3.4-41 133 3.6-20 133 3 Amendment No. 148

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Chapter 3.6 Chapter 3.7 (continued) (continued) 3.6-21 133 3.7-35 133 3.6-22 133 3.7-36 135 3.6-23 133 3.7-37 133 3.7-38 133 3.7-39 133 Chapter 3.7 3.7-40 133 3.7-1 136 3.7-2 133 Chapter 3.8 3.7-3 136 3.7-4 133 3.8-1 133 3.7-5 135 3.8-2 133 3.7-6 133 3.8-3 133 3.7-7 133 3.8-4 133 3.7-8 133 3.8-5 133 3.7-9 133 3.8-6 133 3.7-10 135 3.8-7 133 3.7-11 133 3.8-8 133 3.7-12 133 3.8-9 133 3.7-13 133 3.8-10 133 3.7-14 133 3.8-11 133 3.7-15 133 3.8-12 133 3.7-16 133 3.8-13 133 3.7-17 133 3.8-14 133 3.7-18 133 3.8-15 133 3.7-19 133 3.8-16 133 3.7-20 133 3.8-17 149 3.7-21 133 3.8-18 149 3.7-22 133 3.8-19 133 3.7-23 133 3.8-20 138 3.7-24 133 3.8-21 133 3.7-25 133 3.8-22 133 3.7-26 133 3.8-23 133 3.7-27 133 3.8-24 133 3.7-28 133 3.8-25 149 3.7-29 133 3.8-26 149 3.7-30 133 3.8-27 133 3.7-31 135 3.8-28 133 3.7-32 133 3.8-29 133 3.7-33 133 3.8-30 133 3.7-34 133 3.8-31 133 4 Amendment No. 148

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Chapter 3.8 Chapter 5.0 (continued) (continued) 3.8-32 133 5.0-9 133 3.8-33 149 5.0-10 133 3.8-34 149 5.0-11 133 3.8-35 133 5.0-12 133 3.8-36 133 5.0-13 133 3.8-37 149 5.0-14 133 3.8-38 133 5.0-15 133 5.0-16 133 5.0-17 133 Chapter 3.9 5.0-18 133 5.0-19 133 3.9-1 133 5.0-20 133 3.9-2 133 5.0-21 133 3.9-3 133 5.0-22 133 3.9-4 149 5.0-23 133 3.9-5 133 5.0-24 133 3.9-6 138 5.0-25 142 3.9-7 133 5.0-26 133 3.9-8 149 5.0-27 133 3.9-9 149 5.0-28 135 3.9-10 133 5.0-29 134 3.9-11 149 5.0-30 134 3.9-12 133 5.0-31 133 5.0-32 133 5.0-33 133 Chapter 4.0 5.0-34 133 5.0-35 133 4.0-1 133 5.0-36 133 4.0-2 133 4.0-3 133 Chapter 5.0 5.0-1 133 5.0-2 145 5.0-3 133 5.0-4 133 5.0-5 133 5.0-6 133 5.0-7 144 5.0-8 133 5 Amendment No. 148

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

NOTE -----------------------------

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.3.1.2 ---------------------- NOTE -------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after I THERMAL POWER is Ž>15% RTP.

Compare results of calorimetric heat balance 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months calculation to power range channel output. Adjust power range channel output if calorimetric heat balance calculation results exceed power range channel output by more than +2% RTP.

SR 3.3.1.3 ---------------------- NOTE --------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is _>50% RTP.

Compare results of the incore detector 31 effective full measurements to Nuclear Instrumentation System power days (NIS) AFD. Adjust NIS channel if absolute difference (EFPD) is Ž_2%.

(continued)

CALLAWAY PLANT 3.3-12 Amendment No. 148

July 9, 2002 DIRECTIONS FOR INSERTING CN 02-008 REVISION TO TS BASES Remove and insert pages as listed below. Dashes (---) in the remove or insert column of the directions indicate no actions required.

REMOVE INSERT LIST OF EFFECTIVE PAGES Revision 2f Revision 2g (Total Pages 12) (Total Pages 12)

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BASES LIST OF EFFECTIVE PAGES Page No. Revision Page No. Revision Chapter 3.5 Bases Chapter 3.6 Bases (continued) (continued)

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BASES LIST OF EFFECTIVE PAGES Page No. Revision Page No. Revision Chapter 3.7 Bases Chapter 3.7 Bases (continued) (continued)

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BASES LIST OF EFFECTIVE PAGES Page No. Revision Page No. Revision Chapter 3.7 Bases Chapter 3.8 Bases (continued) (continued)

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BASES LIST OF EFFECTIVE PAGES Page No. Revision Page No. Revision Chapter 3.8 Bases Chapter 3.9 Bases (continued) (continued)

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RTS Instrumentation B 3.3.1 BASES ACTIONS X.1 and X.2 (continued)

Condition X applies to the Environmental Allowance Modifier (EAM) circuitry for the SG Water Level - Low Low trip Function in MODES 1 and 2. With one or more EAM channel(s) inoperable, they must be placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Placing an EAM channel in trip automatically enables the SG Water Level - Low Low (Adverse Containment Environment) bistable for that protection channel, with its higher SG level Trip Setpoint (a higher trip setpoint means a reactor trip would occur sooner). The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is based on Reference 7. Ifthe inoperable channel cannot be placed in the tripped condition within the specified Completion Time, the unit must be placed in a MODE where this Function is not required to be OPERABLE. An additional six hours is allowed to place the unit in MODE 3.

SURVEILLANCE The SRs for each RTS Function are identified by the SRs column of REQUIREMENTS Table 3.3.1-1 for that Function.

A Note has been added stating that Table 3.3.1-1 determines which SRs apply to which RTS Functions.

Note that each channel of process protection supplies both trains of the RTS. When testing Channel I, Train A and Train B must be examined.

Similarly, Train A and Train B must be examined when testing Channel II, Channel III, and Channel IV. The CHANNEL CALIBRATIONs and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies.

SR 3.3.1.1 Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

(continued)

CALLAWAY PLANT B 3.3.1-45 Revision 2f

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.1 (continued)

REQUIREMENTS Agreement criteria are determined by the unit staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

The Frequency is based on operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the LCO required channels.

SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the power range channel output every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Ifthe calorimetric heat balance calculation results exceed the power range channel output by more than +2% RTP, the power range channel is not declared inoperable, but must be adjusted. The power range channel output shall be adjusted consistent with the calorimetric heat balance calculation results if the calorimetric calculation exceeds the power range channel output by more than +2% RTP. If the power range channel output cannot be properly adjusted, the channel is declared inoperable.

If the calorimetic is performed at part-power (<40% RTP), adjusting the power range channel indication in the increasing power direction will assure a reactor trip be!ow the power range high safety analysis limit (SAL) of < 118% RTP in FSAR Table 15.0-4 (Reference 10). Making no adjust to the power range channel in the decreasing power direction due to a part-power calorimetric assures a reactor trip consistent with the safety analyses.

This allowance does not preclude making indicated power adjustments, if desired, when the calorimetric heat balance calculation power is less than the power range channel output. To provide close agreement between indicated power and to preserve operating margin, the power range channels are normally adjusted when operating at or near full power during steady-state conditions. However, discretion must be exercised if the power range channel output is adjusted in the decreasing power direction due to a part-power calorimetric (<40% RTP). This action could introduce a non-conservative bias at higher power levels which could delay an NIS reactor trip until power is above the power range high SAL.

The cause of the non-conservative bias is the decreased accuracy of the calorimetric at reduced power conditions. The primary error contributor to (continued)

CALLAWAY PLANT B 3.3.1-46 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.2 (continued)

REQUIREMENTS the instrument uncertainty for a secondary side power calorimetric measurement is the feedwater flow measurement, which is determined by a AP measurement across a feedwater venturi. While the measurement uncertainty remains constant in AP span as power decreases, when translated into flow the uncertainty increases as a square term. Thus, a 1% flow error at 100% power can approach a 10% flow error at 30% RTP even though the AP error has not changed. To assure a reactor trip below the power range high SAL, the power range neutron flux- high trip setpoint is first set at < 85% RTP prior to adjusting the power range channel output in the decreasing power direction whenever the calorimetric power is > 20% RTP and <40% RTP. To assure a reactor trip below the power range high SAL, the power range neutron flux- high trip setpoint is fir% set a < 70% RTP prior to adjusting the power range channel output in the decreasing power direction whenever the calorimetric power is > 15% RTP and <20% RTR Adjustments in the increasing power direction do not require a prior decrease in the trip setpoint. Following a plant shutdown, it is permissible to reduce the power range neutron flux -- high trip setpoint prior to startup. This would anticipate the potential need for a decreasing power direction adjustment, thereby obviating the need to suspend power escalation for the purpose of first reducing the trip setpoint. Before the power range neutron flux high trip setpoint is re-set to its nominal full power value (< 109% RTP),

the power range channel calibration must be confirmed based on a calorimetric performed at > 40% RTP.

The Note to SR 3.3.1.2 clarifies that this Surveillance is required only if the reactor power is >15% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are allowed for performing the first Surveillance after reaching 15% RTP. A power level of 15% RTP is chosen based on plant stability, i.e., automatic rod control capability (manual rod control is normally used at Callaway) and the turbine generator synchronized to the grid. The 24-hour allowance after increasing THERMAL POWER above 15% RTP provides a reasonable time to attain a scheduled power plateau, establish the requisite conditions, perform the required calorimetric measurement, and make any required adjustments in a controlled, orderly manner and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE for subsequent use.

"TheFrequency of every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is adequate. It is based on unit operating experience, considering instrument reliability and operating history data for instrument drift. Together these factors demonstrate that a difference between the calorimetric heat balance calculation and the (continued)

CALLAWAY PLANT B 3.3.1-47 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.2 (continued)

REQUIREMENTS power range channel output of more than +2% RTP is not expected in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period.

In addition, control room operators periodically monitor redundant indications and alarms to detect deviations in channel outputs.

SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output every 31 EFPD. Ifthe absolute difference is> 2%, the NIS channel is still OPERABLE, but must be readjusted. The excore NIS channel shall be adjusted if the absolute difference between the incore and excore AFD is

> 2%. The purpose of the comparison is to check for differences that result from core power distribution changes that may have occurred since the last required adjustment or incore-excore calibration (SR 3.3.1.6).

Ifthe NIS channel cannot be properly readjusted, the channel is declared inoperable. This Surveillance is performed to verify the f(AI) input to the Overtemperature AT Function.

The Note to SR 3.3.1.3 clarifies that the Surveillance is required only if reactor power is > 50% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillance after reaching 50% RTP. This Note allows power ascensions and associated testing to be conducted in a controlled and orderly manner, at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE for subsequent use. Due to such effects as shadowing from the relatively deep control rod insertion and, to a lesser extent, the axially-dependent radial leakage which varies with power level, the relationship between the incore and excore indications of axial flux difference (AFD) at lower power levels is variable. Thus, it is acceptable to defer the calibration of the excore AFD against the incore AFD until more stable conditions are attained (i.e., withdrawn control rods and a higher power level). The AFD is used as an input to the Overtemperature AT reactor trip function and for assessing compliance with LCO 3.2.3, "AXIAL FLUX DIFFERENCE."

Due to the DNB benefits gained by administratively restricting the power level to 50% RTP, no limits on AFD are imposed below 50% RTP by LCO 3.2.3; thus, the proposed change is consistent with the LCO3.2.3 requirements below 50% RTP. Similarly, sufficient DNB margins are realized through operation below 50% RTP that the intended function of the Overtemperature AT reactor trip function is maintained, even though the excore AFD indication may not exactly match the incore AFD (continued)

CALLAWAY PLANT B 3.3.1-48 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.3 (continued)

REQUIREMENTS indication. Based on plant operating experience, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time frame to limit operation above 50% RTP while completing the procedural steps associated with the surveillance in an orderly manner.

The Frequency of every 31 EFPD is adequate. It is based on unit operating experience, considering instrument reliability and operating history data for instrument drift. Also, the slow changes in neutron flux during the fuel cycle can be detected during this interval.

SR 3.3.1.4 SR 3.3.1.4 is the performance of a TADOT every 31 days on a STAGGERED TEST BASIS. This test shall verify OPERABILITY by actuation of the end devices. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable TADOT of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

The RTB test shall include separate verification of the undervoltage and shunt trip mechanisms. Independent verification of RTBundervoltage and shunt trip Function is not required for the bypass breakers. No capability is provided for performing such a test at power. The independent test for bypass breakers is included in SR 3.3.1.14. The bypass breaker test sha!l include a local manual shunt trip only. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service.

The Frequency of every 31 days on a STAGGERED TEST BASIS is adequate. It is based on industry operating experience, considering instrument reliabi!ity and operating history data.

SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The SSPS is tested every 31 days on a STAGGERED TEST BASIS, using the semiautomatic tester. The train being tested is placed in the bypassed condition, thus preventing inadvertent actuation. Through the semiautomatic tester, all possible logic combinations, with and without (continued)

CALLAWAY PLANT B 3.3.1-49 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.5 (continued)

REQUIREMENTS applicable permissives, are tested for each protection function, including operation of the P-7 permissive which is a logic function only. The Frequency of every 31 days on a STAGGERED TEST BASIS is adequate. It is based on industry operating experience, considering instrument reliability and operating history data.

SR 3.3.1.6 SR 3.3.1.6 is a calibration of the excore channels to the incore channels.

If the measurements do not agree, the excore channels are not declared inoperable but must be calibrated to agree with the incore detector measurements. If the excore channels cannot be adjusted, the channels are declared inoperable. This Surveillance is performed to verify thef(Al) input to the Overtemperature AT Function. Determination of the loop specific vessel AT and Tavg values should be made when performing this calibration, under steady state conditions (ATo and T' [T" for Overpower AT] when at 100% RTP).

A Note modifies SR 3.3.1.6. The Note states thatthis Surveillance is required only if reactor power is_> 75% RTP and that 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after achieving equilibrium conditions with THERMAL POWER> 75% RTP is allowed for performing the first surveillance. Equilibrium conditions are achieved when the core is sufficiently stable at intended operating conditions to perform flux mapping.

The SR is deferred unttil a scheduled testing plateau above 75% RTP is attained during a power ascension. During a typical power ascension, it is usually necessary to control the axial flux difference at lower power levels through control rod insertion. After equilibrium conditions are achieved at the specified power plateau, a flux map must be taken and the required data collected. The data is typically analyzed and the appropriate excore calibrations completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after achieving equilibrium conditions. An additional time allowance of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided during which the effects of equipment failures may be remedied and any required re-testing may be performed.

The allowance of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after equilibrium conditions are attained at the testing plateau provides sufficient time to allow power ascensions and associated testing to be conducted in a controlled and orderly manner at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE for subsequent use.

(continued)

CALLAWAY PLANT B 3.3.1-50 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.6 (continued)

REQUIREMENTS The Frequency of 92 EFPD is adequate. It is based on industry operating experience, considering instrument reliability and operating history data for instrument drift.

SR 3.3.1.7 SR 3.3.1.7 is the performance of a COT every 92 days.

A COT is performed on each required channel to ensure the channel will perform the intended Function. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL OPERATIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

Setpoints must be within the Allowable Values specified in Table 3.3.1-1.

SR 3.3.1.7 is modified by two Notes. Note 1 provides a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months delay in the requirement to perform this Surveillance for source range instrumentation when entering MODE 3 from MODE 2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short time in MODE 3 until the Applicability is exited and SR 3.3.1.7 is no longer required to be performed. If the unit is to be in MODE 3 with the Rod Control System capable of rod withdrawal of one or more rods not fully inserted for > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , this Surveillance must be performed prior to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into MODE 3. Note 2 requires that the quarterly COT for the source range instrumentation shall include verification by observation of the associated permissive annunciator window that the P-6 and P-1 0 interlocks are in their required state for the existing unit conditions.

The Frequency of 92 days is justified in Reference 5.

SR 3.3.1.8 SR 3.3.1.8 is the performance of a COT as described in SR3.3.1.7 and it is modified by the same Note that this test shall include verification that the P-6 and P-1 0 interlocks are in their required state for the existing unit conditions by observation of the associated permissive annunciator' (continued)

CALLAWAY PLANT B 3.3.1-51 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.8 (continued)

REQUIREMENTS window. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL OPERATIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The Frequency is modified by a Note that allows this surveillance to be satisfied if it has been performed within 92 days of the Frequencies prior to reactor startup, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-10, and four hours after reducing power below P-6, as discussed below. The Frequency of "prior to reactor startup" ensures this surveillance is performed prior to critical operations and applies to the source, intermediate and power range low instrument channels. The Frequency of "12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-10" (applicable to intermediate and power range low channels) and "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power below P-6" (applicable to source range channels) allows a normal shutdown to be completed and the unit removed from the MODE of Applicability for this surveillance without a delay to perform the testing required by this surveillance. The Frequency of every 92 days thereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-10, and four hours after reducing power below P-6. The MODE of Applicability for this surveillance is < P-1 0 for the power range low and intermediate range channels and < P-6 for the source range channels. Once the unit is in MODE 3, this surveillance is no longer required. If power is to be maintained < P-10 for more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or

< P-6 for more than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , then the testing required by this surveillance must be performed prior to the expiration of the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limit, as applicable. These time limits are reasonable, based on operating experience, to complete the required testing or place the unit in a MODE where this surveillance is no longer required. This test ensures that the NIS source, intermediate, and power range low channels are OPERABLE prior to taking the reactor critical and after reducing power into the applicable MODE (< P-1 0 or < P-6) for the periods discussed above.

SR 3.3.1.9 SR 3.3.1.9 is the performance of a TADOT and is performed every 92 days, as justified in Reference 5. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable TADOT of a re!ay. This is acceptable because all of the other required contacts of the relay are verified by other Technical (continued)

CALLAWAY PLANT B 3.3.1-52 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.9 (continued)

REQUIREMENTS Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

This SR is modified by a Note that excludes verification of setpoints from the TADOT. Setpoint verification is accomplished during the CHANNEL CALIBRATION.

SR 3.3.1.10 A CHANNEL CALIBRATION is performed every 18 months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the Channel responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the setpoint methodology.

The Frequency of 18 months is based on the assumed calibration interval in the determination of the magnitude of equipment drift in the setpoint methodology SR 3.3.1 10 is mrodified by a Note stating that this test shall include verification that the time constants are adjusted to the prescribed values where applicable. This does not include verification of time delay relays.

These. are verified via response time testing per SR 3.3.1.16. See the discussion Of AT0 'n the Applicable Safety Analyses for the Overtemperatkre LT arid Overpower AT trip functions. Whenever an RTD is replaced in Function 6, 7, or 14.c, the next required CHANNEL CALIBRATION of the RTDs is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

The CHANNEL CALIBRATION of Function 6, Overtemperature AT, includes the axial flux difference penalty circuitry in the 7300 Process Protection System cabinets, but does not include the power range neutron detectors. SR 3.3.1.11 and its Notes 1 and 3 govern the performance and timing of the power range neutron detector plateau voltage verification.

Although not required for any safety function, the CHANNEL CALIBRATION of Function 10, Reactor Coolant Flow-Low, will ensure proper performance and normalization of the RCS flow indicators.

(continued)

CALLAWAY PLANT B 3.3.1-53 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.11 REQUIREMENTS (continued) SR 3.3.1.11 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.10, every 18 months. This SR is modified by three Notes. Note 1 states that neutron detectors are excluded from the CHANNEL CALIBRATION. Neutron detectors are excluded from the CHANNEL CALIBRATION because it is impractical to set up a test that demonstrates and adjusts neutron detector response to known values of the parameter (neutron flux) that the channel monitors. Note 1 applies to the source range proportional counters, intermediate range ion chambers, and power range ion chambers in the Nuclear Instrumentation System (NIS). Note 2 states that this test shall include verification that the time constants are adjusted to the prescribed values where applicable.

Detector plateau curves are obtained, evaluated, and compared to manufacturer's data for the intermediate and power range neutron detectors. The testing of the source range neutron detectors consists of obtaining integral bias curves, evaluating those curves, and comparing the curves to previous data. Note 3 states that the power and intermediate range detector plateau voltage verification is not required to be current until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after achieving equilibrium conditions with THERMAL POWER > 95% RTP. Equilibrium conditions are achieved when the core is sufficiently stable at intended operating conditions to perform a meaningful detector plateau voltage verification. The allowance of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after equilibrium conditions are attained at the testing plateau provides sufficient time to allow power ascension testing to be conducted in a controlled and orderly manner at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE fcr subsequent use. The source range integral bias curves are obtained under the conditions that apply during a plant outage.

The 18 month Frequency is based on past operating experience, which has shown these components usually pass the Surveillance when performed on the 18 month Frequency. The conditions for obtaining the source range integral bias curves and the power and intermediate range detector plateau voltages are described above. The other remaining portions of the CHANNEL CALIBRATIONS may be performed either during a plant outage or during plant operation.

SR 3.3.1.12 Not used.

(continued)

CALLAWAY PLANT B 3.3.1-54 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.13 REQUIREMENTS (continued) SR 3.3.1.13 is the performance of a COT of RTS interlocks every 18 months. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL OPERATIONAL TEST of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.

The Frequency is based on the known reliability of the interlocks and the multichannel redundancy available, and has been shown to be acceptable through operating experience.

SR 3.3.1.14 SR 3.3.1.14 is the performance of a TADOT of the Manual Reactor Trip, the SI Input from ESFAS, and the Reactor Trip Bypass Breaker undervoltage trip mechanisms. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable TADOT of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. This TADOT is performed every 18 months.

The Manual Reactor Trip TADOT shalh independently verify the OPERABILITY of the undervoltage and shunt trip handswitch contacts for both the Reactor Trip Breakers and Reactor Trip Bypass Breakers. The Reactor Trip Bypass Breaker test shall include testing of the automatic undervoltage trip mechanism.

The Frequency is based on the known reliability of the Functions and the multichannel redundancy available, and has been shown to be acceptable through operating experience.

The SR is modified by a Note that excludes verification of setpoints from the TADOT. The Functions affected have no setpoints associated with them.

(continued)

CALLAWAY PLANT B 3.3.1-55 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.15 REQUIREMENTS (continued) SR 3.3.1.15 is the performance of a TADOT of Turbine Trip Functions. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable TADOT of a relay. This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. This TADOT is performed prior to exceeding the P-9 interlock whenever the unit has been in MODE 3. This Surveillance is not required if it has been performed within the previous 31 days. Verification of the Trip Setpoint does not have to be performed for this Surveillance.

Performance of this test will ensure that the turbine trip Function is OPERABLE prior to exceeding the P-9 interlock.

SR 3.3.1,16 SR 3.3.1.16 verifies that the individual channel actuation response times are less than or equal to the maximum values assumed in the accident analysis. Response time verification acceptance criteria are included in Reference 8. No credit was taken in the safety analyses for those channels with response times listed as N.A. No response time testing requirements apply where N.A. is listed in Reference 8. Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the trip setpoint value at the sensor until loss of stationary gripper coil voltage (at which point the rods are free to fall).

The safety analyses include the sum of the following response time components:

(a) Process delay times (e.g., scoop transport delay and thermal lag associated with the narrow range RCS RTDs used in the OTAT, OPAT, and SG low-low Vessel AT (Power-I, Power-2) functions) which are not testable; (b) Sensing circuitry delay time from the time the trip setpoint is reached at the sensor until a reactor trip is generated by the SSPS; (continued)

CALLAWAY PLANT B 3.3.1-56 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.16 (continued)

REQUIREMENTS (c) Any intentional time delay set into the trip circuitry (e.g.,

undervoltage relay time delay, NLL cards (lag, lead/lag, rate/lag) and NPL cards (PROM logic cards for trip time delay) associated with the OTAT, OPAT, and SG low-low level Vessel AT (Power-I, Power-2) trip functions, and NLL cards (lead/lag) associated with the low pressurizer pressure reactor trip function) to add margin or prevent spurious trip signals; (d) For the undervoltage RCP trip function, back EMF delay from the time of the loss of the bus voltage until the back EMF voltage generated by the bus loads has decayed; (e) The time delay for the reactor trip breakers to open; and (f) The time delay for the control rod drive stationary gripper coil voltage to decay and the RCCA grippers to mechanically release making the rods free to fall (i.e., gripper release time measured during the performance of SR 3.1.4.3).

For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time verification is performed with the time constants set at their nominal values. Time constants are verified during the performance of SR 3.3.1.10. The response time may be verified by a series of overlapping tests, or other verification (e.g., Ref. 9 and Ref. 15),

such that the entire response time is verified.

Response time may be verified by actual response time tests in any series of sequenrtial, overlapping, or total channel measurements, or by the summation of allocated sensor, signal processing, and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from:

1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests); (2) inplace, onsite, or offsite (e.g. vendor) test measurements; or (3) utilizing vendor engineering specifications.

WCAP-13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.

WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," provides the basis and methodology for (continued)

CALLAWAY PLANT B 3.3.1-57 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.16 (continued)

REQUIREMENTS using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time.

The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in References 9 and 15 may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.

As appropriate, each channel's response time must be verified every 18 months on a STAGGERED TEST BASIS. Each verification shall include at least one train such that both trains are verified at least once per 36 months. Testing of the final actuation devices (i.e., reactor trip breakers) is included in the verification. Testing of the final actuation devices measures the time delay for the reactor trip breakers to open.

The time delay for the control rod drive stationary gripper coil voltage to decay and the RCCA grippers to mechanically release making the rods free to fall (i.e., gripper release time) is measured during the performance of SR 3.1.4.3 which verifies rod drop time from the beginning of decay of stationary gripper coil voltage. For surveillance testing performance, gripper release time is not included in the reactor trip system instrumentation response time testing due to the difficulty in determining the precise point at which the rods are free to fall. SR 3.1.4.3 specifies a readily qjanttfiable time to use as a separation point for field measurements, i.e., "from the beginning of decay of stationary gripper coil voltage." The rod drop time measurement in SR 3.1.4.3 begins at the time the rod control power cabinet regulator board circuit for a specific rod group is grounded, cidseng the board to reduce the stationary gripper coil current to zero releasing the rod group. This is essentially the same time at which the reactor trip breaker's opening would interrupt current to the stationary gripper coil. The response time definition, "until loss of stationary gripper coil voltage, " is less quantifiable. However, the definition's provision for overlapping testing allows this testing approach since the total response time is determined. The safety analyses are satisfied as long as both surveillances, response time and rod drop time, are met. Some portions of the response time testing cannot be performed during unit operation because equipment operation is required to measure response times. Experience has shown that these components usually pass this Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

CALLAWAY PLANT B 3.3.1-58 Revision 2g

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.16 (continued)

REQUIREMENTS SR 3.3.1.16 is modified by a Note stating that neutron detectors are excluded from RTS RESPONSE TIME testing. This Note is necessary because of the difficulty in generating an appropriate detector input signal.

Excluding the detectors is acceptable because the principles of detector operation ensure a virtually instantaneous response. Response time of the neutron flux signal portion of the channel shall be verified from detector output or input to the first electronic component in the channel.

REFERENCES 1. FSAR, Chapter 7.

2. FSAR, Chapter 15.

3. IELeE-279-1971.

4. 10 CFR 50.49.

5. Callaway OL Amendment No. 17 dated September 8, 1986.

6. Callaway Setpoint Methodology Report, SNP (UE)-565 dated May 1, 1984.

7. Callaway OL Arnendment No. 43 dated April 14, 1989.

8. FSAR Section 16.3, Table 16.3-1.

9. WCA1AF.-13632-.P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996.

10. FSAR Table 15.0-4.

11. WCAP-9226, "Reactor Core Response to Excessive Secondary Steam Releases," Revision 1, January 1978.

12. NRC Generic Letter 85-09 dated May 23, 1985.

13. FSAR Section 15.1.1.

14. RFR - 18637A.

15. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," October 1998.

16. FSAR Section 15.4.6.

CALLAWAY PLANT B 3.3.1-59 Revision 2g

RTS Instrumentation B 3.3.1 Table B 3.3.1-1 (Page 1 of 3)

FUNCTION NOMINAL TRIP SETPOINT (a)

1. Manual Reactor Trip N.A.

2. Power Range Neutron Flux

a. High < 109% RTP

b. Low < 25% RTP

3. Power Range Neutron Flux Rate < 4% RTP with time High Positive Rate constant > 2 sec.

4. Intermediate Range Neutron Flux < 25% RTP

5. Source Range Neutron Flux < 1.0E5 CPS

6. Overtemperature AT See Table 3.3.1-1 Note 1.

7. Overpower AT See Table 3.3.1-1 Note 2.

8. Pressurizer Pressure

a. Low > 1885 psig

b. High < 2385 psig

9. Pressurizer Water Level - High < 92% of instrument span

> 90% of loop minimum

10. Reactor Coolant Flow - Low measured flow (MMF=95,660 gpm)

(continued)

(a) The inequality sign only indicates conservative direction. The as-left value will be within a two-sided calibration tolerance band on either side of the nominal value. This also applies to the Overtemperature AT and Overpower AT Kvalues per Reference 14.

CALLAWAY PLANT B 3.3.1-60 Revision 2g

RTS Instrumentation B 3.3.1 Table B 3.3.1-1 (Page 2 of 3)

FUNCTION NOMINAL TRIP SETPOINT (a)

11. Not used.

12. Undervoltage RCPs >_ 10,584 Vac

13. Underfrequency RCPs > 57.2 Hz

14. Steam Generator (SG) Water Level Low-Low

a. Steam Generator Water Level Low-Low (Adverse

>_ 27.0% of narrow range instrument span I

Containment Environment)

b. Steam Generator Water Level >_ 21.6% of narrow range Low-Low (Normal Containment instrument span Environment)

c. Vessel AT Equivalent including delay timers - Trip Time Delay (1) Vessel AT (Power-I) < Vessel AT Equivalent to 12.41% RTP (with a time delay

< 232 sec.)

(2) Vessel AT (Power- 2) < Vessel AT Equivalent to 22.41% RTP (with a time delay

<122 sec.)

d. Containment Pressure _ 1.5 psig Environmental Allowance Modifier

15. Not used.

(continued)

(a) The inequality sign only indicates conservative direction. The as-left value will be within a two-sided calibration tolerance band on either side of the nominal value. This also applies to the Overtemperature AT and Overpower AT Kvalues per Reference 14.

CALLAWAY PLANT B 3.3.1-61 Revision 2g

RTS Instrumentation B 3.3.1 Table B 3.3.1-1 (Page 3 of 3)

FUNCTION NOMINAL TRIP SETPOINT (a)

16. Turbine Trip

a. Low Fluid Oil Pressure > 598.94 psig

b. Turbine Stop Valve Closure > 1% open

17. Safety Injection (SI) Input from N.A.

Engineered Safety Feature Actuation System (ESFAS)

18. Reactor Trip System Interlocks

a. Intermediate Range Neutron > 1.OE-10 amps Flux, P-6

b. Low Power Reactor Trips Block, N.A.

P-7

c. Power Range Neutron Flux, P-8 < 48% RTP

d. Power Range Neutron Flux, P-9 < 50% RTP

e. Power Range Neutron Flux, 10% RTP P-10

f. Turbine Impulse Pressure, P-13 < 10% Turbine Power

19. Reactor Trip Breakers N.A.

20. Reactor Trip Breaker Undervoltage and N.A.

Shunt Trip Mechanisms

21. Automatic Trip Logic N.A.

(a) The inequality sign only indicates conservative direction. The as-left value will be within a two-sided calibration tolerance band on either side of the nominal value. This also applies to the Overtemperature AT and Overpower AT Kvalues per Reference 14.

CALLAWAY PLANT B 3.3.1-62 Revision 2g

ESFAS Instrumentation B 3.3.2 B 3.3 INSTRUMENTATION B 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation BASES BACKGROUND The ESFAS initiates necessary safety systems, based on the values of selected unit parameters, to protect against violating core design limits and the Reactor Coolant System (RCS) pressure boundary, and to mitigate accidents.

The ESFAS instrumentation is segmented into three distinct but interconnected modules as identified below:

" Field transmitters or process sensors and instrumentation: provide a measurable electronic signal based on the physical characteristics of the parameter being measured;

" Signal processing equipment including 7300 Process Protection System, field contacts, and protection channel sets: provide signal conditioning, bistable setpoint comparison, process algorithm actuation, compatible electrical signal output to protection system devices, and control board/control room/miscellaneous indications; and

" Solid State Protection System (SSPS) including input, logic, and output bays and Balance of Plant (BOP) ESFAS circuitry: initiate the proper unit shutdown or engineered safety feature (ESF) actuation in accordance with the defined logic and based on the bistable outputs from the signal process control and protection system.

Field Transmitters or Sensors To meet the design demands for redundancy and reliability, more than one, and often as many as four, field transmitters or sensors are used to measure unit parameters. In many cases, field transmitters or sensors that input to the ESFAS are shared with the Reactor Trip System (RTS).

In some cases, the same channels also provide control system inputs. To account for calibration tolerances and instrument drift, which are assumed to occur between calibrations, statistical allowances are provided in the Trip Setpoint and Allowable Values. The OPERABILITY of each transmitter or sensor can be evaluated when its "as found" calibration data are compared against its documented acceptance criteria.

(continued)

CALLAWAY PLANT B 3.3.2-1 Revision 0

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