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#REDIRECT [[AEP-NRC-2015-46, License Amendment Request to Adopt TSTF-425-A, Rev. 3, Relocate Surveillance Frequencies to Licensee Control - Risk Informed Technical Specification Task Force Initiative 5B]]
| number = ML15328A452
| issue date = 11/19/2015
| title = Enclosures 6 - 11: CNP Units 1 & 2 TS Bases, TSTF-425 Versus CNP TS Cross-Reference, Proposed No Significance Hazards Consideration, Proposed Inserts and Regulatory Commitments
| author name =
| author affiliation = Indiana Michigan Power Co
| addressee name =
| addressee affiliation = NRC/NRR
| docket = 05000315, 05000316
| license number =
| contact person =
| case reference number = AEP-NRC-2015-46
| package number = ML15328A469
| document type = License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Bases Change
| page count = 324
| project =
| stage = Other
}}
 
=Text=
{{#Wiki_filter:Enclosure 6 to AEP-NRC-2015-46 CNP Unit I TS Bases Pages Marked to Show Proposed Changes SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)
: b. Bank position;c. RCS average temperature;
: d. Fuel burnup based on gross thermal energy generation;
: e. Xenon concentration;
: f. Samarium concentration;
: g. Isothermal temperature coefficient (ITC); and h. Boron penalty (MODES 4 and 5 only).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. The boron penalty must be applied in MODES 4 and 5 since all reactor coolant pumps may be stopped in these MODES.This extra amount of boron ensures that minimum response times are met for the operator to diagnose and mitigate an inadvertent boron dilution event prior to loss of SDM.
.--
c,-e
+Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Chapter 14.3. UFSAR, Section 14.2.5.4. UFSAR, Section 14.1.5.5. 10OCFR 100.Cook Nuclear Plant Unit 1 B3115Rvso o B 3.1.1-5 Revision No. 0 Core Reactivity B 3.1.2*BASES ACTIONS (continued)
B.._1 If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. If the SDM for MODE 3 is not met, then the boration required by SR 3.1.1.1 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 unit systems.SURVEILLANCE REQUIREMENTS SR 3.1.2.1 Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations.
The comparison is made, considering that other core conditions are fixed or stable, including RCS boron concentration, control rod position, RCS average temperature, fuel burnup based on gross thermal energy generation, 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 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. -T-he*reli ~ ~
e
=- Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Chapter 14.Cook Nuclear Plant Unit 1B3125ReionN.0 B3.1.2-5 Revision No. 0 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued) 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._22 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 in which the LCO requirements are not applicable.
To achieve this status,.the unit must be brought to at least MODE 3 within 6 hours.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 unit systems.SURVEILLANCE REQUIREMENTS SR 3.1.4.14a ea I-nsert 2..................
j ..... .....SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2, tripping each control rod would result in radial or axial power tilts, or oscillations.
Exercising each individual control rod ete £-92eays-provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 8 steps will not cause radial or axial power tilts, or oscillations, to occur. 9£ M frq~e edeenn EABLI---f'efdT Between required performances of SR 3.1.4.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable the control rod(s) is considered to be OPERABLE.
At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken.
2.Cook Nuclear Plant Unit I B 3.1.4-7 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.1.4-7 Revision No. 0 Shutdown Bank Insertion Limits B 3.1.5 BASES SURVEILLANCE REQUIREMENTS SR 3.1.5.1 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..,inece=~tad--pG~
~ ead- eu -ba-1 dw ~ ste-t tnyr REFERENCES
: 1. UFSAR, Section 1.4.2.2. UFSAR, Section 1.4.5.3. UFSAR, Section 1.4.6.4. 10OCFR 50.46.5. U FSAR, Chapter 14.-Insert 2 Cook Nuclear Plant Unit 1 B3154Rvso o B 3.1.5-4 Revision No. 0 Control Bank Insertion Limits B 3.1.6 BASES SURVEILLANCE REQUIREMENTS SR 3.1.6.1 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. Verifying theI ECP calculation within 4 hours 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 J td-t aF ~- eJ --Lfher 1  --- Insert 2.
l,-eryiteroim to -ee  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 REFERENCES
: 1. UFSAR, Section 1.4.2.2. UFSAR, Section 1.4.5.3. UFSAR, Section 1.4.6.4. 10 CFR 50.46.5. UFSAR, Chapter 14.~=Insert 2 Cook Nuclear Plant Unit 1 B3165Rvso o B 3.1.6-5 Revision No. 1 PHYSICS TESTS Exceptions
-MODE 2 B 3.1..8 BASES ACTIONS (continued) 531 0 F could violate the assumptions for accidents analyzed in the safety analyses.D.1I If the Required Action and associated Completion Time of Condition C is not met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit 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 unit systems.SURVEILLANCE SR 3.1.8.1 REQUIREMENTS Verification that the RCS lowest loop Tavg is > 531 0 F will ensure that the unit is not operating in a condition that could invalidate the safety analyses.
Verification of the RCS temperature at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the initial conditions of the safety analyses are not violated.SR 3.1.8.2 Verification that the THERMAL POWER is < 5% RTP will ensure that the unit is not operating in a condition that could invalidate the safety analyses.
Ve t-ftinsert-SR 3.1.8.3 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects: a. RCS boron concentration;
: b. Bank position;c. RCS average temperature;
: d. Fuel burnup based on gross thermal energy generation;
: e. Xenon concentration; Cook Nuclear Plant Unit 1 B 3.1.8-6 Revision No. 0 Cook Nuclear Plant Unit 1 B3.1.8-6 Revision No. 0 PHYSICS TESTS Exceptions
-MODE 2 B 3.1.8 BASES SURVEILLANCE REQUIREMENTS (continued)
: f. Samarium concentration;
: g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH);h. Moderator Temperature Defect, when above the POAH; and i.Doppler Defect, when above the POAH.Using the ITC accounts for Doppler reactivity in this calculation when the reactor is subcritical or critical but below the POAH, and the fuel temperature will be changing at the same rate as the RCS.-tagen-elti-ati9-afid-ef-th~e--ewl~r-ebability---ef-afl-aeednt-tiig-wi{,hoitte-i~eq REFERENCES
: 1. 10 CFR 50, Appendix B, Section XI.2. 10 CFR 50.59.3. Regulatory Guide 1.68, Revision 2, August, 1978.4. ANSI/ANS-1 9.6.1-1 997, August 22, 1997.5. WCAP-1 3360-P-A, "Westinghouse Dynamic Rod Worth Measurement Technique," Revision .1, October 1998.6. PA-OSC-0061, "Westinghouse Position Paper on Power Distribution Measurement Requirements for Reload Startup Programs," February 2005.-Insert 2 Cook Nuclear Plant Unit 1 B3187Rvso o B3.1.8-7 Revision No. 1 FQ(Z)B 3.2.1 BASES ACTIONS (continued) 0.1 If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE or condition in which the LCO requirements are not applicable.
This is done by placing the unit in at least MODE 2 within 6 hours.This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.2.1.1 REQUIREMENTS Verification that FC(z) is within its specified limits involves increasing FM(z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain FC~(z) is then compared to its specified limits.If THERMAL POWER has been increased by > 10% RTP since the last determination of FC(z), another evaluation of this factor is required 24 hours after achieving equilibrium conditions at this higher power level (to ensure that values are being reduced sufficiently with power increase to stay within the LCO limits). The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification.
It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ(Z) was last measured.Th-rq
-Insert 2 a4 SR 3.2.1.1 is modified by a Note, which applies during power escalation after a refueling.
The Note states that the Surveillance is not required to be performed until 24 hours after equilibrium conditions at a power level for extended operation are achieved.
This Note allows the unit to startup from a refueling outage and reach the power level for extended operation (normally 100% RTP) prior to requiring performance of the SR. Within 24 hours after equilibrium conditions are reached at the power level for extended operation, the SR must be performed.
Cook Nuclear Plant Unit 1 B 3.2.1-6 Revision No. 0 Cook Nuclear Plant Unit 1 B3.2.1-6 Revision No. 0 FQ(Z)B 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last*verification.
It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which F 0 (Z) was last measured.T ~ Insert 2 tersl T-h- fe4 eneyef-F.=El44..akga ttee-~ i gt i&SR 3.2.1.2 is modified by Note 1, which applies during power escalation after a refueling.
The Note states that the Surveillance is not required to be performed until 24 hours after equilibrium conditions at a power level for extended operation are achieved.
This Note allows the unit to startup from a refueling outage and reach the power level for extended operation (normally 100% RTP) prior to requiring performance of the SR. Within 24 hours after equilibrium conditions are reached at the power level for extended operation, the SR must be performed.
REFERENCES
: 1. 10 CFR 50.46.2. UFSAR, Section 14.2.6.7.3. UFSAR, Section 1.4.5.4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.5. WCAP-1 0216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control (and) F 0 Surveillance Technical Specification," February 1994.Cook Nuclear Plant Unit 1 B3218Rvso o B 3.2.1-8 Revision No. 0 B 3.2.2 BASES ACTIONS (continued)
A.4 Verification that iS within its specified limits after an out of limit occurrence ensures that the cause that led to the FN~)H exceeding its limit is corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the FNAH limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours after THERMAL POWER is > 95% RTP.This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action.B.1 When any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCO requirements are not applicable.
This is done by placing the unit in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging unit systems..SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The value of FN~H iS determined by using the movable incore detector system to obtain a flux distribution map. A data reduction computer program then calculates the maximum value of FNAH from the measured flux distributions.
The measured value of FNH must be multiplied by 1.04 to account for measurement uncertainty before making comparisons to the limit.After each refueling, FNAH must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that FN~H limits are met at the beginning of each fuel cycle.Tf 31S EFFD FI uubu a
--Insert 2 REFERENCES
: 1. UFSAR, Section 14.2.6.7.2. UFSAR, Section 1.4.5.3. 10 CFR 50.46.Cook Nuclear Plant Unit 1B322-ReionN.5 B 3.2.2-5 Revision No. 35 AFD B 3.2.3 BASES SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD as indicated by the NIS excore channels is within the target band. -T-he--S r-ei14a~c-e-Fr-requeac-y-Gf-7--tays i- Insert 2 SR 3.2.3.2
,----Insert 2
ective-1Th power-days-(t -E-F4D)-t1e-ac-c uaift-er-saU
.SR 3.2.3.3 Measurement of the target flux difference is accomplished by taking a flux map when the core is at equilibrium xenon conditions, preferably at high power levels with the control banks nearly withdrawn.
This flux map provides the equilibrium xenon axial power distribution from which the target value can be determined.
The target flux difference varies slowly with core burnup.,A-F-eq~enlyof---E-pp-fterhreiiii g-an1E t er-e,{ter-fiif Insert 2 A Note modifies this SR to allow the predicted beginning of cycle AFD from the cycle nuclear design tO be used to determine the initial target flux difference after each refueling.
REF ERENCES 1. WCAP-8385 (Westinghouse proprietary) and WCAP-8403 (nonproprietary), "Power Distribution Control and Load Following Procedures," Westinghouse Electric Corporation, September 1974.2. UFSAR, Section 7.4.Cook Nuclear Plant Unit 1 B3236Rvso o B3.2.3-6 Revision No. 1 QPTR B 3.2.4 BASES ACTIONS (continued)
Action A.5). The intent of this Note is to have the peaking factor Surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limits and the core returned to power.B.1I If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE or other specified condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging unit systems.SURVEILLANCE SR 3.2.4.1.REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is< 75% RTP and the input from one Power Range Neutron Flux channel is inoperable.
Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1.This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. -T-he.-----
Insert 2 ,f astksit con ,te .nom~ .n lam.. "alal toteoeor in the contro &#xf7;roo...For those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is not required until 12 hours after the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is > 75% RTP.With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded.
Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased.- " " "
* YR-'--- Insert 2 Cook Nuclear Plant Unit 1 B 3.2.4-5 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.2.4-5 Revision No. 0 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.1 Performance of the CHANNEL OH ECK-en-ewe 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.
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.
t-4euftemens-lers-fer Insert 2 SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output e~ei3,-94,-hetw&-.
If the calorimetric exceeds the NIS channel output by > 2% RTP, the NIS is not declared inoperable, but must be adjusted.
If the NIS channel output ca~nnot be properly adjusted, the channel is declared inoperable.
Two Notes modify SR 3.3.1.2. The first Note indicates that the NIS channel output shall be adjusted consistent with the calorimetric results if the absolute difference between the NIS channel output and the calorimetric is > 2% RTP. The second Note clarifies that this Surveillance is required only if reactor power is > 15% RTP and that 12 hours is allowed for performing the first Surveillance after reaching 15% RTP. At lower power levels, calorimetric data are inaccurate.~ ~ Insert 2 h er'lif efcesdmts{-t-4' Cook Nuclear Plant Unit 1 B 3.3.1-38 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output.ev#e~y-If the absolute difference is > 3%, the NIS channel is still OPERABLE, but must be readjusted.
If the 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.Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be adjusted if the absolute difference between the incore and excore AFD is > 3%. Note 2 clarifies that the Surveillance is required only if reactor power is > 15% RTP and that 24 hours is allowed for performing the first Surveillance after reaching 15% RTP.ef tgee-ee~~e~-at-mrtrt~~~rtn l nsert 2 SR 3.3.1.4 SR 3.3.1.4 is the performance of a TADOT evtery!!di&#xa3;et-ays=a 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-Techqnical 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 RTB undervoltage 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 3.3.1.17.
The bypass breaker test shall include a local shunt trip. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service.Cook Nuclear Plant Unit 1 B3313 eiinN.1 B 3.3.1-39 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)4 Insert 2 jtted=a4f-efere~aee=14.
SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The SSPS is tested evr/9.lase using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
T~he-l4-ST-B-4Si-u4~d Insert 2.iH-R-efeireaee41=.
SR 3.3.1.6 SR 3.3.1.6 is the performance of a TADOT and is performed every 92 days on a STAGGERED TEST BASIS. This test applies to the SI Input from ESFAS 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 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.
STAS lEDT-STBA&&=s=Insert 2
SR 3.3.1.7 SR 3.3.1.7 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 the f(AI)input to the Overtemperature AT Function.A Note modifies SR 3.3.1.7. The Note states that this Surveillance is required only if reactor power is > 50% RTP and that 24 hours is allowed for performing the first surveillance after reaching 50% RTP.eX
~ ~ Insert 2 Cook Nuclear Plant Unit 1 B 3.3.1-40 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.8 SR 3.3.1.8 is the performance of a A COT is performed on each required channel to ensure the entire 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 COT 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.The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 8.SR 3.3.1.8 is modified by a Note that provides a 12 hour delay in the requirement to perform this Surveillance for Function 2.b channels after reducing THERMAL POWER below the P-I10 interlock.
The Frequency of 12 hours after reducing power below P-10 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.
SR 3.3.1.9 I=-nsert 2CHANN EL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
Cook Nuclear Plant Unit I1 ..-1ReiinN.1 B 3.3.1-41 Revision .No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)Insert 2 This SR is modified by a Note that states that neutron detectors are excluded from the CHANNEL CALIBRATION.
Changes in power range neutron detector sensitivity are compensated for by normalization of the channel output based on a power calorimetric and flux map performed above 15% RTP (SR 3.3.1.2).SR 3.3.1.10 SR 3.3.1.10 is the performance of a TADOT and is-p@FfGeFF
\ee~ey <=-- Insert 2 e2.2--ays.
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 SR is modified by a Note that excludes verification of relay Setpoints from the TADOT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION.
The Frequency of 92 days is justified in Reference 10.SR 3.3.1.11 SR 3.3.1.11 is the performance of a COT-every--1-84-4ays.
A COT is performed on each required channel to ensure the entire 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 COT 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 Sp~ecificatio~ns tests at least once per refueling interval with applicable extensions.
Setpoints must be within the Allowable Values specified in Table 3.3.1-1.The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
Cook Nuclear Plant Unit 1 ..-2ReiinN.1 B 3.3.1-42 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 8.The Frequency is modified by two Notes. Note 1 provides a 12 hour delay in the requirement to perform this Surveillance for intermediate range instrumentation after reducing THERMAL POWER below the P-10 interlock.
The Frequency of 12 hours after reducing power below P-10 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.
Note 2 provides a 4 hour delay in the requirement to perform this Surveillance for source range instrumentation after THERMAL POWER is reduced below the P-6 interlock.
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 RTBs are open and SR 3.3.1.11 is no longer required to be performed.
If the unit is to be in MODE 3 with the RTBs closed for > 4 hours this Surveillance must be performed prior to4 hours after THERMAL POWER is reduced below the P-6 interlock.Insert 2 SR 3.3.1.12.
N-ip y48m~e e44ay 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
:FeFetee~F4~y~4ae~mh~euftee--
I nse rt 2 dt~ft4the-tep SR 3.3.1.13CHANN EL 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.Cook Nuclear Plant Unit 1 B3314 eiinN.1 B 3.3.1-43 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.ent' 2 SR 3.3.1.14 SR 3.3.1.14 is the performance of a CHANNEL CALIBRATION eiCHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The CHANNEL CALIBRATION for the source range neutron detectors also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION.
Changes in power range neutron detector sensitivity are compensated for by normalization of the channel output based on a power calorimetric and flux map performed above 15% RTP (SR 3.3.1.2).Changes in intermediate range neutron flux detector sensitivity are compensated for by periodically evaluating the compensating voltage setting and making adjustments as neceSsary.
Changes in source range neutron detector sensitivity are compensated for by periodically obtaining the detector plateau or preamp discriminator curves, evaluating those curves, comparing thecurves to the manufacturer's data, and adjusting the channel output as necessary.e 4=-. Insert 2.SR 3.3.1.15 SR 3.3.1.15 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.13, e'~ert--=24--meonhs.
Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the resistance temperature detectors (RTD) sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.Cook Nuclear Plant Unit 1 B 3.3.1-44 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
~ nse rt 2 This SR is modified by a Note that provides a 72 hour delay in the requirement to perform a normalization of the AT channels after THERMAL POWER is > 98% RTP. .The intent of this Note is to maintain reactor power at a nominal 97% RTP to 98% RTP level until the AT normalization is complete before increasing reactor power to 100% RTP.SR 3.3.1.16 SR 3.3.1.16 is the performance of a COT of RTS interlocks 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 COT 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.
oThe-
& Insert 2
~ w4-aop SR 3.3.'1.17 SR 3.3.1.1.7 is the performance of a TADOT of the Manual Reactor Trip (including reactor trip bypass breakers) and RCP Breaker Position.
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.
4e~~rj.2 t s The test shall independently verify the OPERABILITY of the undervoltage and shunt trip mechanisms for the Manual Reactor Trip Funlction for the Reactor Trip Breakers and Reactor Trip Bypass Breakers.
The Reactor Trip Bypass Breaker test shall include testing of the automatic undervoltage trip.Cook Nuclear Plant Unit 1B331-5RvsoN.9 B 3.3.1-45 Revision No. 9 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Insert 2 SR 3.3.1.18 SR 3.3.1.18 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 as described in SR 3.3.1.4, except that this test is performed prior to exceeding the P-8 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-8 interlock.
SR 3.3.1.19 SR 3.3.1.19 verifies that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analysis.
Response time testing acceptance criteria are included in UFSAR, Table 7.2-6 (Ref. 12). 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 to the point at which the equipment reaches the required functional state (i.e., control and shutdown rods fully inserted in the reactor core).For channels that include dynamic transfer Functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer Function set to one, .with the resulting measured response time compared to the appropriate UFSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value, provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.Cook Nuclear Plant Unit 1 B3314 eiinN.1 B 3.3.1-46 Revision No. 17 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Response time may be verified by actual response time tests in any series of sequential, 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) in place, 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," (Ref. 13) 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, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 14) provides the basis and methodology for 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 the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
24mnh~raSTmESTB#t igetefn{ -Insert 2 SR 3.3.1.19 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.The response time testing of the neutron flux signal portion of the channel shall be measured from either the detector output or the input of the first electronic component in the channel.Cook Nuclear Plant Unit 1B33147RvsoN.5 B 3.3.1-47 Revision No. 5 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accurFacies.
SR 3.3.2.1 A CHAN NEL 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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.
tn-xeineta-m.t-~e Insert 2 SR 3.3.2.2 and SR 3.3.2.5 SR 3.3.2.2 is the performance of a This test is a check of the Loss of Voltage Function.
SR 3.3.2.5 is the performance of a TAOT test is a check of the Undervoltage RCP 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 TADOT of a relay. This is acceptable because all of the other required contacts of the relay are verified by otherTechnical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
Each SR is modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and~are verified during CHANNEL CALIBRATION.
Insert 2 Cook Nuclear Plant Unit 1 B 3.3.2-37 Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.3 SR 3.3.2.3 is the performance of an ACTUATION LOGIC TEST.=---he-SP,.S4 tao-tzd orc-c-r;T-02
.using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the master relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERABLE and that there is an intact voltage signal path to the master relay coils. -The-Feaey--ef-iei 2ye-e1.
* Insert 2 4, R~e n ee--4* .SR 3.3.2.4 SR 3.3.2.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. -T-his-es-j~er-feffl
=S=T ASt1. Thfe- Insert 2 SR 3.3.2.6 SR 3.3.2.6 is the performance of a COT. A COT is performed on each required channel to ensure the entire channel will perform the intended Function.
Setpoints must be found within the Allowable Values specified in Table 3.3.1-1. 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 COT 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 difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 6.
-,-- Insert 2 Cook Nuclear Plant Unit 1 .~-8ReiinN.5 B  Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.6 is modified by a Note which applies to the SI Containment Pressure -High, Containment Spray Containment Pressure -High High, Phase B Isolation Containment Pressure -High High, Steam Line Isolation Containment Pressure -High High, and CEQ System Containment Pressure -High Functions.
This Note requires, during the performance of SR 3.3.2.6, the associated transmitters of these Functions to be exercised by applying either a vacuum or pressure to the appropriate side of the transmitter.
Exercising the associated transmitters during the performance of the COT is necessary to ensure Functions 1 .c, 2.c, 3.b.(3), 4.c, and 7.c remain OPERABLE between each CHANNEL CALIBRATION.
SR 3.3.2.7 SR 3.3.2.7 is the performance of a CHANNEL CALIBRATION.-A--
CHAN NEL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.Y Insert 2
~~~f~f ~.
SR 3.3.2.8 SR 3.3.2.8 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contac~t operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function, or is placed in a condition where the relay contact operation can be verified without operation of the equipment.
Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.
t Insert 2.bseo~d .a~d-t~r~t~ingh~y-Eatet-a Cook Nuclear Plant Unit 1 B3323 eiinN.5 B 3.3.2-39 Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.9 SR 3.3.2.9 is the performance of a TADOT. This test is a check of the Manual Initiation Functions, the AFW pump start on trip of all MFW pumps, and the P-4 interlock. me~rm4-ve~=4Te~h 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.
In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.). "T-h~rtenwi~ey-deate, SR 3.3.2.10 SR 3.3.2.10 is the performance of a CHANNEL CALIBRATION.
**--~R-- ~s -v e- -CHAN NEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy.CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
4-Insert 2 eney=e 4-Insert 2 SR 3.3.2.11 SR 3.3.2.11 is the performance of an ACTUATION LOGIC TEST. This SR is applied to the balance of plant actuation logic and relays that do not have the SSPS test circuits installed to utilize the semiautomatic tester or perform the continuity check. All possible logic combinations are tested for Table 3.3.2-1 Functions 6.e and 6.g. Insert 2 SR 3.3.2.12 This SR ensures the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis.Response Time testing acceptance criteria are included in the UFSAR, Cook Nuclear Plant Unit 1 ..-0ReiinN.5 B 3.3.2-40 Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
Table 7.2-7 (Ref. 11). 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, to the point at which the equipment in both trains reaches the required functional state (e.g., pumps at rated discharge pressure, valves in full open or closed position).
For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer functions set to one with the resulting measured response time compared to the appropriate UFSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.Response time may be verified by actual response time tests in any series of sequential, 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) in place, 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," dated January 1996 (Ref. 12), 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 WOAP. Response time verification for other sensor types must be demonstrated by test.WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 13) provides the basis and methodology for 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 the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly Of a transmitter.
1E Y--mfts=n I nse'rt 2 8-of-hfi t p Cook Nuclear Plant Unit I1 ..-1ReiinN.5 B 3.3.2-41 Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) eha~tq4ra.cute ine cee -e-ta-4es i e with-e~aeh-
* hre -Tgrele~-eut4-~eie-vrfet a This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours after reaching 850 psig in the SGs.REFERENCES
: 1. Technical Requirements Manual.2. IEEE-279, "Proposed Criteria for Nuclear Power Plant Protection Systems," August 1968.3. UFSAR, Table 7.2-1.4. UFSAR, Table 14.1-2.5. 10 CFR 50.49.6. WCAP-12741,"Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.7. UFSAR, Chapter 14.8. WCAP-14333-P-A, Revision 1, October 1998.9. WCAP-1 0271-P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," including Supplement 1, May 1986, and Supplement 2, Rev. 1, June 1990.10. WCAP-1 5376, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Intervals and Reactor Trip Breaker Test and Completion Times," October 2000.11. UFSAR, Table 7.2-7.12. WCAP-1 3632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996.13. WCAP-1 4036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.Cook Nuclear Plant Unit 1 B 3.3.2-42 Revision No. 51 Cook Nuclear Plant Unit 1 B 3.3.2-42 Revision No. 51 PAM Instrumentation B 3.3.3 BASES ACTIONS (continued) justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.SURVEILLANCE As noted at the beginning of the SRs, the following SRs apply to each REQUIREMENTS PAM instrumentation Function in Table 3.3.3-1, except where identified in the SR.SR 3.3.3.1 Performance of the CHANN EL CHECKi ree-eveiy4-.ays=ensures that a gross instrumentation failure 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
The Containment Area Radiation (High Range)instrumentation should be compared to similar unit instruments located throughout the unit. When only one channel of the Reactor Coolant Inventory Tracking System is OPERABLE, the RCS Subcooling Margin Monitor and Core Exit Temperature channels may be used for performance of the CHANNEL CHECK of the OPERABLE Reactor Coolant Inventory Tracking System channel.Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, 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. If the channels are within the criteria, it is an indication that the channels are OPERABLE.As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.
T ~ lfc ~
demat~fatse atcHrie ef~lGH 4-ANN--TeLS-- K Insert 2 fantu~g-c-ha~t-Cook Nuclear Plant Unit 1 ..-3Rvso o B 3.3.3-13 Revision No. 0 PAM Instrumentation B 3.3.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.3.2 Deleted SR 3.3.3.3-A4--~IN -e~v B-24aeRt CHAN NEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy.
This SR is modified by a Note that excludes neutron detectors.
For Function 9, the*CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual valve position.
For Functions 15, 16, 17, and 18, whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit Temperature thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing elements.For Functions 20 (Circuit Breaker Status channels) and 24, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual circuit breaker position. Insert 2 REFERENCES
: 1. NRC letter, T. G. Colburn (NRC) to M. P. Alexich (Indiana Michigan-Power Company), "Emergency Response Capability
-Conformance to Regulatory Guide 1.97 Revision 3 for the Dl C. Cook Nuclear Plant, Units 1 and 2," dated December 14, 1990.2. UFSAR, Table 7.8-1.3. Regulatory Guide 1.97, Revision 3, May 1983.4. NUREG-0737, Supplement 1, "TMI Action Items." 5. NRC letter, P. S. Tam (NRC), to M. K. Nazar, (Indiana Michigan Power Company), "Donald C. Cook Nuclear Plant, Units 1 & 2 (DCCNP-1 AND DCCNP-2) -Issuance of Amendments Re: Containment Sump Modifications per Generic Letter 2004-02 (TAC Nos. MD5901 AND MD5902)," dated October 18, 2007.Cook Nuclear Plant Unit 1 ..-4ReiinN.2 B 3.3.3-14 Revision No. 29 Remote Shutdown Monitoring Instrumentation B 3.3.4 BASES ACTIONS (continued)
Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.A.1 Condition A addresses the situation where one or more required Functions of the remote shutdown monitoring instrumentation are inoperable.
The Required Action is to restore the required Function to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.B.1 and B.2 If the Required Action and associated Completion Time of Condition A is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.:3.4.1 REQUI REM ENTS Performance of the CHANN EL CHECK~eaeeeve y.4-dey~s~ensures that a gross failure of instrumentation has not occurred.
A CHANNEL CHECK is normally a comparison of the parameter indicated onone 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. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability.
If the channels are within the criteria, it is an indication that the channels are OPERABLE.
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.Cook Nuclear Plant Unit 1 B 3.3.4-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.3.4-3 Revision No. 0 Remote Shutdown Monitoring Instrumentation B 3.3.4 BASES SURVEILLANCE REQUIREMENTS (continued)
-As specified in the Surveillance, a CHANNEL CHECK is only required for those channels which are normally energized.
TheFqey=e~-8+/-~y nIen I=-nsert 2-HeN SR 3.3.4.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.For the Reactor Trip Breaker Indication Function on the hot shutdown panel, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual reactor trip breaker position.I li; ul L/-'t I In 31 Ill =is-erfe19-.~=Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.3.Cook Nuclear Plant Unit 1 ..- evso o B 3.3.4-4 Revision No. 0 LOP DG Start Instrumentation B 3.3.5 BASES ACTIONS (continued) made inoperable by failure of the LOP DG start instrumentation are required to be entered immediately.
The actions of those LCOs provide for adequate compensatory actions to assure unit safety.SURVEILLANCE REQUIREMENTS SR 3.3.5.1 Performance of the CHANNEL CHECK e~ee-ever-y=-l-2--heur13 ensures that a 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.
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.SR 3.3.5.2 I===nsert 2 SR 3.3.5.2 is the performance of a TADOT. 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.
"-Thi-hes-i 9efFe The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment.
The SRs are modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and are verified during CHANNEL CALIBRATION.
T1,t~ienc--y-is--Insert 2-aece-aele-. Cook Nuclear Plant Unit 1 B 3.3.5-5 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.3.5-5 Revision No. 0 LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.5.3 SR 3.3.5.3 is the performance of a CHANNEL CALIBRATION.
The setpoints, as well as the response to a loss of voltage and a degraded voltage test, shall include a single point verification that the trip occurs within the required time delay.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., .-.- r o, .. , r * .l .r. ..i ...i .-Insert 2 REFERENCES
: 1. UFSAR, Section 8.4.2. UFSAR, Section 8.5.3. UFSAR, Chapter 14.4. WCAP-1 2741, "Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.Cook Nuclear Plant Unit 1 B3356Rvso o B 3.3.5-6 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES ACTIONS (continued)
D..1 Condition D applies to all Containment Purge Supply and Exhaust System Isolation Functions.
If one or more Automatic Actuation Logic and Actuation Relays trains are inoperable, one or more SI Input from ESFAS trains are inoperable, two or more required radiation monitoring channels in a single train are inoperable, or the Required Action and associated Completion Time of Condition A, B, or C are not met, operation may continue provided the containment purge supply and exhaust isolation valves are placed in the closed position immediately.
Placing the containment purge supply and exhaust isolation valves in the closed position accomplishes the safety function of the inoperable trains or channels.SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1 REQUIREMENTS determines which SRs apply to which Containment Purge Supply and Exhaust System Isolation Instrumentation Functions.
SR 3.3.6.1 Performance of the CHANN EL CH ECi Keeeever-y--t-2--heur-s ensures that a 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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.
etaaa I=-nsert 2
* Cook Nuclear Plant Unit 1 B 3.3.6-6 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.3.6-6 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.6.2 SR 3.3.6.2 is the performance of an ACTUATION LOGIC TEST. The train being tested may be placed in the bypass condition, thus preventing actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, may be tested for each protection function.
In addition, the master relay coil may be pulse tested for continuity.
This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. -This
-'-Insert 2 SR 3.3.6.3 SR 3.3.6.3 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity.
Tise--ht~et~5
~ -Isr SR 3.3.6.4 A COT is performed on each required channel to ensure the entire 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 COT 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.
Tl-I"e-l-eqlencjs-15..
Insert 2 bitbaset This test verifies the capability of the instrumentation to provide the Containment Purge Supply and Exhaust System isolation.
The setpoint shall be left consistent with the current unit specific calibration procedure tolerance.
SR 3.3.6.5 SR 3.3.6.5 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation mode is either allowed to function or is placed in a condition where the relay contact operation can be verified without Cook Nuclear Plant Unit 1 B3367Rvso o B 3.3.6-7 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued) operation of the equipment.
Actuation equipment that may not be operated in the design mitigation mode is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.T-he--Fr-eq
'=14aeeepta4b1e Insert 2 SR 3.3.6.6 SR 3.3.6.6 is the performance of a TADOT. This test is a check of the Manual Initiation Function ipeerfer ea--evey4--mefat-h.
Each Manual Initiation Function is tested up to, and including, the master relay coils. 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.
In some instances, the test includes actuation of the end device (i.e., valves cycle).The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no setpoints associated with it.ep 4-Insert 2 SR 3.3.6.7 3-&deg;~~ 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.T 2 REFERENCES
: 1. UFSAR, Section 5.5.3.2. 10OCFR 100.11 3. WCAP-1 5376, Rev. 0, October 2000.Cook Nuclear Plant Unit 1I ..- evso o B 3.3.6-8 Revision No. 0 CREV System Actuation Instrumentation B 3.3.7 BASES ACTIONS (continued) this Completion Time is the same as provided in LCO 3.7.10.B.1.1, B.1.2, and B.2 Condition B applies to the failure of two CREV System Automatic Actuation Logic and Actuation Relays trains in one or more required Functions.
The first Required Action is to place one CREV train in the pressurization/cleanup mode of operation immediately.
This accomplishes the actuation instrumentation Function that may have been lost and places the unit in a conservative mode of operation.
The applicable Conditions and Required Actions of LCO 3.7.10 must also be entered for the CREV train made inoperable by the inoperable actuation instrumentation.
This ensures appropriate limits are placed upon train inoperability as discussed in the Bases for LCO 3.7.10.Alternatively, both trains may be placed in the pressurization/cleanup mode. This ensures the CREV System function is performed even in the presence of a single failure.'C.1 and C.2 Condition C applies when the Required Action and associated Completion Time for Condition A or B have not been met. The unit must be brought to a MODE in which the LCO requirements are not applicable.
To achieve this status, the unit must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS A Note has been added to the SR Table to clarify that Table 3.3.7-1 determines which SRs apply to which CREV System Actuation Instrumentation Functions.
SR 3.3.7.1 SR 3.3.7.1 is the performance of an ACTUATION LOGIC TEST. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the mater relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. -F-is-Insert 2 Cook Nuclear Plant Unit 1 B 3.3.7-3 Revision No. 0 Cook Nuclear Plant Unit I B3.3.7-3 Revision No. 0 CREV System Actuation Instrumentation B 3.3.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.7.2 SR 3.3.7.2 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. Insert 2 SR 3.3.7.3 SR 3.3.7.3 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is-either allowed to function or is placed in a condition where the relay contact operation can be verified without operation of the equipment.
Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.
Insert 2 REFERENCES
: 1. WCAP-1 5376, Rev. 0, October 2000.Cook Nuclear Plant Unit 1 B3374Rvso o B3.3.7-4 Revision No. 0 BOMI B 3.3.8 BASES ACTIONS (continued)
As an alternate to restoring one channel to OPERABLE status within 1 hour (Required Action B.2.1). Required Action B.2.2.1 requires isolation valves for unborated water sources to the Chemical and Volume Control System to be secured to prevent the flow of unborated water into the RCS. In addition, in MODE 5, if the RWST boron concentration is< 2400 ppm and less than the Reactor Coolant System (RCS) boron concentration, the RWST is considered an unborated water source and is required to be isolated from the RCS. Once it is recognized that two source range neutron flux monitoring channels of the BDMI are inoperable, the operators will be aware of the possibility of a boron dilution, and the 1 hour Completion Time is adequate to complete the requirements of Required Action B.2.2. 1.Required Action B.2.2.2 accompanies Required Action B.2.2.1 to verify the SDM according to SR 3.1.1.1 within 1 hour and once per 12 hours thereafter.
This backup action is intended to confirm that no unintended boron dilution has occurred while the BDMI was inoperable, and that the required SDM has been maintained.
The specified Completion Time takes into consideration sufficient time for the initial determination of SDM and other information available in the control room related to SDM.SURVEILLANCE SR 3.3.8.1 REQUIREMENTS Performance of the CHAN NEL CHECK 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 channelsrmonitoring 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.
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.
te e~{ .= Insert 2 eed ~ge~tee~~~ts ihe~lsptays~st~iedw it'ht"he~fl_-redt~-eb'4anies.
Cook Nuclear Plant Unit 1 ..- evso o B 3.3.8-3 Revision No. 0 BDMI B 3.3.8 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.8.2 SR 3.3.8.2 is the performance of a CHANNEL CALIBRATION eve~y-.-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.
The CHANNEL CALIBRATION also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This SR is modified by a Note that states that neutron detectors are excluded from the CHANNEL CALIBRATION.-I nsert 2 REFERENCES
: 1. UFSAR, Section 14.1.5.Cook Nuclear Plant Unit 1 ..- evso o B 3.3.8-4 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued)
POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive.
Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.
ACTIONS A.1 With one or more of the RCS DNB parameters not within LCO limits, action must be taken to restore parameter(s) in order to restore DNB margin and eliminate the potential for violation of the accident analysis.The 2 hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.
B.1 If Required Action A.1 is not met within the associated Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 2 within 6 hours. In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis.
The Completion Time of 6 hours is reasonable to reach the required unit conditions in an orderly manner.SURVEILLANCE REQU IREM ENTS SR 3.4.1.1 r.eeeu 4& e
.--la 2 SR 3.4.1.2--
et4e8
.-Insert 2--F...........
j j Cook Nuclear Plant Unit 1 B3413Rvso o B 3.4.1-3 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.1.3 Verification that the RCS total flow rate is greater than or equal to the limits ensures that the initial condition of the safety analyses are met.
r 4t44 ~ ~ rr gM SR 3.4.1.4 Measurement of RCS total flow rate by performance of a precision calorimetric heat balance ei~c-e-ever~y--4-mer~itC~ialows the installed RCS flow instrumentation to be calibrated and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.~=Insert 2 eqmt9 o fwer4 2 This SR is modified by a Note that allows entry into MODE 1, without having performed the SR, and placement of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 24 hours after > 90% RTP. This exception is appropriate since the heat balance requires the unit to be at a minimum of 90% RTP to obtain the stated RCS flow accuracies.
The Surveillance shall be performed within 24 hours after reaching 90% RTP.REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 1 B3414Rvso o B 3.4.1-4 Revision No. 0 ROS Minimum Temperature for Criticality B 3.4.2 BASES APPLI CABLE SAFETY ANALYSES (continued) criticality limitation provides a small band, 6&deg;F, for critical operation below HZP. This band allows critical operation below HZP during unit startup and does not adversely affect any safety analyses since the MTC is not significantly affected by the small temperature difference between HZP and the minimum temperature for criticality.
The RCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO Compliance with the LCO ensures that the reactor will not be made or maintained critical (keff > 1.0) at a temperature less than a small band below the HZP temperature, which is assumed in the safety analysis.Failure to meet the requirements of this LCO may produce initial conditions inconsistent with the initial conditions assumed in the safety analysis.APPLICABILITY In MODE 1 and MODE 2 with keff -> 1.0, LCO 3.4.2 is applicable since the reactor can only be critical (kerr > 1.0) in these MODES.ACTIONS A.._If the parameters that are outside the limit cannot be restored, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to MODE 2 with keff < 1.0 within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period. The allowed time is reasonable, based on operating experience, to reach MODE 2 with keff < 1.0 in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.4.2.1 RCS loop average temperature is required to be verified at or above 5410&deg; F ur=4eepra-re evrt2~ ~-=-Insert2
~ee19e4eteI el9ee-pef 494~e9~eF=r.eefl~-e~&ie ltAibWfI1UII drIi 4 ypibdIIy ptirfOflTlett'
~freftienet--~i+iea1ity-is-approaohed.
REFERENCES
: 1. UFSAR, Section 14.1.1.Cook Nuclear Plant Unit 1 ..- evso o B 3.4.2-2 Revision No. 0 RCS P/T Limits B 3.4.3 BASES ACTIONS (continued)
Condition C is modified by a Note requiring Required Action 0.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE REQUIREMENTS SR 3.4.3.1 Verification that operation is within limits is required when RCS pressure and temperature conditions are undergoing planned.changes. ThsF-e --
t-te-t ee-iease~n~e--re~afrmre~ve~n ==Insert 2 Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.
This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.
REFERENCES
: 1. WCAP-1 5878, Rev. 0, dated December 2002.2. 10 CFR 50, Appendix G.3. ASME, Boiler and Pressure Vessel Code, Section Ill, Appendix G.4. ASTM E 185-82, July 1982.5. 10 CFR 50, Appendix H.6. Regulatory Guide 1.99, Revision 2, May 1988.7. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.Cook Nuclear Plant Unit 1 ..- evso o B 3.4.3-6 Revision No. 0 RCS Loops -MODES 1 and 2 B 3.4.4 BASES APPLICABILITY (continued)
Operation in other MODES is covered by: LCO 3.4.5, "RCS Loops -MODE 3";LCO 3.4.6, "RCS Loops -MODE 4";LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled";LCO 3.4.8, "RCS Loops -MODE 5, Loops Not Filled";LCO 3.9.4, 'Residual Heat Removal (RHR) and Coolant Circulation
-High Water Level"; and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation
-Low Water Level." ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the unit to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verification e ert42=heir-s hat each RCS loop is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to the DNBR limit. tl tqieney=of-d  44- e-t~-~
ea REFERENCES
: 1. UFSAR, Section 14.1.4===Insert 2 Cook Nuclear Plant Unit 1 B3443Rvso o B3.4.4-3 Revision No. 0 RCS Loops -MODE 3 B 3.4.5 BASES ACTIONS (continued) 0.._1 If one required RCS loop is not in operation, and the Rod Control System is capable of rod withdrawal, the Required Action is to place the Rod Control System in a condition incapable of rod withdrawal (e.g., de-energize all CRDMs by opening the RTBs or de-energizing the motor generator (MG) sets). When the Rod Control System is capable of rod withdrawal, it is postulated that a power excursion could occur in the event of an inadvertent control rod bank withdrawal.
This mandates having the heat transfer capacity of two RCS loops in operation.
If only one loop is in operation, the Rod Control System must be rendered incapable of rod withdrawal.
The Completion Time of 1 hour to defeat the Rod Control System is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period.D.1, D.2, and 0.3 If two required RCS loops are inoperable, or two required RCS loops are not in operation with Rod Control System capable of rod withdrawal, or required RCS loop not in operation with Rod Control System not capable of rod withdrawal, the Rod Control System must be placed in a condition incapable of rod withdrawal (e.g., all CRDMs must be de-energized by opening the RTBs or de-energizing the MG sets). All operations involving introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 must be suspended, and action to restore one of the RCS loops to OPERABLE status and operation must be initiated.
Boron dilution requires forced circulation for proper mixing, and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal.
Suspending operations that would cause the introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 is required to assure c~ontinued safe operation.
With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification eeiy2-h~euis that the required loops are in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.Cook Nuclear Plant Unit 1 ..- evso o B3.4.5-4.Revision No. 0 RCS Loops -MODE 3 B 3.4.5 BASES SURVEILLANCE REQU IREMENTS (continued)
I II~ II.9LJL.11.Jy
~JI It. IIt.JLU.d I.S ~L4III'.JIS.~.I IL *J~JI IJI'~Insert 2 SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side water level is above the lower tap of the SG wide range level instrumentation by >- 420 inches for required RCS loops. If the SG tubes become uncovered, the associated loop may not be capable of providing the heat sink for removal Of the decay heat. The water level can be verified by either the wide range or the narrow range instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. Th--1e-4 tw=Farteqeiiee--i 2 ethr4~a~a4 iees=ef=G&Ieve
.=SIR 3.4.5.3 Verification that each required RCP is OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional ROP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power availability to each required RCP."*===Insert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 1 B3455Rvso o B 3.4.5-5 Revision No. 0 ROS Loops -MODE 4 B 3.4.6 BASES ACTIONS (continued) minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SR 3.4.6.1 This SR requires verification that the. required RCS or RHR loop is in operation and circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. T-he-F teney-of-SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side water level is above the lower tap of the SG wide range level instrumentation by > 420 inches. If the SG U-tubes become uncovered, the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.The water level can be verified by either the wide range or the narrow range level instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met.
w-e~ ef-tber~=Insert 2 SR 3.4.6.3 Verification that each required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump. -FeF~~lee~-7dy~-eaitretreenllInsert 2--==Insert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 1 ..- evso o B 3.4.6-4 Revision No. 0 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES ACTIONS (continued) 0.1 and C.2 If a required RHR loop is not in operation or if no required loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated.
Suspending operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 is required to assure continued safe operation.
With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Times reflect the importance of maintaining operation for heat removal.SURVEILLANCE REQUIREMENTS SR 3.4.7.1 This SR requires verification ever that the required loop is in operation circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
I*-nsert 2 neter-R4pete1le9r-SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side water levels are above the lower tap of the SG wide range level instrumentation by > 420 inches ensures an alternate decay heat removal method via natural circulation in the event that the second RHR loop is not OPERABLE.
The water level can be verified by either the wide range or the narrow range instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. If both RHR loops are OPERABLE, this Surveillance is not needed. ~The ---Insert 2 SR 3.4.7.3 Verification that each required RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by Cook Nuclear Plant Unit I B 3.4.7-4 Revision No. 0 Cook Nuclear Plant Unit 1 B3.4.7-4 Revision No. 0 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued) verifying proper breaker alignment and power available to each required RHR pump. If secondary side water level is above the lower tap of the SG wide range level instrumentation by > 420 inches in at least two SGs.this Surveillance is not needed. T-he-tre lee-f-7--=eye-s--ier ,---Insert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES
: 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation." Cook Nuclear Plant Unit 1 B3475Rvso o B 3.4.7-5 Revision No. 0 RCS Loops -MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE REQUIREMENTS SR 3.4.8.1 This SR requires verification that the required loop is in operation circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. -T-he-Fr-eqtuen~e{-h ie-eu-ie1e4-"
I-nsert 2 SR 3.4.8.2 Verification that each required pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump. T~h fqe -f7d~zi oedeieer-reefteble-ia=vew-efo Insert 2
~
This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 1 B3483Rvso o B 3.4.8-3 Revision No. 0 Pressurizer B 3.4.9 BASES SURVEILLANCE REQUIREMENTS SR 3.4.9.1 This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. T~e ece rse ~
p acev ao-il far 2-t~eeam er-4et eeveac-te.
SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated pressurizer backup heaters are verified to be at their specified capacity.
This may be done by testing the power supply output with the heaters energized. -The Fr ets e utt 2 anc na-z r-Dc -cne REFERENCES
: 1. UFSAR, Chapter 14.2. NUREG-0737, November 1980.Cook Nuclear Plant Unit 1 B3494Rvso o B 3.4.9-4 Revision No. 0 Pressurizer PORVs B 3.4.11 BASES ACTIONS (continued) place the PORV(s) in manual control, this may not be possible for all causes of Condition B entry with PORV(s) inoperable and not capable of being manually cycled (e.g., as a result of failed control power fuse(s) or control switch malfunctions(s))
H.1 and H.2 If any Required Action and associated Completion Time of Condition A, B, C, D, F, F, or G is not met, if three PORVs are inoperable and not capable of being manually cycled, if two PORVs are inoperable and not capable of being manually cycled and one block valve inoperable (for reasons other than to comply with Required Action B.2) in a different line than the inoperable PORVs, or if one PORV is inoperable and not capable of being manually cycled and two block valves are inoperable (for reasons other than to comply with Required Action B.2) in different lines than the inoperable PORV, then the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.4.11.1 REQ U IREM ENTS Block valve cycling verifies that the valve(s) can be opened and closed if needed. -gMCee Insert 2 This SR is modified by a Note, which states that this SR is not required to be performed with the block valve closed in accordance with the Required Actions of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable.
SR 3.4.11.2 SR 3.4.11 .2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR.
c - l~r-met-e
~ ef~ er-fe~mt e-FeqInsert-2<p-~-f- m it ~-'-~l Cook Nuclear Plant Unit 1B341-6RvsoNo0 B 3.4.11-6 Revision No. 0 Pressurizer PORVs B 3.4.11 BASES SURVEILLANCE REQUIREMENTS (continued)
The Note modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. In accordance with Reference 4, administrative controls require this test be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation.
SR 3.4.11.3 Operating the solenoid air control valve associated with each PORV, and the check valves on the air accumulators where applicable, ensures the PORV control system actuates properly when called upon..@9per-ati9g,.exe~a--e-'
r ee=d ot t--rhc-o 2re1 re he~=Insert 2 REFERENCES
: 1. Regulatory Guide 1.32, February 1977.2. UFSAR, Section 14.1.8.3. ASME, Operation and Maintenance Standards and Guides (OM Codes).4. Generic Letter 90-.06, "Resolution of Generic Issue 70,'Power-Operated Relief Valve and Block Valve Reliability,'
and Generic Issue 94, 'Additional Low-Temperature Overpressure for Light-Water Reactors,'
Pursuant to 10 CFR 50.54(f)," June 25, 1990.Cook Nuclear Plant Unit 1B34117RvsoN.0 B 3.4.1 1-7 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued) through the pump control switch being placed in pull to lock and at least one valve in the discharge fl~w path being closed, or at least one valve in the discharge flow path being closed and sealed or locked.In addition, SR 3.4.12.3 is modified by a Note that allows the accumulator discharge isolation valve position to be verified by administrative means.This is acceptable since the valve positi~on was verified prior to deactivating the valve, access to the containment is restricted, and valves are only operated under strict procedural control.-aa-sa-ial-~#-pttH~eenTPom -Insert 2 ettu~{&#xa2; ~ ~ -{SR 3.4.12.4 The required RHR suction relief valve shall be demonstrated OPERABLE by verifying the RHR suction isolation valves are open. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.The RHR suction isolation valves are verified to be opened evefye-tr 2 SR 3.4.12.5 The RCS vent of> 2.0 square inches or a blocked open PORV is proven OPERABLE by verifying its open condition ~--Insert 2
~e19 set~I~b'e-ef=epei9~,vay=e1e&its~Thie=eate&sect;er=y+/-
The passive vent path arrangement must only be open if the vent is being used to satisfy the pressure relief requirements of LCOG 3.4;12.A.2.c.
Cook Nuclear Plant Unit 1 B 3.4.12-11 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.4.12-11 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.12.6 The PORV block valve must be verified open e-v-eJ---_-w.us to provide the flow path for each required PORV to perform its function when actuated.
The valve must be remotely verified open in the main control room. This Surveillance is performed if one or more PORVs satisfy the LCO.The block valve is a remotely controlled, motor operated valve. The power to the valve operator-is not required removed, and the manual operator is not required locked in the inactive position.
Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.f& I nsernt 2 SR 3.4.12.7 Verification that each required emergency air tank bank's pressure is > 900 psig assures adequate air pressure for reliable PORV operation.
With the emergency air supply at -> 900 psig, there will be enough air to support PORV operation for 10 minutes with no operator action upon a loss of control air. T~e-3=ty 4-- Insert SR 3.4.12.8 Performance of a COT is required on each required PORV to verify and, as necessary, adjust its lift setpoint.
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 COT 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 COT will verify the setpoint is within the LCO limit. PORV actuation could depressurize the RCS and is not required.Cook Nuclear Plant Unit 1 B341-2Rvso o B 3.4.12-12 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)
A Note has been added indicating that this SR is not required to be performed until 12 hours after decreasing ROS cold leg temperature to< 2660&deg;F. The COT cannot be performed until in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP setting. The test must be performed within 12 hours after entering the LTOP MODES. 4he- Insert 2-SR 3.4.12.9 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required-vePy -methe~to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input. 4-=.Insert 2 REFERENCES
: 1. 10 CFR 50, Appendix G.2. Generic Letter 88-11.3. ASME, Boiler and Pressure Vessel Code, Section III.4. WCAP-1 3235, "Donald C. Cook Units 1 & 2, Analysis of Low Temperature Overpressurization Mass Injection Events with Pressurizer Steam Bubble and RHR Relief Valve, March 1992;"WCAP-1 2483 Revision 1, "Analysis of Capsule U From the American Electric Power Company D. C. Cook Unit I Reactor Vessel Radiation Surveillance Program, December 2002;" and WCAP-13515, Revision 1, "Analysis of Capsule U From Indiana Michigan Power Company D. C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program, May 2002." 5. 10 CFR 50, Section 50.46.6. 10 CFR 50, Appendix K.7. Generic Letter 90-06.Cook Nuclear Plant Unit 1 B341-3Rvso o B 3.4.12-13 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within thie LCO limits ensures the integrity of the RCPB is maintained.
Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection.
It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an ROS water inventory balance.The ROS water inventory balance must be performed with the reactor at steady state operating conditions.
The Surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed until 12 hours after establishing steady state operation.
The 12 hour allowance provides sufficient time to collect and process all necessary data after stable unit conditions are established.
Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LOG 3.4.15, "RCS Leakage Detection Instrumentation." Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an ROS water inventory balance.T'N 2 SIR 3.4.13.2 This SIR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated.
The primary to secondary LEAKAGE is Cook Nuclear Plant Unit 1 ..35ReiinN.1 B 3.4.13-5 Revision No. 15 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued) measured at room temperature as described in Reference
: 7. Prior to comparison with the 150 gallons per day TS limit, the measured primary to secondary LEAKAGE is multiplied by a volume correction factor of 1.52. The correction factor ensures the offsite dose analyses, which assume primary to secondary leakage is at normal operating temperature and pressure, remain bounding.
The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all of the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours after establishment of steady state operation.
For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.
The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 7).REFERENCES
.1. UFSAR, Section 1.4.3.2. Regulatory Guide 1.45, May 1973.3. UFSAR, Section 14.2.4.4. Letter from Indiana Michigan Power Company (M. W. Rencheck) to the NRC dated October 26, 2000 (Letter C1000-20).
: 5. Letter from NRC (John F. Stang) to Indiana Michigan Power Company (Robert P. Powers), dated November 8, 2000.6. NEI 97-06, 'Steam Generator Program Guidelines." 7. EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines." Cook Nuclear Plant Unit 1 ..36ReiinN.1 B 3.4.13-6 Revision No. 15 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued) potential for an unplanned transient if the Surveillance were performed with the reactor at power.The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.
Therefore, this SR is modified by a Note that states the Surveillance is only required to be performed in MODES 1 and 2. Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance.
SR 3.4.14.2 Verifying that the RHR interlock that prevents the valves from being opened is OPERABLE ensures that RCS pressure will not pressurize the RHR System beyond its design pressure of 600 psig. T-he-A-&deg;-mei&#xb6;
~tIset eet - ecc nditic --aitgeTs mr ncyi REFERENCES-
: 1. 10CFR50.2.
: 2. 10 CFR 50.55a(c).
: 3. WASH-i1400 (NUREG-75/01 4), Appendix V, October 1975.4. Letter from D.G. Eisenhut, NRC, to all LWR licensees, LWR Primary.Coolant System Pressure Isolation Valves, February 23, 1980.5. Letter from S.A.. Varga, NRC, to J. Dolan, Order for Modification of Licenses Concerning Primary Coolant System Pressure Isolation Valves, April 20, 1981.6. Technical Requirements Manual.7. EGG-NTAP-61 75, Inservice Testing of Primary Pressure Isolation Valves, Idaho National Engineering Laboratory, February 1983.8. NRC Safety Evaluation for License Amendment 188.9. ASME, Operation and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 1B341-5RvsoN.0 B 3.4.14-5 Revision No. 0 RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued) atmosphere must be taken to provide alternate periodic information.
The 12-hour interval is sufficient to detect increasing RCS leakage. The Required Action provides 7 days to restore another RCS leakage monitor to OPERABLE status to regain the intended leakage detection diversity.
The 7 day Completion Time ensures that the plant will not be operated in a degraded condition for a lengthy time period.E.1 and E.2 With the containment atmosphere particulate radioactivity monitor and the required containment humidity or containment atmosphere gaseous radioactivity monitor inoperable, the only means of detecting leakage is the containment sump monitor. This Condition does not provide the required diverse means of leakage detection.
The Required Action is to restore either of the inoperable required monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity.
The 30 day Completion Time ensures that the unit will not be operated in a reduced configuration for a lengthy time period.F.1 and F.2 If any Required Action and associated Completion Time of Condition A, B, C, D, or E cannot be met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.G.1 With all three types of required monitors inoperable (i.e., LCO 3.4.15.a, b, and c not met), no automatic means of monitoring leakage are available, and immediate unit shutdown in accordance with LCO 3.0.3 is required.SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is operating properly.
The---
fe I nsert 2 Cook Nuclear Plant Unit 1 B 3.4.15-6 Revision No. 33 Cook Nuclear Plant Unit 1 B 3.4.15-6 Revision No. 33 RCS Leakage Detection Instrumentation B 3.4.15 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. 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 test verifies the alarm se~tpoint and relative accuracy of the instrument string. si~ t-<=. Insert 2-f SR 3.4.15.3, SR 3.4.15.4, and SR 3.4.15.5 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels.
The.calibration verifies the accuracy of the instrument string, including the instruments located inside containment. ecee-f2-afta Insert 2
", "" t' erie+/-~
REFERENCES
: 1. UFSAR, Section 1.4.3.2. Regulatory Guide 1.45, Rev. 0, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.I 3. AEP Letter to NRC, AEP:NRC:0137D, "NRC Generic Letter 84-04;Elimination Of Postulated Pipe Breaks In Primary Main Loops Generic Issue A-2, Asymmetric Blowdown Loads On PWR Primary Systems Request. For License Condition Deletion," dated September 10, 1984.4. NRC Letter to AEP, "Generic Letter 84-04, Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," dated November 22, 1985.5. UFSAR, Section 4.2.7 6. WCAP-15435, Rev. 1, Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for D.C. Cook Units 1 and 2 Nuclear Power Plant, August 2000.Cook Nuclear Plant Unit 1 ..57ReiinN.3 B 3.4.15-7 Revision No. 33 RCS Specific Activity B 3.4.16 BASES ACTIONS. (continued)
B.1 If any Required Action and associated Completion Time of Condition A is not met, if the DOSE EQUIVALENT I-131 is in the unacceptable region of Figure 3.4.16-1, or if gross specific activity of the reactor coolant is not within limit, the reactor must be brought to MODE 3 with RCS average temperature
< 5000&deg;F within 6 hours. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 below 500&deg;F from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.4.16.1 SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at--eeet--aee-evefy-While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in gross specific activity.Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions.
T-he -F
<- Insert 2 SR 3.4.16.2 This Surveillance requires the verification that the reactor coolant DOSE EQUIVALENT 1-131 specific activity is within limit. This Surveillance is accomplished by performing an isotopic analysis of a reactor coolant sample. This Surveillance is performed in MODE I only to ensure iodine remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur. ,T-he-44ey=Fr~eqteney~is-ig ae~i atfe The Frequency, between 2 and 6 hours after a power change > 15% RTP within a 1 hour period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.SR 3.4.16.3 A radiochemical analysis for determination is required evei--484-daye-with the unit operating in MODE 1 equilibrium conditions.
The determination directly relates to the LCO and is required to verify unit~=Insert 2 Cook Nuclear Plant Unit I B 3.4.16-4 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.4.16-4 Revision No. 0 RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued) operation within the specified gross activity LCO limit. The analysis for IF is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. T]--e-F-eele Insert 2~ef,4ae This SR has been modified by a Note that indicates sampling is not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours. This ensures that the radioactive materials are at equilibrium so the analysis for F is representative and not skewed by a crud burst or other similar abnormal event.REFERENCES
: 1. 100CFRI100.11.
: 2. UFSAR, Section 14.2.4.Cook Nuclear Plant unit 1 ..65Rvso o B 3.4.16-5 Revision No. 0 Accumulators B 3.5.1 BASES ACTIONS (continued)
D.__I If more than one accumulator is inoperable, the unit is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE REQUIREMENTS SR 3.5.1.1 Each accumulator isolation valve should be verified to be fully open everd 4=2aeus. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions.
4I=F-r-eqie1ey-ie-eber-ieid..
e ree~ -thr- ~ ~ rr4~~.e~~=
--Insert 2 SR 3.5.1.2 and SR 3.5.1.3-4.e~l-e&#xa3;je, borated water volume and nitrogen cover pressure are verified for each accumulator. eirte ,,=-Insert 2=8-L .Fi
@,s4 ejq ..eT44-4 e.
u r-F-r-eq e=a f=e SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator~e since the static design of the accumulators limits the ways in which the concentration can be changed.Thy-I==nsert 2 1-.h" .,~
Sampling the affected accumulator within 6 hours after a volume increase of 13 ft 3 will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), because the water contained in the RWST is within the accumulator boron concentration requirements.
This is consistent with the recommendation of NUREG-1366 (Ref. 4).Cook Nuclear Plant Unit 1 ..- evso o B3.5.1-6 Revision No. 0 Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.5 Verification power is removed from each accumulator isolation valve operator when the RCS pressure is > 2000 psig ensures that an active failure could not result in the closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. -Sieeteweir-ie-r~eiae eaei-a~I14s4ei4-tRve
<= Insert 2 This SR allows power to be supplied to the motor operated isolation valves when ROS pressure is < 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns.
REFERENCES
: 1. UFSAR, Section 14.3.2. 10 CFR 50.46.3. WCAP-1 5049-A, "Risk-Informed Evaluation of an Extension to Accumulator Completion Times," Rev. 1, April 1999.4. NUREG-1 366, February 1990.Cook Nuclear Plant Unit 1 ..- evso o B3.5.1-7.Revision No. 0 EGOS -Operating B 3.5.2 BASES ACTIONS (continued) 0.1_Condition A is applicable with one or more ECOS trains inoperable.
The allowed Completion Time of Required Action A.1 is based on the assumption that at least 100% of the EGGS flow equivalent to a single OPERABLE ECGS train is available.
An inoperable RHR or SI pump concurrent with a closed cross-tie valve in the affected system will result in less than 100% of the EGGS flow equivalent to a single OPERABLE EGGS train because there will be flow to only two RGS loops. With less than 100% of the EGOS flow equivalent to a single OPERABLE EGGS train available, the facility is in a condition outside of the accident analyses.
Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE REQUIREMENTS SR 3.5.2.1 Verification of proper valve position ensures that the flow path from the EGOS pumps to the RCS is maintained.
Misalignment of these valves could render both EGGS trains inoperable.
Securing these valves in position by locking out control power ensures that they cannot change position as a result of an active failure or be inadvertently misaligned.
These valves are of the type, described in Reference 9, that can disable the function of both EGGS trains and invalidate the accident analyses.
A--<==Insert 2 I m Ill Ifl ................................
r................
]SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the EGGS flow paths provides assurance that the proper flow paths will exist for EGGS operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation.
Rather, it involves verification that those valves capable of being mispositioned are in the correct position. yi 1ite-<te4h U LAAU ~..UJ ~ ~JtJt....
L4L~.,%.A LII *~ALI LILA 11111 *I~~flI LUI.IU L' LIL'* ~I.* LI*, LII*~ LU.r' LUW~*-Insert 2wal-h<tg~~ram-xe-e Cook Nuclear Plant Unit 1 B 3.5.2-7 Revision No. 24 Cook Nuclear Plant Unit 1 B 3.5.2-7 Revision No. 24 ECCS -Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.3 Verifying that each ECCS pump's developed head at the flow test point is greater than or equal to the required developed head ensures that ECCS pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of ECCS pump performance required by the ASME OM Code (Ref. 10). Since the ECCS pumps cannot be tested with flow through the normal ECCS flow paths, they are tested on recirculation flow (RHR and SI pumps) or normal charging flow path (centrifugal charging pumps). This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance.
The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.5.2.4 and SR 3.5.2.5 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
T-49 cod-e e=on Insert 2-4atr-t-ee t--
~ SR 3.5.2.6 Proper throttle valve position is necessary for proper ECCS performance.
These valves have stops to allow proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This Surveillance verifies the mechanical stop of each listed ECCS throttle valve is in the correct position., Insert 2*Fe~ _ * ...... .sdtbhU.atJ rrn -e .4 .-i Cook Nuclear Plant Unit I B 3.5.2-8 Revision No. 24 Cook Nuclear Plant Unit 1 B 3.5.2-8 Revision No. 24 ECCS -Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.7 Periodic inspections of the containment sump suction inlets ensure that they are unrestricted and stay in proper operating condition.
This Surveillance verifies that the sump suction inlets are not restricted by debris and the suction inlet strainers show no evidence of structural distress, such as openings or gaps, which would allow debris to bypass the strainers.
@re--Insert 2 ,I Ae4e-su ffieieatt~e.eteet~abmeer REFERENCES
: 1. UFSAR, Section 1.4.7.2. 10 CFR 50.46.3. UFSAR, Section 14.3.1.4. UFSAR, Section 14.3.2.5. UFSAR, Section 14.2.4.6. UFSAR, Section 14.2.5.7. UFSAR, Section 14.3.4.8. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.9. IE Information Notice No. 87-01 10. ASME, Operations and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 1 B 3.5.2-9 Revision No. 24 Cook Nuclear Plant Unit 1 B 3.5.2-9 Revision No. 24 RWST B 3.5.4 BASES ACTIONS (continued) brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.5.4.1 The RWST borated water temperature should be verified e e4h r-s to be within the limits assumed in the accident analyses band..T-hi
-,-- Insert 2 wtandea h-aceh gh=e-(X-1e4eee.
SR 3.5.4.2 The RWST water volume should be verified ~eiv--dy to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation.
Siiee449e-RN&#xa5;ST-iiit --= Insert'2 4se~rlj--~-r1elrtceb-mtf t e,e-xe-ier~c-e.
SR 3.5.4.3 The boron concentration of the RWST should be verified be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized.
REFERENCES
: 1. UFSAR, Section 6.2.2.2. UFSAR, Section 14.3.<-=-=Insert 2 Cook Nuclear Plant Unit 1 ..- evso o B 3.5.4-5 Revision No. 0 Seal Injection Flow B 3.5.5 BASES APPLICABILITY In MODES 1, 2, and 3, the seal injection flow resistance limit is dictated by ECCS flow requirements, Which are specified for MODES 1, 2, 3, and 4. The seal injection flow resistance limit is not applicable for MODE 4 and lower, however, because high seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES. Therefore, RCP seal injection flow resistance must be limited in MODES 1, 2, and 3 to ensure adequate ECCS performance.
ACTIONS A.1 With the seal injection flow resistance not within its limit, the amount of charging flow available to the RCS may be reduced. Under this condition, action, must be taken to restore the flow resistance to within its limit. The operator has 4 hours from the time the flow resistance is known to not be within the limit to correctly position the manual valves and thus be in compliance with the accident analysis.
The Completion Time minimizes the potential exposure of the unit to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow resistance within limits. This time is conservative with respect to the Completion Times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.
B.1 and B.2 When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated.
The Completion Time of 6 hours for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators.
Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.
SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Verification~wevey8+/-ye-that the seal injection flow resistance is within the limit ensures that the ECCS injection flows stay within the safety analysis.
A differential pressure is established between the charging header and the RCS, and the total seal injection flow is verified to be within the limit determined in accordance with the ECCS safety analysis.The flow resistance shall be > 0.227 ft/g pm 2.Cook Nuclear Plant Unit 1 B 3.5.5-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.5.5-3 Revision No. 0 Seal Injection Flow B 3.5.5 BASES SURVEILLANCE REQUIREMENTS (continued)
The seal injection flow resistance, RSL, is determined from the following expression:
RSL =2.31 (PcHP-PsI)/Q 2 where: PCHP = charging pump header pressure (psig);= 2148 psig (low pressure operation) or 2300 psig (high pressure operation);
and Q = total seal injection flow (gpm).eeaset {with, Pre k
2-e~tteftey-l9~e-lfevcfl tc~ bo accoptablo
~As noted, the Surveillance is not required to be performed until 4 hours after the pressurizer pressure has stabilized within a _+ 20 psig range of normal operating pressure.
The pressurizer pressure requirement is specified since this configuration will produce the required pressure.conditions necessary to assure that the manual valves are set correctly.
The pressurizer pressure indications are averaged to determine whether the appropriate pressure has been achieved.
The exception is limited to 4 hours to ensure that the Surveillance is timely.REFERENCES
: 1. UFSAR, Section 14.3.1.2. UFSAR, Section 14.3.2.3. UFSAR, Section 14.2.4.4. UFSAR, Section 14.2.5.Cook Nuclear Plant Unit I1 ..- evso o B 3.5.5-4 Revision No. 0 Containment Air Locks B 3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued) air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by the Containment Leakage Rate Testing Program.The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria which is applicable to SR 3.6.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C containment leakage rate.SR 3.6.2.2 The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY.
Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment.
Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur.
t~ a e~i4tek_~~,.a4a~e~ck Insert 2e dte~- e~~-r h~~,si 2--et.~-~~a-~as~eates~~prem L;4 -- REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 14.2.6.3. UFSAR, Section 5.7.4. 10 CFR 50, Appendix J, Option B.Cook Nuclear Plant Unit 1 B3626Rvso o B3.6.2-6 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES ACTIONS (continued) locked,.sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned.
Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is small.D.1 and D.2 If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.3.1 This SR ensures that the containment purge supply and exhaust valves are closed as required or, if open, open for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable.
The SR is not required to be met when the containment purge valves are open for the reasons stated. The valves may be opened for pressure control, ALARA or air-quality considerations for personnel entry, or for Surveillances or mafntenance activities that require the valves to be open. The containment purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. , =.=Insert 2 SR 3.6.3.2 This SR requires verification that each containment isolation manual valve and blind flange located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. This SR does not require any testing or valve .manipulation.
Rather, it involves verification that those containment isolation valves outside containment and capable of being mispositioned are in the correct position.
Insert 2 Cook Nuclear Plant Unit 1 B 3.6.3-7 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued) aa~ae-a~efauem nd wa h~4e.l~~vl ee ~~-e--ece--c-p~-es The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time the valves are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.SR 3.6.3.3 This SR requires verification that each containment isolation manual.valve and blind flange located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. For containment isolation valves inside containment, the Frequency of "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time they are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.This Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4, for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.Cook Nuclear Plant Unit 1 B3638Rvso o B3.6.3-8 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.3.4 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY.
The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses.The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.3.5 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under .administrative controls.
th
-Insert 2 ff~rmt u~re~erte-e~test&#xa2;ta f~ ~ iewt+h ce-t wr.-4#ftrje'~~re~s cAe 4e~ela-&#xa2;e-2-fett-rltep a
REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 14.2.6.3. UFSAR, Section 5.4.1 and Table 5.4-1.Cook Nuclear Plant Unit 1 B3639Rvso o B 3.6.3-9 Revision No. 0 Containment Pressure B 3.6.4 BASES S URVElILLAN CE REQ U IREM ENTS SR 3.6.4.1 Verifying that containment pressure is within limits ehsures that unit operation remains within the limits assumed in the containment analysis.-Thqq .ea -ee <Insert 2 REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 5.2.2.2.3. 10 CFR 50, Appendix K.Cook Nuclear Plant Unit 1 B3643Rvso o B 3.6.4-3 Revision No. 0 Containment Air Temperature B 3.6.5 BASES ACTIONS A.1 When containment average air temperature in the upper or lower compartment is not within the limit of the LCO, the average air temperature in the affected compartment must be restored to within limits within 8 hours. This Required Action is necessary to return operation to within the bounds of the containment analysis.
The 8 hour Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.B.1 and B.2 If the containment average air temperature cannot be restored to within its limits within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.5.1 and SR 3.6.5.2 Verifying that containment average air temperature is within the LCO limits ensures that containment operation remains within the limits assumed for the containment analyses.
In order to determine the containment average air temperature, an average is calculated using measurements taken at locations within the containment selected to provide a representative sample of the oV~erall containment atmosphere.
In the upper compartment, two locations at a nominal elevation of 712 ft 0 inches and a third location at a nominal elevation of 624 ft 10 inches are used and an arithmetic average taken. In the lower compartment, a volume weighted average temperature is calculated whereby the volume fraction for each of the various areas of lower containment is multiplied by the representative temperature, utilizing one or more temperature instruments, in that volume; In this way the temperatures are "weighted" according to the volume fraction.
These weighted temperatures are then summed to determine the Weighted Average Temperature for Lower Containment.
REFERENCES
: 1. UFSAR, Section 14.3.4.2. 10 CFR 50.49.,--Insert 2 Cook Nuclear Plant Unit 1 B3653Rvso o B 3.6.5-3 Revision No. 7 Containment Spray System B 3.6.6 BASES ACTIONS A.1I With one containment spray train inoperable, the affected train must be restored to OPERABLE status within 72 hours. The components in this degraded condition are capable of providing 100% of the heat removal and iodine removal needs after an accident.
The 72 hour Completion Time was developed taking into account the redundant heat removal and iodine removal .capabilities afforded by the OPERABLE train and the low probability of a DBA occurring during this period.B.1 and B.2 If the affected containment spray train cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 84 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.SURVEILLANCE REQUIREMENTS SR 3.6.6.1 Verifying the correct alignment of manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position since they were verified in the correct position prior to being secured. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation.
Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned, are in the correct position.
2 SR 3.6.6.2 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 5).Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect Cook Nuclear Plant Unit 1B36.5ReionN.2 B 3.6.6-5 Revision No. 32 Containment Spray System B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued) incipient failures by indicating abnormal performance.
The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.6.3 and SR 3.6.6.4 These SRs require verification that each automatic containment spray valve actuates to its correct position and each containment spray pump starts upon receipt of an actual or simulated containment spray actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls..Th-4m t=Frfuae~qae-Ih-4 2$uetae-~rh.e~lesht.pt~~ga ue~
These Surveillances include a Note that states that in MODE 4, only the manual portion of the actuation signal is required.
This is acceptable since the automatic portion of the actuation signal is not required to be OPERABLE by ITS 3.3.2, "Engineered Safety Features Actuation System (ESFAS) Instrumentation." SR 3.6.6.5 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections.
This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded.
The event based surveillance frequency following maintenance that could result in nozzle blockage was chosen because this passive portion of the system is not susceptible to service induced degradation.
REFERENCES
: 1. UFSAR, Section 1.4.7.2. UFSAR, Section 14.3.4.3. 10 CFR 50.49.4. 10 CFR 50, Appendix K.5. ASME, Operation and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 1B3666ReionN.2 B 3.6.6-6 Revision No. 32 Spray Additive System B 3.6.7 BASES ACTIONS A.1 If the Spray Additive System is inoperable, it must be restored to OPERABLE within 72 hours. The pH adjustment of the Containment Spray System flow for corrosion protection and iodine retention enhancement is reduced in this condition.
The 72 hour Completion Time takes into account the redundant flow path capabilities and the low probability of the worst case DBA occurring during this period. In addition, if the Containment Spray System is available, it would remove some iodine from the containment atmosphere in the. event of a DBA.B.1 and B.2 If the Spray Additive System cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be* brought to at least MODE 3 within 6 hours and to MODE 5 within 84 hours. The allowed Completion Time of 6 hours is reasonable, based* on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems. The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for the release of radioactive material from the Reactor Coolant System is reduced in MODE 3.SURVEILLANCE REQUIREMENTS SR 3.6.7.1 Verifying the correct alignment of Spray Additive System manual, power operated, and automatic valves in the spray additive flow path provides assurance that the system is able to provide additive to the Containment Spray System in the event of a DBA. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation.
Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned are in the correct position. .Isr SR 3.6.7.2 To provide effective iodine retention, the containment spray must be an alkaline solution.
Since the RWST contents are normally acidic, the volume of the spray additive tank must provide a sufficient volume of spray additive to adjust pH for all water injected.
This SR is performed to verify the availability of sufficient NaOH solution in the Spray Additive System. ~ w Insert 2 Cook Nuclear Plant Unit 1I ..- evso o B 3.6.7-3 Revision No. 0 Spray Additive System B 3.6.7 BASES SURVEILLANCE REQUIREMENTS (continued) f#t,-e SR 3.6.7.3 This SR provides verification (by chemical analysis) of the NaOH concentration in the spray additive tank and is sufficient to ensure that the spray solution being injected into containment is at the correct pH level.
2 SR 3.6.7.4 This SR provides verification that each automatic valve in the Spray Additive System flow path actuates to its correct position.
This Surveillance is not required for valves that are locked, sealed; or otherwise secured in the required position under administrative controls.4ere-e-tee
=9pr 1e~aePa eet-a4 c.g ae4ea~~-~~li~~-~ta~it I==nsert 2 SR 3.6.7.5 To ensure that the correct pH level is established in the borated water solution provided by the Containment Spray System, the flow rate in the Spray Additive System is verified once every 5 years.. This SR provides assurance that the correct amount of NaOH will be metered into the flow path upon Containment Spray System initiation.
The test is performed by verifying the flow rate from the spray additive tank test line to each Containment Spray System train with each containment spray pump operating in the recirculation mode. Insert 2 el REFERENCES
: 1. UFSAR, Chapter 14.3.5.9.Cook Nuclear Plant Unit 1 B3674Rvso o B 3.6.7-4 Revision No. 0 DIS B 3.6.9 BASES ACTIONS (continued)
Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.9.1 This SR confirms that >- 34 of 35 hydrogen ignitors can be successfully energized in each train. The ignitors are simple resistance elements.Therefore, energizing provides assurance of OPERABILITY.
The allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise redundancy in that region, the containment regions are interconnected so that ignition in one region would cause burning to progress to the others (i.e., there is overlap in each hydrogen ignitor's effectiveness between regions). -R Insert 2 SR 3.6.9.2 This SR confirms that the two inoperable hydrogen ignitors allowed by SR 3.6.9.1 (i.e., one in each train) are not in the same containment region.
ee~l~bt-teei -eH 9e~ ~ et 4 t -e1 re==Insert 2 SR 3.6.9.3 A more detailed functional test is performed to verify system OPERABILITY.
Each ignitor is visually examined to ensure that it is clean and that the electrical circuitry is energized.
All ignitors, including normally inaccessible ignitors, are visually checked for a glow to verify that they are energized.
Additionally, the surface temperature of each ignitor is measured to be > 1700&deg;F to demonstrate that a temperature sufficient for ignition is achieved.
: 2. FSA, ectieon dto5.8.a-a p~==Insert 2 Cook Nuclear Plant Unit 1B369-ReionN.4 B3.6.9-4 Revision No. 44 CEQ System B 3.6.10 BASES LCO In the event of a DBA, one train of the CEQ System is required to provide the minimum air recirculation for heat removal and hydrogen mixing assumed in the safety analyses.
To ensure this requirement is met, two trains of the CEQ System must be OPERABLE.
This will ensure that at least one train will operate, assuming the worst case single failure occurs.APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the CEQ System.Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the CEQ System is not required to be OPERABLE in these MODES.ACTIONS A.il If one of the trains of the CEQ System is inoperable, it must be restored to OPERABLE status within 72 hours. The components in this degraded condition are capable of providing 100% of the flow and hydrogen skimming needs after an accident.
The 72 hour Completion Time was developed taking into account the redundant flow and hydrogen skimming* capability of the OPERABLE CEQ System train and the low probability of a DBA occurring in this period.B.1 and B.2 If the CEQ System train cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQU IREM ENTS SR 3.6.10.1 Verifying that each CEQ System fan starts on an actual or simulated actuation signal, after a-delay > 108 seconds and < 132 seconds, and operates for > 15 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly.
It also ensures that blockage, fan and/or motor failure, or excessive vibration can be detected for corrective action. -Td-he- Irnsert 2 Cook Nuclear Plant Unit 1B36103RvsoN.0 B 3.6.10-3 Revision No. 0 CEQ System B 3.6.10 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR has been modified by a Note that states that this Surveillance is only required to be met in MODES 1, 2, and 3. This allowance is necessary since the specified delay (i.e., > 108 seconds and< 132 seconds) is only applicable to the automatic actuation signal (i.e., Containment Pressure -High), which is only required to be OPERABLE in MODES 1, 2, and 3. In addition, LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," requires the CEQ System Manual Initiation Function to be OPERABLE in MODE 4 and requires the performance of a TADOT every 24 months. This requirement will ensure the Manual Initiation Function can actuate the required equipment in MODE 4.SR 3.6.10.2 Verifying, with the return air fan discharge backdraft damper locked closed and the fan motor energized, the static pressure between the fan discharge and the backdraft damper is > 4.0 inches water gauge confirms one operating condition of the fan. This test is indicative of overall fan motor performance.
Such tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance.
Thae~Insert 2 ee s-e-a19l-t-he-twe-tai9f-a dt {sbte.SR 3.6.10.3 Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exist when the fan is started. By applying the correct counterweight, the damper operation can be confirmed.
+he~Fe~r rg2 de-~-heme 2 S R 3.6.10.4 Verifying the OPERABILITY of the motor operated valve in the hydrogen skimmer header provides assurance that the proper flow path will exist when the valve receives an actuation signal. Thde-urltenjieie-w'
.---Insert 2 t Cook Nuclear Plant Unit 1 B 3.6.10-4 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.6.10-4 Revision No. 0 Ice Bed B 3.6.11 BASES APPLICABILITY In MODES 1, 2, 3, and 4, a OBA could cause an increase in containment pressure and temperature requiring the operation of the ice bed.Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ice bed is not required to be OPERABLE in these MODES.ACTIONS A.1 If the ice bed is inoperable, it must be restored to OPERABLE status within 48 hours. The Completion Time was developed based on operating experience, which confirms that due to the very large mass of stored ice, the parameters comprising OPERABILITY do not change appreciably in this time period. If a degraded condition is identified, even for temperature, with such a large mass of .ice it is not possible for the degraded condition to significantly degrade further in a 48 hour period.Therefore, 48 hours is a reasonable amount of time to correct a degraded condition before initiating a shutdown.B.1 and B.2 If the ice bed cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.11.1 Verifying that the maximum temperature of the ice bed is<27&deg;F ensures that the ice is kept well below the melting point. Tfhe
___ Insert 2 waqaete-~ra4g-exe~a~ih rP, Cook Nuclear Plant Unit 1 B 3.6.11-4 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.6.11-4 Revision No. 0 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued)
The total ice mass and individual radial zone ice mass requirements defined in this Surveillance, and the minimum ice mass per basket requirement defined by SR 3.6.11.3, are the minimum requirements for OPERABILITY.
Additional ice mass beyond the SRs is maintained to address sublimation.
This sublimation allowance is generally applied to baskets in each radial zone, as appropriate, at the beginnirng of an operating cycle to ensure sufficient ice is available at the end of the-operating cycle for the ice condenser to perform its intended design function.
2 SR 3.6.11.3 Verifying that each selected sample basket from SR 3.6.11.2 contains at least 600 lbs of ice in the as-found (pre-maintenance) condition ensures that a significant localized degraded mass condition is avoided.This SR establishes a per basket limit to ensure any ice mass degradation is consistent with the initial conditions of the DBA by not significantly affecting the containment pressure response.
Reference 2 provides insights through sensitivity runs that demonstrate that the containment peak pressure during a DBA is not significantly affected by the ice mass in a large localized region of baskets being degraded below the required safety analysis mean, when the radial zone and total ice mass requirements of SR 3.6.11.2 are satisfied.
Any basket identified as containing less than 600 lbs of ice requires appropriately entering ACTION A for an inoperable ice bed due to the potential that it may represent a significant condition adverse to quality.As documented in Reference 2, maintenance practices actively manage individual ice basket mass above the required safety analysis mean for each radial zone. Specifically, each basket is serviced to keep its ice mass above 1132 lbs for- zone A, 1132 lbs for zone B, and 1132 lbs for zone C. If a basket sublimates below the safety analysis mean value, this instance is identified within the CNP corrective action program, including evaluating maintenance practices to identify the cause and correct any deficiencies.
These maintenance practices provide defense in depth beyond compliance with the ice bed Surveillance Requirements by limiting the occurrence of individual baskets with ice mass less than the required safety analysis mean.Cook Nuclear Plant Unit 1 B 3.6.11-6 Revision No. 0 Ice Bed B 3.6. 11 BASES SURVEILLANCE REQUIREMENTS (continued) accumulation on lattice frames and wall panels. The flow area through the ice basket support platform is not a more restrictive flow area because it is easily accessible from the lower plenum and is maintained clear of ice accumulation.
There is no mechanistically credible method for ice to accumulate on the ice basket support platform during unit operation.
Plant and industry experience has shown that the vertical flow area through the ice basket support platform remains clear of ice accumulation that could produce blockage.
Normally only a glaze may develop or exist on the ice basket support plafform which is not significant to blockage of flow area. Additionally, outage maintenance practices provide measures to clear the ice basket support plafform following maintenance activities of any accumulation of ice that could block flow areas.Frost buildup or loose ice is not to be considered as. flow channel blockage, whereas attached ice is considered blockage of a flow channel.Frost is the solid form of water that is loosely adherent, and can be brushed off with the open hand.2 SR 3.6.11.5 This SR ensures that a representative sampling of ice baskets, which are relatively thin walled, perforated cylinders, have not been degraded by wear, cracks, corrosion, or other damage. The SR is designed around a full-length inspection of a sample of baskets, and is intended to monitor the effect of the ice condenser environment on ice baskets. The groupings defined in the SR (two baskets in each azimuthal third of the ice bed) ensure that the sampling of baskets is reasonably distributed.
The Frequency of 40 months for a visual insPection of the structural soundness of the ice baskets is based on engineering judgment and considers such factors as the thickness of the basket walls relative to corrosion rates expected in their service environment and the results of the long term ice storage testing. <==-Insert 2 SIR 3.6.11.6 Verifying the chemical composition of the stored ice ensures that the stored ice has a boron concentration
> 1800 ppm and < 2300 ppm as sodium tetraborate and a high pH, > 9.0 and < 9.5 at 25&deg;C, in order to meet the requirement for borated water when the melted ice is used in the ECCS recirculation mode of operation.
Additionally, the minimum boron concentration limit is used to assure reactor subcriticality in a post LOCA environment, while the maximum boron concentration limit is used as the bounding value in the hot leg switchover timing calculation (Ref. 4). This is accomplished by obtaining at least 24 ice samples. Each sample is taken approximately one foot from the top of the ice of each randomly Selected ice basket in each ice condenser bay. The SR is modified by a Cook Nuclear Plant Unit 1 ..18ReiinN.4 B 3.6.11-8 Revision No. 43 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued)
Note that arrows the boron concentration and pH value obtained from averaging the individual samples' analysis results to satisfy the requirements of the SR. If either the average boron concentration or average pH value is outside their prescribed limit, then entry into Condition A is required.
Sodium tetraborate has been proven effective in maintaining the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine, although the removal of iodine from the containment atmosphere by the sodium tetraborate is not assumed in the accident analysis.
This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment Spray System fluids in the recirculation mode of operation.
The Frequency of 54 months is intended to be consistent with the expected length of three fuel cycles, and was developed considering these facts: a. Long term ice storage tests have determined that the chemical composition of the stored ice is extremely stable;b. There are no normal operating mechanisms that decrease the boron concentration of the stored ice, and pH remains within a 9.0-9.5 range when boron concentrations are above approximately 1200 ppm;c. Operating experience has demonstrated that meeting the boron concentration and pH requirements has never been a problem; and d. Someone would have to enter the containment to take the sample, and, if the unit is at power, that person would receive a radiation dose.
2 SR 3.6.11.7 This SR ensures that initial ice fill and any subsequent ice additions meet the boron concentration and pH requirements of SR 3.6.11.6.
The SR is modified by a Note that allows the chemical analysis to be performed on either the liquid or resulting ice of each sodium tetraborate solution prepared.
If ice is obtained from offsite sources, then chemical analysis data must be obtained for the ice supplied.Cook Nuclear Plant Unit 1 B361- eiinN.4 B 3.6.11-9 Revision No. 43 Ice Condenser Doors B 3.6.12 BASES ACTIONS (continued)
C.1 If Required Action B.1 or B.2 and associated Completion Time is not met, the doors must be restored to OPERABLE status and closed positions within 48 hours. The 48 hour Completion Time is based on the fact that, with the very large mass of ice involved, it would not be possible for the temperature to increase to the melting point and a significant amount of ice to melt in a 48 hour period. The 48 hour Completion Time is also consistent with the ACTIONS of LCO 3.6.11, "Ice Bed." 0.1 and 0.2 With any Required Action and associated Completion Time of Condition A or C not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SIR 3.6.12.1 Verifying that the inlet doors are in their closed positions makes the operator aware of an inadvertent opening of one or more doors. ,The 2 Frete 4 t ~
if w th-tt~rf SR 3.6.12.2 Verifying, by visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris provides assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed.
eeei9=e1 j1reefij 4.Rkei
~ t"lre~rblt-Ym--Bw eer====Insert 2 Cook Nuclear Plant Unit 1B36125RvsoN.0 B 3.6.12-5 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.3 Verifying, by visual inspection, that the top deck doors are in place and not obstructed provides assurance that the doors are performing their function of keeping warm air out of the ice condenser during normal operation, and would not be obstructed if called upon to open in response to a DBA. T -re 1ecye r
I p ,e-,h~gtce~t.te-e~4e<---Insert 2 SR 3.6.12.4 Verifying, by visual inspection, that the ice condenser inlet doors are not impaired by ice, frost, or debris provides assurance that the doors are free to open in the event of a DBA. T-he-=e sr 2 SR 3.6.12.5 Verifying the opening torque of the inlet doors provides assurance that no doors have become stuck in the closed position.
The value of 675 in-lb is based on the design opening pressure on the doors of 1.0 lb/ft 2.lFiT~hii-,=.=
Insert 2 cT-a1b 1 tnltO U13 f rQ4ieJ4~weveb,-eseauhed e~r-~e41es9-whieh-dees-Ret--aI~w r ia u ~ tter-Ree e-e-tre.-Be~eeib wr e alyer4e~r-rebl~tu d~~e~ew4, Cook Nuclear Plant Unit 1 B 3.6.12-6 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.6.12-6 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.6 The torque test Surveillance ensures that the inlet doors have not developed excessive friction and that the return springs are producing a door return torque within limits. The torque test consists of the following:
: 1. Verify that the torque, T(OPEN), required to cause opening motion at the 400 open position is -< 195 in-Ib;2. Verify that the torque, T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 400 open position is > 78 in-lb; and 3. Calculate the frictional torque, T(FRICT) = 0.5 {T(OPEN) -T(CLOSE)}, and verify that the T(FRICT) is -< 40 in-lb.T(OPEN) is known as the "door opening torque" and is equal to the nominal door torque plus a frictional torque component.
T(CLOSE) is defined as the "door closing torque" and is equal to the nominal door torque minus a frictional torque component.
The purpose of the friction and return torque Specifications is to ensure that, in the event of a small break LOCA or SLB, all of the 24 door pairs open uniformly.
This assures that, during the initial blowdown phase, the steam and water mixture entering the lower compartment does not pass through part of the ice condenser, depleting the ice there, while bypassing the ice in other bays. ;R~eaP ye~.-mRhq~ae-ah 2
.-e #
Cook Nuclear Plant Unit 1B36127RvsoN.0 B 3.6.12-7 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.7 Verifying the OPERABILITY of the intermediate deck doors provides assurance that the intermediate deck doors are free to open in the event of a OBA. The verification consists of visually inspecting the intermediate doors for structural deterioration, verifying free movement of the vent assemblies, and ascertaining free movement of each door when lifted with the applicable force shown below: Door.Liftinqi Force a. Adjacent to crane wall b. Paired with door adjacent to crane wall c. Adjacent to containment wall d. Paired with door adjacent to containment wall< 37.4 lb<_ 33.8 lb<31.8 lb_<31.0 lbe~j-efh&#xf7; Insert 2 iaerooco t-r).I&sect; REFERENCES
: 1. UFSAR, Section 14.3.4.Cook Nuclear Plant Unit 1B3.128evsoN.0 B 3.6.12-8 Revision No. 0 Divider Barrier Integrity B 3.6.13 BASES ACTIONS (continued) 0.1 and 0.2.If divider barrier integrity cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.1 3.1 Verification, by visual inspection, that all personnel access doors and equipment hatches between the upper and lower containment compartments are closed provides assurance that divider barrier integrity is maintained prior to the reactor being taken from MODE 5 to MODE 4.This SR is necessary because many of the doors and hatches may have been opened for maintenance during the shutdown.SR 3.6.13.2 Verification, by visual inspection, that the personnel access door and equipment hatch seals, sealing surfaces, and alignments are acceptable provides assurance that divider barrier integrity is maintained.
This inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.13.2 is required for each door or hatch that has been opened, prior to the final closure. Some doors and hatches may not be opened for long periods of time. -t Inser-t 2 be~uranc-athat-t-ee teqeas-t--edee=e~ef~p1e~eIder-ae4 m+traa s'e~e e4a~i-=f l~4rPa -~- -44 Ai ae Jeeuewaf),a~-
~ramcjprecP4a~~a SR 3.6.13.3 Verification, by visual inspection, after each opening of a personnel access door or equipment hatch that it has been closed makes the operator aware of the importance of closing it and thereby provides additional assurance that divider barrier integrity is maintained while in.applicable MODES.Cook Nuclear Plant Unit 1 B 3.6.13-4 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.6.13-4 Revision No. 0 Divider Barrier Integrity B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.13.4 Conducting periodic physical property tests on divider barrier seal test coupons provides assurance that the seal material has not degraded in the containment environment, including the effects of irradiation with the reactor at power. The required tests include a tensile strength test and a test for elongation..T aye-4me 4e4.4sek~a.estee-ef4 ae~t SR 3.6.13.5 Visual inspection of the seal around the perimeter provides assurance that the seal is properly secured in place, such that the total divider barrier bypass area is less than or equal to the design basis limit of 7 square feet. -ToFoucz'o 2
h~~~- ~eer tesall m
* eur ..tsface~eece~ceeietlqet d i --Insert 2=-'nsert 2 aee epueb ~ya REFERENCES
: 1. UFSAR, Section 14.3.4.1.3.1.3.
: 2. UFSAR, Section 14.3.4.1.3.1.1.e Cook Nuclear Plant Unit 1 ..35ReiinN.4 B 3.6.13-5 Revision No. 46 Containment Recirculation Drains B 3.6.14 BASES ACTIONS (continued)
C.1I If one CEQ fan room drain is inoperable, 1 hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the containment analysis.
The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1,"Containment," which requires that the containment be restored to OPERABLE status within 1 hour.D. 1 If one flow path in the flood-up overflow wall is inoperable, 1 hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the containment analysis.
The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires that containment be restored to OPERABLE status within 1 hour.E.1 and E.2 If the affected drain(s) cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.14.1 and SR 3.6.14.2 Verifying the OPERABILITY of the required refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.14.2 confirms that the required refueling canal drain blind flanges have been removed and that the required drains are clear of any obstructions that could impair their functioning.
In addition to debris near the drains, attention must be given to any debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains. This verification is performed by SR 3.6.14.1, which requires verification that there is no debris present in the upper containment or refueling canal that could obstruct the required refueling canal drains. SR 3.6.14.1 and SR 3.6.14.2 must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the blind flanges have been removed and that no debris that could impair the drains was deposited during the time the canal was filled. Ra-adle~te4-. er#p-eef to 4-=--Insert 2 Cook Nuclear Plant Unit 1 ..44ReiinN.1 B 3.6.14-4 Revision No. 18 Containment Recirculation Drains B 3.6.14 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.14.3 Verifying the OPERABILITY of the ice condenser floor drains ensures that they will be able to perform their functions in the event of a DBA.Inspecting the drain valve disk ensures that the valve is performing its function of sealing the drain line from warm air leakage into the ice condenser during normal operation, yet will open if melted ice fills the line following a DBA. Verifying that the drain lines are not obstructed ensures their readiness to drainwater from the ice condenser.
<=-l-nsert 2ee e~e~e~gsue~~~~steaee&#xa2;gt ex ae4t 4&#xa2;s -i~e 4 -e --
efrm~qF~---4ehF-eeee e --
elaa-lr gewre SR 3.6.14.4 and SR 3.6.14.5 Verifying the operability of the CEQ fan room drains ensures that they will be able to perform their function in the event of a DBA. SR 3.6.14.4 confirms that the required drains are clear of any obstructions.
In addition to debris near the drains, attention must be given to debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains.SR 3.6.14.4 must be performed before entering MODE 4 from MODE 5 and after personnel entry into a CEQ fan room while in MODES 1 through 4. This frequency was developed considering such factors as the location of the drains, and the absence of personnel traffic in the vicinity of the drains. The SR is modified by a Note. The Note indicates that only the CEQ fan room that has been entered need be inspected if the SR is being performed due to personnel entry in MODES I through 4. The Note precludes unnecessarily requiring inspection of both CEQ fan rooms if only one has been entered. SR 3.6.14.5 confirms that the CEQ fan room debris interceptors are installed and free of structural distress.
SR 3.6.14.5 also confirms that the flow opening at the pipe tunnel sump is not obstructed.
The 24 month frequency was developed considering such factors as the location and the design of the debris interceptors and flow opening.Cook Nuclear Plant Unit I B 3.6.14-5 Revision No. 18 Cook Nuclear Plant Unit 1 B 3.6.14-5 Revision No. 18 SGSVs B 3.7.2 BASES SURVEILLANCE REQUIREMENTS (continued)
The Frequency is in accordance with the Inservice Testing Program.This test is conducted in MODE 3 with the unit at operating temperature and pressure.
This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.
SR 3.7.2.2 This SR verifies that each SGSV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage. Th thes ei99pfie pei~eiffa REFERENCES
: 1. UFSAR, Section 10.2.2. UFSAR, Section 14.2.5.3. 10OCFR 100.11.4. Technical Requirements Manual 5. ASME, Operations and Maintenance Standards and Guides (OM Codes).
2 Cook Nuclear Plant Unit 1B3724ReionN.4 B 3.7.2-4 Revision No. 34 MFIVs and MFRVs B 3.7.3 BASES ACTIONS (continued) 0.1I With both the MFIV and MFRV inoperable in the same flow path, there is no redundant system to operate automatically and perform the required safety function.
Under these conditions, the affected flow path must be isolated within 8 hours. This action returns the system to the condition where at least one valve in each flow path is performing the required safety function.
The 8 hour Completion Time is reasonable, based on operating experience, to complete the actions required to close the MFIV or MFRV, or otherwise isolate the affected flow path.0.1 and 0.2 If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours and in MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.3.1 and SR 3.7.3.2 REQUIREMENTS These SRs verify that the closure time of each MFIV and MFRV is within the limit given in Reference 2 and is within that assumed in the accident and transient analyses.
The valve(s) may also be tested to more restrictive requirements in accordance with the Inservice Testing Program.The Frequency for this SR is in accordance with the Inservice Testing Program.SR 3.7.3.3 This SR verifies that each MFIV and MFRV can close on'an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage.
ra~~~
2 REFERENCES
: 1. UFSAR, Section 10.5.1.2.2. Technical Requirements Manual Cook Nuclear Plant Unit 1B3734ReionN.4 B3.7.3-4 Revision No. 34 SG PORVs B 3.7.4 BASES ACTIONS (continued)
C.1 and C.2 If the SG PORV(s) cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance upon steam generator for heat removal, within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.D.1 If one or more required SG PORVs are inoperable in MODE 4, the unit is in a degraded condition with reduced safety related means to cool the unit.to RHR entry conditions following an event, and the possibility of no means for conducting a cooldown with nonsafety related equipment since the condenser may be unavailable for use with the Steam Dump.System.
The seriousness of this condition requires that action be started immediately to restore the inoperable SG PORV(s) to OPERABLE status.SURVEILLANCE REQ U IREM ENTS SR 3.7.4.1 To perform a controlled cooldown of the RCS, the SG PORVs must be able to be opened remotely and throttled through their full range. This SR ensures that the SG PORVs are tested through a full control cycle nt ',cast ,e~ee--pr2-=me~ths.
Performance of inservice testing or use of an SG PORV during a unit cooldown may satisfy this requirement.
Insert 2 e
piL REFERENCES
: 1. UFSAR, Section 10.2.2.2. UFSAR, Section 14.2.4.Cook Nuclear Plant Unit 1 B 3.7.4-3 Revision No. 0 AFW System B 3.7.5 BASES ACTIONS (continued)
E.1 In MODE 4, either the reactor coolant pumps or the RHR loops can be used to provide forced circulation.
This is addressed in LCO 3.4.6, "RCS Loops -MODE 4." With one required AFW train inoperable, action must be taken to immediately restore the inoperable train to OPERABLE status. The immediate Completion Time is consistent with LCO 3.4.6.SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the required AFW System water and steam supply flow paths provides assurance that the proper flow paths will exist for AFW operation.
Verification of the AFW System water supply flow path includes both the suction (either a flow path from the CST or the Essential Service Water (ESWV) System) and discharge flow paths. This SR does.not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the-correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.The SR is modified by a Note that states one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AEW mode of operation, provided it is not otherwise inoperable.
This exception allows the system to be out of its normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s) inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, and these manual operations are an accepted function of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained.
T4 qtereyq-baee~f-ea~aef~g~~l 4-==Insert 2 ee fr-et=
Cook Nuclear Plant Unit 1 ..- evso o B3.7.5-7 Revision No. 0 AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.5.2.Verifying that each required AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 2).Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance.
Performance of inservice testing discussed in the ASME OM Code (Ref. 2) (only required at 3 month intervals) satisfies this requirement.
This SR is modified by a Note indicating that the SR should be deferred for the turbine driven AFW pump until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test at entry into MODE 3. At 850 psig, there is sufficient pressure to perform the test.SIR 3.7.5.3 This SIR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.T-he---24,-n' I Inse rt 2 The SR is modified by two Notes. Note 1 states that one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not otherwise inoperable.
This exception allows the AFW train(s) to be out of normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s) inoperable.
Cook Nuclear Plant Unit I B 3.7.5-8 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.7.5-8 Revision No. 0 AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued)
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, these manual operations are an accepted condition of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained.
Note 2 states that the SR is only required to be met in MODES 1, 2, and 3. It is not required to be met in MODE 4 since the AFW train is only required for the purposes of removing decay heat when the SG is relied upon for heat removal. The operation of the AFW train is by manual means and automatic startup of the AFW train is not required.SR 3.7.5.4 This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an ESFAS by demonstrating that each AEW pump starts automatically on an actual or simulated actuation signal in MODES 1, 2, and 3. Th- " ao on th Insert 2 we This SR is modified by three Notes. Note 1 indicates that the SR may be deferred for the turbine driven AFW pump until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test at entry into MODE 3. At 850 psig, there is sufficient steam pressure to perform the test. Note 2 states that one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not otherwise inoperable.
This exception allows the AFW train(s) to be out of normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s)inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, these manual operations are an accepted condition of the AFW System. OPERABILITY (i.e., the intended safety function)continues to be maintained.
Note 3 states that the SR is only required to be met in MODES 1, 2, and 3. It is not required to be met in MODE 4 since the AFW train is only required for the purposes of removing decay heat when the SG is relied upon for heat removal. The operation of the AFW train is by manual means and automatic startup of the AFW train is not required.Cook Nuclear Plant Unit 1 B3759Rvso o B 3.7.5-9 Revision No. 0 CST B 3.7.6 BASES ACTIONS (continued) adequate to ensure the backup auxiliary feedwater supply continues to be available.
The 7 day Completion Time is reasonable, based on an OPERABLE backup auxiliary feedwater supply being available, and the low probability of an event occurring during this time period requiring the CST.B.1 and B.2 If any Required Action and associated Completion Time cannot be met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance on the steam generator for heat removal, within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.7.6.1 This SR verifies that the CST contains, the required volume of cooling water. T h-yit~-peei eaee1he--,=
Insert 2 tee #,i~ae4 -ety eare REFERENCES
: 1. UFSAR, Section 10.5.2.2. UFSAR, Chapter 14.Cook Nuclear Plant Unit 1 ..- eiinN.2 B 3.7.6-3 Revision No. 26 CCW System B 3.7.7 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "'RCS Loops -MODE 4," be entered if an inoperable CCW train results in an inoperable RHR loop.This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function.The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period.B.1 and B.2 If the CCW train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions-from full power conditions inl an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.7.1 REQUIREMENTS T~his SR is modified by a Note indicating that the isolation of the CCW flow to individual components may render those components inoperable but does not affect the OPERABILITY of the CCW System.Verifying the correct alignment for manual, power operated, and automatic valves in the CCW flow path provides assurance that the proper flow paths exist for CCW operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.T-he e~~er~t~geti e Insert 2 wi-h th roo
~re-e d-rer Cook Nuclear Plant Unit 1 ..- evso o B 3.7.7-3 Revision No. 0 COW System B 3.7.7 BASES SURVEILLANCE REQUI REMENTS (continued)
SR 3.7.7.2 This SR verifies proper automatic operation of the CCW valves on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation.
This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. _______e____e_
lr~eee
:""Insert 2-anad tho ~ e SR 3.7.7.3 This SR verifies proper automatic operation of the COW pumps on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation.
T ~
Insert 2a-p~ euaea4thf~t~4aereutret-aet f-erat -a-9ee --
shwP-h{.es ee{suua~~e r'i~eeveaeeep-aef~reri.9 fllifty-~standIpe-fif.-
REFERENCES
: 1. UFSAR, Section 9.5.2. UFSAR, Table 9.5-3.REFERENCES
: 1. UFSAR, Section 9.5.2. UFSAR, Table 9.5-3.Cook Nuclear Plant Unit 1 ..- evso o B 3.7.7-4 Revision No. 0 ESW System B 3.7.8 BASES SURVEILLANCE REQUIREMENTS (continued) rather, it involves verification that those valves capable of being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
.I--nsert 2 SR 3.7.8.2 This SR verifies proper automatic operation of the ESW valves on an actual or simulated actuation signal. The ESW System is a normally operating system that cannot be fully actuated as part of normal testing.This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.u~# ~~etm" ' Insert SR 3.7.8.3 This SR verifies proper automatic operation of the required ESW pumps on an actual or simulated actuation
*lSwi'Nat tho-4=ertc e-qaay4,<--==Ilnsert 2 REFERENCES
: 1. UFSAR, Section 9.8.3.2. UFSAR, Section 9.8.3.2.3. UFSAR, Section 9.5.2.Cook Nuclear Plant Unit 1 ..- evso o B 3.7.8-4 Revision No. 0 UHS B 3.7.9 BASES SURVEILLANCE REQUIREMENTS (continued) determining the UHS temperature is averaging the available operating circulating water pumps discharge temperatures.
Insert 2 i--
r~a m -- 1 ......... --r REFERENCES
: 1. UFSAR, Section 10.6.2.2. UFSAR, Table 9.8-5.3. Regulatory Guide 1.27, Revision 2, January 1976.4. MD-12-ESW-1 06-N Assessment of Increased Lake Water Temperature on Safety Related and Non-Safety Related Systems.Cook Nuclear Plant Unit I1 ..- eiinN.5 B3.7.9-3 Revision No. 52 CREV System B 3.7.10 BASES ACTIONS (continued)
An alternative to Required Action E.1 is to immediately suspend activities that could result in a release of radioactivity that might require isolation of the CRE (Required Action E.2). This places the unit in a condition where the LCO does not apply. This does not preclude the movement of fuel to a safe position.F.1 During movement of irradiated fuel assemblies in the containment, auxiliary building, or Unit 2 containment, with two CREV trains inoperable, or with one or more CREV System trains inoperable due to inoperable CRE boundary, action must be taken immediately to suspend activities that could result in a release of radioactivity that requires isolation of the ORE. This places the unit in a condition that minimizes the accident risk.This does not preclude the movement of fuel to a safe position.G.1 If both CREV trains are inoperable in MODE 1, 2, 3, or 4 for reasons other than an inoperable ORE boundary or filter unit (i.e., Conditions B and C), the CREV System may not be capable of performing the intended function and the unit is in a condition outside the accident analyses.Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly.
As the environment and normal operating conditions on this system are not too severe, testing each  provides an adequate check of this system. Operating the CREV train, with flow through the HEPA filter and charcoal adsorber train, for> 15 minutes demonstrates the function of the CREV train. -Th -- Insert 2-e TGEI -eb-t heq 4pmeRP~ta FieeJn-ay.-
SR 3.7.10.2 This SR verifies that the required CREV System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum and maximum flow rate, and the physical properties of the activated charcoal.
Specific test Frequencies and additional information are discussed in detail in the VFTP.Cook Nuclear Plant Unit 1 B 3.7.10-6 Revision No. 23 Cook Nuclear Plant Unit 1 B 3.7.10-6 Revision No. 23 CREV System B 3.7.10 BASES SURVEILLANCE REQU IREM ENTS (continued)
SR 3.7.10.3 This SR verifies that each CREV train starts and operates on an actual or simulated actuation signal. The only actuation signal necessary to be verified is the Safety Injection (SI) signal, since the Control Room Radiation
-High signal is not assumed in the accident analysis.
A Note has been included that states the Surveillance is only required to be met in MODES 1, 2, 3, and 4, since these are the MODES the SI signal is assumed to start the CREV trains. The CREV trains are assumed to be manually started during a fuel handling accident.
GOfae-tij-extperde~ee Insert 2 pe~:rm ~
s-u eh-a-wy4 -er4 yhe-eae R-E --V-et9-S-'fet~jeetie*R-si&sect;1aekff SR 3.7.10.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 rem TEDE, the CRE occupants are protected from smoke, and analyses demonstrate that the CREV System is not needed to prevent exceeding hazardous~chemical limits. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences.
When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered. Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident.Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 4) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 5). These compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref. 6). Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status. There are no SRs to verity CREV System operability for hazardous chemicals or smoke.Cook Nuclear Plant Unit 1 B371- eiinN.2 B 3.7.10-7 Revision No. 23 CRAG System B 3.7.11 BASES ACTIONS (continued) 0.1 and 0.2 During movement of irradiated fuel, if the inoperable CRAG train cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CRAC train must be placed in operation immediately.
This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that active failures will be readily detected.An alternative to Required Action 0.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room (Required Action C.2). This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.D. 1 During movement of irradiated fuel assemblies, with two CRAG trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.E. 1 If both CRAG trains are inoperable in MODE 1, 2, 3, or 4, the CRAG System may not be capable of performing its intended function.Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.7.11.1 and SR 3.7.11.2 REQUI REM ENTS These SRs verify that the heat removal capability of each CRAG train is sufficient to maintain control room air temperature
< 850&deg;F.
-Insert 2~ 4.t -tf~~et~c-~gca lg-{a~t ee~hst~ir-pvertis-dr-fhe -pe~~~-lyvrfe-a1s3a~ta-wt4e d e e. -Fr.eeueia c-y-4ef-t GRE-System.
REFERENCES
: 1. UFSAR, Section 9.10.Cook Nuclear Plant Unit 1B3.113RvsoN.0 B 3.7.11-3 Revision No. 0 ESF Ventilation System B 3.7.12 BASES ACTIONS A.1 With one ESF Ventilation train inoperable, action must be taken to restore OPERABLE status within 7 days. During this time, the remaining OPERABLE train is adequate to perform the ESF Ventilation System function.The 7 day Completion Time is appropriate because the risk contribution is less than that for the ECCS (72 hour Completion Time), and this system is not a direct support system for the ECCS. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.
B.1 If the ESF enclosure boundary is inoperable, the ESF Ventilation trains cannot perform their intended functions.
Actions must be taken to restore an OPERABLE ESF enclosure boundary within 24 hours. During the period that the ESF enclosure boundary is inoperable, appropriate compensatory measures consistent with the intent, as applicable, of GDC 19, 60, 64 and 10 CFR Part 100 should be utilized to protect plant personnel from potential hazards. Preplanned measures should be-available to address these concerns for intentional and unintentional entry- -into the condition.
The 24 hour Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures.
The 24 hour Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the ESF enclosure boundary.C.1 and C.2 If the ESF Ventilation train or ESF enclosure boundary cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly.
t-ader9I96-epe-aagc-e*dlte~s Insert 23t Cook Nuclear Plant Unit 1B37123RvsoN.0 B 3,7.12-3 Revision No. 0 ESF Ventilation System B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued)
V-a4a~at-~ -apata~ E SR 3.7.12.2 This SR verifies that the required ESF Ventilation System testing is performed in accordance with the Ventilation Filter Testing Program (V FTP). The VFTP includes testing HEPA filter performance, charcoal adsorbers efficiency, minimum and maximum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
Specific test Frequencies and additional information are discussed in detail in the VFTP.SR 3.7.12.3 This SR verifies that each ESF Ventilation train starts and operates on an actual or simulated actuation signal. One ESF Ventilation train is normally operating with flow bypassing the charcoal adsorber section.This test confirms that each train, when in standby, starts upon receipt of a Containment Pressure -High High signal and that the exhaust flow can be directed through the entire filter unit including the HEPA filter and charcoal adsorber section. Insert 2 SR 3.7.12.4 This SR verifies the integrity of the ESF enclosure.
The ability of the ESF enclosure to maintain a negative pressure, with respect to potentially uncontaminated adjacent areas, is periodically tested to verify proper functioning of the ESF Ventilation System. During the post accident mode of operation, the ESF Ventilation System is designed to maintain a slight negative pressure in the ESF enclosure, with respect to adjacent areas, at a flowrate -< 22,500 cfm to prevent unfiltered leakage. T-the- Insert 2aSTA E-- 4s Cook Nuclear Plant Unit 1B37.24RvsoN.0 B 3.7.12-4 Revision No. 0 FHAEV System B 3.7.13 BASES APPLICABILITY During movement of irradiated fuel in the auxiliary building, the FHAEV System is required to be OPERABLE to alleviate the consequences of a fuel handling accident.In MODE 1, 2, 3, 4, 5, or 6, the FHAEV Systemis not required to be OPERABLE since the FHAEV System is only credited during a fuel handling accident in the auxiliary building.ACTIONS LCO 3.0.3 is not applicable while in MODE 5 or 6. However, since irradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operations.
Entering LCO 3.0.3, while in MODE 1, 2, 3, or 4 would require the unit to be shutdown unnecessarily.
A.1 When the required FHAEV train is inoperable or not in operation during movement of irradiated fuel assemblies in the auxiliary building, action must be taken to place the unit in a condition in which the LCO does not apply. Action must be taken immediately to suspend movement of irradiated fuel assemblies in the auxiliary building.
This does not preclude* the movement of fuel to a safe position.SURVEILLANCE REQUIREMENTS SR 3.7.13.1 This SR requires verification-every4j2-heu~s~that the required FHAEV train is operating with flow through the filter unit, including the HEPA filter and charcoal adsorber section. Verification includes fan status and also verifies that each charcoal bypass damper is closed. T-he-F~eq*ei~ey-ef~-
Insert 2 SR 3.7.13.2 Standby systems should be checked periodically to ensure that they.function properly.
As the environmental and normal operating conditions on this system are not severe, testing each provides an adequate check on this system.Operating the required FHAEV train, with flow through the HEPA filter and charcoal adsorber train, for > 15 minutes demonstrates the function of the system. e+eil ~--f-th- I.nsert 2 Cook Nuclear Plant Unit 1 B 3.7.13-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.7.13-3 Revision No. 0 FHAEV System B 3.7.13 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.13.3 This SR verifies that the required FHAEV System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum and maximum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
Specific test frequencies and additional information are discussed in detail in the VFTP.SR 3.7.13.4 This SR verifies that the required FHAEV train actuates on an actual or simulated actuation signal. The test must verify that the signal automatically shuts down each of the Fuel Handling Area Supply Air System fans.
~Insert 2 T-h e-ert-e--Ffeqe~ec--y4-i each -ac-~e--fe4-a-r-eib;#vsandoJ4nt.
SR 3.7.13.5 This SR verifies the integrity of the auxiliary building enclosure.
The ability of the pool storage area to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the FHAEV train. During the accident mode of operation, the FHAEV train is designed to maintain a slight negative pressure in the FHAEV train, to prevent unfiltered leakage. The FHAEV train is designed to maintain a pressure > 0.125 inches of vacuum water gauge with respect to atmospheric pressure at a flow rate of< 27,000 cfm. Insert 2r#t REFERENCES
: 1. UFSAR, Section 9.9.3.2.2. UFSAR, Section 14.2.1.3. 10CFR100.Cook Nuclear Plant Unit 1 ..34ReiinN.2 B 3.7.13-4 Revision No. 26 Fuel Storage Pool Water Level B 3.7.14 BASES ACTIONS A...1 When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring.
When the fuel storage pool water level is lower than the required level, the movement of irradiated fuel assemblies in the fuel storage pool is immediately suspended to a safe position.
This action effectively precludes the occurrence of a fuel handling accident.
This does not preclude movement of a fuel assembly to a safe position.Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations.
Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event of a fuel handling accident.
The water level in the fuel storage pool must be checked periodically.
T-he- ~ -e~-t4et- nst 2 REFERENCES
: 1. UFSAR, Section 9.7.2.2. UFSAR, Section 9.4.3. UFSAR, Section 14.2.1.4. 10CFR 100.11.Cook Nuclear Plant Unit 1 B371- eiinN.2 B 3.7.14-2 Revision No. 26 Fuel Storage Pool Boron Concentration B 3.7.15 BASES LCO LCO The fuel storage pooi boron concentration is required to be > 2400 ppm.The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference
: 2. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel storage pool.APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool, until a complete spent fuel storage pool verification has been performed following the last movement of fuel assemblies in the spent fuel storage pool. This LCO does not apply following the verification, since the verification would confirm that there are no misloaded fuel assemblies.
With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.ACTIONS A.1, A.2.1, and A.2.2 When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.This is most efficiently achieved by immediately suspending the movement of fuel assemblies.
The initiation of action to restore the concentration of boron to within limit occurs simultaneously with suspending movement of fuel assemblies.
Alternatively, beginning a verification of the fuel storage pool fuel locations, to ensure proper locations of the fuel, can be performed.
However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored.
This does not preclude movement of a fuel assembly to a safe position.The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply. If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable.
If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.
Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE REQ U IREM ENTS SR 3.7.15.1 This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed.
The9--da-y-F-f-eaieIycs-aie~-3pep~ete Insert 2-fpe ~tri-&#xa2;l oer-eucq4~-a -she f4ifne. " Cook Nuclear Plant Unit 1B37152RvsoN.0 B 3.7.15-2 Revision No. 0 Secondary Specific Activity B 3.7.17 BASES ACTIONS A.1 and A.2 Specific activity of the secondary coolant exceeding the allowable value is an indication of a problem in the ROS and contributes to increased post accident doses. If the secondary specific activity is not within limits, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.17.1 REQ U IREM ENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis.
A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases.
It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. e fee~~ase-r4e~tc4P
<=.- Insert 2 f -appr-epfiete-aet~iei94-e ho.tako,-to mift ~  REFERENCES
: 1. 10OCFR 100.11.2. 10 CFR 50, Appendix A, G DC 19.3. UFSAR, Section 14.2.7.Cook Nuclear Plant Unit 1B371-3RvsoN.0 B 3.7.17-3 Revision No. 0 AC Sources -Operating B 3.8.1 BASES ACTIONS (continued) degraded level, any further losses in the AC electrical power system will cause a loss of function.
Therefore, no additional time is justified for continued operation.
The unit is required by LCO 3.0.3 to commence a controlled shutdown.SURVEILLANCE REQUIREMENTS The AC sources are designed to permit inspection and testing of all important areas and features, especially those that have a standby function, in accordance with Plant Specific Design Criterion (PSDC) 39 (Ref. 8). Periodic component tests are supplemented by extensive functional tests during refueling outages (under simulated accident conditions).
The SRs for demonstrating the OPERABILITY of the DGs are in accordance with the recommendations of Regulatory Guide 1.9 (Ref. 3), Regulatory Guide 1.108 (Ref. 9), Regulatory Guide 1.137 (Ref. 10), and IEEE Standard 387-1995 (Ref. 11) as addressed in the applicable SR discussion.
Where the SRs discussed herein specify voltage and frequency tolerances, the following is applicable.
The minimum steady state output voltage of 3910 V is 94% of the nominal 4160 V output voltage. This value allows for voltage drop to the terminals of 4160 V motors whose minimum operating voltage is specified as 90% or 3740 V. It also allows for voltage drops to motors and other equipment down through the 120 V level where the minimum operating voltage is also usually specified as 90% of nameplate rating. The specified maximum steady state output voltage of 4400 V is equal to the maximum operating voltage specified for 4000 V motors. It ensures that for a lightly loaded distribution system, the voltage at the terminals of 4000 V motors is no more than the maximum rated operating voltages.
The specified minimum and maximum steady state frequencies of the DG are 59.4 Hz and 60.5 Hz, respectively.
These values ensure .the ESF pumps can achieve adequate fluid flow to meet their safety and accident mitigation functions.
SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that the required qualified offsite circuits are OPERABLE, and that appropriate independence of offsite circuits is Isr maintained.
<" Ine"-
.
Cook Nuclear Plant Unit I B 3.8.1-16 Revision No. 41 Cook Nuclear Plant Unit 1 B 3.8.1-16 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SIR 3.8.1.2 and SR 3.8.1.8 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.
To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 1 for SR 3.8.1.2 and Note for SR 3.8.1.8) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and followed by a warmup period prior to loading.For the purposes of SR 3.8.1.2 and SR 3.8.1.8 testing, the DGs are started from standby conditions.
Standby conditions for a DG means that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations.
In order to reduce stress and wear on diesel engines, the manufacturer recommends a modified start in which the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 2.SR 3.8.1.8 requires DG starts from standby conditions and achieves required voltage and frequency within 10 seconds. The 10 second start requirement supports the assumptions of the design basis LOCA analysis in the UFSAR, Section 14.3 (Ref. 5).The 10 second start requirement is not applicable to SIR 3.8.1.2 (see Note 2 of SR 3.8.1.2) when a modified start procedure as described above is used. If a modified start is not used, the 10 second start requirement of SR 3.8.1.8 applies.Since SR 3.8.1.8 requires a 10 second start, it is more restrictive than SIR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2.addition, the DG is required to maintain proper voltage and frequency limits after steady state is achieved.
The voltage and frequency limits are normally achieved within 10 seconds. The time for the DG to reach steady state operation, unless the modified DG start method is employed, is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance.
T-he-3-ely-e-ffS -..4.2-e-n{=wi-teat--w{
.4-.. Insert 2 Cook Nuclear Plant Unit 1 B3811 eiinN.4 B 3.8.1-17 Revision No. 41 AC Sources -. Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.3 Consistent with Regulatory Guide 1.9 (Ref. 3), this Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads 90% to 100% of the continuous rating of the 0G. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.The 0.8 value is the design rating of the machine, while the 1.0 is an operational goal to ensure circulating currents are minimized.
The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections being required in order to maintain OG reliability.eil~ee4 I nse rt 2 This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized.
Note 2 states that momentary transients, because of changing bus loads, do not invalidate this test. Note 3 indicates that this Surveillance should be conducted on only one Unit 1 DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.
Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
SR 3.8.1.4 This SR provides verification that the level of fuel oil in the day tank is above the level at which fuel oil is automatically added. The level is expressed as an equivalent volume in gallons, of which 31.4 gallons is unusable (due to tank geometry and vortexing considerations) and 70 gallons is usable, and is selected to ensure adequate fuel oil for greater than 15 minutes of DG operation at full load.Cook Nuclear Plant Unit I B 3.8.1-18 Revision No. 41 Cook Nuclear Plant Unit 1 B 3.8.1-18 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
T-r #~~py <- Insert 2 I~ u..... ..... F ....SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from each fuel oil day tank oee~ve ytl 4 Mays eliminates the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainrierht in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by " ......bacteria.
Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. T-he--S vei4Ifee=-Feileneies--fe-est-abtished-ty
.==. Ir Ri s~ -e-e~ h rave eee-ca4ee4f4his4Iv
-,eiteariee.
SR 3.8.1.6 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available.
While the system design requirements provide for two engine start cycles from each of the two air start receivers associated with each DG without recharging, only one start sequence is required to meet the OPERABILITY requirements (since the accident analysis assumes the DG starts on the first attempt).The pressure specified in this SR reflects the lowest value at which one DG start can be accomplished with one air start receiver.nsert 2
~ 2 SR 3.8.1.7 This Surveillance demonstrates that each required fuel oil transfer pump (one per fuel oil transfer system) operates automatically and transfers fuel oil from its associated storage tank to its associated day tank. This is required to support continuous operation of standby power sources. This Cook Nuclear Plant Unit 1 B 3.8.1-19 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.Tho
.v ueaev- Insert 2... ...O M ...... J ... ..SR 3.8.1.9 Automatic transfer of each 4.16 kV emergency bus power supply from normal auxiliary circuit to the preferred offsite circuit and the manual alignment to the alternate required offsite circuit demonstrates the OPERABILITY of the required offsite circuit to power the shutdown Ioa(T-44---m @rth-F-r~enev oef th"+S-u'~ia-es4base-ei~aerc=e the ds..---Insert 2 t-e8ho-ec -hcu ath4hese.eei drmd{-e2-q4eref fre4-a-etia ntia As noted (Note 1Ito SR 3.8.1.9), SR 3.8.1.9.a is only required to be met when the auxiliary source is supplying the onsite electrical power subsystem.
This is acceptable since the preferred offsite source would be supplying the onsite electrical power subsystem and a transfer would not be necessary.
SR 3.8.1.10 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding a predetermined frequency and while maintaining a specified margin to the overspeed trip. Voltage and frequency are also verified to reach steady state conditions within 2 seconds. For this unit, the single load for each DG is 600 kW. This Surveillance may be accomplished by: a. Tripping the UG output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while paralleled to offsite power, or while solely supplying the bus; or Cook Nuclear Plant Unit 1 ..-0ReiinN.4 B 3.8.1-20 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
: b. Tripping its associated single largest post-accident load with the DG solely supplying the bus.Consistent with Regulatory Guide 1.9 (Ref. 3), the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the overspeed trip setpoint, or 15% above nominal speed, whichever is lower. This corresponds to 64.4 Hz, which is the nominal speed plus 75% of the difference between nominal speed and the overspeed trip setpoint.The time, voltage, and frequency tolerances specified in this SR are derived from Regulatory Guide 1.9 (Ref. 3) recommendations for response during load sequence intervals.
The2 seconds specified is equal to approximately 60% of the 3.49 second load sequence interval associated with sequencing of the largest load. The voltage and frequency specified are consistent with the design range of the equipment powered by the 0G. SR 3.8.1.10.a corresponds to the maximum frequency excursion, while SR 3.8.1.10.b and SR 3.8.1.10.c are steady state voltage and frequency values to which the system must recover following load rejection.
The- Nt-~eeey-ie 4asee-er+
a-hs-Insert 2 eopo9e{uafrp rt1S R-w vhenrpl~efof r-ed-at=theQ'rnn F-ru --
This SR is modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems.This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Cook Nuclear Plant Unit I B 3.8.1-21 Revision No. 41 Cook Nuclear Plant Unit 1 B 3.8.1-21 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Credit may be taken for unplanned events that satisfy this SR. Credit may be taken for unplanned events that satisfy this SR.Note 2 ensures that the OG is tested under load conditions that are as close to design basis conditions as possible.
When synchronized with offsite power, testing should be performed at a power factor of < 0.86.This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions.
Under certain conditions, however, Note 2 allows the Surveillance to be conducted at a power factor other than _< 0.86. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to -< 0.86 results in voltages on the emergency busses that are too high.Under these conditions, the power factor should be maintained as close as practicable to 0.86 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the DG excitation levels needed to obtain a power factor of 0.86 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained as close as practicable to 0.86 without exceeding the DG excitation limits.SR 3.8.1.11 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.8, this Surveillance demonstrates the DG capability to reject a full load (90% to 100% of the DG continuous rating) without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping.
This Surveillance ensures proper engine generator load response under the simulated test conditions.
This test simulates the loss of the total connected load that the OG experiences following a full load rejection and verifies that the OG does not trip upon loss of the load. These acceptance criteria provide for OG damage protection.
While the DG is not expected to experience this transient during an event and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated.Insert 2 t t er reure --~e-fr This SR has been modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could Cook Nuclear Plant Unit 1 ..-2ReiinN.4 B 3.8.1-22 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) the DG to automatically achieve the required voltage and frequency within the specified time.The DG autostart time of 10 seconds is derived from requirements of the accident analysis to respond to a design basis large break LOCA. The Surveillance should be continued for a minimum, of 5 minutes in order to demonstrate that all starting transients have decayed and stability is achieved.The requirement to verify the connection and power supply of permanent and autoconnected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation.
For instance, Emergency Core Cooling Systems (ECCS) injection valves are not desired to be stroked open, or centrifugal charging trains are not capable of being operated at full flow, or residual heat removal (RHR)trains performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation.
In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG systems to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified..Th--eu 4m~h aeae j~e4jC In sert 2 This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4, is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Cook Nuclear Plant Unit 1 B3812 eiinN.4 B 3.8.1-24 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.13 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.5, this Surveillance demonstrates that the DO automatically starts and achieves the required voltage and frequency within the specified time (10 seconds)from the design basis actuation signal (ESE actuation signal). In addition, the DO is required to maintain proper voltage and frequency limits after steady state is achieved.
The voltage and frequency limits are normally achieved within 10 seconds. The time for the DG to reach the steady state voltage and frequency limits is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance.
The DO is required to operate for > 5 minutes. The 5 minute period provides sufficient time to demonstrate stability.
SR 3.8.1.13.d and SR 3.8.1.13.e ensure that permanently connected loads and emergency loads are energized from the offsite electrical power system on an ESF signal without loss of offsite power.The requirement to verify the connection of permanent and auto-connected loads is intended to satisfactorily show the relationship of these loads to the DO loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation.
For instance, ECCS injection valves are not desired to be stroked open, or centrifugal charging trains are not capable of being operated at full flow, or RHR trains performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation.
In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DO system to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.T4 e Insert 2 Qpra4n-~-e4 t Cook Nuclear Plant Unit 1 B 3.8.1-25 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR is modified by two Notes. The reason for Note 1 s to minimize wear and tear on the DGs during testing. For the purpose of this testing,*the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.14 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.12, this Surveillance demonstrates that DG noncritical protective functions (e.g., low lube oil pressure) are bypassed on a loss of voltage signal or an ESF actuation test signal. The noncritical trips are bypassed during DBAs and provide an alarm on an abnormal engine condition.
This alarm provides the operator with sufficient time to react appropriately.
The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the 0G.
2a S--hn1~
~dd4 The SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DG from service. This Cook Nuclear Plant Unit 1 .:-6ReiinN.4 B 3.8:1-26 Revision No. 41 AC Sources -Operating B 3.8.1 B]ASES SURVEILLANCE REQUIREMENTS (continued) restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERAB]ILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERAB]ILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.15 This Surveillance demonstrates the DGs can start and run continuously at full load capability (90% to 100% of the OG continuous rating) for an interval of not less than 8 hours. The run duration of 8 hours is consistent with IEEE Standard 387-1995 (Ref. 11). The DG starts for this Surveillance can be performed either from standby or hot conditions.
The provisions for prelubricating and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.The load band is provided to avoid routine overloading of the DG.Routine overloading may result in more frequent teardown inspections being required in order to maintain DG reliability.tt g == Insert 2*SR-whe-e e-24-met-h-F-r-eqtuenaey-T-her-efer-ertbe--
..-
This Surveillance is modified by three Notes. Note 1 states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the power factor limit will not invalidate the test. The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in Cook Nuclear Plant Unit 1 ..-7ReiinN.4[] 3.8.1-27 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR. Note 3 ensures that the DG is tested under load conditions that are as close to design basis conditions as possible.
When synchronized with offsite power, testing should be performed at a power factor of < 0.86. This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions.
Under certain conditions, however, Note 3 allows the Surveillance to be conducted as a power factor other than < 0.86. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to < 0.86 results in voltages on the emergency busses that are too high.Under these conditions, the power factor should be maintained as close as practicable to 0.86 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the OG excitation levels needed to obtain a power factor of 0.86 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained close as practicable to 0.86 without exceeding the DG excitation limits.SR 3.8.1.16 This Surveillance demonstrates that the diesel engine can restart from a hot condition, such as subsequent to shutdown from normal Surveillances, and achieve the required voltage and frequency within 10 seconds. The 10 second time is derived from the requirements of the accident analysis to respond to a design basis large break LOCA. The 2 Fr-gt c " -e-ey eei3- gmet~.SRwel ~
e Cook Nuclear Plant Unit 1 ..-8ReiinN.4 B 3.8.1-28 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
* This SR is modified by two Notes. Note 1 ensures that thetest is performed with the diesel sufficiently hot. The load band is provided to avoid routine overloading of the DG. Routine overloads may result in more frequent teardown inspections being required in order to maintain DG reliability.
The requirement that the diesel has operated for at least 2 hours at full load conditions prior to performance of this Surveillance is based on operating experience for achieving hot conditions.
Momentary transients due to changing bus loads do not invalidate this test. Note 2 allows all OG starts to be preceded by an engine prelube period to minimize wear and tear on the diesel during testing.SR 3.8.1.17 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.1 1, this Surveillance ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and the DG can be returned to ready-to-load status when offsite power is restored.
It also ensures that the auto-start logic is reset to allow the DG to reload if a subsequent loss of offsite power occurs. The DG is considered to be in ready-to-load status when the DG is running at rated speed and voltage, with the DG output breaker open.
* iat, Insert 2 This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1, 2, 3, or 4.Cook Nuclear Plant Unit I B 3.8.1-29 Revision No. 41 Cook Nuclear Plant Unit 1 B 3.8.1-29 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.18 Under accident conditions loads are sequentially connected to the bus by the individual time delay relays. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs or RATs (as applicable) due to high motor starting currents.Verifying the load sequencer time within plus or minus 5% of its required value ensures that sufficient time exists for the DG to restore frequency and voltage and RATs to restore voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated.
Reference 4 provides a summary of the automatic loading of emergency buses.
~
t Insert 2 i~-e e a This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these.outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 1 B3813 eiinN.4 B 3.8.1-30 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.19 In the event of a DBA coincident with a loss of offsite power, the OGs are required to supply the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded.This Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.12, during a loss of offsite power actuation test signal in conjunction with an ESF actuation signal. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
-.--Insert 2eefrm ta-usuIasswn SRw iii e fj-ii~
il 1~uI~ .TI ,-efithe-Fr~eueney This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for D~s. The reason for Note 2 is that the performance of the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit I1 ..-1ReiinN.4 B 3.8.1-31 Revision No. 41 AC Sources -Operating B 3.8.1 ASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.20 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing that involves connecting the DG to its test load resistor bank, and the DG will automatically reset to ready to load operation if a ESF actuation signal is received during operation in the test mode. Ready to load operation is defined as the DG running at rated speed and voltage with the DG output breaker open.The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.13.
The intent in the requirement associated with SR 3.8.1 .20.b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified..T Insert 2 This SR is modified by two Notes. Note 1 states that this Surveillance is only required to be met when the applicable DG is connected to its test load resistor bank. This is allowed since the test mode override only functions when the DG is connected to its associated test load resistor bank. When the OG is not connected to its associated test load resistor bank, the feature is not necessary; thus the Surveillance is not required to be met under this condition.
The reason for Note 2 is that performing the Surveillance would remove a required DG from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider.
the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied' together or operated Cook Nuclear Plant Unit 1 B3813 eiinN.4 B 3.8.1-32 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SIR.SR 3.8.1.21 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing and the DG will automatically reset to ready to load operation if a LOCA actuation signal is received during operation in the test mode. Ready to load operation is defined as the DG running at rated speed and voltage with the DG output breaker open.The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.13.
The intent in the requirement associated with SR 3.8.1.21.b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.nsert 2@eea~toa=atm 4J~e ~r4 (; 4~.e4e4a--a~heaht.eee~aa JaetFrel re 4e~ et F~q ~ -eeaelt)4 a This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DG from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit~safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation Cook Nuclear Plant Unit 1 ..-3ReiinN.4 B 3.8.1-33 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
The~se shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.22 This Surveillance demonstrates that the DG starting independence has not been compromised.
Also, this Surveillance demonstrates that each engine can achieve proper speed within the specified time when the DGs are started simultaneously.,e~-ee~i mmaa~ne=-
Insert 2 This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
SR 3.8.1.23 With the exception of this Surveillance, all other Surveillances of this Specification (SR 3.8.1.1 through SR 3.8.1.22) are applied to Unit I sources. This Surveillance is provided to direct that appropriate Surveillances for the required Unit 2 AC sources are governed by the*applicable Unit 2 Technical Specifications.
Performance of the applicable Unit 2 Surveillances will satisfy the Unit 2 requirements as well as satisfy this Unit 1 Surveillance Requirement.
Exceptions are noted to the Unit 2 SRs of LCO 3.8.1. SR 3.8.1.9.b is not required to be met since only one offsite circuit is required to be OPERABLE.
SR 3.8.1.13, SR 3.8.1.14 (ESF actuation signal portion only), SR 3.8.1.19, SR 3.8.1.20, and SR 3.8.1.21 are not required to be met because the ESF actuation signal is not required to be OPERABLE.
SR 3.8.1.22 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE.The Frequency required by the applicable Unit 2 SR also governs performance of that SR for Unit 1.Cook Nuclear Plant Unit 1 ..-4ReiinN.4 B 3.8.1-34 Revision No. 41 Diesel Fuel Oil B 3.8.3 BASES SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load.The 7 day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.e -Insert 2 SR 3.8.3.2 The tests listed below are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion.
If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage tank(s). The tests, limits, and applicable ASTM Standards are as follows: a. Sample the new fuel oil in accordance with ASTM D4057-81 (Ref. 5);b. Verify that the sample has: (1) when tested in accordance with ASTM D1298-80 (Ref. 5) an absolute specific gravity at 60/60&deg; F of > 0.82 and _< 0.88, an API gravity at 60&deg;F of -> 30&deg; and < 400, an API gravity of within 0.3 degrees at 60&deg;F when compared to the supplier's certificate, or a specific gravity of within 0.0016 at 60/600 when compared to the supplier's certificate; (2) a kinematic viscosity at 40&deg;C of > 1.9 centistokes and -< 4.1 centistokes or Saybolt viscosity at 100 0 F of-> 32.6 and <40.1, if gravity was not determined by comparison with supplier's certification, when tested in accordance with ASTM 975-81 (Ref. 5); and (3) a flash point of > 125&deg;F when tested in accordance with ASTM D975-81 (Ref. 5); and c. Verify that the new fuel oil has a clear and bright appearance with proper color when tested in accordance with ASTM D4176-82 (Ref. 5).Failure to meet any of the above limits is cause for rejecting the new fuel oil, but does not represent a failure to meet the LCO concern since, the fuel oil is not added to the storage tanks.Following the initial new fuel oil sample, the fuel oil is analyzed within 31 days following addition of the new fuel oil to the fuel oil storage tank(s)to establish that the other properties specified in Table 1 of Cook Nuclear Plant Unit 1 B3.8.3-4 Revision No. 0 Diesel Fuel Oil B 3.8.3 BASES SURVEILLANCE REQUIREMENTS (continued)
ASTM D975-81 (Ref. 6) are met for new fuel oil when tested in accordance with ASTM D975-81 (Ref. 5), except that the analysis for sulfur may be performed in accordance with ASTM D2622-82 (Ref. 5).The 31 day period is acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an immediate effect on DG operation.
This Surveillance ensures the availability of high quality fuel oil for the DGs.Fuel oil degradation during long term storage shows up as an increase in particulate, due mostly to oxidation.
The presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause fouling of filters and fuel oil injection equipment, however, which can cause engine failure.Particulate concentrations should be determined in accordance with ASTM D2276-83, Method A (Ref. 5). This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/I. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing.The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.
SR 3.8.3.3 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must*have a water environment in order to survive. Removal of water from the fuel storage tanks-~e~c -ve'c~3-1--Eleys eliminates the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, and contaminated fuel oil, and from breakdown of the fuel oil by bacteria.
Frequent checking for and removal of accumulated water minimizes fouling and provides data. regarding the watertight integrity of the fuel oil system. The Surveillance Frequencies are established by Regulatory Guide 1.137 (Ref. 2). This SR is for preventive maintenance.
The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during performance of the Surveillance.
.-=Insert 2 Cook Nuclear Plant Unit 1 ..- evso o B3.8.3-5 Revision No. 0 DC Sources -Operating B 3.8.4 BASES ACTIONS (continued)
E.1 If one or both required Unit 2 Train A and Train B DC electrical power subsystems are inoperable, the associated ESW train(s) are not capable of performing their intended function.
Immediately declaring the affected supported feature, e.g., ESW train, inoperable allows the ACTIONS of LCO 3.7.8 to apply appropriate limitations on continued reactor operation.
SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the battery chargers, which support the ability of the batteries to perform their intended function.
Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state while supplying the continuous steady state loads of the associated DC subsystem.
On float charge, battery cells will receive adequate current to optimally charge the battery. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the minimum float voltage established by the battery manufacturer (2.20 Vpc or 255.2 VDC at the battery terminals of the Train A and Train B batteries and 2.20 Vpc or 257.4 VDC for the Train N battery).
This voltage maintains the battery plates in a condition that supports maintaining the grid life (expected to be approximately 20 years). -s~eieer-te 4,====-I nse rt 2 SR 3.8.4.2 This SR verifies the design capacity of the battery chargers.
According to Regulatory Guide 1.32 (Ref. 9), the battery charger supply is recommended to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences.
The minimum required amperes and duration ensure that these requirements can be satisfied.
This SR requires that each Train A and Train B required battery charger be capable of supplying
> 300 amps at > 250 VDC for > 4 hours and the Train N battery charger is capable of supplying
> 25 amps at > 250 VDC for > 4 hours. The ampere requirements are based on the output rating of the chargers.
The voltage requirements are based on the charger voltage Cook Nuclear Plant Unit 1 B3.8.4-7 Revision No. 0 DC Sources -Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued) level after a response to a loss of AC power. The time period is sufficient to detect significant charger failures.
ai Insert 2
~.t~ dh~
~ ~  SR 3.8.4.3 A battery service test is a special test of the battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The battery charger must be disconnected throughout the performance of the battery service test. The discharge rate and test length should correspond to the design duty cycle requirements as specified in the applicable design documents.
<.=-=- I nse rt 2 j~g
~ rif9el rec a-h ,6me~atsuaIe~steRMe~
~rm This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu .of a service test.The reason for Note 2 is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for- the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced.
This. assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 1 B 3.8.4-8 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE SR 3.8.6.1 REQUIREMENTS Verifying battery float current while on float charge is used to determine the state of charge of the battery. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a charged state.The float current requirements are based on the float current indicative of a charged battery. Use of float current to determine the state of charge of the battery is consistent with IEEE-450 (Ref. 1).
Insert 2ece~rete*
~~e-=e ~re~ e e(R This SR is modified by a Note that states the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.8.4.1. When this float voltage is not maintained the Required Actions of LCO 3.8.4 ACTION A are being taken, which provide the necessary and appropriate verifications of the battery condition.
Furthermore, the float current limit of 2 amps is established based on the nominal float voltage value and is not directly applicable when this voltage is not maintained.
SR 3.8.6.2 and SR 3.8.6.5 Optimal long term battery performance is obtained by maintaining a float voltage greater than or equal to the minimum established design limits provided by the battery manufacturer, which corresponds to 257.5 VDC for a 116 cell battery and 259.7 VDC for a 117 cell battery at the battery terminals, or 2.22 Vpc. This provides adequate over-potential, which limits the formation of lead sulfate and self discharge, which could eventually render the battery inoperable.
Float voltages in this range or less, but greater than 2.07 Vpc, are addressed in Specification 5.5.15." SRs 3.8.6.2 and 3.8.6.5 require verification that the cell float voltages are equal to or greater than the short term absolute minimum voltage of 2.07 V. T- Insert 2
=--SR 3.8.6.3 The limit specified for electrolyte level (i.e., greater than or equal to the low level mark) ensures that the plates suffer no physical damage and maintains adequate electron transfer capability.
T:he e -Insert 2 Cook Nuclear Plant Unit 1 ..- evso o B3.8.6-5 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.6.4 This Surveillance verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (i.e., 60&deg;F for the Train A and Train B 250 VDC batteries and 45&deg;F for the Train N 250 VDC battery).
Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the design requirements.
Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity.
Insert 2 SR 3.8.6.6 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.6.6;however, only the modified performance discharge test may be used to satisfy the battery service test requirements of SR 3.8.4.3.A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity.
Initial conditions for the modified performance discharge test should be identical to those specified for a performance discharge test as specified in IEEE-450 (Ref. 1).It may consist of just two rates: for instance the one minute rate for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the modified performance discharge test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test must remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.. Currently, the modified performance discharge test is performed by testing the battery using the service test profile for the first 4 hours followed by the performance discharge test profile for the Cook Nuclear Plant Unit 1 ..- evso o B 3.8.6-6 Revision No. O Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued) remainder of the test. This method has been determined by the system engineer and the battery manufacturer to be an acceptable modified performance test procedure, and is consistent with IEEE-450 (Ref. 1).The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 1) and IEEE-485 (Ref. 3). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements..
Furthermore, the battery is sized to meet the assumed duty cycle loads when the battery design capacity reaches this 80% limit. Insert 2 TaeS~i4 I f th e battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity > 100% of the manufacturer's ratings. Degradation is indicated, according to IEEE-450 (Ref. 1), when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is below 90% of the manufacturer's rating.The 12 month and 60 month Frequencies are consistent with the recommendations in IEEE-450 (Ref. 1). The 24 month Frequency is derived from the recommendations of IEEE-450 (Ref. 1).This SR is modified by a Note. The reason for the Note is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial*Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1,2, 3 or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 1 ..- evso o B 3.8.6-7 Revision No. 0 lnverters
-Operating B 3.8.7 BASES'ACTIONS (continued) inverter inoperabilit~y.
This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail. When the 120 VAC vital bus is powered from its regulated 600/120 VAC transformer, it is relying upon interruptible AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the 120 VAC vital buses is the preferred source for powering instrumentation trip setpoint devices.B.1 With two inverters in the same train inoperable, the remaining inverters are capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in one of the two remaining inverters could result in the minimum ESE functions not being supported.
Therefore, one of the inverters must be restored to OPERABLE status within 6 hours.The 6 hour Completion Time is consistent with that allowed for an inoperable RTS train and an inoperable ESFAS train, since the inverters support the 120 VAC vital buses, which in turn support the RTS and ESFAS trains.C.1 and C.2 If the Train A or Train B inverter(s) cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and the 120 VAC vital buses energized from the associated inverter.
Each inverter must be connected to its associated 250 VDC bus. The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation of the RTS and ESFAS connected to the 120 VAC vital buses.
.Insert 2 r-
-<ivrtmm f4rc4e Cook Nuclear Plant Unit 1 B 3.8.7-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.8.7-3 Revision No. 0 Inverters
-Shutdown B 3.8.8 BASES SURVEILLANCE REQU IREMENTS SR 3.8.8.1 This Surveillance verifies that the required inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter.
The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation connected to the AC vital buses. HT-e-7=lay F-eecrtast t-r-re ~~-a~ t ~'e Insert 2 REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 1 B3884Rvso o B 3.8.8-4 Revision No. 0 Distribution Systems -Operating B 3.8.9 BASES SURVEILLANCE REQUIREMENTS SR 3.8.9.1 This Surveillance verifies that the required AC, DC, and 120 VAC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment.
The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus.The verification of proper voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these buses. ,-Rhe-7-=tay Freuaye{e-a--ccea--erela~et@b~tf ~ f
# tP==Insert 2 REFERENCES
: 1. Safety Guide 6, March 1971.2. UFSAR, Chapter.14.
: 3. Regulatory Guide 1 .93, December 1974.Cook Nuclear Plant Unit 1B3.9-0RvsoN.0 B 3.8.9-10 Revision No. 0 Distribution Systems -Shutdown B 3.8.10 BASES SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the AC, DC, and 120 VAC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized.
The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. *REFERENCES
: 1. UFSAR, Chapter 14.----Insert 2 Cook Nuclear Plant Unit 1B381-4RvsoN.0 B 3.8.10-4 Revision No. 0 Boron Concentration B 3.9.1 BASES ACTIONS (continued)
Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position.
Operations that individually add limited positive reactivity (e.g., temperature fluctuations from inventory addition or temperature control fluctuations), but when combined with all other operations affecting core reactivity (e.g., intentional boration) result in overall net negative reactivity addition, are not precluded by this action.A.3 In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately.
In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied.
The only requirement is to restore the boron concentration to its required value as soon as possible.
In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.
Once actions have been initiated, they must be continued until the boron concentration is restored.
The restoration time depends on the amount of*boron that must be injected to reach the required concentration.
SURVEILLANCE REQUIREMENTS SR 3.9.1.1 and SR 3.9.1.2 These SRs ensure that the coolant boron concentration in the RCS, and connected portions of the refueling canal and the refueling cavity, is within the COLR limits. The boron concentration is determined periodically and prior to re-connecting portions of the refueling canal and the refueling cavity to the RCS, by chemical analysis.= ---Insert 2 The -Frcqucncy aeae The SR 3.9.1.2 Frequency of once within 72 hours prior to connecting the refueling canal and refueling cavity to the RCS ensures that if any dilution activity has occurred while the cavity and canal were disconnected from the RCS, correct boron concentration is verified prior to communication with the RCS.REFERENCES
: 1. UFSAR, Section 1.4.5.Cook Nuclear Plant Unit 1 B3913Rvso o B 3.9.1-3 Revision No. 0 Nuclear Instrumentation B 3.9.2 BASES ACTIONS (continued) since CORE ALTERATIONS and positive reactivity additions are not to be made, the core reactivity, condition is stabilized until the source range neutron flux monitors are OPERABLE.
This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists.The Completion Time of once per 12 hours is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified.
The 12 hour Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.C. 1 With no audible count rate OPERABLE, prompt and definite indication of a boron dilution event, consistent with the assumptions of the safety analysis, is lost. In this situation, the boron dilution event may not be detected quickly enough to assure sufficient time is available for operators to manually isolate the unborated water source and stop the dilution prior to the loss of SHUTDOWN MARGIN. Therefore, action must be taken to prevent an inadvertent boron dilution event from occurring.
This is accomplished by isolating all the unborated water flow paths to the Reactor Coolant System. Isolating these flow paths ensures that an inadvertent dilution of the reactor coolant boron concentration is prevented.
The Completion Time of "'Immediately" assures a prompt response by operations and requires an operator to initiate actions to isolate an affected flow path immediately.
Once actions are initiated, they must be continued until all the necessary flow paths are isolated or the circuit is restored to OPERABLE status.SURVEILLANCE SR 3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is normally a comparison of the parameter indicated on one channel to a similar parameter on another channel. It is based on the assumption that the two indication channels should be consistent with core conditions.
Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.
* h-Freq -Insert 2ei si- Cook Nuclear Plant Unit 1 B 3.9.2-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.9.2-3 Revision No. 0 Nuclear Instrumentation B 3.9.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.2.2 SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION ey.CHANNEL CALIBRATION is a complete check of the instrument loop, except the detector.
The CHANNEL CALIBRATION for the Westinghouse source range neutron flux monitors also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. In addition, the CHANNEL CALIBRATION includes verification of the audible count rate function for the required monitor. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. 1 1-t Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Section 14.1.5.Cook Nuclear Plant Unit 1 B3924Rvso o B3.9.2-4 Revision No. 0 Containment Penetrations B 3.9.3 BASES SURVEILLANCE REQUIREMENTS SR 3.9.3.1 This Surveillance demonstrates that each of the containment penetrations is in its required status. The LCO 3.9.3.c.2 status requirement, which requires penetrations to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System, can be verified by ensuring each required valve operator is capable of closing automatically if needed. This Surveillance does not require cycling of the valves since this is performed at the appropriate frequency in accordance with SR 3.9.3.2.The S'ir!eillaPce is performed every 7 ~C.~ :...,.-....-Insert 2 co- pto f el an4 ejeF ati-a s.- Thi s She-ae anc nuc -th- t.n* .lcsoG SR 3.9.3.2 This Surveillance demonstrates that each required containment purge supply and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. T-he-2A44e, et4~h -Insert 2-Ffe -e de{-as-s-em gmt]-9e l J The SR is modified by a Note stating that this Surveillance is not required to be met for valves in isolated penetrations.
The LCO provides the option to close penetrations in lieu of requiring automatic actuation capability.
REFERENCES
: 1. UFSAR, Section 14.2.1.5.Cook Nuclear Plant Unit 1 B3934Rvso o B 3.9.3-4 Revision No. 1 RHR and Coolant Circulation
-High Water Level B 3.9.4 BASES SURVEILLANCE REQUIREMENTS SR 3.9.4.1 This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core..Te-h teyie-ef'.
-Insert 2 ,l#-aursss~iatrc.-aie48h-lw-e~fm up.m.iagO-R:R-~m REFERENCES
: 1. UFSAR, Section 9.3.2.Cook Nuclear Plant Unit 1 B3944Rvso o B 3.9.4-4 Revision No. 0 RHR and Coolant Circulation
-Low Water Level B 3.9.5 BASES ACTIONS (continued)
B.2 If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation.
Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.
B.3, B.4. and B.5 If no RHR is in operation, the following actions must be taken: a. The equipment hatch must be closed and secured with four bolts;b. One door in each air lock must be closed; and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System.With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions stated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.The Completion Time of 4 hours allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. Tr-he 4.= Insert 2m Cook Nuclear Plant Unit 1 ..- evso o B3.9.5-3 Revision No. 0 RHR and Coolant Circulation
-Low Water Level B 3.9.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to the required pump. T4he- e-r f-7 e~e~eeidereed-r-easebie~ar3le-iv=-
Insert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
REFERENCES
: 1. UFSAR, Section 9.3.2.Cook Nuclear Plant Unit 1 ..- evso o B3.9.5-4 Revision No. 0 Refueling Cavity Water Level B 3.9.6 BASES ACTIONS A.1 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.SURVEILLANCE REQ U IREM ENTS SR 3.9.6.1 Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 1).Z~eFrelec--f-.
rsist~-ete-aiera-ugeta4
~ ~ ~=Insert 2I REFERENCES
: 1. UFSAR, Section 14.2.1.2. 10CFR 100.10.Cook Nuclear Plant Unit 1 B 3.9.6-2 Revision No. 26 Enclosure 7 to AEP-NRC-2015-46 CNP Unit 2 TS Bases Pages Marked to Show Proposed Changes SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)
: b. Bank position;c. RCS average temperature;
: d. Fuel burnup based on gross thermal energy generation;
: e. Xenon concentration;
: f. Samarium concentration;
: g. Isothermal temperature coefficient (ITC); and h. Boron penalty (MODES 4 and 5 only).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. The boron penalty must be applied in MODES 4 and 5 since all reactor coolant pumps may be stopped in these MODES.This extra amount of boron ensures that minimum response times are met for the operator to diagnose and mitigate an inadvertent boron dilution event prior to loss of SDM.T" o e--u-ic of 24-sour5-c4baacd on .--=lnsert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Chapter 14.3. UFSAR, Section 14.2.5.4. UFSAR, Section 14.1.5.5. 10OCFR 100.Cook Nuclear Plant Unit 2 B3115Rvso o B 3.1.1-5 Revision No. 0 Core Reactivity B 3.1.2 BASES ACTIONS (continued)
B.1I If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE in which, the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. If the SDM for MODE 3 is not met, then the boration required by SR 3.1.1.1 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 unit systems.SURVEILLANCE REQUIRE MENTS SR 3.1.2.1 Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations.
The comparison is made, considering that other core conditions are fixed or stable, including RCS boron concentration, control rod position, RCS average temperature,.fuel burnup based on gross thermal energy generation, xenon concentration, and samarium concentration.
The Surveillance is performed prior to entering MODE I 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 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.
T-he Insert 2 re4el db~ 3 EfPD fcwiig-tbeiinti  after-eatoring MODE 1, icQs n he-slowl~srateofre
'(Qrf, AF, t.) f -rptft4aeleat-ief--efe~e~aey.
REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Chapter 14.Cook Nuclear. Plant Unit 2 B3125Rvso o B3.1.2-5 Revision No. 0 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued) 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 in which the LCO requirements are not applicable.
To achieve this status, the unit must be brought to at least MODE 3 within 6 hours.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 unit systems.SURVEILLANCE SR 3.1.4.1 REQ U IREM ENTS'Vf ~ {
Insert 2 iseeaa4euheity-pq~exlae-te4-h-e-&aewr4a~
e.-eeeattef SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2, tripping each control rod would result in radial or axial power tilts, or oscillations.
Exercising each individual control rod ayes- provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 8 steps will not cause radial or axial power tilts, or oscillations, to occur. TFhe- .,-.- Insert 2rrnte-e-s e ~ r f~~ e Between required performances of SR 3.1.4.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable the control rod(s) is considered to be OPERABLE.
At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken..Cook Nuclear Plant Unit 2 B3147Rvso o B 3.1.4-7 Revision No. 0 Shutdown Bank Insertion Limits B 3.1.5 BASES SURVEILLANCE REQUIREMENTS SR 3.1.5.1 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.Sio tho s iah e r.a--Relea-fsalfuet-a~
--~4 REFERENCES
: 1. UFSAR, Section 1.4.2.2. UFSAR, Section 1.4.5.3. UFSAR, Section 1.4.6.4. 10 CFR 50.46.5. UFSAR, Chapter 14.Insert 2 Cook Nuclear Plant Unit 2 B3154Rvso o B 3.1.5-4 Revision No. 0 Control Bank Insertion Limits B 3.1.6 BASES SURVEILLANCE REQUIREMENTS SR 3.1.6.1 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. Verifying the ECP calculation within 4 hours 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 Insert 2 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 CO LR.
Insert 2 ,cbee4a-S6,2.
REFERENCES
: 1. UFSAR, Section 1.4.2.2. UFSAR, Section 1.4.5.3. UFSAR, Section 1.4.6.4. 10 CFR 50.46.5. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2 B3165Rvso o B 3.1,6-5 Revision No. 1 PHYSICS TESTS Exceptions
-MODE 2 B 3.1.8 BASES ACTIONS (continued) 531 &deg;F could violate the assumptions for accidents analyzed in the safety analyses.D. 1 If the Required Action and associated Completion Time of Condition C is not met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit 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 unit systems.SURVEILLANCE REQUIRE MENTS SR 3.1.8.1 Verification that the RCS lowest loop Tavg is > 531&deg;0 F will ensure that the unit is not operating in a condition that could invalidate the safety analyses.
Verification of the RCS temperature at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the initial conditions of the safety analyses are not violated.SR 3.1.8.2 Verification that the THERMAL POWER is _< 5% RTP will ensure that the unit is not operating in a condition that could invalidate the safety analyses.
-"ictc o h T ER O"Rc c-FrPtueqiey~f 2 4-a4:ij~lealeee-,
SR 3.1.8.3 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects: a. RCS boron concentration;
: b. Bank position;c. RCS average temperature;
: d. Fuel burnup based on gross thermal energy generation;
: e. Xenon concentration; Cook Nuclear Plant Unit 2 B3186Rvso o B3.1.8-6 Revision No. 0 PHYSICS TESTS Exceptions
-MODE 2 B 3.1.8 BASES SURVEILLANCE REQUIREMENTS (continued)
: f. Samarium concentration;
: g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH);h. Moderator Temperature Defect, when above the POAH; and i. Doppler Defect, when above the POAH.Using the ITC accounts for Doppler reactivity in this calculation when the reactor is subcritical or critical but below the POAH, and the fuel temperature will be changing at the same rate as the RCS.-Th seaeF4ega~~ei I nsert 2 reqttiI'ed4ber-efi-eeneeontatiea-e "e~wp itie oeeeuwr4I~wileu4he-r.eqEi,44ed-SDM.
REFERENCES
: 1. 10 CFR 50, Appendix B, Section Xl.2. 10 CFR 50.59.3. Regulatory Guide 1.68, Revision 2, August, 1978.4. ANSI/ANS-19.6.1-1997, August 22, 1997.5. WCAP-1 3360-P-A, "Westinghouse Dynamic Rod Worth Measurement Technique," Revision 1, October 1998.6. PA-OSC-0061, "Westinghouse Position Paper on Power Distribution Measurement Requirements for Reload Startup Programs," February 2005.Cook Nuclear Plant Unit 2 B3187Rvso o B3.1.8-7 Revision No. 1 FQ(Z)B 3.2.1 BASES ACTIONS (continued)
C. I If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE or condition in which the LCO requirements are not applicable.
This is done by placing the unit in at least MODE 2 within 6 hours.This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.2.1.1 REQUI REM ENTS Verification that FC(z) is within its specified limits involves increasing FM(z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain F&(Z). is then compared to its specified limits.If THERMAL POWER has been increased by > 10% RTP since the last determination of FC~(z), another evaluation of this factor is required 24 hours after achieving equilibrium conditions at this higher power level (to ensure that FC~(z) values are being reduced sufficiently with power increase to stay within the LCO limits). The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification.
It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ(Z) was last measured.
----Insert 2
~ ~ e e-n~l eet-re~hrHe~a-s-e~~daaerlrcw
~ ~ S-p i isan'ifr=-T--p SR 3.2.1.1 is modified by a Note, which applies during power escalation after a refueling.
The Note states that the Surveillance is not required to be performed until 24 hours after equilibrium conditions at a power level for extended operation are achieved.
This Note allows the unit to startup from a refueling outage and reach the power level for extended operation (normally 100% RTP) prior to requiring performance of the SR. Within 24 hours after equilibrium conditions are reached at the power level for extended operation, the SR must be performed.
Cook Nuclear Plant Unit 2 B 3.2.1-6 Revision No. 0 Cook Nuclear Plant Unit 2 B3.2.1-6 Revision No. 0 FQ(Z)B 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification.
It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ(Z) was last measured.The
~~eltetor r
2
---eG~i ys4elaet-eiert~--as-fe a~.-~e-trlae wth4e-S eki4 4 SR 3.2.1.2 is modified by Note 1, which applies during power escalation after a refueling.
The Note states that the Surveillance is not required to be performed until 24 hours after equilibrium conditions at a power level for extended operation are achieved.
This Note allows the unit to startup from a refueling outage and reach the power level for extended operation (normally 100% RTP) prior to requiring performance of the SR. Within 24 hours after equilibrium conditions are reached at the power level for extended operation, the SR must be performed.
REFERENCES
: 1. 10 CFR 50.46.2. UFSAR, Section 14.2.6.1.2.
: 3. .UFSAR, Section 1.4.5.4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.5. WCAP-1 0216-P-A, Rev. tA, "Relaxation of Constant Axial Offset Control (and) F 0 Surveillance Technical Specification," February 1994.Cook Nuclear Plant Unit 2 B3218Rvso o B 3.2.1-8 Revision No. 0 FNH B 3.2.2 BASES ACTIONS (continued).A.4 Verification that FNAH is within its specified limits after an out of limit occurrence ensures that the cause that led to the FNAH exceeding its limit is corrected, and that subsequent operation proceeds within the LCOG limit. This Action demonstrates that the FNH limit is within the LOG limits prior to exceeding 50% RTP, again prior- to exceeding 75% IRTP, and within 24 hours after THERMAL POWER is > 95% IRTP.This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action.B.1 When any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCOG requirements are not applicable.
This is done by placing the unit in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SIR 3.2.2.1 REQUIREMENTS The value of FN~H is determined by using the movable incore detector system to obtain a flux distribution map. A data reduction computer program then calculates the maximum value of FNAH from the measured flux distributions.
The measured value of FNH must be multiplied by 1.04 to account for measurement uncertainty before making comparisons to the FNAH limit.After each refueling, must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that FNH limits are met at the beginning of each fuel cycle.
atte--pe-wer-dist-ibti{+hrn 4-Insert 2 4-s-rqt~eyi y-s1e-esehr-t-49ght-at-t49e-F{-eait~aet-b-xeeeded-f f-elper-a-iem-REFERENCES
: 1. UFSAIR, Section 14.2.6.1.2.
: 2. UFSAIR, Section 1.4.5.3. 10 CFIR 50.46.Cook Nuclear Plant Unit 2B3225ReionN.4 B 3.2.2-5 Revision No. 34 AFD B 3.2.3 BASES SURVEILLANCE REQUIREMENTS SR 3.2.3.1 This Surveillance verifies that the AFD as indicated by the NIS excore channels is within the target band.
td~ys "*=--lnsert 2-by-t9e-pree-e-c-emputer.-Fur-term~er-er
,---dev~aieon f~e-AFBfrom t~he4arg et-baald-t~et--i-r~et-aer-lmed-s4heute-be SR 3.2.3.2-===lInsert 2-c-h aaeje~t--ay-ec-c-uf-i4e4gret-1dfu--lf4er-eesT~
ataper~ed-eue4e SR 3.2.3.3 Measurement of the target flux difference is accomplished by taking a flux map when the core is at equilibrium xenon conditions, preferably at high power levels with the control banks nearly withdrawn.
This flux map provides the equilibrium xenon axial power distribution from which the target value can be determined.
The target flux difference varies slowly with core burnup.---re~e.Je----4 Insert 2.,--a+~ur t fr-f A Note modifies this SR to allow the predicted beginning of cycle AFD from the cycle nuclear design to be used to determine the initial target flux difference after each refueling.
REFERENCES
: 1. WCAP-8385 (Westinghouse proprietary) and WCAP-.8403 (nonproprietary), "Power Distribution Control and Load Following Procedures,'
Westinghouse Electric Corporation, September 1974.2. UFSAR, Section 7.4.Cook Nuclear Plant Unit 2 B3236Rvso o B 3.2.3-6 Revision No. 1 QPTR B 3.2.4 BASES ACTIONS (continued)
Action A.5). The intent of this Note is to have the peaking factor Surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limits and the core returned to power.B.1 If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE or other specified condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging unit systems.SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is< 75% RTP and the input from one Power Range Neutron Flux channel is inoperable.
Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1.This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. Insert 2* -eu~ye--tyse4s-~ -eetehr~e For those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is not required until 12 hours after the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is > 75% RTP.With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded.
Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased.
Pe~feFmiag-GR--3.2?.4 7
Insert 2.e---ersp~vasaae~t-~tmtv-en-e-~uighf-~
Cook Nuclear Plant Unit 2 B3245Rvso o B3.2.4-5 Revision No. 0 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.1 Performance of the CHANNEL CHECK once every 12 hours 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.
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.T s 21-rsl~ied-eheffe~el.
SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output every 24 hours. If the calorimetric exceeds the NIS channel output by > 2% RTP, the NIS is not declared inoperable, but must be adjusted.
If the NIS channel output cannot be properly adjusted, the channel is declared inoperable.
Two Notes modify SR 3.3.1.2. The first Note indicates that the NIS channel output shall be adjusted consistent with the calorimetric results if the absolute difference between the NIS channel output and the calorimetric is > 2% RTP. The second Note clarifies that this Surveillance is required only if reactor power is -> 15% RTP and that 12 hours is allowed for performing the first Surveillance after reaching 15% RTP. At lower power levels, calorimetric data are inaccurate.
<--- Insert 2 drs-ael-exceeds-2%
ii ny-24-hetw-perod.-~-
Cook Nuclear Plant Unit 2 B3313 eiinN.1 B 3.3.1:38 Revision No. 17 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) 4Rtiea ieiasa44-alae-detec-t deviaie e-i9-ehe9ael--ei~tlit5.
SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output evefy'.,S1=fP If the absolute difference is > 3%, the NIS channel is still OPERABLE, but must be readjusted."If the 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.Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be adjusted if the absolute difference between the incore and excore AFD is > 3%. Note 2 clarifies that the Surveillance is required only if reactor power is > 15% RTP and that 24 hours is allowed for performing the first Surveillance after reaching 15% RTP.T4 re~
e-a, 4-Insert 2 eie---aget
~ec eeera8atrue tya t uSR 3.3.1.4 SR 3.3.1.4 is the performance of a TADOT~ee-vy=4a-y62--eai-a.
'SAGE- E3FBA1S 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 RTB undervoltage 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 3.3.1.17.
The bypass breaker test shall include a local shunt trip. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service.Cook Nuclear Plant Unit 2 B3313 eiinN.1 B 3.3.1-39 Revision No. 17 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
T~eF--~Ray-m-%A---S~
4--Insert 2
SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The SSPS is tested ev ~ y-r.a8 using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
'T~e 02id-y3 Insert 2 SR 3.3.1.6 SR 3.3.1.6 is the performance of a TADOT and is performed every 92 days on a STAGGERED TEST BASIS. This test applies to the SI Input from ESFAS 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 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.Ie RE~-TEgTs-
,--I nsert 2 jii-trhe~n--Refefeine=1 SR 3.3.1.7 SR 3.3.1.7 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 the f(AI)input to the Overtemperature AT Function.A Note modifies SR 3.3.1.7. The Note states that this Surveillance is required only if reactor power is > 50% RTP and that 24 hours is allowed for performing the first surveillance after reaching 50% RTP.e ~ -
_ Insert 2 Coox Nuclear Plat Unit 2 B 3.3.-40 Revision Nto~d t .1 Cook Nuclear Plant Unit 2 B 3.3.1-40 Revision No. 17 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUI REM ENTS (continued)
SR 3.3.1.8 SR 3.3.1.8 is the performance of a COT every 92 days.A COT is performed on each required channel to ensure the entire 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 COT 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.The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 8.SR 3.3.1.8 is modified by a Note that provides a 12 hour delay in the requirement to perform this Surveillance for Function 2.b channels after reducing THERMAL POWER below the P-10 interlock.
The Frequency of 12 hours after reducing power below P-10 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.
-Insert 2 SR 3.3.1.9CHAN NEL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
Cook Nuclear Plant Unit 2 B 3.3.1-41 Revision No. 16 Cook Nuclear Plant Unit 2 B 3.3.1-41 Revision No. 16 RTS Instrumentation
*B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
P-te-asu'+~ea f ---Insert 2 This SR is modified by a Note that states that neutron detectors are excluded from the CHANNEL CALIBRATION.
Changes in power range neutron detector sensitivity are compensated for by normalization of the channel output based on a power calorimetric and flux map performed above 15% RTP (SR 3.3.1.2).SR 3.3.1.10 SR 3.3.1.10 is the performance of a TADOT and <,---Insert 2A 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 SR is modified by a Note that excludes verification of relay setpoints from the TAD OT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION.
The Frequency of 92 days is justified in Reference 10.SR 3.3.1.11 SR 3.3.1.11 is the performance of a COT e A COT is performed on each required channel to ensure the entire 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 COT 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.The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
Cook Nuclear Plant Unit 2 B 3.3.1-42 Revision No. 16 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 8.The Frequency is modified by two Notes. Note 1 provides a 12 hour delay in the requirement to perform this Surveillance for intermediate range instrumentation after reducing THERMAL POWER below the P-I10 interlock.
The Frequency of 12 hours after reducing power below P-10 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.
Note 2 provides a 4 hour delay in the requirement to perform this Surveillance for source range instrumentation after THERMAL POWER is reduced below the P-6 interlock.
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 RTBs are open and SR 3.3.1.11 is no longer required to be performed.
If the unit is to be in MODE 3 with the RTBs closed for > 4 hours this Surveillance must be performed prior to 4 hours after THERMAL POWER is reduced below the P-6 interlock.
--& Insert 2 SR 3.3.1.12 CHAN NEL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
* T h --e f e e -f l s I a e L t e a s m ~ ~ e -m l 8 e I n s e rt 2 SR 3.3.1.13 CHAN NEL 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.Cook Nuclear Plant Unit 2 B3314 eiinN.1 B 3.3.1-43 Revision No. 16 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.-t
-Insert 2 SR 3.3.1.14 SR 3.3.1.14 is the performance of a CHANNEL CALlBRATION~everyi-
.2--ment4he.
CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The CHANNEL CALIBRATION for the source range neutron detectors also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION.
Changes in power range neutron detector sensitivity are compensated for by normalization of the channel output based on a power calorimetric and flux map performed above 15% RTP (SR 3.3.1.2).Changes in intermediate range neutron flux detector sensitivity are compensated for by periodically evaluating the compensating voltage setting and making adjustments as necessary.
Changes in source range neutron detector sensitivity are compensated for by periodically obtaining the detector plateau or preamp discriminator curves, evaluating those curves, comparing the curves to the manufacturer's data, and adjusting the channel output as necessary.--Insert 2 SR 3.3.1.15 SR 3.3.1.15 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.13, .ever Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the resistance temperature detectors (RTD) sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.CookNucearPlan Unt 2B 3..1-4 Rvisin N. 1 Cook Nuclear Plant Unit 2 B 3.3.1-44 Revision No. 16 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
-Insert 2 This SR is modified by a Note that provides a 72 hour delay in the requirement to perform a normalization of the AT channels after THERMAL POWER is > 98% RTP. The intent of this Note is to maintain reactor power at a nominal 97% RTP to 98% RTP level until the AT normalization is complete before increasing reactor power to 100% RTP.SR 3.3.1.16 SR 3.3.1.16 is the performance of a COT of RTS interlocks every--r4-ner~he.
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 COT 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.
* h-e~ry tekea-elt-f4ere~e--~eh
<"-I nsert 2~ em-h~4erseetqe a e e-t re e t~@ -xj 4e e.SR 3.3.1.17 SR 3.3.1.17 is the performance of aTADOT of the Manual Reactor Trip (including reactor trip bypass breakers) and RCP Breaker Position.
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 test shall independently verify the OPERABILITY of the undervoltage and shunt trip mechanisms for the Manual Reactor Trip Function for the Reactor Trip Breakers and Reactor Trip Bypass Breakers.
The Reactor Trip Bypass Breaker test shall include testing of the automatic undervoltage trip.Cook Nuclear Plant Unit 2 B3314 eiinN.1 B 3.3.1-45 Revision No. 16 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The-F- ee-sIas~-nekawnreaiityeJ eF s -Insert 2 aec-efftaee4-ue-e fper-at~r-&sect;-ex-perier1ee7.
SR 3.3.1.18 SR 3.3.1.18 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 as described in SR 3.3.1.4, except that this test is performed prior to exceeding the P-8 interlock whenever the unit has been in MODE 3. This Surveillance is not required if it has been I 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-8 interlock.
SR 3.3.1.19 SR 3.3.1.19 verifies that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analysis.
Response time testing acceptance criteria are included in UFSAR, Table 7.2-6 (Ref. 12). 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 to the point at which the equipment reaches the required functional state (i.e., control and shutdown rods fully inserted in the reactor core).For channels that include dynamic transfer Functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer Function set to one, with the resulting measured response time compared to the appropriate UFSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value, provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.Cook Nuclear Plant Unit 2 B 3.3.1-46 Revision No. 18 Cook Nuclear Plant Unit 2 B 3.3.1-46 Revision No. 18 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Response time may be verified by actual response time tests in any series of sequential, 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) in place, 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," (Ref. 13) 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, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 14) provides the basis and methodology for 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 the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
4e~ems~-e~i
-Insert 2-h4at ee4te~vceqqcaeliateete w4 th ~ nft-sela-~~-tPetacewe-ef -tth&#xa3; ee-va-eue~~
ac-eept-able-ff~-are{9ai~t-y-etea9dj eiat SR 3.3.1.19 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.The response time testing of the neutron flux signal portion of the channel shall be measured from either the detector output or the input of the first electronic component in the channel.Cook Nuclear Plant Unit 2 B 3.3.1-47 Revision No. 5 Cook Nuclear Plant Unit 2 B 3.3.1-47 Revision No. 5 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
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.Insert 2 SR 3.3.2.2 and SR 3.3.2.5 SR 3.3.2.2 is the performance of a
,-a This test is a check of the Loss of Voltage Function.
SR 3.3.2.5 is the performance of a test is a check of the Undervoltage ROP 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 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.
Each SR is modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and are verified during CHANNEL CALIBRATION. "T Frg,,nc o-f-- 3,3.2.2---
Insert 2 0 I h'noe ru nnr inorhln Arnn~rfIn rv-nerlodrinri.SR 3.3.2.5 is juctificd in Rcfrnc0 SR 3.3.2.3 SR 3.3.2.3 is the performance of an ACTUATION LOGIC TEST. --T-he-." apgy-c tesaede$eT-9 da, using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the master relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERABLE and that there is an intact voltage signal path to the master relay coils. Th rouny-f--r 2daso Insert 2
! O.Cook Nuclear Plant Unit 2B33237RvsoN.0 B 3.3.2-37 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.4 SR 3.3.2.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. "T-is--test-le GERa TT 30I. Te Insert 2 Ref e'rellee-fr SR 3.3.2.6 SR 3.3.2.6 is the performance of a COT. A COT is performed on each required channel to ensure the entire channel will perform the intended Function.
Setpoints must be found within the Allowable Values specified in Table 3.3.1-1. 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 COT 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 difference between the current "as found" values .and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current Unit specific setpoint methodology.
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 6.nserti 2 SR 3.3.2.6 is modified by a Note which applies to the SI Cdntainment Pressure -High, Containment Spray Containment Pressure -High High, Phase B Isolation Containment Pressure -High High, Steam Line Isolation Containment Pressure -High High, and CEQ System Containment Pressure -High Functions.
This Note requires, during the performance of SR 3.3.2.6, the associated transmitters of these Functions to be exercised by applying either a vacuum or pressure to the appropriate side of the transmitter.
Exercising the associated transmitters during the performance of the COT is necessary to ensure Functions 1 .c, 2.c, 3.b.(3), 4.c, and 7.c remain OPERABLE between each CHANNEL CALIBRATION.
Cook Nuclear Plant Unit 2B332-8RvsoN.0 B 3.3.2-38 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.7 SR 3.3.2.7 is the performance of a CHANNEL CALIBRATION.:h-A--G=AN--A-BATINispF~e CHAN NEL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.-Th--~ee
.Insert 2-a ~ti-h-e~iFtIe~e l v-SR 3.3.2.8 SR 3.3.2.8 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function, or is placed in a condition where the relay contact operation can be verified without operation of the equipment.
Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.
.Insert 2-and-er-atg-hI~stof-at SR 3.3.2.9 SR 3.3.2.9 is the performance of a TADOT. This test is a check of the Manual Initiation Functions, the AFW pump start on trip of all MFW pumps, and the P-4 interlock.
It is pcrorc ....r, ..... month,,,,.
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 Cook Nuclear Plant Unit 2 B 3.3.2-39 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.3.2-39 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) extensions.
In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.). Th-eqeeyi-aeeEtefat l-'nsert 2 SR 3.3.2.10 SR 3.3.2.10 is the performance of a CHANNEL CALIBRATION. -A-erord 4-- eI-tths. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy.CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
TheF a4hsit~eit-ethedete#.
SR 3.3.2.11 SR 3.3.2.11 is the performance of an ACTUATION LOGIC TEST. This SR is applied to the balance of plant actuation logic and relays that do not have the SSPS test circuits installed to utilize the semiautomatic tester or perform the continuity check. All possible logic combinations are tested for Table 3.3.2-1 Functions 6.e and 6.g. TCs-refee-e-4mrth4~ r- --
* retie1ti-y--a SR 3.3.2.12 This SR ensures the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis.Response Time testing acceptance criteria are included in the UFSAR, Table 7.2-7 (Ref. 11). 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, to the point at which the equipment in both trains reaches the required functional state (e.g., pumps at rated discharge pressure, valves in full open or closed position).
Cook Nuclear Plant Unit 2 B 3.3.2-40 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.3.2-40 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer functions set to one with the resulting measured response time compared to the appropriate UFSAR response time. Alternlately, the response time test can be performed with the time constants set to their nominal value provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.Response time may be verified by actual response time tests in any series of sequential, 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) in place, 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," dated January 1996 (Ref. 12), 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, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 13) provides the basis and methodology for 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 the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
_N_--_____-__-___te -eer,- nsert 2 S8TA CLEDT he oehef'~-__-he-f'on-v
.~a _J.--e-f ~~t e Cook Nuclear Plant Unit 2 B 3.3.2-41 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.3.2-41 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) efa4~mn s--m This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours after reaching 850 psig in the SGs.REFERENCES
: 1. Technical Requirements Manual.2. IEEE-279, "Proposed Criteria for Nuclear Power Plant Protection Systems," August 1968.3. UFSAR, Table 7.2-1.4. UFSAR, Table 14.1.0-4.5. 10 CFR 50.49.6. WCAP-12741, "Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.7. UFSAR, Chapter 14.8. WCAP-14333-P-A, Revision 1, October 1998.9. WCAP-1 0271-P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," including Supplement 1, May 1986, and Supplement 2, Rev. 1, June 1990.10. WCAP-1 5376, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Intervals and Reactor Trip Breaker Test and Completion Times," October 2000.11. UFSAR, Table 7.2-7.12. WCAP-1 3632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996.13. WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.Cook Nuclear Plant Unit 2B332-2RvsoN.0 B 3.3.2-42 Revision No. 0 PAM Instrumentation B 3.3.3 SURVEILLANCE As noted at the beginning of the SRs, the following SRs apply to each REQUIREMENTS PAM instrumentation Function in-Table 3.3.3-1, except where identified in the SR.SR 3.3.3.1 Performance of the CHANNEL CHECK @e ieevy--31-day~s.ensures that a gross instrumentation failure has not occurred.
A CHANNEL CHECK is BASES sURvEILLANCE REQUIREMENTS (continued) 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
The Containment Area Radiation (High Range)instrumentation should be compared to similar unit instruments located throughout the unit. When only one channel of the Reactor Coolant Inventory Tracking System is OPERABLE, the RCS Subcooling Margin Monitor and Core Exit Temperature channels may be used for performance of the CHANNEL CHECK of the OPERABLE Reactor Coolant Inventory Tracking System channel.Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, 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. If the channels are within the criteria, it is an indication that the channels are OPERABLE.As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.
~ B-ee-Ng-e#
2 K c-heauaels..
Cook Nuclear Plant Unit 2 B3331 eiinN.1 B3.3.3-13 Revision No. 16 PAM Instrumentation B 3.3.3 BASES SURVEILLANCE REQUIREMENTS (continued).SR 3.3.3.2 .Deleted SR 3.3.3.3 A C NN pe~er r-r--4-Ph
: e. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy.
This SR is modified by a Note that excludes neutron detectors.
For Function 9, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual valve position.
For Functions 15, 16, 17, and 18, whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit Temperature thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing elements.For Functions 20 (Circuit Breaker Status channels) and 24, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual circuit breaker position..T-he-24--mie~h 2* F-r e~ gexp~~.I REFERENCES 1.NRC letter, T. G. Colburn (NRC) to M. P. Alexich (Indiana Michigan Power Company), 'Emergency Response Capability
-Conformance to Regulatory Guide 1.97 Revision 3 for the D. C. Cook Nuclear Plant, Units 1 and 2," dated December 14, 1990.2. UFSAR, Table 7.8-1.3. Regulatory Guide 1.97, Revision 3, May 1983.4. NUREG-0737, Supplement 1, "TMI Action Items." 5. NRC letter, P.S.Tam (NRC), to M. K. Nazar, (Indiana Michigan Power Company), "Donald C. Cook Nuclear Plant, Units 1 & 2 (DCCNP-1 and DCCNP-2) -Issuance of Amendments Re: Containment Sump Modifications per Generic Letter 2004-02 (TAC Nos. MD5901 and MD5902)," dated October 18, 2007.Cook Nuclear Plant Unit 2 B3331 eiinN.2 B 3.3.3-14 Revision No. 29 Remote Shutdown Monitoring Instrumentation B 3.3.4 BASES ACTIONS (continued)
Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.A.1 Condition A addresses the situation where one or more required Functions of the remote shutdown monitoring instrumentation are inoperable.
The Required Action is to restore the required Function to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.B.1 and B.2 If the Required Action and associated Completion Time of Condition A is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.3.4.1 REQUIREMENTS Performance of the CHANNEL CHECK' ree-evty&~4ys ensures that a 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. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability.
If the channels are within the criteria, it is an indication that the channels are OPERABLE.
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.Cook Nuclear Plant Unit 2 B3343Rvso o B 3.3.4-3 Revision No. 0 Remote Shutdown Monitoring Instrumentation B 3.3.4 BASES SURVEILLANCE REQUIREMENTS (continued)
As specified in the Surveillance, a CHANNEL CHECK is only required for those channels which are normally energized.
Trhe-F-r 4ass~se.e~
2 d SR 3.3.4.2 CHANNEL CALIBRATION is a complete~check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.For the Reactor Trip Breaker Indication Function on the hot shutdown panel, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual reactor trip breaker position.~FFte=F1~e~uoF-" ~f 24 i-~ ~ci tI ,s ie-baee~-eI9-ej3eiat419pe~ieI~ee.
"*-l'nsert 2 REFERENCES.
: 1. UFSAR, Section 1.4.3.Cook Nuclear Plant Unit 2 B3344Rvso o B 3.3.4-4 Revision No. 0 LOP DG Start Instrumentation B 3.3.5 BASEIS ACTIONS (continued) made inoperable by failure of the LOP DG start instrumentation are required to be entered immediately.
The actions of those LCOs provide for adequate compensatory actions to assure unit safety.SURVEILLANCE REQUIREMENTS SR 3.3.5.1.Performance of the CHANNEL CHECK enc-e-ever--t2--heur-ensures that a 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.
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.
le-a~je1~e-e e4"a ~ {t 2 SR 3.3.5.2 SR 3.3.5.2 is the performance of a TADOT. 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.
* e -lra4eZ ,~ The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment.
The SRs are modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and are verified during CHANNEL CALIBRATION.
T-he-Preq~e-ne-is-4 tbd idcts he ra~eaaMLtaa
--w~~ea~-asae~hw4 aee-epta .l]e419gt-~ef~
atip-eferie rtee.---Insert 2 Cook Nuclear Plant Unit 2 B3355Rvso o B 3.3.5-5 Revision No. 0 LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.5.3 SR 3.3.5.3 is the performance of a CHANNEL CALIBRATION.
The setpoints, as well as the response to a loss of voltage and a degraded voltage test, shall include a single point verification that the trip occurs within the required time delay.
e99 Aery4 .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.~Insert 2 REFERENCES
: 1. UFSAR, Section 8.4.2. UFSAR, Section 8.5.3. UFSAR, Chapterl14.
: 4. WCAP-1 2741, "Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.Cook Nuclear Plant Unit 2 B3356Rvso o B 3.3.5-6 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES ACTIONS (continued)
D.1 Condition D applies to all Containment Purge Supply and Exhaust System Isolation Functions.
If one or more Automatic Actuation Logic and Actuation Relays trains are inoperable, one or more SI Input from ESFAS trains are inoperable, two or more required radiation monitoring channels in a single train are inoperable, or the Required Action and associated Completion Time of Condition A, B, or C are rnot met, operation may continue provided the containment purge supply and exhaust isolation valves are placed in the closed position immediately.
Placing the containment purge supply and exhaust isolation valves in the closed position accomplishes the safety function of the inoperable trains or channels.SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1 REQUIREMENTS determines which SRs apply to which Containment Purge Supply and Exhaust System Isolation Instrumentation Functions.
SR 3.3.6.1 Performance of the CHANNEL CHECK ensures that a 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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.-Insert 2t~~-s--fe--
t b~-nf-r t --efea~i-d -i rm Cook Nuclear Plant Unit 2 B 3.3.6-6 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.3.6-6 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.6.2 SR 3.3.6.2 is the performance of an ACTUATION LOGIC TEST. The train being tested may be placed in the bypass condition, thus preventing actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, may be tested for each protection function.
In addition, the master relay coil may be pulse tested for continuity.
This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. This-t%'t- TEST l Insert 2 SR 3.3.6.3 SR 3.3.6.3 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity.
T-his-tes'-is- -Isr SR 3.3.6.4 A COT is performed ever-y-1-84-da-y on each required channel to ensure the entire 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 COT 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.
Hr-rqtee-is Insert 2-fsdet--tei*Jaee~ra~gexe-ece This test verifies the capability of the instrumentation to provide the Containment Purge Supply and Exhaust System isolation.
The setpoint shall be left consistent with the current unit specific calibration procedure tolerance.
SR 3.3.6.5 SR 3.3.6.5 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation mode is either allowed to function or is placed in a condition where the relay contact operation can be verified without Cook Nuclear Plant Unit 2 B3367Rvso o B3.3.6-7 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued) operation of the equipment.
Actuation equipment that may not be operated in the design mitigation mode is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.
SR 3.3.6.6 SR 3.3.6.6 is the performance of a TADOT. This test is a check of the Manual Initiation Function Each Manual Initiation Function is tested up to, and including, the master relay coils. 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.
In some instances, the test includes actuation of the end device (i.e., valves cycle).The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no setpoints associated with it.l-nsert 2 I. fl~ -IIfl~i, t 9 LAL.* It,) I.) UI~L,~Sfl~
Li UUI**-i--nsert 2 I -.
I J ,l,. IiL Ib c I\ IU ho '..U Lnn L to .w b i I I LU II_, L ep~ee~at~ie~ee-,.
SR 3.3.6.7 rc ~r- 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.-fiienkC iu "z dze "" -It Insert 2 REFERENCES
: 1. UFSAR, Section 5.5.3.2. 10 CFR 100.11 3. WCAP-15376, Rev. 0, October 2000.Cook Nuclear Plant Unit 2 B3368Rvso o B3.3.6-8 Revision No. 0 CREV System Actuation Instrumentation B 3.3.7 BASES ACTIONS (continued) this Completion Time is the same as provided in LCO 3.7.10.B.1.1, B.1.2, and B.2 Condition B applies to the failure of two CREV System Automatic Actuation Logic and Actuation Relays trains in one or more required Functions.
The first Required Action is to place one CREV train in the pressurization/cleanup mode of operation immediately.
This accomplishes the actuation instrumentation Function that may have been lost and places the unit in a conservative mode of operation.
The applicable Conditions and Required Actions of LCO 3.7.10 must also be entered for the CREV train made inoperable by the inoperable actuation instrumentation.
This ensures appropriate limits are placed upon. train inoperability as discussed in the Bases for LCO 3.7.10.Alternatively, both trains may be placed in the pressurization/cleanup mode. This ensures the CREV System function is performed even in the presence of a single failure.C.1 and C.2 Condition C applies when the Required Action and associated Completion Time for Condition A or B have not been met. The unit must be brought to a MODE in which the LCO requirements are not applicable.
To achieve this status, the unit must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS A Note has been added to the SR Table to clarify that Table 3.3.7-1 determines which SRs apply to which CREV System Actuation Instrumentation Functions.
SR 3.3.7.1 SR 3.3.7.1 is the performance of an ACTUATION LOGIC TEST. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the mater relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. 1~iTT-h- 2* trvi~~
-4f d4--ee-ac4--
Cook Nuclear Plant Unit 2 B3.3.7-3 Revision No. 0 CREV System Actuation Instrumentation B 3.3.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.7.2 SR 3.3.7.2 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil[. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity.
T-hi&-t~e--t nset SR 3.3.7.3 SR 3.3.7.3 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function or is placed in a condition where the relay contact operation can be verified without operation of the equipment.
Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.Taitsq~e~-aeee~-4mas cetat -e --taiitya -Insert 2 REFERENCES
: 1. WCAP-1 5376, Rev. 0, October 2000.Cook Nuclear Plant Unit 2 B3374Rvso o B3.3.7-4 Revision No. 0 BDMI B 3.3.8 BASES ACTIONS (continued)
As an alternate to restoring one channel to OPERABLE status within 1 hour (Required Action B.2.1). Required Action B.2.2.1 requires isolation valves for unborated water sources to the Chemical and Volume Control System to be secured to prevent the flow of unborated water into the RCS. In addition, in MODE 5, if the RWST boron concentration is< 2400 ppm and less than the Reactor Coolant System (RCS) boron concentration, the RWST is considered an unborated water source and is required to be isolated from the RCS. Once it is recognized that two source range neutron flux monitoring channels of the BDMI are inoperable, the operators will be aware of the possibility of a boron dilution, and the 1 hour Completion Time is adequate to complete the requirements of Required Action B.2.2. 1.Required Action B.2.2.2 accompanies Required Action B.2.2.1 to verify the SDM according to SR 3.1.1.1 within 1 hour and once per 12 hours thereafter.
This backup action is intended to confirm that no unintended boron dilution has occurred while the BDMI was inoperable, and that the required SDM has been maintained.
The specified Completion Time takes into consideration sufficient time for the initial determination of 8DM and other information available in the control room related to SDM.SURVEILLANCE SR 3.3.8.1 REQUIREMENTS Performance of the CHANNEL CHECK ef~ee-evieriy--24teire 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.
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.T-e
,,---I nsert 2 c ELiCl WCI( uap-_T-h ei4&e-fe r-al, ee~ee Cook Nuclear Plant Unit 2 B3383Rvso o B3.3.8-3 Revision No. 0 BDMI B 3.3.8 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.8.2 SR 3.3.8.2 is the performance of a CHANNEL CALIBRATION
'veiy-* CHANNEL CALIBRATION is a complete check of the instrument ioop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.
The CHANNEL CALIBRATION also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This SR is modified by a Note that states that neutron detectors are excluded from the CHANNEL CALIBRATION.-Th-Ftl ~ re~e l'--nsert 2 REFERENCES
: 1. UFSAR, Section 14.1.5.Cook Nuclear Plant Unit 2 B3384Rvso o B3.3.8-4 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued)
POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP.. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive.
Also, since they represent trarnsients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.
ACTIONS A.1 With one or more of the RCS DNB parameters not within LCO limits, action must be taken to restore parameter(s) in order to restore DNB margin and eliminate the potential for violation of the accident analysis.The 2 hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.
B.1_If Required Action A.1 is not met within the associated Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 2 within 6 hours. In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis.
The Completion Time of 6 hours is reasonable to reach the required unit conditions in an orderly manner.SURVEILLANCE REQUIREMENTS SR 3.4.1.1r-eq ea edi f rdt eq 2-heu i I tt, .., I I r * ,- I I I .- i-Insert 2 SR 3.4.1.2;Ve4fca~ ,fa~
ti ~ -nge-- LRunu's-'~a'f'-r i dti"c " " 1l~ *d~y~-n e-'e ---e---~-~ ~ thsI nshw " -Isr pi-aecetio
-to b c~aereu ~
~ ~ ~
Cook Nuclear Plant Unit 2 B3413Rvso o B 3.4.1-3 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.1.3 Verification that the RCS total flow rate is greater than or equal to the limits ensures that the initial condition of the safety analyses are met.Tae-l42-h -ati ~er.~4r-e 2 SR 3.4.1.4 Measurement of RCS total flow rate by performance of a precision calorimetric heat balance allows the installed RCS flow instrumentation to be calibrated and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.Thet=Ftequ -oenet'~ey=frrl24a.tle ew a1 Insert 2 ee4e--e-ac--e#
ep~ebe.This SR is modified by a Note that allows entry into MODE 1, without having performed the SR, and placement.
of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 24 hours after > 90% RTP. This exception is appropriate since the heat balance requires the unit to be at a minimum of 90% RTP to obtain the stated RCS flow accuracies.
The Surveillance shall be performed within 24 hours after reaching 90% RTP.REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2 B3414Rvso o B 3.4.1-4 Revision No. 0 RCS Minimum Temperature for Criticality B 3.4.2 BASES APPLI CABLE SAFETY ANALYSES (continued) criticality limitation provides a small band, 6&deg;F, for critical operation below HZP. This band allows critical operation below HZP during unit startup and does not adversely affect any safety analyses since the MTC is not significantly affected by the small temperature difference between HZP and the minimum temperature for criticality.
The RCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO Compliance with the LCO ensures that the reactor will not be made or maintained critical (keff > 1.0) at a temperature less than a small band below the HZP temperature, which is assumed in the safety analysis.Failure to meet the requirements of this LCO may produce initial conditions inconsistent with the initial conditions assumed in the safety analysis.APPLICABILITY In MODE 1 and MODE 2 with keff> 1.0, LCO 3.4.2 is applicable since the reactor can only be critical (kerr > 1.0) in these MODES.ACTIONS A.1 If the parameters that are outside the limit cannot be restored, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to MODE 2 with keff < 1.0 within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period. The allowed time is reasonable, based on operating experience, to reach MODE 2 with keff < 1.0 in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.4.2.1 RCS loop average temperature is required to be verified at or above 5410 &deg;F e
*,-- Insert 24 e fl UT Thrra,.J,.., I LI,.JA I, ~ U~ V %..
REFERENCES
: 1. UFSAR, Section 14.1.1.Cook Nuclear Plant Unit 2 B3422Rvso o B 3.4.2-2 Revision No. 0 RCS P/T Limits B 3.4.3 BASES ACTIONS (continued)
Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE SR 3.4.3.1 REQ U IREM ENTS Verification that operation is within limits is required eve~ ia0=h~tes when RCS pressure and temperature conditions are undergoing planned changes. This qeaie~ey sc-~dre-rsre eavew-ef-t41eN~tfel
<-nsert 2 r-eeeef.abte-t4ine:
Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition give~n in the relevant plant procedure for ending the activity is satisfied.
This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.
REFERENCES
: 1. WCAP-15047, Rev. 2, dated May 2002.2. 10 CFR 50, Appendix G.3. ASME, Boiler and Pressure Vessel Code, Section I11, Appendix G.4. ASTM E 185-82, July 1982.5. 10 CFR 50, Appendix H.6. Regulatory Guide 1.99, Revision 2, May 1988.7. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.Cook Nuclear Plant Unit 2 B3436Rvso o B3.4.3-6 Revision No. 0 RCS Ldops -MODES 1 and 2 B 3.4.4 BASES APPLICABILITY (continued)
Operation in other MODES is covered by: LCO 3.4.5, "RCS Loops -MODE 3";LCO 3.4.6, "RCS Loops -MODE 4";LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled";LCO 3.4.8, "RCS Loops -MODE 5, Loops Not Filled";LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation
-High Water Level"; and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation
-Low Water Level." ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the unit to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.4.4.1 This SR requires verification evei-y 4-2-heu~s that each RCS loop is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to the DNBR limit. T4he-Fr-eqel'aey-4f
--- Insert 2 I--- -~r r-- ----- --.... ... r" r- ..............
REFERENCES
: 1. UFSAR, Section 14.1.Cook Nuclear Plant Unit 2 B3443Rvso o B3.4.4-3 Revision No. 0 RCS Loops -MODE 3 B 3.4.5 BASES ACTIONS (continued) 0.1 If one required ROS loop is not in operation, and the Rod Control System is capable of rod withdrawal, the Required Action is to place the Rod Control System in a condition incapable of rod withdrawal (e.g., de-energize all CRDMs by opening the RTBs or de-energizing the motor generator (MG) sets). When the Rod Control System is capable of rod withdrawal, it is postulated that a power excursion could occur in the event of an inadvertent control rod bank withdrawal.
This mandates having the heat transfer capacity of two RCS loops in operation.
If only one loop is in operation, the Rod Control System must be rendered incapable of rod withdrawal.
The Completion Time of 1 hour to defeat the Rod Control System is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period.D.1. D.2. and D.3 If two required RCS loops are inoperable, or two required RCS loops are not in operation with Rod Control System capable of rod withdrawal, or required RCS loop not in operation with Rod Control System not capable of rod withdrawal, the Rod Control System must be placed in a condition incapable of rod withdrawal (e.g., all CRDMs must be de-energized by opening the RTBs or de-energizing the MG sets). All operations involving introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 must be suspended, and action to restore one of the RCS loops to OPERABLE status and operation must be initiated.
Boron dilution requires forced circulation for proper mixing, and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal.
Suspending operations that would cause the introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 is required to assure continued safe operation.
With coolant added without forced circulation, unmixed coolant could be introdtqced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification evey--12.-heutffs that the required loops are in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.Cook Nuclear Plant Unit 2 B3454Rvso o B3.4.5-4 Revision No. 0 RCS Loops -MODE 3 B 3.4.5 BASES SURVEILLANCE REQUIREMENTS (continued) p h I I.1 r.1, ~ tA I ..J~I..J ..JL4 I~ ~
SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side water level is above the lower tap of the SG wide range level instrumentation by > 418.77 inches for required RCS loops. If the SG tubes become uncovered, the associated loop may not be capable of providing the heat sink for removal of the decay heat. The water level can be verified by either the wide range or the narrow range instruments.
A narrow range level instrument.
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. T-he--2-heuf-Fe~lt~eaey~s-eons-ider-ed-adlequiate I-nsert~ejleer-a3=r--te f-661e'te1'.
SR 3.4.5.3 II-[ -- r L-]-II Verification that each required RCP is OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat'removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power availability to each required RCP.Inser-t This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 2 B 3.4.5-5 Revision No. 0 RCS Loops -MODE 4 B 3.4.6 BASES ACTIONS (continued) minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SR 3.4.6.1 This SR requires verification evet,-24-2--eew that the required RCS or RHR loop is in operation and circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The-Peq1tey~ef rs a~ t-t --aqelr memrie-- q petfo-maiaee
.-."---nsert 2 SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side water level is above the lower tap of the SG wide range level instrumentation by >418.77 inches.If the SG U-tubes become uncovered, the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.The water level can be verified by either the wide range or the narrow range level instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. T-e--htfF-etee4 ire-'~ele wf --Insert 2 SR 3.4.6.3 Verification that each required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump. 4n4we-thr-giit-a~ec-a--~-vie~ea e 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 2 B3464Rvso o B 3.4.6-4 Revision No. 0 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES ACTIONS (continued) 0.1 and 0.2 If a required RHR loop is not in operation or if no required loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated.
Suspending operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 is required to assure continued safe operation.
With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum S.DM maintains acceptable margin to subcritical operations.
The immediate Completion Times reflect the importance of maintaining operation for heat removal.SURVEILLANCE REQUIREMENTS SR 3.4.7.1 This SR requires verification-ever-y4-2-heii1-s that the required loop is in operation circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
.*---'nsert 2 SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side water levels are above the lower tap of the SG wide range level instrumentation by > 418.77 inches ensures an alternate decay heat removal method via natural circulation in the event that the second RHR loop is not OPERABLE.
The water level can be verified by either the wide range or the narrow range instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. If both RHR loops are OPERABLE, this Surveillance is not needed. T-he-t-2-h~eF~eeley-is-
-Insert 2 rea+eae-#hertreh -lvh SR 3.4.7.3 Verification that each required RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by Cook Nuclear Plant Unit 2 B 3.4.7-4 Revision No. 0 Cook Nuclear Plant Unit 2 B3.4.7-4 Revision No. 0 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued) verifying proper breaker alignment and power available to each required RHR pump. If secondary side water level is above the lower tap of the SG wide range level instrumentation by > 418.77 inches in at least two SGs, this Surveillance is not needed. Th-Fr~eeitef9y-ef-7-daye-ts- Insert 2-~~e~~4 e- -thm,-e a-hae-beIs-sewf4Nc o c-pabc-lct .e This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES
: 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation." Cook Nuclear Plant Unit 2 B3475Rvso o B3.4.7-5 Revision No. 0 RCS Loops -MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE REQ UIRE MENTS SR 3.4.8.1 This SR requires verification
#ever-A2_-heiwrs that the required loop is in operation circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
"--Insert 2 SR 3.4.8.2 Verification that each required pump is OPERABLE ensures that an.additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump.
f-7da-" "~ .ciiod "c~Il in ic:of- .l-nsert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 2 B3483Rvso o B 3.4.8-3 Revision No. 0 Pressurizer B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level.
pafaet &c-,-hi~q--T-he42e-ieter-i#at4e-hx b-cathwnby j 1 I n sert 2ryasscssI
'-e-a:-dea-e--
SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated pressurizer backup heaters are verified to be at their specified capacity.
This may be done by testing the power supply output with the heaters energized.
T,-e l---nsert 2 F F4 ths dee- te-r REFERENCES
: 1. UFSAR, Chapter 14..2. NUREG-0737, November 1980.Cook Nuclear Plant Unit 2 B3494Rvso o B 3.4.9-4 Revision.
No. 0 Pressurizer PORVs B 3.4.11 BASES ACTIONS (continued) place the PORV(s) in manual control, this may not be possible for all causes of Condition B entry with PORV(s) inoperable and not capable of being manually cycled (e.g., as a result of failed control power fuse(s) or control switch malfunctions(s))
H.1 and H.2 If any Required Action and associated Completion Time of Condition A, B, C, D, E, F, or G is not met, if three PORVs are inoperable and not capable of being manually cycled, if two PORVs are inoperable and not capable of being manually cycled and one block valve inoperable (for reasons other than to comply with Required Action B.2) in a different line than the inoperable PORVs, or if one PORV is inoperable and not capable of being manually cycled and two block valves are inoperable (for reasons other than to comply with Required Action B.2) in different lines than the inoperable PORV, then the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block valve cycling verifies that the valve(s) can be opened and closed if needed.
2 This SR is modified by a Note, which states that this SR is not required to be performed with the block valve closed in accordance with the Required Actions of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable.
SR 3.4.11.2 SR 3.4.1 1.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. @#raige-e~~-qa 2@s~er-ti-mebt--tthoe eF Cook Nuclear Plant Unit 2 B 3.4.11-6 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.4.11-6 Revision No. 0 Pressurizer PORVs B 3.4.11 BASES SURVEILLANCE REQUIREMENTS (continued)
The Note modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. In accordance with Reference 4, administrative controls require this test be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation.
SR 3.4.11.3 Operating the solenoid air control valve associated with each PORV, and the check valves on the air accumulators where applicable, ensures the PORV control system actuates properly when called upon. ep-a~in 2 T4- en thelu Fcr-wcg-uee 1 1 e-e#4he~h eF REFERENCES
: 1. Regulatory Guide 1.32, February 1977.2. UFSAR, Section 14.1.8.3. ASME, Operation and Maintenance Standards and Guides (OM Codes).4. Generic Letter 90-06, "Resolution of Generic Issue 70,'Power-Operated Relief Valve and Block Valve Reliability,'
and Generic Issue 94, 'Additional Low-Temperature Overpressure for Light-Water Reactors,'
Pursuant to 10 CFR 50.54(f)," June 25, 1990.Cook Nuclear Plant Unit 2B34117RvsoNo0 B 3.4.11-7 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued) through the pump control .switch being placed in pull to lock and at least one valve in the discharge flow path being closed, or at least one valve in the discharge flow path being closed and sealed or locked.In addition, SR 3.4.12.3 is modified by a Note that allows the accumulator discharge isolation valve position to be verified by administrative means.This is acceptable since the valve position was verified prior to deactivating the valve, access to the containment is restricted, and valves are only operated under strict procedural control.T--h
--Insert 2 4 4~f~e SR 3.4.12.4 The required RHR suction relief valve shall be demonstrated OPERABLE by verifying the RHR suction isolation valves are open. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.The RHR suction isolation valves are verified to be opened evePfy-1-2 I .I r -Insert 2 SR 3.4.12.5 The ROS vent of> 2.0 square inches or a blocked open PORV is proven OPERABLE by verifying its open condition .eithef:
2 e~-=G9z-evcr; 1-2z4y eufcr-aa ef~t vaeq cJ 1 -ee~et The passive vent path arrangement must only be open if the vent is being used to satisfy the pressure relief requirements of LCO 3.4.12.A.2.c.
Cook Nuclear Plant Unit 2 B 3.4.12-11 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.4.12-11 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.12.6 The PORV block valve must be verified open t6 provide the flow path for each required PORV topgerform its function when actuated.
The valve must be remotely verified open in the main control room. This Surveillance is performed if one~or more PORVs satisfy the LCO.The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required removed, and the manual operator is not required locked in the inactive position.
Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.
ho 72hor reunc e in viw -ethe... l--nsert 2 I~a~=ee6fftet9-.&sect; e....SR 3.4.12.7 Verification eveiy-34-aey-s-.that each required emergency air tank bank's pressure is > 900 psig assures adequate air pressure for reliable PORV operation.
With the emergency air supply at > 900 psig, there will be enough air to support PORV operation for 10 minutes with no operator action upon a loss of control air. T~e-he----y-F~r t~eykeesht-e-c-eeeee~a tieee-e~vrle-te~-tee 2 SR 3.4.12.8 Performance of a COT is required eve~y-34-days-on each required PORV to verify and, as necessary, adjust its lift setpoint.
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 COT 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 COT will verify the setpoint is within the LCO limit. PORV actuation could depressurize the RCS and is not required.Cook Nuclear Plant Unit 2 B341-2Rvso o B 3.4.12-12 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)
A Note has been added indicating that this SR is not required to be performed until 12 hours after decreasing RCS cold leg temperature to_< 299&deg;F. The COT cannot be performed until in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP setting. The test must be performed within 12 hours after entering the LTOP MODES. -Fh ,- Iner 2*
-vrFs F-v beJi#tit  SR 3.4.12.9 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required-ee to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input. .=-Insert 2 REFERENCES
: 1. 10 CFR 50, Appendix G.2. Generic Letter 88-11.3. ASME, Boiler and Pressure Vessel Code, Section II1.4. WCAP-1 3235, "Donald C. Cook Units I & 2, Analysis of Low Temperature Overpressurization Mass Injection Events with Pressurizer Steam Bubble and RHR Relief Valve, March 1992;"WCAP-12483.
Revision 1, "Analysis of Capsule U From the American Electric Power Company D. C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program, December 2002;" and WCAP-13515, Revision 1, "Analysis of Capsule U From Indiana Michigan Power Company D. C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program, May 2002." 5. 10 CFR 50, Section 50.46.6. 10 CFR 50, Appendix K.7. Generic Letter 90-06.Cook Nuclear Plant Unit 2 B341-3Rvso o B 3.4.12-13 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued) operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation." Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.eeee ---nsert 2 SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary, to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated.
The primary to secondary LEAKAGE is measured at room temperature as described in Reference
: 5. Prior to comparison with the 150 gallons per day TS limit, the measured primary to secondary LEAKAGE is multiplied by a volume correction factor of 1.52. The correction factor ensures the offsite dose analyses, which assume primary to secondary leakage is at normal operating temperature and pressure, remain bounding.
The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all of the primary to~secondary LEAKAGE should be conservatively assumed to be from one SG.The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours after establishment of steady state operation.
For ROS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.Cook Nuclear Plant Unit 2 B 3.4.13-5 Revision No. 15 Cook Nuclear Plant Unit 2 B 3.4.13-5 Revision No. 15 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued) th ivIc
* F;-qn ef -Ie b~. e_4 nterv~ake4I..e4
-t~r4ma -e~ee~i-rare Insert 2 4m4fcgas The pri mary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5).REFERENCES
: 1. UFSAR, Section 1.4.3.2. Regulatory Guide 1.45, May 1973.3. UFSAR, Section 14.2.4.4. NEl 97-06, "Steam Generator Program Guidelines." 5. EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines." Cook Nuclear Plant Unit 2 B341- eiinN.1 B 3.4.13-6 Revision No. 12 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued) potential for an unplanned transient if the Surveillance were performed with the reactor at power.The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.
Therefore, this SR is modified by a Note that states the Surveillance is only required to be performed in MODES 1 and 2. Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance.
SR 3.4.14.2 Verifying that the RHR interlock that prevents the valves from being opened is OPERABLE ensures that RCS pressure will not pressurize the RHR System beyond its design pressure of 600 psig.
Insert 2 REFERENCES
: 1. 10 CFR 50.2.2. 10 CFRS50.55a(c).
: 3. WASH-i1400 (NUREG-75/01 4), Appendix V, October 1975.4. Letter from D.G. Eisenhut, NRC, to all LWR licensees, LWR Primary Coolant System Pressure Isolation Valves, February 23, 1980.5. Letter from S.A. Varga, NRC, to J. Dolan, Order for Modification of Licenses Concerning Primary Coolant System Pressure Isolation Valves, April 20, 1981.6. Technical Requirements Manual.7. EGG-NTAP-61 75, Inservice Testing of Primary Pressure Isolation Valves, Idaho National Engineering Laboratory, February 1983.8. NRC Safety Evaluation for License Amendment 174.9. ASME, Operation and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 2B34145RvsoN.0 B 3.4.14-5 Revision No. 0 RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued)
Completion Time ensures that the plant will not be operated in a degraded configuration for a lengthy time period.I E.1 and E.2 With the required containment atmosphere radioactivity monitor and the containment humidity monitor inoperable, the only means of detecting leakage is the containment sump monitor. This Condition does not provide the required diverse means of leakage detection.
The Required Action is to restore either of the inoperable required monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity.
The 30 day Completion Time ensures that the unit will not be operated in a reduced configuration for a lengthy time period.F.1 and F.2 If any Required Action and associated Completion Time of Condition A, B, C, 0, or E cannot be met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.G._1_With all three types of required monitors inoperable (i.e., LCO 3.4.15.a, b, and c not met), no automatic means of monitoring leakage are available, and immediate unit shutdown in accordance with LCO 3.0.3 is required.SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is operating properly. iset SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. 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 Cook Nuclear Plant Unit 2 B 3.4.15-6 Revision No. 32 RCS Leakage Detection Instrumentation B 3.4.15 BASES SURVEILLANCE REQUIREMENTS (continued) 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 test verifies the alarm setpoint and relative accuracy of the instrument string. The SR 3.4.15.3.
SR 3.4.15.4.
and SR 3.4.15.52 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels.
The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.
T-hareequeney-ef-94--me&#xb6;9-hs
~=-=insert2 ety=ai~d=eper.a4Ai9fre~~
~r~tei9~r~
A9e~-fr=evei9=that
=eeee~e.REFERENCES 1." UFSAR, Section 1.4.3.2. Regulatory Guide 1.45, Rev. 0, "Reactor Coolant Pressure Boundary Leakage Detection System," May 1973.3. AEP Letter to NRC, AEP:NRC:0137D, "NRC Generic Letter 84-04;Elimination Of Postulated Pipe Breaks In Primary Main Loops Generic Issue A-2, Asymmetric Blowdown Loads On PWR Primary Systems Request For License Condition Deletion," dated September 10, 1984.4. NRC Letter to AEP, "Generic Letter 84-04, Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," dated November 22, 1985.5. UFSAR, Section 4.2.7 Cook Nuclear Plant Unit 2 B341- eiinN.3 B 3.4.15-7 RCS Specific Activity B 3.4.16 BASES ACTIONS (continued)
B.1 If any Required Action and associated Completion Time of Condition A is not met, if the DOSE EQUIVALENT 1-131 is in the unacceptable region of Figure 3.4.16-1, or if gross specific activity of the reactor coolant is not within limit, the reactor must be brought to MODE 3 with RCS average temperature
< 500&deg;F within 6 hours. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 below 5000.F from full power conditions in an orderly manner and-without challenging unit systems.SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at-least--enee-every-While basically a
measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in gross specific activity.Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions.Th Insert 2u-e d U ing-t4-f e~ie.SR 3.4.16.2 This Surveillance requires the verification that the reactor coolant DOSE EQUIVALENT 1-131 specific activity is within limit. This Surveillance is accomplished by performing an isotopic analysis of a reactor coolant sample. This Surveillance is performed in MODE 1 only to ensure iodine remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur..Th--
y~rqem~ Insert 2
~ ~ gr aci.yi4aa The Frequency, between 2 and 6 hours after a power change >- 15% RTP within a 1 hour period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.SR 3.4.16.3 A radiochemical analysis for determination is required with the unit operating in MODE 1 equilibrium conditions.
The determination directly relates to the LCO and is required to verify unit Cook Nuclear Plant Unit 2B34164RvsoN.0 B 3.4.16-4 Revision No. 0 RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued) operation within the specified gross activity LCO limit. The analysis for is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. rqieerty "-'-=-sert 2 This SR has been modified by a Note that indicates sampling is not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours. This ensures that the radioactive materials are at equilibrium so the analysis for IE is representative and not skewed by a crud burst or other similar abnormal event.REFERENCES
: 1. 10OCFRI100.11.
: 2. UFSAR, Section 14.2.4.Cook Nuclear Plant Unit 2B3.165RvsoN.0 B 3.4.16-5 Revision No. 0 Accumulators B 3.5.1 BASES ACTIONS (continued) reach the required plant conditions from full power conditions in an orderly manner and without challenging unit systems.0.1 If more than one accumulator is inoperable, the unit is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE REQUIREMENTS SR 3.5.1.1 Each accumulator isolation valve should be verified to be fully open @evey.A-2--heuis.-,.
This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the IRCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions.
ide~ed .SR 3.5.1.2 and SIR 3.5.1.3-Insert 2borated water volume and nitrogen cover pressure are verified for each accumulator.
T-i-Frq~fey
~ ade aene~e-tri~---O e-ue e-uu4yaiw~eeeqe4 ehae-~fr4et-~-eee-O~ra~ge#4ch  Fr ~ e :-mn-eet n-fe*'--Insert 2/SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulatorei the static design of the accumulators limits the ways in which the concentration can be changed.Th-daen eee r < Insert 2
~ , Sampling the affected accumulator within 6 hours after a volume increase of 13 ft 3 will identify whether inleakage has caused a reduction in boron concentration to below the required limit. lt is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (IRWST), because the water contained in the RWST is within the accumulator boron concentration requirements.
This is consistent with the recommendation of NUIREG-1366 (iRef. 4).Cook Nuclear Plant Unit 2B3516ReionN.0 B3.5.1-6 Revision No. 30 Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.5 Verification every 31 days that power is removed from each accumulator isolation valve operator when the RCS pressure is > 2000 psig ensures that an active failure could not result inthe closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. Sincospowor,,ec i
2 This SR allows power to be supplied to the motor operated isolation valves when ROS pressure is < 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns.
REFERENCES
: 1. UFSAR, Section 14.3.2. 10 CFR 50.46.3. WCAP-1 5049-A, "Risk-Informed Evaluation of an Extension to Accumulator Completion Times," Rev. 1, April 1999.4. NUREG-1 366, February 1990.Cook Nuclear Plant Unit 2 B3517Rvso o B3.5.1-7 Revision No. 0 ECCS -Operating B 3.5.2 BASES ACTIONS (continued) train available, the facility is in a condition outside of the accident analyses.
Therefore, LCO 3.0.3 must be en~tered immediately.
SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained.
Misalignment of these valves could render both ECCS trains inoperable.
Securing these valves in position by locking out control power ensures that they cannot change position as a result of an active failure or be inadvertently misaligned.
These valves are of the type, described in Reference 9, that can disable the function of both ECCS trains and invalidate the accident analyses..A-f}s 2 SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation.
Rather, it involves verification that those valves capable of being mispositioned are in the correct position.
The -rfun ~ f4 Ch v
<--- Insert 2 SR 3.5.2.3 Verifying that each ECCS pump's developed head at the flow test point is greater than or equal to the required developed head ensures that ECCS pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of ECCS pump performance required by the ASME OM Code (Ref. 10). Since the ECCS pumps cannot be tested with flow through the normal ECCS flow paths, they are tested on recirculation flow (RHR and SI pumps) or normal charging flow path (centrifugal charging pumps). This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect Cook Nuclear Plant Unit 2B3527ResinN.0 B3.5.2-7 Revision No. 30 EGOS -Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued) incipient failures by indicating abnormal performance.
The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.5.2.4 and SR 3.5.2.5 These Surveillances demonstrate that each automatic ECOS valve actuates to the required position on an actual or simulated SI signal and that each EGOS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
T~he- s--a -- -~c, du.i
"-vcl~n0 -e-per@efmedec-sas--s%{ae Ilnsert 2 SR 3J.5.2.6 Proper throttle valve position is necessary for proper EGOS performance.
These valves have stops to allow proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This Surveillance verifies the mechanical stop of each listed EGOS throttle valve is in the correct position.
T~he a{Ie4t Fe ~ -bseSEsZ-3.-i5.2.5.-B SR 3.5.2.7 Periodic inspections of the containment sump suction inlets ensure that they are unrestricted and stay in proper operating condition.
This Surveillance verifies that the sump suction inlets are not restricted by debris and the suction inlet strainers show no evidence of structural distress, such as openings or gaps, which would allow debris to bypass the strainers..-Phe-24nrot-F-reqtr~~
~ ,J the-need-toffave a-t...esmt-rl -T-UF- uny-hsbe-ft teee
-Insert 2 Ilnsert 2 REFERENCES
: 1. UFSAR, Section 1.4.7.2. 10 CFR 50.46.3. UFSAR, Section 14.3.1.4. UFSAR, Section 14.3.2.Cook Nuclear Plant Unit 2 B 3.5.2-8 Revision No. 16 Cook Nuclear Plant Unit 2 B 3.5.2-8 Revision No. 16 RWST B 3.5.4 BASES ACTIONS (continued) brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.5.4.1 The RWST borated water temperature should be verified everj24iei~sssle .__T_ee~~,dfeett et~~~ I'-nsernt 2
er~qeacltbetr~
SR 3.5.4.2 The RWST water volume should be verified be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. q.ex1perie19~e:
4'=-Ilnsert 2 SR 3.5.4.3 The boron concentration of the RWST should be verified ,evefiWy==dys to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized.
* ecepet*q~epheug-per-amieitr~ree.
*REFERENCES
: 1. UFSAR, Section 6.2.2.2. UFSAR, Section 14.3.Cook Nuclear Plant Unit 2 B3545Rvso o B3.5.4-5 Revision No. 0 Seal Injection Flow B 3.5.5 BASES APPLICABILI.TY In MODES 1, 2, and 3, the seal injection flow resistance limit is dictated by ECCS flow requirements, which are specified for MODES 1, 2, 3, and 4. The seal injection flow resistance limit is not applicable for MODE 4 and lower, however, because high seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES. Therefore, RCP seal injection flow resistance must be limited in MODES 1, 2, and 3 to ensure adequate ECCS performance.
ACTIONS A.1 With the seal injection flow resistance not within its limit, the amount of charging flow available to the RCS may be reduced. Under this condition, action must be taken to restore the flow resistance to within its limit. The operator has 4 hours from the time the flow resistance is known to not be within the limit to correctly position the manual valves and thus be in compliance with the accident analysis.
The Completion Time minimizes the potential exposure of the unit to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow resistance within limits. This time is conservative with respect to the Completion Times of other ECOS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.
B.1 and B.2 When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated.
The Completion Time of 6 hours for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal coold own rates, and does not challenge plant safety systems or operators.
Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.
SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Verification e -.=dys that the seal injection flow resistance is within the limit ensures that the ECCS injection flows stay within the safety analysis.
A differential pressure is established between the charging header and the RCS, and the total seal injection flow is verified to be within the limit determined in accordance with the ECCS safety analysis.The flow resistance shall be >- 0.227 ftlgpm 2.Cook Nuclear Plant Unit 2 B 3.5.5-3 Revision No. 0 Cook Nuclear Plant Unit 2 B3.5.5-3 Revision No. 0 Seal Injection Flow B 3.5.5 BASES SURVEILLANCE REQUIREMENTS (continued)
The seal injection flow resistance, RSL, is determined from the following expression:
RSL = 2.31 (PoHP-PsI)/Q 2 where: PCHP = charging pump header pressure (psig);= 2300 psig (high pressure operation);
and Q =total seal injection flow (gpm).Sdz.j3 i3 bc3cd 3n ongin&cring judgrnor.t and iz eet~tsi~ste*wit~
~tl 1 e~ EGOS v;l~c OelIaic~ F~qu~.1 ci~. T[-.~~ bo acczptabl c ~2 As noted, the Surveillance is not required to be performed until 4 hours after the pressurizer pressure has stabilized within a + 20 psig range of normal operating pressure.
The pressurizer pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly.
The pressurizer pressure indications are averaged to determine whether the appropriate pressure has been achieved.
The exception is limited to 4 hours to ensure that the Surveillance is timely.REFERENCES
: 1. UFSAR, Section 14.3.1.2. UFSAR, Section 14.3.2.3. UFSAR, Section 14.2.4.4. UFSAR, Section 14.2.5.Cook Nuclear Plant Unit 2 B3554Rvso o B 3.5.5-4 Revision No. 0 Containment Air Locks B 3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued) air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by the Containment Leakage Rate Testing Program.The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria Which is applicable to SR 3.6.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C containment leakage rate.SR 3.6.2.2 The air Jock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY.
Thus, the door interlock feature supports containment OPERABILITY while the air Jock is being, used for personnel transit in and out of the containment.
Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur.
S-s- t u-' ro cr trj-;-4nd erl- eK -iv t; f.i~f&ti~  me seerr-nt-nner-en ve~el en ir a trict-aelhrieo-te2$ly p U.J.) LI IL, ....JLII U L.IIILAI IL,'.., III,. IL,,* ~JtI I US II ILU.J 1.11. t.r SI ELI 111.1111 L.LJL'..,I
*Uy. S*a4-lec-ke.
8et REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 14.2.6.3. UFSAR, Section 5.7.4. 10 CFR 50, Appendix J, Option B.Cook Nuclear Plant Unit 2 B3626Rvso o B 3.6.2-6 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES ACTIONS (continued) locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned.
Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is small.D.1 and D.2 If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.6.3.1 REQUI REM ENTS This SR ensures that the containment purge supply and exhaust valves are closed as required or, if open, open for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable.
The SR is not required to be met when the containment purge valves are open for the reasons stated. The valves may be opened for pressure control, ALARA or air quality considerations for personnel entry, or for Surveillances or maintenance activities that require the valves to be open. The containment purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. T- re~/ -ie 2 SR 3.6.3.2 This SR requires verification that each containment isolation manual valve and blind flange located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. This SR does not require any testing or valve manipulation..
Rather, it involves verification that those containment isolation valves outside containment and capable of being mispositioned are in the correct position. 2 Cook Nuclear Plant Unit 2 B3637Rvso o B3.6.3-7 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time the valves are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since acdess to these areas is typically restricted during MODES 1, 2, 3, and 4 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.SR 3.6.3.3 This SR requires verification that each containment isolation manual valve and blind flange located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design, limits. For containment isolation valves inside containment, the Frequency of "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time they are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.This Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4, for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.Cook Nuclear Plant Unit 2 B 3.6.3-8 Revision No. 0 Cook Nuclear Plant Unit 2 B3.6.3-8 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.3.4 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY.
The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses.The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.3.5 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. ~ teey~-ae~-
ne~'l~~~-~~~rve~--
p g-'Insert 2 REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 14.2.6.3. UFSAR, Section 5.4.1 and Table 5.4-1.Cook Nuclear Plant Unit 2 B3639Rvso o B3.6.3-9 Revision No. 0 Containment Pressure B 3.6.4 BASES SURVEILLANCE REQ U IREM ENTS SR 3.6.4.1 Verifying that containment pressure is within limits ensures that unit operation remains within the limits assumed in the containment analysis.1hz 12ie hour-eee
~R R &#xa2; 2 REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 5.2.2.2.3. 10 CFR 50, Appendix K.Cook Nuclear Plant Unit 2 B3643Rvso o B 3.6.4-3 Revision No. 0 Containment Air Temperature 1B 3.6.5 BASES ACTIONS A.1I When containment average air temperature in the upper or lower compartment is not within the limit of the LCOG, the average air temperature in the affected compartment must be restored to within limits within 8 hours. This Required Action is necessary to return operation to within the bounds of the containment analysis.
The 8 hour Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.B.1 and B.2 If the containment average air temperature cannot be restored to within its limits within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.5.1 and SR 3.6.5.2 Verifying that containment average air temperature is within the LCOG limits ensures that containment operation remains within the limits assumed for the containment analyses.
In order to determine the containment average air temperature, an average is calculated using measurements taken at locations within the containment selected to provide a representative sample of the overall containment atmosphere.
In the upper compartment, two locations at a nominal elevation of 712 ft o inches and a third location at a nominal elevation of 624 ft 10 inches are used and an arithmetic average taken. In the lower compartment, a volume weighted average temperature is calculated whereby the volume fraction for each of the various areas of lower containment is multiplied by the representative temperature, utilizing one or more temperature instruments, in that volume. In this way the temperatures are "weighted" according to the volume fraction.
These weighted temperatures are then summed to determine the Weighted Average Temperature for Lower Containment.
=-Insert 2 REFERENCES
: 1. U0FAR50Sctin49.34
: 2. 100FR5O.49.
Cook Nuclear Plant Unit 2 B3653Rvso o 13 3.6.5-3 Revision No. 8 Containment Spray System B 3.6.6 BASES ACTIONS A.1 With one containment spray train inoperable, the affected train must be restored to OPERABLE status within 72 hours. The components in this degraded condition are capable of providing 100% of the heat removal and iodine removal needs after an accident.
The 72 hour Completion Time was developed taking into account the redundant heat removal and iodine removal capabilities afforded by the OPERABLE train and the low probability of a DBA occurring during this period.B.1 and B.2 If the affected containment spray train cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 84 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Verifying the correct alignment of manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position since they were verified in the correct position prior to being secured. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation.
Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned, are in the correct position.
2 SR 3.6.6.2 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 5).Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect Cook Nuclear Plant Unit 2B366-ReionN.0 B3.6.6-5 Revision No. 30 Containment Spray System B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued) incipient failures by indicating abnormal performance.
The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.6.3 and SR 3.6.6.4 These SRs require verification that each automatic containment spray valve actuates to its correct position and each containment spray pump starts upon receipt of an actual or simulated containment spray actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative.
controls.
T'he e 4sJte 2.... i4y u au t , 4 These Surveillances include a Note that states that in MODE 4, only the manual portion of the actuation signal is required.
This is acceptable since the automatic portion of the actuation signal is' not required to be OPERABLE by ITS 3.3.2, "Engineered Safety Features Actuation System (ESFAS) Instrumentation." SR 3.6.6.5 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections.
This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded.
The event based surveillance frequency following maintenance that could result in nozzle blockage was chosen because this passive portion of the system is not susceptible to service induced degradation.
REFERENCES
: 1. UFSAR, Section 1.4.7.2. UFSAR, Section 14.3.4.3. 10 CFR 50.49.4. 10 CFR 50, Appendix K.5. ASME, Operation and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 2B366-ReionN.1 B3.6.6-6 Revision No. 31 Spray Additive System B 3.6.7 BASES ACTIONS A.1j If the Spray Additive System is inoperable, it must be restored to OPERABLE within 72 hours. The pH adjustment of the Containment Spray System flow for corrosion protection and iodine retention enhancement is reduced in this condition.
The 72 hour Completion Time takes into account the redundant flow path capabilities and the low probability of the worst case DBA occurring during this period. In addition, if the Containment Spray System is available, it would remove some iodine from the containment atmosphere in the event of a DBA.B.1 and B.2 If the Spray Additive System cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 84 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems. The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for the release of radioactive material from the Reactor Coolant System is reduced in MODE 3.SURVEILLANCE REQUIREMENTS SR 3.6.7.1 Verifying the correct alignment of Spray Additive System manual, power operated, and automatic valves in the spray additive flow path provides assurance that the system is able to provide additive to the Containment Spray System in the event of a DBA. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation.
Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned are in the correct position. Insert 2 SR 3.6.7.2 To provide effective iodine retention, the containment spray must be an alkaline solution.
Since the RWST contents are normally acidic, the volume of the spray additive tank must provide a sufficient volume of spray additive to adjust pH for all water injected.
This SR is performed to verify the availability of sufficient NaOH solution in the Spray Additive System.
"=--'Insert 2 Cook Nuclear Plant Unit 2 B 3.6.7-3 Revision No. 0 Spray Additive System B 3.6.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.7.3 This SR provides verification (by chemical analysis) of the NaOH concentration in the spray additive tank and is sufficient to ensure that the spray solution being injected into containment is at the correct pH level.4 ~et4~ t
.
b SR 3.6.7.4 This SR provides verification that each automatic valve in the Spray Additive System flow path actuates to its correct position.
This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.T4 i Insert 22
* ec pa D-rc Womsateepet.
SR 3.6.7.5 To ensure that the correct pH level is established in the borated water solution provided by the ContainmentSpray System, the flow rate in the Spray Additive System is verified once every 5 years. This SR provides assurance that the correct amount of NaOH will be metered into the flow path upon Containment Spray System initiation.
The test is performed by verifying the flow rate from the spray additive tank test line to each Containment Spray System train with each containment spray pump operating in the recirculation mode.
Insert 2 REFERENCES
: 1. UFSAR, Chapter 14.3.5.9.Cook Nuclear Plant Unit 2 B3674Rvso o B3.6.7-4 Revision No. 0 DIS B 3.6.9 BASES ACTI ONS B.1 Condition B is one containment region with no OPERABLE hydrogen ignitor. Thus, while in Condition B, or in Conditions A and B simultaneously, there would always be ignition capability in the adjacent containment regions that would provide redundant capability by flame propagation to the region with no OPERABLE ignitors.Required Action B.1 calls for the restoration of one hydrogen ignitor in each region to OPERABLE status within 7 days. The 7 day Completion Time is based on the same reasons given under Required Action A. 1.C..11 If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCO does not apply. This is done by placing the unit in at least MODE 3 within 6 hours. The allowed.Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.9.1 This SR confirms that >- 34 of 35 hydrogen ignitors (or > 33 ignitors if allowed by footnote) can be successfully energized in each train. The ignitors are simple resistance elements.
Therefore, energizing provides assurance of OPERABILITY.
The allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise redundancy in that region, the containment regions are interconnected so that ignition in one region would cause burning to progress to the others (i.e., there i~s~overlap in each hydrogen ignitor's effectiveness between regions).
Tt
,, Insert 2 SR 3.6.9.2 This SR confirms that the two inoperable hydrogen ignitors allowed by SR 3.6.9.1 (i.e., one in each train) are not in the same containment region. -Insert 2 Footnote:
For the remainder of Fuel Cycle 18, or until the next entry into a MODE which allows replacement of the affected ignitors, DIS Train B can still perform its safety function and may be considered OPERABLE with one lower containment Phase 2 Power Supply ignitor inoperable and with one lower containment Phase 3 Power Supply ignitor inoperable (Reference 3).Cook Nuclear Plant Unit 2 B 3.6.9-4 Revision No. 25 DIS B 3.6.9 BASES SURVEILLANCE REQUIREMENTS SR 3.6.9.3 A more detailed functional test is performed every 24 months to verify system OPERABILITY.
Each ignitor is visually examined to ensure that it is clean and that the electrical circuitry is energized.
All ignitors, including normally inaccessible ignitors, are visually checked for a glow to verify that they are energized.
Additionally, the surface temperature of each ignitor (see footnote) is measured to be > 1700&deg; F to demonstrate that a temperature sufficient for ignition is achieved.
The-e.~-F-r-e41eey
<=l'nsert 2
~e~~eio eaie~ts-ueu'eli paee4i tae-eeneliide~tebe' vi--- ............
.... J ....... r REFERENCES
: 1. 10 CFR 50.44.2. UFSAR, Section 5.8.3. Letter from T. Beltz, NRC, to J. Jensen, l&M, "Donald C. Cook Nuclear Plant, Unit 2 -Issuance of Exigent Amendment Re: The Containment Distributed Ignition System (TAC No. ME3129)," dated February 4, 2010.Footnote:
For the remainder of Fuel Cycle 18, or until the next entry into a MODE which allows replacement of the affected ignitors, DIS Train B can still perform its safety function and may be considered OPERABLE with one lower containment Phase 2 Power Supply ignitor inoperable and with one lower containment Phase 3 Power Supply ignitor inoperable (Reference 3).Cook Nuclear Plant Unit 2 B 3.6.9-5 Revision No. 39 Cook Nuclear Plant Unit 2 B 3.6.9-5 Revision No. 39 CEQ System B 3.6.10 BASES LCO In the event of a DBA, one train of the CEQ System is required to provide the minimum air recirculation for heat removal and hydrogen mixing assumed in the safety analyses.
To ensure this requirement is met, two trains of the CEQ System must be OPERABLE.
This will ensure that at least one train will operate, assuming the worst case single failure occurs.APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the CEQ System.Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the CEQ System is not required to be OPERABLE in these MODES.ACTIONS A.1 If one of the trains of the CEQ System is inoperable, it must be restored to OPERABLE status within 72 hours. The components in this degraded condition are capable of providing 100% of the flow and hydrogen skimming needs after an accident.
The 72 hour Completion Time was developed taking into account the redundant flow and hydrogen skimming capability of the OPERABLE CEQ System train and the low probability of a DBA occurring in this period.B.1 and B.2 If the CEQ System train cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.10.1 Verifying that each CEQ System fan starts on an actual or simulated actuation signal, after a delay > 108 seconds and < 132 seconds, and operates for > 15 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly.
It also ensures that blockage, fan and/or motor failure, or excessive vibration can be detected for corrective action. ifhe. Insert 2 ari~ elaac~~~~le Cook Nuclear Plant Unit 2B361-3RvsoN.0 B 3.6.10-3 Revision No. 0 C EQ System B 3.6.10 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR has been modified by a Note that states that this Surveillance is only required to be met in MODES 1, 2, and 3. This allowance is necessary since the specified delay (i.e., > 108 seconds and< 132 seconds) is only applicable to the automatic actuation signal (i.e., Containment Pressure -High), which is only required to be OPERABLE in MODES 1, 2, and 3. In addition, LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," requires the CEQ System Manual Initiation Function to be OPERABLE in MODE 4 and requires the performance of a TADOT every 24 months. This requirement will ensure the Manual Initiation Function can actuate the required equipment in MODE 4.SR 3.6.10.2 Verifying, with the return air fan discharge backdraft damper locked closed and the fan motor energized, the static pressure between the fan discharge and the backdraft damper is > 4.0 inches water gauge confirms one operating condition of the fan. This test is indicative of overall fan motor performance.
Such tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance..T4h.
.Preq -etst ge r''tsfrs ia-e~
eia 2 SR 3.6.10.3 Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exist when the fan is started. By applying _the correct counterweight, the damper operation can be confirmed.
The Insert 2d ~~rte~e-a~ae
~
Qp ae e.SR 3.6.10.4 Verifying the OPERABILITY of the motor operated valve in the hydrogen skimmer header provides assurance that the proper flow path will exist when the valve receives an actuation signal. Th9 ~zya*-Frequlen-ywes Insert 2 Cook Nuclear Plant Unit 2 B 3.6.10-4 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.6.10-4 RevisionNo.
0 Ice Bed B 3.6.11 BASES APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice bed.Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ice bed is not required to be OPERABLE in these MODES.ACTIONS A.1 If the ice bed is inoperable, it must be restored to OPE'RABLE status within 48 hours. The Completion Time was developed based on operating experience, which confirms that due to the very large mass of stored ice, the parameters comprising OPERABILITY do not change appreciably in this time period. If a degraded condition is identified, even for temperature, with such a large mass of ice it is not possible for the degraded condition to significantly degrade further in a 48 hour period.Therefore, 48 hours is a reasonable amount of time to correct a degraded condition before initiating a shutdown.B.1 and B.2 If the ice bed cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQU IREMENTS SR 3.6.11.1 Verifying that the maximum temperature of the ice bed is < 27&deg;F ensures that the ice is kept well below the melting point. -T-he-4-2=le -e e y~ea--yo se t taj~
a aeesig4 a~m rfao .--
u l f er-ater44 nJter-ia e-Svetem Cook Nuclear Plant Unit 2 B 3.6.11-4 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.6.11-4 Revision No. 0 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued)
The total ice mass and individual radial zone ice mass requirements defined in this Surveillance, and the minimum ice mass per basket requirement defined by SR 3.6.11.3, are the minimum requirements for OPERABILITY.
Additional ice mass beyond the SRs is maintained to address sublimation.
This sublimation allowance is generally applied to baskets in each radial zone, as appropriate, at the beginning of an operating cycle to ensure sufficient ice is available at the end of the operating cycle for the ice condenser to perform its intended design function.:F~'rI~ny'4=et~aelae'~e~e~g <-==I nserti 2 tye~-u .v*Ra+I9teiiae4.
SR 3.6.11.3 Verifying that each selected sample basket from SR 3.6.11.2 contains at least 600 lbs of ice in the as-found (pre-maintenance) condition ensures that a significant localized degraded mass condition is avOided.This SR establishes a per basket limit to ensure any ice mass degradation is consistent with the initial conditions of the DBA by not significantly affecting the containment pressure response.
Reference 2 provides insights through sensitivity runs that~demonstrate that the containment peak pressure during a DBA is not significantly affected by the ice mass in a large localized region of baskets being degraded below the required safety analysis mean, when the radial zone and total ice mass requirements of SR 3.6.11.2 are satisfied.
Any basket identified as containing less than 600 lbs of ice requires appropriately entering ACTION A for an inoperable ice bed due to the potential that it may represent a significant condition adverse to quality.As documented in Reference 2, maintenance practices actively manage individual ice basket mass above the required safety analysis mean for each radial zone. Specifically, each basket is serviced to keep its ice mass above 1132 lbs for zone A, 1132 lbs for zone B, and 1132 lbs for zone C. If a basket sublimates below the safety analysis mean value, this instance is identified within the CNP corrective action program, including evaluating maintenance practices to identify the cause and correct any deficiencies.
These maintenance practices provide defense in depth beyond compliance with the ice bed Surveillance Requirements by limiting the occurrence of individual baskets with ice mass less than the required safety analysis mean.Cook Nuclear Plant Unit 2 B 3.6.11-6 Revision No. 0 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued) the ice basket support platform is not a more restrictive flow area because it is easily accessible from the lower plenum and is maintained clear of ice accumulation.
There is no mechanistically credible method for ice to accumulate on the ice basket support platform during unit operation.
Plant and industry experience has shown that the vertical flow area through the ice basket support platform remains clear of ice accumulation that could produce blockage.
Normally only a glaze may develop or exist on. the ice basket support platform which is not significant to blockage of flow area. Additionally, outage maintenance practices provide measures to clear the ice basket support platform following maintenance activities of any accumulation of ice that could block flow areas.Frost buildup or loose ice is not to be considered as flow channel blockage, whereas attached ice is considered blockage of a flow channel.Frost is the solid form of water that is loosely adherent, and can be brushed off with the open hand. =-.-Insert 2 SR 3.6.11.5 This SR ensures that a representative sampling of ice baskets, which are relatively thin walled, perforated cylinders, have not been degraded by wear, cracks, corrosion, or other damage. The SR is designed around a full-length inspection of a sample of baskets, and is intended to monitor the effect of the ice condenser environment on ice baskets. The groupings defined in the SR (two baskets in each azimuthal third of the ice bed) ensure that the sampling of baskets is reasonably distributed.
The Frequency of 40 months for a visual inspection of the structural soundness of the ice baskets is based on engineering judgment and considers such factors as the thickness of the basket walls relative to corrosion rates expected in their service environment and the results of the long term ice storage testing. "Iflnsert 2 SR 3.6.11.6*Verifying the chemical composition of the stored ice ensures that the stored ice has a boron concentration
> 1800 ppm and <2300 ppm as sodium tetraborate and a high pH, > 9.0 and < 9.5 at 25&deg;C, in order to meet the requirement for borated water when the melted ice is used in the ECCS recirculation mode of operation.
Additionally, the minimum boron concentration limit is used to assure reactor subcriticality in a post LOCA environment, while the maximum boron concentration limit is used as the bounding value in the hot leg switchover timing calculation (Ref. 4). This is accomplished by obtaining at least 24 ice samples. Each sample is taken approximately one foot from the top of the ice of each randomly selected ice basket in each ice condenser bay. The SR is modified by a Note that allows the boron concentration and pH value obtained from averaging the individual samples' analysis results to satisfy the B 3.6.11-8 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued) requirements of the SR. If either the average baron concentration or average pH value is outside their prescribed limit, then entry into Condition A is required.
Sodium tetraborate has been proven effective in maintaining the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine, although the removal of iodine from the containment atmosphere by the sodium tetraborate is not assumed in the'accident analysis.
This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment Spray System fluids in the recirculation mode of operation.
The Frequency of 54 months is intended to be consistent with the.expected length of three fuel cycles, and was developed considering these facts: a. Long term ice storage tests have determined that the chemical composition of the stored ice is extremely stable;b. There are no normal operating mechanisms that decrease the boron concentration of the stored ice, and pH remains within a 9.0-9.5 range when boron concentrations are above approximately 1200 ppm;c. Operating experience has demonstrated that meeting the boron concentration and pH requirements has never been a problem; and d. Someone would have to enter the containment to take the sample, and, if the unit is at power, that person would receive a radiation dose. <---Insert 2 SR 3.6.11.7 This SR ensures that initial ice fill and any subsequent ice additions meet the boron concentration and pH requirements of SR 3.6.11.6.
The SR is modified by a Note that allows the chemical analysis to be performed on either the liquid or resulting ice of each sodium tetraborate solution prepared.
If ice is obtained from offsite sources, then chemical analysis data must be obtained for the ice supplied.Cook Nuclear Plant Unit 2 B361- eiinN.4 B 3.6.11-9 Revision No. 47 Ice Condenser Doors B 3.6.12 BASES ACTIONS (continued)
C.1I If Required Action B.1 or B.2 and associated Completion Time is not met, the doors must be restored to OPERABLE status and closed positions within 48 hours. The 48 hour Completion Time is based on the fact that, with the very large mass of ice involved, it would not be possible for the temperature to increase to the melting point and a significant amount of ice to melt in a 48 hour period. The 48 hour Completion Time is also consistent with the ACTIONS of LCO 3.6.11, "Ice Bed." D.1 and D.2 With any Required Action and associated Completion Time of Condition A or C not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed.Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.6.12.1 REQUIREMENTS Verifying that the inlet doors are in their closed positions makes the nnr~n =\t-rf in ,~rlklrnt nc~rninn n~f nr mnrn C1Cnnr_ ,:#1 2 ftlie ovreolesier-Meniir~-Systei.
SR 3.6.12.2 Verifying, by visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris provides assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed. l=- nsert 2............
....... ..... ..........
... .... i ..... j --
d Cook Nuclear Plant Unit 2B36125evsoN.0 B 3.6.12-5 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.3 Verifying, by visual inspection, that the top deck doors are in place and not obstructed provides assurance that the doors are performing their function of keeping warm air out of the ice condenser during normal operation, and would not be obstructed if called upon to open in response to a DBA. T4FFf~aye-2d-ss~sde-ae~@j~~.v t, onideed o -afie4jia~
de --
fle~pth~ vei9a4#aer-wer-e-#ebFite4.
SR 3.6.12.4 Verifying, by visual inspection, that the ice condenser inlet doors are not impaired by ice, frost, or debris provides assurance that the doors are free to open in the event of a DBA. T~feueqi gexeiae-
* te-r-meet-tiei-aeeep-a e-eri {e-i~a-Beeaus
....hrd~t, v~i~-- ~ ~4e4e-~{4-f
~wrep~gnM r4m ~
SR 3.6.12.5 2 Insert z Verifying the opening torque of the inlet doors provides assurance that no doors have become stuck in the closed position.
The value of 675 in-lb is based on the design opening pressure on the doors of 1 .0 lb/ft 2..r-hi*
--
ne4keywvr ts-fta-ee-dsif~hcl~sni
"=-nsert 2 weter-c- e a e dt~e~ t-g-e-e).---
hierat dej- wIfrhe ~t a4 Cook Nuclear Plant Unit 2B36126RvsoN.0 B 3.6.12-6 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.6 The torque test Surveillance ensures that the inlet doors have not developed excessive friction and that the return springs are producing a door return torque within limits. The torque test consists of the following:
: 1. Verify that the torque, T(OPEN), required to cause opening motion at the 400 open position is < 195 in-Ib;2. Verify that the torque, T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 400 open position is > 78 in-Ib; and 3. Calculate the frictional torque, T(FRICT) = 0.5 {T(OPEN) -T(CLOSE)}, and verify that the T(FRICT) is -< 40 in-lb.T(OPEN) is known as the "door opening torque" and is equal to the nominal door torque plus a frictional torque component.
T(CLOSE) is defined as the "door closing torque" and is equal to the nominal door torque minus a frictional torque component.
The purpose of the friction and return torque Specifications is to ensure that, in the event of a small break LOCA or SLB, all of the 24 door pairs open uniformly.
This assures that, during the initial blowdown phase, the steam and water mixture entering the lower compartment does not pass through part of the ice condenser, depleting the ice there, while bypassing the ice in other bays..T-he-=Feaene-M-I4h-ela -ePi-t-he~~ eea~--4 ajsete~r 4----nsert 2 aeee.Rc-e-er4ter~-Beeau~eejh4"-a~aterH94h-e-t1efi91e Cook Nuclear Plant Unit 2B36127RvsoN.0 B 3.6.12-7 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.7 Verifying the OPERABILITY of the intermediate deck doors provides assurance that the intermediate deck doors are free to open in the event of a DBA. The verification consists of visually inspecting the intermediate doors for structural deterioration, verifying free movement of the vent assemblies, and ascertaining free movement of each door when lifted with the applicable force shown below: Door Liftincj Force a. Adjacent to crane wall b. Paired with door adjacent to crane wall c. Adjacent to containment wall d. Paired with door adjacent to containment wall< 37.4 lb< 33.8 lb.< 31.8 lb-<31.0 lb
~ -Bs~d-n-tn -=
ef rsaeetyi -F-eaae~a--ete--ee-4 ti~ -e~-cefst eb ie 4kie =Insert 2 insead-weogt4' REFERENCES
: 1. UFSAR, Section 14.3.4.Cook Nuclear Plant Unit 2B3.128RvsoN.0 B 3.6.12-8 Revision No. 0 Divider Barrier Integrity B 3.6.13 BASES ACTIONS (continued)
C.1 and C.2 If divider barrier integrity cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.1 3.1 Verification, by visual inspection, that all personnel access doors and equipment hatches between the upper and lower containment compartments are closed provides assurance that divider barrier integrity is maintained prior to the reactor being taken from MODE 5 to MODE 4.This SR is necessary because many of the doors and hatches may have been opened for maintenance during the shutdown.SR 3.6.13.2 Verification, by visual inspection, that the personnel access door and equipment hatch seals, sealing surfaces, and alignments are acceptable provides assurance that divider barrier integrity is maintained.
This inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.13.2 is required for each door or hatch that has been opened, prior to the final closure. Some doors and hatches may not be opened for long periods of time. T$he -4hat-use-r-esiieat-Fat-er-iets-ia-t*Ie-seale-mu~st e5ac  SR 3.6.1 3.3 Verification, by visual inspection, after each opening of a personnel access door or equipment hatch that it has been closed makes the operator aware of the importance of closing it and thereby provides additional assurance that divider barrier integrity is maintained while in applicable MODES.
2 Cook Nuclear Plant Unit 2B36134RvsoN.0 B 3.6.1 3-4 Revision No. 0 Divider Barrier Integrity B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.13.4 Conducting periodic physical property tests on divider barrier seal test coupons provides assurance that the seal material has not degraded in the containment environment, including the effects of irradiation with the reactor at power. The required tests include a tensile strength test and a test for elongation. tswseeeet co .iorn .cuch qy fthesa~aer Ilnsert 2~GdrtcGG8~ibUGfAheSe61e5Rd5b5eflee~f41fl9eiT*fehiity e4he=i~4t conditionc noodo~4e~pei~m the SR. Operating experience
~pei~f~ri~d at tho 24 ~concludod to bo ~SR 3.6.13.5 Visual inspection of the seal around the perimeter provides assurance thatthe seal is properly secured in place, such that the total divider barrier bypass area is less than or equal to the design basis limit of 7'tsueh~i449etiet-s-a 9-ef4 el-a9isee-tief-rt6-f seeIre4h~e
~ee1a94e-tflteeilift neee-efom-he-fR.
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~Insert 2 2-epat REFERENCES
: 1. UFSAR, Section 14.3.4.1.3.1.3.
: 2. UFSAR, Section 14.3.4.1.3.1.1.e Cook Nuclear Plant Unit 2 B361- eiinN.4 B 3.6.13-5 Revision No. 44 Containment Recirculation Drains B 3.6.14 BASES ACTIONS (continued)
C.1I If one CEQ fan room drain is inoperable, 1 hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the containment analysis.
The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1,"Containment," which requires that the containment be restored to OPERABLE status within 1 hour.D.1I If one flow path in the flood-up overflow wall is inoperable, 1 hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operations to within the bounds of the containment analysis..The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires that containment be restored to OPERABLE status within 1 hour.E.1 and E.2 If the affected drain(s) cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.6.14.1 and SR 3.6.14.2 REQUIREMENTS Verifying the OPERABILITY of the required refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.14.2 confirms that the required refueling canal drain blind flanges have been removed and that the required drains are clear of any obstructions that could impair their functioning.
In addition to debris near the drains, attention must be given to any debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains. This verification is performed by SR 3.6.14.1, which requires verification that there is no debris present in the upper containment or refueling canal that could obstruct the required refueling canal drains. SR 3.6.14.1 and SR 3.6.14.2.must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the blind flanges have been removed and that no debris that could impair the drains was deposited during the time the canal was filled.
-Insert 2 aywetvtf dcsd g-fee--s-e.-
ssi i-- -~e-~-ia -h -b v~e~~~- -~ -~~i~sadth-~~
ayMh Cook Nuclear Plant Unit 2 B 3.6.14-4 Revision No. 16 Cook Nuclear Plant Unit 2 B3.6.14-4 Revision No. 16 Containment Recirculation Drains B 3.6.14 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.14.3 Verifying the OPERABILITY of the ice condenser floor drains ensures that they will be able to perform their functions in the event of a DBA.Inspecting the drain valve disk ensures that the valve is performing its function of sealing the drain line from warm air leakage into the ice condenser during normal operation, yet will open if melted ice fills the line following a DBA. Verifying that the drain lines are not obstructed ensures their readiness to drain water from the ice condenser.
T&#xa2;-e4-8--melit-h t4e-be-ac-eept-able
~ feerBeu e .hjhr4.a~e~th-~eatf.
~-as.~~ $eelae.- r-ayJn-i ~ a Insert 2 SR 3.6.14.4 and SR 3.6.14.5 Verifying the operability of the CEQ fan room drains ensures that they will be able to perform their function in the event of a DBA. SR 3.6.14.4 confirms that the required drains are clear of any obstructions.
In addition to debris near the drains, attention must be given to debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains.SR 3.6.14.4 must be performed before entering MODE 4 from MODE 5 and after personnel entry into a CEQ fan room while in MODES 1 through 4. This frequency was developed considering such factors as the location of the drains, and the absence of personnel traffic in the vicinity of the drains. The SR is modified by a Note. The Note indicates that only the CEQ fan room that has been entered need be inspected if the SR is being performed due to personnel entry in MODES 1 through 4. The Note precludes unnecessarily requiring inspection of both CEQ fan rooms if only one has been entered. SR 3.6.14.5 confirms that the CEQ fan room debris interceptors are installed and free of structural distress.SR 3.6.14.5 also confirms that the flow opening at the lower containment sump is not obstructed.
The 24 month frequency was developed considering such factors as the location and the design of the debris interceptors and flow opening.Cook Nuclear Plant Unit 2 B 3.6.14-5 Revision No. 16 Cook Nuclear Plant Unit 2 B 3.6.14-5 Revision No. 16 SGSVs B 3.7.2 BASES SURVEILLANCE REQUIREMENTS (continued)
The Frequency is in accordance with the Inservice Testing Program.This test is conducted in MODE 3 with the unit at operating temperature and pressure.
This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.
SR 3.7.2.2 This SR verifies that each SGSV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage..;hI-q ofes Insert 2 t=h i ! tI -iii !sd-r-e q &#xb6;ene-yl -aeefa-Blelrom-a REFERENCES
: 1. UFSAR, Section 1,0.2.2. UFSAR, Section 14.2.5.3. 10CFR 100.1.1.4. Technical Requirements Manual 5. ASME, Operations and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 2B3724ReionN.3 B 3.7.2-4 Revision No. 33 MFIVs and MFRVs B 3.7.3 BASES ACTIONS (continued)
C. l With both the MFIV and MFRV inoperable in the same flow path, there is no redundant system to operate automatically and perform the required safety function.
Under these conditions, the affected flow path must be isolated within 8 hours. This action returns the system to the condition where at least one valve in each flow path is performing the required safety function.
The 8 hour Completion Time is reasonable, based on operating experience, to complete the actions required to close the MFIV or MFRV, or otherwise isolate the affected flow path.0.1 and 0.2 If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours and in MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.3.1 and SR 3.7.3.2 REQUIREMENTS These SRs verify that the closure time of each MFIV and MFRV is within the limit given in Reference 2 and is within that assumed in the accident and transient analyses.
The valve(s) may also be tested to more restrictive requirements in accordance with the Inservice Testing Program.The Frequency for this SR is in accordance with the Inservice Testing Program.SR 3.7.3.3 This SR verifies that each MFIV and MFRV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage.TheR-efienylfev9erh---i9--evpes.--qp et~ei e~eas<= Insert 2 shw-ft, ~ -ae4ec-f REFERENCES
: 1. UFSAR, Section 10.5.1.2.2. Technical Requirements Manual Cook Nuclear Plant Unit 2B3734ReionN.3 B3.7.3-4 Revision No. 33 SG PORVs B 3.7.4 BASES ACTIONS (continued) 0.1 and C.2 If the SG PORV(s) cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in' which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance upon steam generator for heat removal, within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.0.1 If one or more required SG PORVs are inoperable in MODE 4, the unit is in a degraded condition with reduced safety related means to cool the unit to RHR entry conditions following an event, and the possibility of no means for conducting a cooldown with nonsafety related equipment since the condenser may be unavailable for use with the Steam Dump System.The seriousness of this condition requires that action be started immediately to restore the inoperable SG PORV(s) to OPERABLE status.SURVEILLANCE SR 3.7.4.1 REQUIREMENTS To perform a controlled cooldown of the RCS, the SG PORVs must be able to be opened remotely and throttled through their full range. This SR ensures that the SG PORVs are tested through a full control on~e-pe--4-.iae~hs.
Performance of inservice testing or use of an SG PORV during a unit cooldown may satisfy this requirement. -Gpea Insert 2-S~~lFe-hmef~e-th-4mFhF-fu -Freunc s cctab1e-f~em -Fe iMiy~tapea-.
REFERENCES
: 1. UFSAR, Section 10.2.2.2. UFSAR, Section 14.2.4.Cook Nuclear Plant Unit 2 B3743Rvso o B 3.7.4-3 Revision No. 0 AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued) initiation without declaring the train(s) inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, and these manual operations are an accepted function of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained.
Th4 Y'efee~Jae~negaera ti se nsert 2 wfhtho~eeea-et"t~evrrig~le-p~te a sfs.eer-r~eetw.a4.p e sitieIa .SR 3.7.5.2 Verifying that each required AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 2).Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance.
Performance of inservice testing discussed in the ASME OM Code (Ref. 2) (only required at 3 month intervals) satisfies this requirement.
This SR is modified by a Note indicating that the SR should be deferred for the turbine driven AFW pump until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test at entry into MODE 3. At 850 psig, there is sufficient pressure to perform the test.SR 3.7.5.3 This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.nsert 2t a.S~i~a  Cook Nuclear Plant Unit 2 B3758Rvso o B3.7.5-8 Revision No. 0 AFW System B 3.7.5 BASES SURVEILLANCE REQUI REM ENTS (continued)
The SR is modified by two Notes. Note 1 states that one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not otherwise inoperable.
This exception allows the AFW train(s) to be out of normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s) inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, these manual operations are an accepted condition of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained.
Note 2 states that the SR is only required to be met in MODES 1, 2, and 3. It is not required to be met in MODE 4 since the AFW train is only required for the purposes of removing decay heat when the SG is relied Upon for heat removal. The operation of the AFW train is by manual means and automatic startup of the AFW train is not required.SR 3.7.5.4 This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an ESFAS by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal in MODES 1, 2, and 3. Th ' o.t rqetaey ede19ae 2 This SR is modified by three Notes. Note I indicates that the SR may be deferred for the turbine driven AFW pump until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test at entry into MODE 3. At 850 psig, there is sufficient steam pressure to perform the test. Note 2 states that one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not Otherwise inoperable.
This exception allows the AFW train(s) to be out of normal standby alignment and temporarily incapable of auitomatic initiation without declaring the train(s)inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, these manual operations are an accepted condition of the AFW System. OPERABILITY (i.e., the intended safety function)Cook Nuclear Plant Unit 2 B3759Rvso o B3.7.5-9 Revision No. 0 CST B 3.7.6 BASES ACTIONS (continued) adequate to ensure the backup auxiliary feedwater supply continues to be available.
The 7 day Completion Time is reasonable, based on an OPERABLE backup auxiliary feedwater supply being available, and the low probability of an event occurring during this time period requiring the CST.B.1 and B.2 If any Required Action and associated Completion Time cannot be met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance on the steam generator for heat removal, within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR verifies that the CST contains the required volume of cooling water.
S-t eteae~ratyex' e~ nsert 2 ,ivrt an a~rse ldva -nfeC-- REFERENCES'
: 1. UFSAR, Section 10.5.2.2. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2B3.6-ReionN.2 B3.7.6-3 Revision No. 26 CCW System B 3.7.7 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "ROS Loops -MODE 4," be entered if an inoperable CCW train results in an inoperable RHR loop.This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this condition, the remaining OPERABLE COW train is adequate to perform the heat removal function.The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period.B.1 and B.2 If the COW train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.7.1 REQU IREM ENTS This SR is modified by a Note indicating that the isolation of the COW flow to individual components may render those components inoperable but does not affect the OPERABILITY of the COW System.Verifying the correct alignment for manual, power operated, and automatic valves in the COW flow path provides assurance that the proper flow paths exist for COW operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.:Fhee-l 2)e ~ ~ e Cook Nuclear Plant Unit 2 B3773Rvso o B3.7.7-3 Revision No. 0 CCW System B 3.7.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.7.2 This SR verifies proper automatic operation of the CCW valves on an actual or simulated actuation signal. The COW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation.
This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. -Th e n~el4 2~~~e~d ~ Itosa-feydf8aut oaaea~~ e-ee-~-4r.napan~-rasetf ue~a we m4.w
~
shw~a-4ee m retsuual~s-4eu-elaew a
er=fr-Feq
~ueR ey-is' eiht.SR 3.7.7.3 This SR verifies proper automatic operation of the COW pumps on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal -nsert 2 te euog~ne~eeeti~e-ar-~aadt-asin i hs shw-44tJ s -a~e4-naAFr~fue f ~eF--~ee~aeeept~be4f.e n--reiabyit9tI-pee~into.-
REFERENCES
: 1. UFSAR, Section 9.5.2. UFSAR, Table 9.5-3.Cook Nuclear Plant Unit 2 B3774Rvso o B 3.7.7-4 Revision No. 0 ESW System B 3.7.8 BASES SURVEILLANCE REQUIREMENTS (continued) rather, it involves verification that those valves capable of being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.T-he SR 3.7.8.2 jageRt~-ri 4oi~oa ee".R@-av-e-a a-r-2 This SR verifies proper automatic operation of the ESW valves on an actual or simulated actuation signal. The ESW System is a normally operating system that cannot be fully actuated as part of normal testing.This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.Srn 2 .11 r~:2 Iii :i ri.-,1--..-J.-~-'----.---.-I---..-.............
-r Insert tmeth~e~r+th uP8aa84.airtf ewr-ee-e4ihhereee-{.ra4 44 Insert ~+I rt I ] JSJ IJ E2 gJ J AI&#xa3;J k4JLLJJ J JJJJ T-hef~r~t9e-Freqiiei~ey~is-eepabte-fna-ehb1iat~y-standpein+f.
SR 3.7.8.3 This SR verifies proper automatic operation of the required ESW pumj on an actual or simulated actuation signal.
~rt~~ u p dr ee4 2-e~-Fr~m-ps--Insert 2 REFERENCES
: 1. UFSAR, Section 9.8.3.2. UFSAR, Section 9.8.3.2.3. UFSAR, Section 9.5.2.Cook Nuclear Plant Unit 2 B3784Rvso o B 3.7.8-4 Revision No. 0 UHS B 3.7.9 BASES SURVEILLANCE REQUIREMENTS (continued) determining the UHS temperature is averaging the available operating circulating water pumps discharge 2~-pra~-x-e~
Nver4 REFERENCES
: 1. UFSAR, Section 10.6.2.2. UFSAR, Table 9.8-5.3. Regulatory Guide 1.27, Revision 2, January 1976.4. MD-12-ESW-1 06-N Assessment of Increased Lake Water Temperature on Safety Related and Non-Safety Related Systems.Cook Nuclear Plant Unit 2B3.9-ReionN.5 B3.7.9-3 Revision No. 50 CREV System B 3.7.10 BASES ACTIONS (continued)
An alternative to Required Action E.1 is to immediately suspend activities that could result in a release of radioactivity that might require isolation of the ORE (Required Action E.2). This places the unit in a condition where the LCO does not apply. This does not preclude the movement of fuel to a safe position.F.1 During movement of irradiated fuel assemblies in the containment, auxiliary building, or Unit 1 containment, with two CREV trains inoperable, or with one or more CREV System trains inoperable due to inoperable ORE boundary, action must be taken immediately to suspend activities that could result in a release of radioactivity that requires isolation of the ORE. This places the unit in a condition that minimizes the accident risk.This does not preclude the movement of fuel to a safe position.G.1 If both CREV trains are inoperable in MODE 1, 2, 3, or 4 for reasons other than an inoperable ORE boundary or filter unit (i.e., Conditions B and C), the CREV System may not be capable of performing the intended function and the unit is in a condition outside the accident analyses.Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly.
As the environment and normal operating conditions on this system are not too severe, testing each trai n-cnc cvcry 02 dayz provides an adequate check of this system. Operating the CREV train, with flow through the HEPA filter and charcoal adsorber train, for> 15 minutes demonstrates the function of the CREV train.
Insert 2 SR 3.7.10.2 This SR verifies that the required CREV System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum and maximum flow rate, and the physical properties of the activated charcoal.
Specific test Frequencies and additional information are discussed in detail in the VFTP.Cook Nuclear Plant Unit 2 B 3.7.10-6 Revision No. 23 Cook Nuclear Plant Unit 2 B 3.7.10-6 Revision No. 23 CREV System B 3.7.10 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.10.3 This SR verifies that each CREV train starts and operates on an actual or simulated actuation signal. The only actuation signal necessary to be verified is the Safety Injection (SI) signal, since the Control Room Radiation
-High signal is not assumed in the accident analysis.
A Note has been included that states the Surveillance is only required to be met in MODES 1, 2, 3, and 4, since these are the MODES the SI signal is assumed to start the CREV trains. The CREV trains are assumed to be manually started during a fuel handling accident.
Insert 2 i~~e~e-th-2ma-Fel -fret elec4--
ie=-- ~s-u{e 4er+ha4 --~ bJtyt SR 3.7.10.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 rem TEDE, the CRE occupants are protected from smoke, and analyses demonstrate that the CREV System is not needed to prevent exceeding hazardous chemical limits. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences.
When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered. Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within, the licensing basis habitability limits for the occupants following an accident.Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 4) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 5). These compensatory measures may also be used as mitigating actions as required by Required Action B3.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref. 6). Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status. There are no SRs to verify CREV System operability for hazardous chemicals or smoke.Cook Nuclear Plant Unit 2 B 3.7.10-7 Revision No. 23 Cook Nuclear Plant Unit 2 B 3.7.10-7 Revision No. 23 CRAC System B 3.7.11 BASES ACTIONS (continued) 0.1 and 0.2 During movement of irradiated, fuel, if the inoperable CRAC train cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CRAC train must be placed in operation immediately.
This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that active failures will be readily detected.An alternative to Required Action 0.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room (Required Action 0.2). This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.D.__1 During movement of irradiated fuel assemblies, with two CRAC trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.E.1 If both CRAC trains are inoperable in MODE 1, 2, 3, or 4, the CRAC System may not be capable of performing its intended function.Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.7.11.1 and SR 3.7.11.2 REQUIREMENTS These SRs verify that the heat removal capability of each CRAC train is sufficient to maintain control room air temperature
< 850&deg;F. -T-he=-2-heuf
-Insert 2 thedC- -Ae~te of tewa a-imt-exhee weAer4
* l~e, e 1-ee Fr-eeu fv-etbeGR  REFERENCES
: 1. UFSAR, Section 9.10.Cook Nuclear Plant Unit 2B37113RvsoN.0 B 3.7.11-3 Revision No. 0 ESF Ventilation System B 3.7.12 BASES ACTIONS A.1 With one ESF Ventilation train inoperable, action must be taken to restore OPERABLE status within 7 days. During this time, the remaining* OPERABLE train is adequate to perform the ESF Ventilation System function.The 7 day Completion Time is appropriate because the risk contribution is less than that for the ECCS (72 hour Completion Time), and this system is not a direct support system for the ECCS. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.
B.1 If the ESF enclosure boundary is inoperable, the [SF Ventilation trains cannot perform their intended functions.
Actions must be taken to restore an OPERABLE ESF enclosure boundary within 24 hours. During the period that the ESF enclosure boundary is inoperable, appropriate compensatory measures consistent with the intent, as applicable, of GDC 19, 60, 64 and 10 CFR Part 100 should be utilized to protect plant personnel from potential hazards. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition.
The 24 hour Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures.
The 24 hour Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the [SF enclosure boundary.C.1 and C.2 If the ESF Ventilation train or [SF enclosure boundary cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. wrivreimeiat-.an4-er~eaI--eper-at]
~ceei~t4eB&&#xf7;nsr pe~ga Cook Nuclear Plant Unit 2B3.123RvsoN.0 B 3.7.12-3 Revision No. 0 ESF Ventilation System B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued)SR 3.7.12.2 This SR verifies that the required ESF Ventilation System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorbers efficiency, minimum and maximum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
Specific test Frequencies and additional information are discussed in detail in the VFTP.SR 3.7.12.3 This SR verifies that each ESF Ventilation train starts and operates on an actual or simulated actuation signal. One [SF Ventilation train is normally operating with flow bypassing the charcoal adsorber section.This test confirms that each train, when in standby, starts upon receipt of a Containment Pressure -High High signal and that the exhaust flow can be directed through the entire filter unit including the HEPA filter and charcoal adsorber section. Q#ler-at~g Iner 2 oempeae to rvil~ele-erfer-eel~t-t4I-e 2-.mPhF,~qu- .e:e4r SR 3.7.12.4 This SR verifies the integrity of the ESF enclosure.
The ability of the ESF enclosure to maintain a negative pressure, with respect to potentially uncontaminated adjacent areas, is periodically tested to verify proper functioning of the [SF Ventilation System. During the post accident mode of operation, the [SF Ventilation System is designed to maintain a slight negative pressure in the [SF enclosure, with respect to adjacent areas, at a flowrate < 22,500 cfm to prevent unfiltered leakage. *-Insert 2 Frelu Cook Nuclear Plant Unit 2B3.124RvsoN.0 B 3.7.12-4 Revision No. 0 FHAEV System B 3.7.13 BASES APPLICABILITY During movement of irradiated fuel in the auxiliary building, the FHAEV System is required to be OPERABLE to alleviate the consequences of a fuel handling accident.In MODE 1, 2, 3, 4, 5, or 6, the FHAEV System is not required to be OPERABLE since the FHAEV System is only credited during a fuel handling accident in the auxiliary building.ACTIONS LCO 3.0.3 is not applicable while in MODE 5 or 6. However, since irradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operations.
Entering LCO 3.0.3, while in MODE 1, 2, 3, or 4 would require the unit to be shutdown unnecessarily.
A.11 When the required FHAEV train is inoperable or not in operation during movement of irradiated fuel assemblies in the auxiliary building, action must be taken to place the unit in a condition in which the LCO does not apply. Action must be taken immediately to suspend movement of irradiated fuel assemblies in the auxiliary building.
This does not preclude the movement of fuel to a safe position.SURVEILLANCE REQUIREMENTS SR 3.7.13.1 This SR requires verification every -l2hu's that the required FHAEV train is operating with flow through the filter unit, including the HEPA filter__, and charcoal adsorber section. Verification includes fan status and also verifies that each charcoal bypass damper is closed. -ThoFrq uccy Insert 2 4&#xa3;her4' fi~-es te- E}  SR 3.7.13.2 Standby systems should be checked periodically to ensure that they function properly.
As the environmental and normal operating conditions on this system are not severe, testing each train once "er"-9 2-y.provides an adequate check on this system.Operating the required FHAEV train, with flow through the HEPA filter and charcoal adsorber train, for > 15 minutes demonstrates the function of the system. T-he--2-Efay-FFequeF1~-~seeth4mV~t Insert 2 Cook Nuclear Plant Unit 2B37133RvsoN.0 B 3.7.13-3 Revision No. 0 FHAEV System B 3.7.13 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.13.3 This SR verifies that the required FHAEV System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA-filter performance, charcoal adsorber efficiency, minimum and maximum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
Specific test frequencies and additional information are discussed in detail in the VFTP.SR 3.7.13.4 This SR verifies that the required FHAEV train actuates on an actual or simulated actuation signal. The test must verify that the signal automatically shuts down each of the Fuel Handling Area Supply Air System fans. @fe 2 us.at~s4h-S~i~reNhaeeeta-~-4-~
~ ~ y T- f --eqecyi-acef~t-~
SR 3.7.13.5 This SR verifies the integrity of the auxiliary building enclosure.
The ability of the pool storage area to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the FHAEV train. During the accident mode of operation, the FHAEV train is designed to maintain a slight negative pressure in the FHAEV train, to prevent unfiltered leakage. The FHAEV train is designed to maintain a pressure > 0.125 inches of vacuum water gauge with respect to atmospheric pressure at a flow rate of<27,000 cfm. *'"-Insert 2 REFERENCES
: 1. UFSAR, Section 9.9.3.2.2. UFSAR, Section 14.2.1.3. 10 CFR 100.Cook Nuclear Plant Unit 2 B371- eiinN.2 B 3.7.13-4 Revision No. 26 Fuel Storage Pool Water Level B 3.7.14 BASES ACTIONS A.1j When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring.
When the fuel storage pool water level is lower than the required level, the movement of irradiated fuel assemblies in the fuel storage pool is immediately suspended to a safe position.
This action effectively precludes the occurrence of a fuel handling accident.
This does not preclude movement of a fuel assembly to a safe position.Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations.
Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE REQUIREMENTS SR 3.7.14.1 This SR verifies sufficient fuel storage pool water is available in the event of a fuel handling accident.
The water level in the fuel storage pool must be checked periodically.
T~-7 Insertc-ys-sr c~-t-le-ypate-ed~~-ada--e-p~beae-rp
{a ex-per-iefc-e.
REFERENCES
: 1. UFSAR, Section 9.7.2.2. UFSAR, Section 9.4.3. UFSAR, Section 14.2.1.4. 10OCFR100.11.
Cook Nuclear Plant Unit 2 B371- eiinN.2 B 3.7.14-2 Revision No. 26 Fuel Storage Pool Boron Concentration B 3.7.15 BASES SURVEILLANCE REQUIREMENTS (continued) accidents are fully addressed. -T-lhe-7--eey-r-eqluerey-4e-appr-epfr-iae Insert 2 be
~t- -pe4a~ 4t~4tshQ-p eriG eJ-of-fme.
REFERENCES
: 1. Double contingency principle of ANSI Ni16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1 .2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).2. UFSAR, Section 9.7.2.Cook Nuclear Plant Unit 2 B371- eiinN.2 B 3.7.15-3 Revision No. 24
*. Secondary Specific Activity B 3.7.17 BASES ACTIONS A.1 and A.2 Specific activity of the secondary coolant exceeding the allowable value is an indication of a problem in the RCS and contributes to increased post accident doses. If the secondary specific activity is not within limits, the unit must be placed in.a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The all owed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.17.1 REQUIREMENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis.
A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases.
It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE..TC1o-day-F-queny-isb&~H =etitre-f.-f -rbe-al-tuv-v -U e L-X&1imiit.
REFERENCES
: 1. 10CFR100.11.
: 2. 10 CFR 50, Appendix A, G DC 19.3. UFSAR, Section 14.2.7.Cook Nuclear Plant Unit 2B37.-3RvsoN.0 B 3.7.17-3 Revision No. 0 AC Sources -Operating B3 3.8.1 BASES ACTIONS (continued) continued operation.
The unit is required by LCO 3.0.3 to commence a controlled shutdown.SURVEILLANCE REQUIREMENTS The AC sources are designed to permit inspection and testing of all important areas and features, especially those that have a standby function, in accordance with Plant Specific Design Criterion (PSDC) 39 (Ref. 8). Periodic component tests are supplemented by extensive functional tests during refueling outages (under simulated accident conditions).
The SRs for demonstrating the OPERABILITY of the DGs are in accordance with the recommendations of Regulatory Guide 1 .9 (Ref. 3), Regulatory Guide 1.108 (Ref. 9), Regulatory Guide 1.137 (Ref. 10), and IEEE Standard 387-1995 (Ref. 11) as addressed in the applicable SR discussion.
Where the SRs discussed herein specify voltage and frequency tolerances, the following is applicable.
The minimum steady state output voltage of 3910 V is 94% of the nominal 4160 V output voltage. This value allows for voltage drop to the terminals of 4160 V motors whose minimum operating voltage is specified as 90% or 3740 V. It also allows for voltage drops to motors and other equipment down through the 120 V level Where the minimum operating voltage is also usually specified as 90% of nameplate rating. The specified maximum steady state output voltage of 4400 V is equal to the maximum operating voltage specified for 4000 V motors. It ensures that for a lightly loaded distribution system, the voltage at the terminals of 4000 V motors is no more than the maximum rated operating voltages.
The specified minimum and maximum steady state frequencies of the DG are 59.4 Hz and 60.5 Hz, respectively.
These values ensure the ESF pumps can achieve adequate fluid flow to meet their safety and accident mitigation functions.
SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that the required qualified offsite circuits are OPERABLE, and that appropriate independence of offsite circuits is maintained.
T1te-7 s
2 ne~kt-~h~ewtet4ee~rae~i4jaaef~ stats-i-is~ayede-eon-tfeeo t-r~eom.SR 3.8.1.2 and SR 3.8.1.8 These SRs help to ensure the availability of the standby electrical power Cook Nuclear Plant Unit 2 t3 3.8.1-16 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.
To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 1 for SR 3.8.1.2 and Note for SR 3.8.1.8) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and followed by a warmup period prior to loading.For the purposes of SR 3.8.1.2 and SR 3.8.1.8 testing, the DGs are started from standby conditions.
Standby conditions for a DG means that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations.
In order to reduce stress and wear on diesel engines, the manufacturer recommends a modified start in which the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 2.SR 3.8.1.8 requires that, at a 184 day Frequency, the DG starts from standby conditions and achieves required voltage and frequency within 10 seconds. The 10 second start requirement supports the assumptions of the design basis LOCA analysis in the UFSAR, Section 14.3 (Ref. 5).The 10 second start requirement is not applicable to SR 3.8.1.2 (see Note 2 of SR 3.8.1.2) when a modified start procedure as described above is used. If a modified start is not used, the 10 second start requirement of SR 3.8.1.8 applies.Since SR 3.8.1.8 requires a 10 second start, it is more restrictive than SR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2.In addition, the DG is required to maintain proper voltage and frequency limits after steady state is achieved.
The voltage and frequency limits are normally achieved within 10 seconds. The time for the OG to reach steady state operation, unless the modified OG start method is employed, is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance.risert 2
.-- ---te yfrS sar teir 4n-e#ts~ -ers~etwtaGe W Frqte~eel MT Cook Nuclear Plant Unit 2 B3811 eiinN.4 B 3.8.1-17 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.3 Consistent with Regulatory Guide 1.9 (Ref. 3), this Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads 90% to 100% of the continuous rating of the 0G. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.The 0.8 value is the design rating of the machine, while the 1.0 is an operational goal to ensure circulating currents are minimized.
The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections being required in order to maintain DG reliability.
They--refueey-fer-4-91J-at~viIaee-is--esist i fflrR-egtjfato 2
This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized.
Note 2 states that momentary transients, because of changing bus loads, do not invalidate this test. Note 3 indicates that this Surveillance should be conducted on only one Unit 2 DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.
Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
SR 3.8.1.4 This SR provides verification that the level of fuel oil in the day tank is above the level at which fuel oil is automatically added. The level is expressed as an equivalent volume in gallons, of which 31.4 gallons is unusable (due to tank geometry and vortexing considerations) and 70 gallons is usable, and is selected to ensure adequate fuel oil for greater than 15 minutes of DG operation at full load.T-he-fulet-oiHs~il-~nee~ew -lever -aer-s-ar-p Fevided-ai9fae-fy ete~er-tef-weul-beeee-awef-af~ya-r-fe-uesefse-ed-tfl19e-eil i&-psfie.nsert 2 SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must Cook Nuclear Plant Unit 2 B3811 eiinN.4 B 3.8.1-18 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) have a water environment in order to survive. Removal of water from each fuel oil day tank eliminates .the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria.
Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. T-h~~~e-iP1le~ee7 Insert 2 SR 3.8.1.6 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available.
While the system design requirements provide for two engine start cycles from each of the two air start receivers associated with each DG without recharging, only one start sequence is required to meet the OPERABILITY requirements (since the accident analysis assumes the OG starts on the first attempt).The pressure specified in this SR reflects the lowest value at which one DG start can be accomplished with one air start receiver.T-lormc to al r-h-~ tth -ap .raiitoryt, SR 3.8.1.7 This Surveillance demonstrates that each required fuel oil transfer pump (one per fuel oil transfer system) operates automatically and transfers fuel oil from its associated storage tank to its associated day tank. This is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.T~eF~qer --
e -.G@&-re~eds-t4e-th-est4a-re4if=met6-fer-3un~sc--i~a~ie-it-a fln-- 4.-Insert 22 Cook Nuclear Plant Unit 2 B3811 eiinN.4 B 3.8.1-19 Revision No. 41 AC Sources -Operating B 3.8.1 I BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.9 Automatic transfer of each 4.16 kV emergency bus power supply from the normal auxiliary circuit to the preferred offsite circuit and the manual alignment to the alternate required offsite circuit demonstrates the OPERABILITY of the required offsite circuit to power the shutdown loads.~ Insert 2 As noted (Note 1Ito SR 3.8.1.9), SR 3.8.1.9.a is only required to be met when the auxiliary source is supplying the onsite electrical power subsystem.
This is acceptable since the preferred offsite source would be supplying the onsite electrical power subsystem and a transfer would not be necessary.
SR 3.8.1.10 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding a predetermined frequency and while maintaining a specified margin to the overspeed trip. Voltage and frequency are also verified to reach steady state conditions within 2 seconds. For this unit, the single load for each DG is 600 kW. This Surveillance may be accomplished by: a. Tripping the DG output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while paralleled to offsite power, or while solely supplying the bus; or b. Tripping its associated single largest post-accident load with the DG solely supplying the bus.Consistent with Regulatory Guide 1.9 (Ref. 3), the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the overspeed trip setpoint, or 15% above nominal speed, whichever is lower. This corresponds to 64.4 Hz, which is the nominal speed pius 75% of the difference between nominal speed and the overspeed trip setpoint.The time. voltagqe, and frequency tolerances specified in this SR are CookNuclear Plant Unit 2 B 3.8.1-20 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) derived from Regulatory Guide 1.9 (Ref. 3) recommendations for response during load sequence intervals.
The 2 seconds specified is equal to approximately 60% of the 3.49 second load sequence interval associated with sequencing of the largest load. The voltage and frequency specified are consistent with the design range of the equipment powered by the DG. SR 3.8.1.10.a corresponds to the maximum frequency excursion, while SR 3.8.1.10O.b and SR 3.8.1.'10O.c are steady state voltage and frequency values to which the system must recover following load rejection. uen qency-is -ei -i eaiera j.d 4 _4i-e ~-o e-&gex9eiacehs4 w 2 F-reR heneyfT-he 1~ 4 -et4ie-F=F ueai4l le4t-e4be-a--eleptel~e-This SR is modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR. Credit may be taken for unplanned events that satisfy this SR.Note 2 ensures that the DG is tested under load conditions that are as close to design basis conditions as possible.
When synchronized with offsite power, testing should be performed at a power factor of-< 0.86.This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions.
Under certain conditions, however, Note 2 allows the Surveillance to be conducted at a power factor other than _< 0.86. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to < 0.86 results in voltages on the emergency busses that are too high.Cook Nuclear Plant Unit 2 B 3.8.1-21 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Under these conditions, the power factor should be maintained as close as practicable to 0.86 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the DG excitation levels needed to obtain a power factor of 0.86 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained as close as practicable to 0.86 without exceeding the DG excitation limits.SR 3.8.1.11 Consistent with Regulatory Guide 1 .9 (Ref. 3), paragraph C.2.2.8, this Surveillance demonstrates the DG capability to reject a full load (90% to 100% of the DG continuous rating) without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping.
This Surveillance ensures proper engine generator load response under the simulated test conditions.
This test simulates the loss of the total connected load that the DG experiences following a full load rejection and verifies that the DG does not trip upon loss of the load. These acceptance criteria provide for DG damage protection.
While the DG is not expected to experience this transient during an event and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated.Insert 2 This SR has been modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could cause perturbation to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the.potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Cook Nuclear Plant Unit 2 B 3.8.1-22 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) loaded without undue hardship or potential for undesired operation.
For instance, Emergency Core Cooling Systems (ECCS) injection valves are not desired to be stroked open, or centrifugal charging trains are not capable of being operated at full flow, or residual heat removal (RHR)trains performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation.
In lieu of actual demonstration of connection and loading of loads, testing that adequately shows t[he capability of the DG systems to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.,T-he' r-eY i M e. ey-H.l M I-MI LL h AL -I M --. a e.-e I-IeMgIeerM-II ULMIL~. IMIiMg-jM gereLI ,--ki " l"-nsert 2This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the D.Gs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4, is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing
*OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.13 Consistent with Regulatory Guide 1 .9 (Ref. 3), paragraph C.2.2.5, this Surveillance demonstrates that the DG automaticallv starts and achieves Cook Nuclear Plant Unit 2 B 3.8.1-24 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) the required voltage and frequency within the specified time (10 seconds)from the design basis actuation signal (ESF actuation signal). In addition, the DG is required to maintain proper voltage and frequency limits after steady state is achieved.
The voltage and frequency limits are normally achieved within 10 seconds. The time for the DG to reach the steady state voltage and frequency limits is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance.
The DG is required to operate for > 5 minutes. The 5 minute period provides sufficient time to demonstrate stability.
SR 3.8.1.13.d and SR 3.8.1.13.e ensure that permanently connected loads and emergency loads are energized from the offsite electrical power system on an ESF signal without loss of offsite power.The requirement to verify the connection of permanent and auto-connected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation.
For instance, ECCS injection valves are not desired to be stroked open, or centrifugal charging trains are not capable of being operated at full flow, or RHR trains performing a decay heat removal function are not desired to be realigned to the ECOS mode of operation.
In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.=l'eFrqee'f2"etssasde-niern-~eet i Insert 2.h eeee~f-fi--ali~
4 ~ i 44ebe- se-,sewe-that-the -uee e11y-aes-4 SRw al re- e-tte2- ~4--eeUecy.--
d eei This SR is modified by two Notes. The reason for Note I is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing Cook Nuclear Plant Unit 2 B3812 eiinN.4 B 3.8.1-25 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.14 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph 0.2.2.12, this Surveillance demonstrates that DG noncritical protective functions (e.g., low lube oil pressure) are bypassed on a loss of voltage signal or an ESE actuation test signal. The noncritical trips are bypassed during DBAs and provide an alarm on an abnormal engine condition.
This alarm provides the operator with sufficient time to react appropriately.
The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG.
4t Insert 2 Onp er-t~ge-xn
~c-e -ha-siw-h ewt-that-eserptet--euereu~elltylass-t49e The SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DG from service. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when the~y are tied together or operated independently for the Surveillance:
Cook Nuclear Plant Unit 2 B 3.8.1-26 Revision Nol 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.15 This Surveillance demonstrates the DGs can start and run continuously at full load capability (90% to 100% of the DG continuous rating) for an interval of not less than 8 hours. The run duration of 8 hours is consistent with IEEE Standard 387-1995 (Ref. 11). The DG starts for this Surveillance can be performed either from standby or hot conditions.
The provisions for prelubricating and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.The load band is provided to avoid routine overloading of the 0G.Routine overloading may result in more frequent teardown inspections being required in order to maintain DG reliability.
~ ~ rgee4gj~~,ta~gt Insert 2 This Surveillance is modified by three Notes. Note 1 statesthat momentary transients due to changing bus loads do not invalidate this test.Similarly, momentary power factor transients above the power factor limit will not invalidate the test. The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems.This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with Cook Nuclear Plant Unit 2 B 3.8.1-27 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR. Note 3 ensures that the DG is tested under load conditions that are as close to design basis conditions as possible.
When synchronized with offsite power, testing should be performed at a power factor of < 0.86. This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions.
Under certain conditions, however, Note 3 allows the Surveillance to be conducted as a power factor other than < 0.86. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to - 0.86 results in voltages on the emergency busses that are too high.Under these conditions, the power factor should be maintained as close as practicable to 0.86 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the DG excitation levels needed to obtain a power factor of 0.86 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained close as practicable to 0.86 without exceeding the DG excitation limits.SR 3.8.1.16 This Surveillance demonstrates that the diesel engine can restart from a hot condition, such as subsequent to shutdown from normal Surveillances, and achieve the required voltage and frequency within 10 seconds. The 10 second time is derived from the requirements of the.accident analysis to respond to a design basis large break LOCA. --T-he- lsr This SR is modified by two Notes. Note 1 ensures that the test is performed with the diesel sufficiently hot. The load band is provided to avoid routine overloading of the DG. Routine overloads may result in more frequent teardown inspections being required in order to maintain DG reliability.
The requirement that the diesel has operated for at least 2 hours at full load conditions prior to performance of this Surveillance is based on operating experience for achieving hot conditions.
Momentary transients due to changing bus loads do not invalidate this test. Note 2 allows all DG starts to be preceded by an engine prelube period to minimize wear and tear on the diesel during testing.Cook Nuclear Plant Unit 2 B 3.8.1-28 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.17 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.1 1, this Surveillance ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and the DG can be returned to ready-to-load status when offsite power is restored.
It also ensures that the auto-start logic is reset to allow the DG to reload if a subsequent loss of offsite power occurs. The DG is considered to be in ready-to-load status when the OG is running at rated speed and voltage, with the DG output breaker open.* ~ g Insert 2 This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1, 2, 3, or 4.Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.18 Under accident conditions loads are sequentially connected to the bus by the individual time delay relays. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs or RATs (as applicable) due to high motor starting currents.Verifying the load sequencer time within plus or minus 5% of its required value ensures that sufficient time exists for the DG to restore frequency and voltage and RATs to restore voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated.
Reference 4 provides a summary of the Cook Nuclear Plant Unit 2 B 3.8.1-29 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) automatic loading of emergency buses.
ta~g n se rt 2eyey-d -c4 el4te-ee-aec-epttae-fr--el-a-r~el1-abiPity&#xb6;ai1
.This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the~avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.19 In the event of a DBA coincident With a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, ROS, and containment design limits are not exceeded.This Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.12, during a loss of offsite power actuation test signal in conjunction with an ESF actuation signal. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.-Th a 4gree4jj~m~-tkfg nsert 2 Cook Nuclear Plant Unit 2 B 3.8.1-30 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR is modified by two Notes. The reason for Note I is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for DGs. The reason for Note 2 is that the performance of the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This a~ssessmnent shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.20 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing that involves connecting the DG to its test load resistor bank, and the DG will automatically reset to ready to load operation if a ESE actuation signal is received during operation in the test mode. Ready to load operation is defined as the DG running at rated speed and voltage with the DG output breaker open.The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.13.
The intent in the requirement associated with SR 3.8.1.20.b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that Cook Nuclear Plant Unit 2 B 3.8.1-31 Revision No. 41 AC Sources -Operating 8 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) adequately shows the capability of the emergency loads to perform these fun ctions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.Insert 2 This SR is modified by two Notes. Note I states that this Surveillance is only required to be met when the applicable DG is connected to its test load resistor bank. This is allowed since the test mode override only functions when the DG is connected to its associated test load resistor bank. When the DG is not connected to its associated test load resistor bank, the feature is not necessary; thus the Surveillance is not required to be met under this condition.
The reason for Note 2 is that performing the Surveillance would remove a required DG from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope'with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.21 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing and the DG will automatically reset to ready to load operation if a LOCA actuation signal is received during operation in the test mode. Ready to load operation is defined as the DG running at rated speed and voltage Cook Nuclear Plant Unit 2 B 3.8.1-32 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUI REMENTS (continued) with the DG output breaker open.The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.13.
The intent in the requirement associated with SR 3.8.1.21 .b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.-Th thae Saeg~e~a-urva e-i -in tecc.~ w~a4
=-=---nsert 2 irr,-,-r,-, n,ini, ',.-nrdu .4 cr1*r -, ,-r'nn+-, li o -,-nI k]i- c'4.A ,nin This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DG from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determinethat unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.22 This Surveillance demonstrates that the DG starting independence has*not been compromised.
Also, this Surveillance demonstrates that each engine can achieve proper speed within the specified time when the DGs are started simultaneously.
Cook Nuclear Plant Unit 2 B 3.8.1-33 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
T-ho10-ycarF-r,,g
...ncyi-c-ea~et~h~e-r~ees-at eg&e Insert 2 This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
SR 3.8.1.23 With the exception of this Surveillance, all other Surveillances of this Specification (SR 3.8.1.1 through SR 3.8.1.22) are applied to Unit 2 sources. This Surveillance is provided to direct that appropriate Surveillances for the required Unit I AC sources are governed by the applicable Unit 1 Technical Specifications.
Performance of the applicable Unit 1 Surveillances will satisfy the Unit I requirements as well as satisfy this Unit 2 Surveillance Requirement.
Exceptions are noted to the Unit 1 SRs of LCO 3.8.1. SR 3.8.1 .9.b is not required to be met since only one offsite circuit is required to be OPERABLE.
SR 3.8.1.13, SR 3.8.1.14 (ESF actuation signal portion only), SR 3.8.1.19, SR 3.8.1.20, and SR 3.8.1.21 are not required to be met because the ESF actuation signal is not required to be OPERABLE.
SR 3.8.1.22 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE.The Frequency required by the applicable Unit 1 SR also governs performance of that SR for Unit 2.As noted (Note I to SR 3.8.1.23), if Unit 1 is in MODE 5 or 6, or moving irradiated fuel assemblies, SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, SR 3.8.1.14 through SR 3.8.1.17, and SR 3.8.1.18 are not required to be performed.
This ensures that this Unit 2 SR will not require a Unit 1 SR to be performed, when the Unit I Technical Specifications exempts performance of a Unit 1 SR (however, as stated in the Unit I SR 3.8.2.1 Note 1, while performance of an SR exempted, the SR must still be met).As noted (Note 2 to SR 3.8.1.23), SR 3.8.1.9.a is only required to be met when the auxiliary source is supplying the Unit I electrical power distribution subsystem since the preferred offsite source is required to support Unit 2 operations.
REFERENCES
: 1. Atomic Energy Commission Proposed General Design Criterion 39, July 1967.2. UFSAR, Section 8.3.Cook Nuclear Plant Unit 2 B 3.8.1-34 Revision No. 4 Diesel Fuel Oil B 3.8.3 BASES SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load.The 7 day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.f .=-Insert 2 SR 3.8.3.2 The tests listed below are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion.
If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage tank(s). The tests, limits, and applicable ASTM Standards are as follows: a. Sample the new fuel oil in accordance with ASTM 04057-8 1 (Ref. 5);b. Verify that the sample has: (1) when tested in accordance with ASTM D1298-80 (Ref. 5) an absolute specific gravity at 60/60&deg;F of> 0.82 and -< 0.88, an API gravity at 60&deg;F of >- 300 and _< 400, an API gravity of within 0.3 degrees at 60&deg;F when compared to the supplier's certificate, or a specific gravity of within 0.0016 at 60/600 when compared to the supplier's certificate; (2) a kinematic viscosity at 40&deg;C of >- 1.9 centistokes and <- 4.1 centistokes or Saybolt viscosity at 1 00&deg;F of>_ 32.6 and < 40.1, if gravity was not determined by comparison with supplier's certification, when tested in accordance with ASTM 975-81 (Ref. 5); and (3) a flash point of_ 125&deg;F when tested in accordance with ASTM 0975-81 (Ref. 5); and c. Verify that the new fuel oil has a clear and bright appearance with proper color when tested in accordance with ASTM 04176-82 (Ref. 5).Failure to meet any of the above limits is cause for rejecting the new fuel oil, but does not represent a failure to meet the LCO concern since the fuel oil is not added to the storage tanks.Following the initial new fuel oil sample, the fuel oil is analyzed within 31 days following addition of the new fuel oil to the fuel oil storage tank(s)to establish that the other properties specified in Table I of Cook Nuclear Plant Unit 2 B 3.8.3-4 Revision No. 0 Diesel Fuel Oil B 3.8.3 BASES SURVEILLANCE REQUIREMENTS (continued)
ASTM D975-81 (Ref. 6) are met for new fuel oil when tested in accordance with ASTM 0975-81 (Ref. 5), except that the analysis for sulfur may be performed in accordance with ASTM D2622-82 (Ref. 5).The 31 day period is acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an immediate effect on OG operation.
This Surveillance ensures the -availability of high quality fuel oil for the DGs.Fuel oil degradation during long term storage shows up as an increase in particulate, due mostly to oxidation.
The presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause fouling of filters and fuel oil injection equipment, however, which can cause engine failure.Particulate concentrations should be determined in accordance with ASTM 02276-83, Method A (Ref. 5). This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/I. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing.The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.
SR 3.8.3.3 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel storage tanks eiaee-e er-y.-4=-a.ys-eliminates the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, and contaminated fuel oil, and from breakdown of the fuel oil by bacteria.
Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequencies are established by Regulatory Guide 1.137 (Ref. 2). This SR is for preventive maintenance.
The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during performance of the Surveillance.
Cook Nuclear Plant Unit 2 B 3.8.3-5 Revision No. 0 Cook Nuclear Plant Unit 2 B3.8.3-5 Revision No. 0 DC Sources -Operating B 3.8.4 BASES ACTIONS (continued)
E. 1 If one or both required Unit 1 Train A and Train B DC electrical power subsystems are inoperable, the associated ESW train(s) are not capable of performing their intended function.
Immediately declaring the affected supported feature, e.g., ESW train, inoperable allows the ACTIONS of LCO 3.7.8 to apply appropriate limitations on continued reactor operation.
SURVEILLANCE REQUIREMENTS SR 3.8.4.1 Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the battery chargers, which support the ability of the batteries to perform their intended function.
Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state while supplying the continuous steady state loads of the associated DC subsystem.
On float charge, battery cells will receive adequate current to optimally charge the battery. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the minimum float voltage established by the battery manufacturer (2.20 Vpc or 255.2 VDC at the battery terminals of the Train A and Train B batteries and 2.20 Vpc or 257.4 VDC for the Train N battery).
This voltage maintains the battery plates in a condition that supports maintaining the grid life (expected to be approximately 20 years). The Ne7d~---eu~-~--~e-tM--e~re-~t 1 -'nsert 2........ ... \ ...... l" SR 3.8.4.2 This SR verifies the design capacity of the battery chargers.
According to Regulatory Guide 1 .32 (Ref. 9), the battery charger supply is recommended to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences.
The minimum required amperes and duration ensure that these requirements can be satisfied.
This SR requires that each Train A and Train B required battery charger be capable of supplying
> 300 amps at > 250 VDC for > 4 hours and the Train N battery charger is capable of supplying
> 25 amps at > 250 VDC for > 4 hours. The ampere requirements are based on the output rating of the chargers.
The voltage requirements are based on the charger voltage Cook Nuclear Plant Unit 2 B 3.8.4-7 Revision No. 0 Cook Nuclear Plant Unit 2 B3.8.4-7 Revision No. 0 DC Sources -Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued) level after a response to a loss of AC power. The time period is sufficient to detect significant charger failures.*
n 2 SR 3.8.4.3 A battery service test is a special test of the battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The battery charger must be disconnected throughout the performance of the battery service test. The discharge rate and test length should correspond to the design duty cycle requirements as specified in the applicable design documents.
isrei+-u eHt-eea c-2 .---T-her-efer-e,-t-he-Fr-e This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu of a service test.The reason for Note 2 is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 2 B3848Rvso o B3.8.4-8 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS SR 3.8.6.1 Verifying battery float current while on float charge is used to determine the state of charge of the battery. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a charged state.The float current requirements are based on the float current indicative of a charged battery. Use of float current to determine the state of charge of the battery is consistent with IEEE-450 (Ref. 1)..'TheW-eay-Fr-eqteree?-s.nr osrA''
* hah-eemra~eseq This SR is modified by a Note that states the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.8.4.1. When this float voltage is not maintained the Required Actions of LCO 3.8.4 ACTION A are being taken, which provide the necessary and appropriate verifications of the battery condition.
Furthermore, the float current limit of 2 amps is established based on the nominal float voltage value and is not directly applicable when this voltage is not maintained.
SR 3.8.6.2 and SR 3.8.6.5 I=-nsert 2 Optimal long term battery performance is obtained by maintaining a float voltage greater than or equal to the minimum established design limits provided by the battery manufacturer, which corresponds to 257.5 VDC for a 116 cell battery and 259.7 VDC for a 117 cell battery at the battery terminals, or 2.22 Vpc. This provides adequate over-potential, which limits the formation of lead sulfate and self discharge, which could eventually render the battery inoperable.
Float voltages in this range or less, but greater than 2.07 Vpc, are addressed in Specification 5.5.15.SRs 3.8.6.2 and 3.8.6.5 require verification that the cell float voltages are equal to or greater than the short term absolute minimum voltage of 2.07 V. The r ee~han9-e-2 y, cczh-- e <--"nsert 2 SR 3.8.6.3 The limit specified for electrolyte level (i.e., greater than or equal to the low level mark) ensures that the plates suffer no physical damage and maintains adequate electron transfer capability.
T-hf-Pr~equriyi 4 l--nsert 2 Cook Nuclear Plant Unit 2 B 3.8.6-5 Revision No. 0 Cook Nuclear Plant Unit 2 B3.8.6-5 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.6.4 This Surveillance verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (i.e., 60&deg;F for the Train A and Train B 250 VDC batteries and 45&deg;F for the Train N 250 VDC battery).
Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the design requirements.
Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity.
The e(ke.t) Insert 2 SR 3.8.6.6 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.6.6;however, only the modified performance discharge test may be used to satisfy the battery service test requirements of SR 3.8.4.3.A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity.
Initial conditions for the modified performance discharge test should be identical to those specified for a performance discharge test as specified in IEEE-450 (Ref. 1).It may consist of just two rates; for instance the one minute rate for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the modified performance discharge test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test must remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test. Currently, the modified performance discharge test is performed by testing the battery using the service test profile for the first 4 hours followed by the performance discharge test profile for the Cook Nuclear Plant Unit 2 B3.8.6-6 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued) remainder of the test. This method has been determined by the system engineer and the battery manufacturer to be an acceptable modified performance test procedure, and is consistent with I EEE-450 (Ref. 1).The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 1) and IEEE-485 (Ref. 3). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.
Furthermore, the battery is sized to meet the assumed duty cycle loads when the battery design capacity reaches this 80% limit. net2 1r-ts4-ar-a{e t If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity -> 100% of the manufacturer's ratings. Degradation is indicated, according to IEEE-450 (Ref. 1), when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is below 90% of the manufacturer's rating.The 12 month and 60 month Frequencies are consistent with the recommendations in IEEE-450 (Ref. 1). The 24 month Frequency is derived from the recommendations of IEEE-450 (Ref. 1).This SR is modified by a Note. The reason for the Note is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3 or4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 2 B3.8.6-7 Revision No. 0 Inverters
-Operating B 3.8.7 BASES ACTIONS (continued) inverter inoperability.
This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail. When the 120 VAC vital bus is powered from its regulated 600/1 20 VAC transformer, it is relying upon interruptible AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the 120 VAC vital buses is the preferred source for powering instrumentation trip setpoint devices.B.._11 With two inverters~in the same train inoperable, the remaining inverters are capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in one of the two remaining inverters could result in the minimum ESF functions not being supported.
Therefore, one of the inverters must be restored to OPERABLE status within 6 hours.The 6 hour Completion Time is consistent with that allowed for an inoperable RTS train and an inoperable ESFAS train, since the inverters support the 120 VAC vital buses, which in turn support the RTS and ESFAS trains.C.1 and C.2 If the Train A or Train B inverter(s) cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.8.7.1 This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and the 120 VAC vital buses energized from the associated inverter.
Each inverter must be connected to its associated 250 VDC bus. The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation of the RTS and ESFAS connected to the 120 VAC vital buses.
~ Insert 2-r Cook Nuclear Plant Unit 2 B3873Rvso o B3.8.7-3 Revision No. 0 Inverters
-Shutdown B 3.8.8 BASES SURVEILLANCE REQ U IREM ENTS SR 3.8.8.1 This Surveillance verifies that the required inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter.
The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation connected to the AC vital buses. -T-e-+ay-.F-eue-~~seeeeumte--~~l~-~a 2 REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2 B3884Rvso o B 3.8.8-4 Revision No. 0 Distribution Systems -Operating B 3.8.9 BASES SURVEILLANCE REQ UI REM ENTS SR 3.8.9.1 This Surveillance verifies that the required AC, DC, and 120 VAC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment.
The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus.The verification of proper voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these buses.
2~beye~eR~=i9~
UllUUIU REFERENCES
: 1. Safety Guide 6, March 1971.2. UFSAR, Chapter 14.3. Regulatory Guide 1.93, December 1974.Cook Nuclear Plant Unit 2B38910RvsoN.0 B 3.8.9-10 Revision No. 0 Distribution Systems -Shutdown B 3.8.10 BASES ACTIONS (continued) power is not available.
It is therefore required to declare the FHAEV System inoperable.
Since the Unit 1 AC electrical power distribution subsystem only affects the FHAEV System, the associated portions of the FHAEV System are declared inoperable and the applicable ACTIONS of LCO 3.7.13, "Fuel Handling Area Exhaust Ventilation (FHAEV) System," are entered.SURVEILLANCE REQUIREMENTS SR 3.8.10.1 This Surveillance verifies that the AC, DC, and 120 VAC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized.
The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. Since the Unit 1 AC electrical power distribution subsystem only*affects the FHAEV System, the SR is modified by a Note that specifies the SR is not required to be met for the Unit 1 AC electrical power distribution subsystem when the associated FHAEV System is not required to be OPERABLE per LCO 3.7.13. Te-d i-Fr-eiqsery ,ak"='-Insert 2 REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2 B 3.8.10-4 Revision No. 0 Boron Concentration B 3.9.1 BASES ACTIONS (continued)
Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position.
Operations that individually add limited positive reactivity (e.g., temperature fluctuations from inventory addition or temperature control fluctuations), but when combined with all other operations affecting core reactivity (e.g., intentional boration) result in overall net negative reactivity addition, are not precluded by this action.A.3 In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately.
In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied.
The only requirement is to restore the boron concentration to its required value as soon as possible.
In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.
Once actions have been initiated, they must be continued until the boron concentration is restored.
The restoration time depends on the amount of boron that must be injected to reach the required concentration.
SURVEILLANCE REQUIREMENTS SR 3.9.1.1 and SR 3.9.1.2 These SRs ensure that the coolant boron concentration in the RCS, and connected portions of the refueling canal and the refueling cavity, is within the COLR limits. The boron concentration is determined periodically and prior to re-connecting portions of the refueling canal and the refueling cavity to the RCS, by chemical analysis...-.........
J.-..'--'-...-............................-
sa le-s.T-he.-t~9eruefy49-s13a.ec.r-eaereat4eg e-xper-e5'1tat e
e-adequ~e~e-.
The SR 3.9.1.2 Frequency of once within 72 hours prior to connecting the refueling canal and refueling cavity to the RCS ensures that if any dilution activity has occurred while the cavity and canal were disconnected from the ROS, correct boron concentration is verified prior to communication with the RCS.-Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.Cook Nuclear Plant Unit 2 B3913Rvso o B 3.9.1-3 Revision No. 0 Nuclear Instrumentation B 3.9.2 BASES ACTIONS (continued) since CORE ALTERATIONS and positive reactivity additions are not to be made, the core reactivity condition is stabilized until the source range neutron flux monitors are OPERABLE.
This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists.The Completion Time of once per 12 hours is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified.
The 12 hour Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.C.1 With no audible count rate OPERABLE, prompt and definite indication of a boron dilution event, consistent with the assumptions of the safety analysis, is lost. In this situation, the boron dilution event may not be detected quickly enough to assure sufficient time is available for operators to manually isolate the unborated water source and stop the dilution prior to the loss of SHUTDOWN MARGIN. Therefore, action must be taken to prevent an inadvertent boron dilution event from occurring.
This is accomplished by isolating all the unborated water flow paths to the Reactor Coolant System. Isolating these flow paths ensures that an inadvertent dilution of the reactor coolant boron concentration is prevented.
The Completion Time of "Immediately" assures a prompt response by operations and requires an operator to initiate actions to isolate an affected flow path immediately.
Once actions are initiated, they must be continued until all the necessary flow paths are isolated or the circuit is restored to OPERABLE status.SURVEILLANCE SR 3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is normally a comparison of the parameter indicated on one channel to a similar parameter on another channel. It is based on the assumption that the two indication channels should be consistent with core conditions.
Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.
The ~~Ilnsert 2 Cook Nuclear Plant Unit 2 B 3.9.2-3 Revision No. 0 Cook Nuclear Plant Unit 2 B3.9.2-3 Revision No. 0 Nuclear Instrumentation B 3.9.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.2.2 SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION-e-ver-y
* CHANNEL CALIBRATION is a complete check of the instrument loop, except the detector.
The CHANNEL CALIBRATION for the Westinghouse source range neutron flux monitors also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. In addition, the CHANNEL CALIBRATION includes verification of the audible count rate function for the required monitor. This SR is modified b~y a Note stating that neutron detectors are .excluded from the CHANNEL CALIBRATION. j.s -&#xa2;e el,{-e
"-Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Section 14.1.5.REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Section 14.1.5.Cook Nuclear Plant Unit 2 B3924Rvso o B 3.9.2-4 Revision No. 0 Containment Penetrations B 3.9.3 BASES SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations is in its required status. The LCO 3.9.3.c.2 status requirement, which requires penetrations to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System, can be verified by ensuring each required valve operator is capable of closing automatically if needed. This Surveillance does not require cycling of the valves since this is performed at the appropriate frequency in accordance with SR 3.9.3.2.-e1 i-e .ete nsert 2
~ t.q -sleeme~r thJe -~~-e.-- pe
--htr~ae-~-~ap-rae
-r-4-le-~SR 3.9.3.2 This Surveillance demonstrates that each required containment purge supply and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. Th24rmr-'--nsert 2 Frgoc "anan ,~s~e* r-equl-- n. LC .. ,-6eeit-aRwe-Pure--St~pp]
a t- l? ! t I ....... II? _ _ [__" ........ The SR is modified by a Note stating that this Surveillance is not required to be met for valves in isolated penetrations.
The LCO provides the option to close penetrations in lieu of requiring automatic actuation capability.
REFERENCES
: 1. UFSAR, Section 14.2.1.5.Cook Nuclear Plant Unit 2 B3934Rvso o B3.9.3-4 Revision No. 1 RHR and Coolant Circulation
-High Water Level B 3.9.4 BASES SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core.
2 4-2-ee-s --ei ,-eef&sect;1er~ewieterg e-~ pdf~e~tple-aerfttr
, REFERENCES
: 1. UFSAR, Section 9.3.2.Cook Nuclear Plant Unit 2 B3944Rvso o B 3.9.4-4 Revision No. 0 RHR and Coolant Circulation
-Low Water Level B 3.9.5 BASES ACTIONS (continued)
B.._22 If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation.
Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.
B.3, B.4, and B.5 If no RHR is in operation, the following actions must be taken: a. The equipment hatch must be closed and secured with four bolts;b. One door in each air lock must be closed; and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System.With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions s'tated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.The Completion Time of 4 hours allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.SURVEILLANCE REQUIREMENTS SR 3.9.5.1.This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal anid boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. T-heee-f--efsi-ufei I --nsertI 2 p
*mat 4a4~--R-Se-r4  Cook Nuclear Plant Unit 2 B 3.9.5-3 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.9.5-3 Revision No. 0 RHR and Coolant Circulation
-Low Water Level B 3.9.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to the required pump.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.Insert 2 REFERENCES
: 1. UFSAR, Section 9.3.2.Cook Nuclear Plant Unit 2 B3954Rvso o B 3.9.5-4 Revision No. 0 Refueling Cavity Water Level B 3.9.6 BASES.ACTIONS A.1 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.SURVEILLANCE SR 3.9.6.1 REQ U IREM ENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 1).REFERENCES
: 1. UFSAR, Section 14.2.1.2. 10 CFR 100.10.2 Cook Nuclear Plant Unit 2B39.-ReionN.2 B3.9.6-2 Revision No, 26 Enclosure 8 to AEP-NRC-2015-46 TSTF-425 (NUREG-1431) versus CNP TS Cross- Reference Technical Specification Section Title TSTF-425 CNP CNP____________________________________
______Unit I Unit 2 Reactivity Controls Systems / Shutdown Margin 3.1.1 3.1.1.1 3.1.1.1 3.1.1.1 Core Reactivity 3.1.2 3.1.2.1 3.1.2.1 3.1.2.1 Rod Group Alignment Limits 3.1.4 3.1.4.1 3.1.4.1 3.1.4.1 3.1.4.2 3.1.4.2 3.1.4.2 Shutdown Bank Insertion Limits 3.1.5 3.1.5.1 3.1.5.1 3.1.5.1 Conro Bnk nsrtonLimts3..63.1.6.2 3.1.6.2 3.1.6.2 3.1.6.3 3.1.6.3 3.1.6.3 PHYSICS TESTS Exceptions
-MODE 2 3.1.8 3.1.8.2 3.1.8.1 3.1.8.1 3.1.8.3 3.1.8.2 3.1.8.2 3.1.8.4 3.1.8.3 3.1.8.3 FQ(Z) (CAOC-Fxy Methodology) 3.2.1A _______ 3.2.1.1 3.2.1.2 F 0 (Z) (RAOC-W(Z)
Methodology)3.2.1 B 3.2.1.1 3.2.1.2 FQ(Z) (CAOC-W(Z)
Methodology)3.2.1 C 3.2.1.1 3.2.1.1 3.2.1.1 3.2.1.2 3.2.1.2 3.2.1.2 FNAH 3.2.2 3.2.2.1 3.2.2.1 3.2.2.1 AFD (CAOC Methodology) 3.2.3A 3.2.3.1 3.2.3.1 3.2.3.1 3.2.3.2 3.2.3.2 3.2.3.2 3.2.3.3 3.2.3.3 3.2.3.3 AFD (RAOC Methodology) 3.2.3B 3.2.3.1 QPTR 3.2.4 3.2.4.1 3.2.4.1 3.2.4.1 3.2.4.2 3.2.4.2 3.2.4.2 RTS Instrumentation 3.3.1 3.3.1.1 3.3.1.1 3.3.1.1 3.3.1.2 3.3.1.2 3.3.1.2 3.3.1.3 3.3.1.3 3.3.1.3 3.3.1.4 3.3.1.4 3.3.1.4 3.3.1.5 3.3.1.5 3.3.1.5 3.3.1.6 3.3.1.6 3.3.1.6 3.3.1.7 3.3.1.7 3.3.1.7 3.3.1.8 3.3.1.8 3.3.1.9 3.3.1.9 3.3.1.8 3.3.1.11 3.3.1.11 3.3.1.9 3.3.1.10 3.3.1.10 3.3.1.12 3.3.1.12 3.3.1.10 3.3.1.13 3.3.1.13 3.3.1.11 3.3.1.14 3.3.1.14 3.3.1.12 3.3.1.15 3.3.1.15 3.3.1.13 3.3.1.16 3.3.1.16 3.3.1.14 3.3.1.17 3.3.1.17 3.3.1.16 3.3.1.19 3.3.1.19 ESFAS Instrumentation 3.3.2 3.3.2.1 3.3.2.1 3.3.2.1 3.3.2.2 3.3.2.2 3.3.2.2 3.3.2.3 3.3.2.3_____________________________________________
3.3.2.3_______________________________________
3.3.2.4 3.3.2.4 3.3.2.4 Enclosure 8 to AEP-NRC-2015-46 Pg Page 2 Technical Specification Section Title TSTF-425 CNP CNP Uniti Unit2 3.3.2.5 3.3.2.6 3.3.2.6 3.3.2.6 3.3.2.8 3.3.2.8 3.3.2.7 3.3.2.5 3.3.2.5 3.3.2.7 3.3.2.7 3.3.2.8 3.3.2.9 3.3.2.9 3.3.2.9 3.3.2.10 3.3.2.10 3.3.2.11 3.3.2.11 3.3.2.10 3.3.2.12 3.3.2.12 PAM Instrumentation 3.3.3 3.3.3.1 3.3.3.1 3.3.3.1 S3.3.3.2 3.3.3.3 3.3.3.3 Remote Shutdown System 3.3.4 3.3.4.1 3.3.4.1 3.3.4.1 3.3.4.2...
3.3.4.3 3.3.4.2 3.3.4.2 3.3.4.4 --LOP DG StartInstrumentation 3.3.5 3.3.5.1 3.3.5.1 3.3.5.1 3.3.5.2 3.3.5.2 3.3.5.2 3.3.5.3 3.3.5.3 3.3.5.3 Containment Purge and ExhaustlIsolation Instrumentation 3.3.6 3.3.6.1 3.3.6.1 3.3.6.1 3.3.6.2 3.3.6.2 3.3.6.2 3.3.6.3 3.3.6.3 3.3.6.3 3.3.6.4 3.3.6.5 3.3.6.6 3.3.6.4 3.3.6.4 3.3.6.7 3.3.6.5 3.3.6.5 3.3.6.8 3.3.6.6 3.3.6.6 3.3.6.9 3.3.6.7 3.3.6.7 CREES ActuationlInstrumentation 3.3.7 3.3.7.1...
3.3.7.2 3.3.7.3 3.3.7.1 3.3.7.1 3.3.7.4 3.3.7.2 3.3.7.2 3.3.7.5 3.3.7.6 .....3.3.7.7 3.3.7.3 3.3.7.3 3.3.7.8 3.3.7.9 FBACS ActuationlInstrumentation 3.3.8 3.3.8.1 --3.3.8.2 3.3.8.3 3.3.8.4 ---3.3.8.5 BDPS 3.3.9 3.3.9.1 3.3.8.1 3.3.8.1 3.3.9.2 3.3.9.3 3.3.8.2 3.3.8.2 RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4.1.1 3.4.1.1 3.4.1.1 3.4.1.2 3.4.1.2 3.4.1.2 3.4.1.3 3.4.1.3 3.4.1.3 3.4.1.4 3.4.1.4 3.4.1.4 RCS Minimum Temperature for Criticality 3.4.2 3.4.2.1 3.4.2.1 3.4.2.1 RCS P/T Limits 3.4.3 3.4.3.1 3.4.3.1 3.4.3.1 RCS Loops -MODES 1 and 2 -3.4.4 3.4.4.1 3.4.4.1 3.4.4.1 Enclosure 8 to AEP-NRC-2015-46 Pg Page 3 Technical Specification Section Title TSTF-425 CNP CNP Unit I Unit 2 RCS Loops -MODE 3 -3.4,5 3.4.5.1 3.4.5.1 3.4.5.1 3.4.5.2 3.4.5.2 3.4.5.2 3.4.5.3 3.4.5.3 3.4.5.3 RCS Loops -MODE 4 -3.4.6 3.4.6.1 3.4.6.1 3.4.6.1 3.4.6.2 3.4.6.2 3.4.6.2 3.4.6.3 3.4.6.3 3.4.6.3 RCS Loops -MODE 5, Loops Filled 3.4.7 3.4.7.1 3.4.7.1 3.4.7.1 3.4.7.2 3.4.7.2 3.4.7.2 3.4.7.3 3.4.7.3 3.4.7.3 RCS Loops -MODE 5, Loops Not Filled 3.4.8 3.4.8.1 3.4.8.1 3.4.8.1 3.4.8.2 -3.4.8.2 3.4.8.2 Pressurizer 3.4.9.1 3.4.9.1 3.4.9.1 3.4.9.2 3.4.9.2 3.4.9.2 3.4.9.3 Pressurizer PORVs 3.4.11 3.4.11.1 3.4.11.1 3.4.11.1______________________________________
3.4.11.2 3.4.11.2 3.4.11.2 3.4.11.3 3.4.11.3 3.4.11.3 3.4.11.4 LTOP System 3.4.I2 __________________
3.4.12.1 3.4.12.1 3.4.12.1 3.4.12.2 3.4.12.2 3.4.12.2_______________________________________
3.4.12.3 3.4.12.3 3.4.12.3 3.4.12.4 3.4.12.4 3.4.12.4-3.4.12.5 3.4.12.5 3.4.12.5 3.4.12.6 3.4.12.6 3.4.12.6 3.4.12.7 3.4.12.7 3.4.12.7 3.4.12.8 3.4.12.8 3.4.12.8 3.4.12.9 3.4.12.9 3.4.12.9 RCS Operational LEAKAGE 3.4.13 3.4.13.1 3.4.13.1 3.4.13.1_______________________________________
3.4.13.2 3.4.13.2 3.4.13.2 RCS PIV Leakage 3.4.14 3.4.14.1 3.4.14.2 3.4.14.2 3.4.14.2 3.4.14.3 RCS Leakage Detection Instrumentation 3.4.15 3.4.15.1 3.4.15.1 3.4.15.1 3.4.15.2 3.4.15.2 3.4.15.2 3.4.15.3 3.4.15.3 3.4.15.3 3.4.15.4 3.4.15.4 3.4.15.4 3.4.15,5... 3.4.15.5 3.4.15.5 RCS Specific Activity 3.4.16 3.4.16.1 3.4.16.1 3.4.16.1 3.4.16.2 3.4.16.2 3.4.16.2 3.4.16.3 3.4.16.3 3.4.16.3 RCS Loop Isolation Valves 3.4.17 3.4.17.1 RCS Loops -Test Exceptions 3.4.19 3.4.19.1 Accumulators 3.5.1 3.5.1.1 3.5.1.1 3.5.1.1 3.5.1.2 3.5.1.2 3.5.1.2 3.5.1.3 3.5.1.3 3.5.1.3 3.5.1.4 3.5.1.4 3.5.1.4 3.5.1.5 3.5.1.5 3.5.1.5 ECCS -Operating 3.5.2 3.5.2.1 3.5.2.1 3.5.2.1 Enclosure 8 to AEP-NRC-2015-46 Page 4 Technical Specification Section Title TSTF-425 CNP CNP Uniti1 Unit 2_______________________________________
3.5.2.2 3.5.2.2 3.5.2.2___________________________________________
3.5.2.3 3.5.2.4 3.5.2.5 3.5.2.4 3.5.2.4 3.5.2.6 3.5.2.5 3.5.2.5 3.5.2.7 3.5.2.6 3.5.2.6 3.5.2.8 3.5.2.7 3.5.2.7 RWST 3.5.4 3.5.4.1 3.5.4.1 3.5.4.1________________________________________
3.5.4.2 3.5.4.2 3.5.4.2________________________________________
3.5.4.3 3.5.4.3 3.5.4.3 Seal Injection Flow 3.5.5 3.5.5.1 3.5.5.1 3.5.5.1 BIT 3.5.6 3.5.6.1 .....--_____________________________________________
3.5.6.2 3.5.6.3 Containment Air Locks (Atmospheric, Subatmospheric, Ice 3.6.2.2 3.6.2.2 3.6.2.2 Condenser, and Dual) 3.6.2 Containment Isolation Valves (Atmospheric, Subatmospheric, Ice 3.6.3.1 Condenser, and Dual) 3.6.3 ____3.6.3.2 3.6.3.1 3.6.3.1 3.6.3.3 3.6.3.2 3.6.3.2 3.6.3.5 3.6.3.6 3.6.3.7. ..3.6.3.8 3.6.3.5 3.6.3.5 3.6.3.9 --3.6.3.10 Containment Pressure (Atmospheric, Dual, and Ice Condenser) 3.6.4A.1 3.6.4.1 3.6.4.1 3.6.4A Containment Pressure (Subatmospheric) 3.6.4B 3.6.4B.1 ...Containment Air Temperature (Atmospheric and Dual) 3.6.5A 3.6.5A. 1 Containment Air Temperature (Ice Condenser) 3.6.5B 3.6.5B.1 3.6.5.1 3.6.5.1 3.6.5B.2 3.6.5.2 3.6.5.2 Containment Air Temperature (Subatmospheric) 3.6.5SC 3.6.5C. 1 Containment Spray and Cooling Systems (Atmospheric and Dual) 3.6.6A.1 --3.6.6A 3.6.6A.2 --3.6.6A.3 3.6.6A.4 --3.6.6A.5 3.6.6A.6 3.6.6A.7 3.6.6A.8 Containment Spray and Cooling Systems (Atmospheric and Dual) 3.6.6B.1 --3.6.6B 3.6.6B.2 3.6.6B.3 ---3.6.6B.5 3.6.6B.6 --3.6.6B.7 3.6.6B.8 Enclosure 8 to AEP-NRC-2015-46 Pg Page 5 Technical Specification Section Title TSTF-425 CNP CNP Unit I Unit 2 Containment Spray and Cooling Systems (Ice Condenser) 3.6.60 3.6.6C.1 3.6.6.1 3.6.6.1________________________________________
3.6.6C.3 3.6.6.3 3.6.6.3________________________________________
3.6.6C.4 3.6.6.4 3.6.6.4 3.6.60.5 QS System (Subatmospheric) 3.6.6D 3.6.6D.1 .....-3.6.6D.3 .....3.6.6D.4 3.6.6D.5 RS System (Subatmospheric) 3.6.6E 3.6.6E.1 3.6.6E.2 3.6.6E.3 .....__3.6.6E.6 3.6.6E.7 Spray Addititve System (Atmospheric, Subatmospheric, Ice 3.6.7.1 3.6.7.1 3.6.7.1 Condenser, and Dual) 3.6.7 ____3.6.7.2 3.6.7.2 3.6.7.2 3.6.7.3 3.6.7.3 3.6.7.3 3.6.7.4 3.6.7.4 3.6.7.4 3.6.7.5 3.6.7.5 3.6.7.5 Shield building (Dual and Ice Condenser) 3.6.8 3.6.8.1 3.6.8.2 3.6.8.4 HMS (Atmospheric, Ice Condenser, and Dual) 3.6.9 3.6.9.1 3.6.9.2 3.6.9.3...
HIS (Ice Condenser) 3.6.10 3.6.10.1 3.6.9.1 3.6.9.1 3.6.10.2 3.6.9.2 3.6.9.2 3.6.10.3 3.6.9.3 3.6.9.3 lCS (Atmospheric and Subatmospheric) 3.6.11 3.6.11.1 3.6.11.3 3.6.11.4 SBAC (Dal nd ce Cndeser 3..133.6.13.1 3.6.13.3 3.6.13.4 3.6.13.5.AR (ce onensr)3.6143.6.14.1 3.6.10.1 3.6.10.1 3.6.14.2 3.6.10.2 3.6.10.2 3.6.14.3 3.6.10.3 3.6.10.3 3.6.14.4 3.6.10.4 3.6.10.4 Ice Bed (Ice Condenser) 3.6.15 3.6.15.1 3.6.11.1 3.6.11.1 3.6.15.2 3.6.11.2 3.6.11.2 3.6.15.3 3.6.11.3 3.6.11.3 3.6.15.4 3.6.11.4 3.6.11.4 3.6.15.5 3.6.11.6 3.6.11.6 3.6.15.6 3.6.11.5 3.6.11.5 Ice Condenser Doors (Ice Condenser) 3.6.16 3.6.16.1 3.6.12.1 3.6.12.1 3.6.16.2 3.6.12.2 3.6.12.2 3.6.16.3 3.6.12.4 3.6.12.4 3.6.16.4 3.6.12.5 3.6.12.5 Enclosure 8 to AEP-NRC-2015-46 Pg Page 6 Technical Specification Section Title TSTF-425 CNP CNP Unit1I Unit 2 3.6.16.5 3.6.12.6 3.6.12.6 3.6.16.6 3.6.12.7 3.6.12.7 3.6.16.7 3.6.12.3 3.6.12.3 Divider BarrierlInterit (Ice Condenser) 3.6.17 3.6.17.2 3.6.13.2 3.6.13.2 3.6.17.4 3.6.13.4 3.6.13.4 3.6.17.5 3.6.13.5 3.6.13.5 Containment Recirculation Drains (Ice Condenser) 3.6.18 3.6.18.1 3.6.14.1 3.6.14.1 3.6.18.1 ..........
3.6.18.2 3.6.14.3 3.6.14.3 MSIVs 3.7.2 3.7.2.2 3.7.2.2 3.7.2.2 MFIVs and MFRVs and [Associated Bypass Valves] 3.7.3 3.7.3.2 3.7.3.3 3.7.3.33.7.4 3.7.4.1 3.7.4.1 3.7.4.1 3.7.4.2 AFW System 3.7.5 3.7.5.1 3.7.5.1 3.7.5.1 3.7.5.3 3.7.5.3 3.7.5.3 3.7.5.4 3.7.5.4 3.7.5.4 CST 3.7.6 3.7.6.1 3.7.6.1 3.7.6.1 CCW System 3.7.7 3.7.7.1 3.7.7.1 3.7.7.1________________________________________
3.7.7.2 3.7.7.2 3.7.7.2________________________________________
3.7.7.3 3.7.7.3 3.7.7.3 SWS 3.7.8 3.7.8.1 3.7.8.1 3.7.8.1 3.7.8.2 3.7.8.2 3.7.8.2 3.7.8.3 3.7.8.3 3.7.8.3 UHS 3.7.9 3.7.9.1 3.7.9.2 3.7.9.1 3.7.9.1 3.7.9.3 3.7.9.4 .....-CREFS 3.7.10 3.7.10.1 3.7.10.1 3.7.10.1 3.7.10.3 3.7.10.3 3.7.10.3 3.7.10.4 3.7.11.1 3.7.11.1 CREATCS 3.7.11 3.7.11.1 3.7.11.2 3.7.11.2 ECCS PREACS 3.7.12 3.7.12.1 3.7.12.1 3.7.12.1 3.7.12.3 3.7.12.3 3.7.12.3 3.7.12.4 3.7.12.4 3.7.12.4 3.7.12.5... 3.7.13.1 3.7.13.1 FBACS 3.7.13 3.7.13.1 3.7.13.2 3.7.13.2 3.7.13.3 3.7.13.4 3.7.13.4 3.7.13.4 3.7.13.5 3.7.13.5 3.7.13.5 PREACS 3.7.14 3.7.14.1 --3.7.14.3 3.7.14.4 3.7.14.5 Fue Strag Pol ate Lvel3.7153.7.15.1 3.7.14.1 3.7.14.1[Fuel Storage Pool Boron Concentration]
3.7.16 3.7.16.1 3.7.15.1 3.7.15.1 AC ouce -Opeatng3..13.7.18.1.
3.7.17.1 3.7.17.1 3.8.1.1 3.8.1.1 3.8.1.1 Enclosure 8 to AEP-NRC-2015-46 Pg Page 7 Technical Specification Section Title TSTF-425 CNP CNP______ Uniti1 Unit 2 3.8.1.3 3.8.1.3 3.8.1.3 3.8.1.4 3.8.1.4 3.8.1.4______________________________________
3.8.1.5 3.8.1.5 3.8.1.5______________________________________
3.8.1.6 3.8.1.7 3.8.1.7______________________________________
3.8.1.7 3.8.1.8 3.8.1.8______________________________________
3.8.1.8 3.8.1.9 3.8.1.9 3.8.1.9 3.8.1.10 3.8.1.10___________________________________
3.8.1.10 3.8.1-.11 3.8.1.11_______________________________________
-3.8.1.11 3.8.1.12 3.8.1.12_________________________________________
3.8.1.12 3.8.1.13 3.8.1.13_______________________________________
3.8.1.13 3.8.1.14 3.8.1.14_______________________________________
3.8.1.14 3.8.1.15 3.8.1.15_______________________________________
3.8.1.15 3.8.1.16 3.8.1.16_______________________________________
3.8.1.16 3.8.1.17 3.8.1.17 3.8.1.17 3.8.1.21 3.8.1.21 3.8.1.18 3.8.1.18 3.8.1.18 3.8.1.19 3.8.1.19 3.8.1.19 3.8.1.20 3.8.1.22 3.8.1.22 3.8.1.20 3.8.1.20 Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 3.8.3.1 3.8.3.1 3.8.3.1 3.8.3.2.3.8.3.4 3.-8.1.6 3.8.1.6 3.8.3.5 3.8.3.3 3.8.3.3 DC Sources -0perating 3.8.4 3.8.4.1 3.8.4.1 3.8.4.1_______________________________________
3.8.4.2 3.8.4.2 3.8.4.2 3.8.4.3 3.8.4.3 3.8.4.3 Battery Parameters 3.8.6 3.8.6.1 3.8.6.1 3.8.6.1 3.8:6.2 3.8.6.2 3.8.6.2 3.8.6.3 3.8.6.3 3.8.6.3 3.8.6.4 3.8.6.4 3.8.6.4 3.8.6.5 3.8.6.5 3.8.6.5 3.8.6.6 3.8.6.6 3.8.6.6 Inverters
-Operating 3.8.7 3.8.7.1 3.8.7.1 3.8.7.1 Inverters
-Shutdown 3.8.8 3.8.8.1 3.8.8.1 3.8.8.1 Distribution Systems -Operating 3.8.9 3.8.9.1 3.8.9.1 3.8.9.1 Distribution Systems -Shutdown 3.8.10 3.8.10.1 3.8.10.1 3.8.10.1 Boron Concentration 3.9.1 3.9.1.1 3.9.1.1 3.9.1.1[Unborated Water Source Isolation Valves] 3.9.2 3.9.2.1 Nuclear Instrumentation 3.9.3 3.9.3.1 3.9.2.1 3.9.2.1_______________________________________
3.9.3.2 3.9.2.2 3.9.2.2 Containment Penetrations 3.9,4 3.9.4.1 3.9.3.1 3.9.3.1 3.9.4.2 3.9.3.2 3.9.3.2 RHR and Coolant Circulation
-High Water Level 3.9.5 3.9.5.1 3.9.4.1 3.9.4.1 RHR and Coolant Circulation
-Low Water Level 3.9.6 3.9.6.1 3.9.5.1 3.9.5.1 3.9.6.2 3.9.5.2 3.9.5.2 Refueling Cavity Water Level 3.9.7 3.9.7.1 3.9.6.1 3.9.6.1 Programs and Manuals 5.5 5.5.18 5.5.17 5.5.17 Enclosure 9 to AEP-NRC-2015-46 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION Description of Amendment Request: This amendment request involves the adoption of approved changes to the improved standard technical specifications (ISTS) Westinghouse Plants, NUREG-1431, to allow relocation of specific technical specification (TS) surveillance frequencies to a licensee-controlled program. The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090850642) related to the Relocation of Surveillance Frequencies to Licensee Control -RITSTF Initiative 5b and are described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).The proposed changes are consistent with U. S. Nuclear Regulatory Commission (NRC)-approved industry/TSTF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -RITSTF Initiative 5b." The proposed changes relocate surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP). The changes are applicable to licensees using probabilistic risk guidelines contained in NRC-approved Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. 071360456).
Basis for proposed no significant hazards consideration (NSHC): As required by 10 CFR 50.91 (a), the Indiana Michigan Power Company, licensee for Donald C. Cook Nuclear Power Plant Units 1 and 2, analysis of the issue of NSHC for adoption of TSTF-425, Revision 3 is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:
No.The proposed changes relocate the specified frequencies for-periodic surveillance requirements (SR) to licensee control under a new SFCP. Surveillance frequencies are not an initiator to any accident previously evaluated.
As a result, the probability of any accident previously evaluated is not significantly increased.
The systems and components required by the TS for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the SRs, and be capable of performing any mitigation function assumed in the accident analysis.
As a result, the consequences of any accident previously evaluated are not significantly increased.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:
No.No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.
In addition, the changes do not impose any new or different requirements.
The changes do not alter assumptions made in the safety analysis.
The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.Consequently, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Does the proposed amendment involve a significant reduction in a margin of safety?Response:
No.The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SS0s), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies.
Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, I&M will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP. NEI 04-10, Revision 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk'increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.Therefore, the proposed changes do not involve a significant reduction in a margin of safety.Based on the above, Indiana Michigan Power Company concludes that the proposed amendment presents NSHC under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.
Enclosure 10 to AEP-NRC-2015-46 Proposed Inserts N ote that Insert 1 and Inset 2 capitalization and punctuation is varied based on the use in each specific surveillance requirement.
j INSERT 1 In accordance with the Surveillance Frequency Control Program INSERT 2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Enclosure 11 to AEP-NRC-2015-46 REGULATORY COMMITMENTS The following table identifies an action committed to by Indiana Michigan Power Company (I&M)in this document.
Any other actiOns discussed in this submittal represent intended or planned actions by I&M. They are described to the U. S. Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.
All commitments discussed in this table are one-time commitments.
Commitment Scheduled Completion Date____ ___ ___ ___ ___ ____ ___ ___ ___ ___ ____ ___ ___ ___ ___(if applicable)
Implement the resolution for the following Supporting Requirements Prior to program from Enclosure 9, Technical Adequacy Justification Table: implementation SY-BI0, HR-G4, DA-C15, IFSN-A16, IFSN-A17, IFEV-A8, and I FQU-A3 Re-integrated the Fire Probabilistic Risk Assessment (PRA) model Prior to program with the. Internal Events PRA model. implementation The Computer Aided Fault Tree Analysis (CAFTA) plafform will be Prior to program implemented for the Internal Events PRA Model of Record. implementation Enclosure 6 to AEP-NRC-2015-46 CNP Unit I TS Bases Pages Marked to Show Proposed Changes SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)
: b. Bank position;c. RCS average temperature;
: d. Fuel burnup based on gross thermal energy generation;
: e. Xenon concentration;
: f. Samarium concentration;
: g. Isothermal temperature coefficient (ITC); and h. Boron penalty (MODES 4 and 5 only).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. The boron penalty must be applied in MODES 4 and 5 since all reactor coolant pumps may be stopped in these MODES.This extra amount of boron ensures that minimum response times are met for the operator to diagnose and mitigate an inadvertent boron dilution event prior to loss of SDM.
.--
c,-e
+Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Chapter 14.3. UFSAR, Section 14.2.5.4. UFSAR, Section 14.1.5.5. 10OCFR 100.Cook Nuclear Plant Unit 1 B3115Rvso o B 3.1.1-5 Revision No. 0 Core Reactivity B 3.1.2*BASES ACTIONS (continued)
B.._1 If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. If the SDM for MODE 3 is not met, then the boration required by SR 3.1.1.1 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 unit systems.SURVEILLANCE REQUIREMENTS SR 3.1.2.1 Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations.
The comparison is made, considering that other core conditions are fixed or stable, including RCS boron concentration, control rod position, RCS average temperature, fuel burnup based on gross thermal energy generation, 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 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. -T-he*reli ~ ~
e
=- Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Chapter 14.Cook Nuclear Plant Unit 1B3125ReionN.0 B3.1.2-5 Revision No. 0 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued) 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._22 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 in which the LCO requirements are not applicable.
To achieve this status,.the unit must be brought to at least MODE 3 within 6 hours.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 unit systems.SURVEILLANCE REQUIREMENTS SR 3.1.4.14a ea I-nsert 2..................
j ..... .....SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2, tripping each control rod would result in radial or axial power tilts, or oscillations.
Exercising each individual control rod ete &#xa3;-92eays-provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 8 steps will not cause radial or axial power tilts, or oscillations, to occur. 9&#xa3; M frq~e edeenn EABLI---f'efdT Between required performances of SR 3.1.4.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable the control rod(s) is considered to be OPERABLE.
At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken.
2.Cook Nuclear Plant Unit I B 3.1.4-7 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.1.4-7 Revision No. 0 Shutdown Bank Insertion Limits B 3.1.5 BASES SURVEILLANCE REQUIREMENTS SR 3.1.5.1 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..,inece=~tad--pG~
~ ead- eu -ba-1 dw ~ ste-t tnyr REFERENCES
: 1. UFSAR, Section 1.4.2.2. UFSAR, Section 1.4.5.3. UFSAR, Section 1.4.6.4. 10OCFR 50.46.5. U FSAR, Chapter 14.-Insert 2 Cook Nuclear Plant Unit 1 B3154Rvso o B 3.1.5-4 Revision No. 0 Control Bank Insertion Limits B 3.1.6 BASES SURVEILLANCE REQUIREMENTS SR 3.1.6.1 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. Verifying theI ECP calculation within 4 hours 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 J td-t aF ~- eJ --Lfher 1  --- Insert 2.
l,-eryiteroim to -ee  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 REFERENCES
: 1. UFSAR, Section 1.4.2.2. UFSAR, Section 1.4.5.3. UFSAR, Section 1.4.6.4. 10 CFR 50.46.5. UFSAR, Chapter 14.~=Insert 2 Cook Nuclear Plant Unit 1 B3165Rvso o B 3.1.6-5 Revision No. 1 PHYSICS TESTS Exceptions
-MODE 2 B 3.1..8 BASES ACTIONS (continued) 531 0 F could violate the assumptions for accidents analyzed in the safety analyses.D.1I If the Required Action and associated Completion Time of Condition C is not met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit 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 unit systems.SURVEILLANCE SR 3.1.8.1 REQUIREMENTS Verification that the RCS lowest loop Tavg is > 531 0 F will ensure that the unit is not operating in a condition that could invalidate the safety analyses.
Verification of the RCS temperature at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the initial conditions of the safety analyses are not violated.SR 3.1.8.2 Verification that the THERMAL POWER is < 5% RTP will ensure that the unit is not operating in a condition that could invalidate the safety analyses.
Ve t-ftinsert-SR 3.1.8.3 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects: a. RCS boron concentration;
: b. Bank position;c. RCS average temperature;
: d. Fuel burnup based on gross thermal energy generation;
: e. Xenon concentration; Cook Nuclear Plant Unit 1 B 3.1.8-6 Revision No. 0 Cook Nuclear Plant Unit 1 B3.1.8-6 Revision No. 0 PHYSICS TESTS Exceptions
-MODE 2 B 3.1.8 BASES SURVEILLANCE REQUIREMENTS (continued)
: f. Samarium concentration;
: g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH);h. Moderator Temperature Defect, when above the POAH; and i.Doppler Defect, when above the POAH.Using the ITC accounts for Doppler reactivity in this calculation when the reactor is subcritical or critical but below the POAH, and the fuel temperature will be changing at the same rate as the RCS.-tagen-elti-ati9-afid-ef-th~e--ewl~r-ebability---ef-afl-aeednt-tiig-wi{,hoitte-i~eq REFERENCES
: 1. 10 CFR 50, Appendix B, Section XI.2. 10 CFR 50.59.3. Regulatory Guide 1.68, Revision 2, August, 1978.4. ANSI/ANS-1 9.6.1-1 997, August 22, 1997.5. WCAP-1 3360-P-A, "Westinghouse Dynamic Rod Worth Measurement Technique," Revision .1, October 1998.6. PA-OSC-0061, "Westinghouse Position Paper on Power Distribution Measurement Requirements for Reload Startup Programs," February 2005.-Insert 2 Cook Nuclear Plant Unit 1 B3187Rvso o B3.1.8-7 Revision No. 1 FQ(Z)B 3.2.1 BASES ACTIONS (continued) 0.1 If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE or condition in which the LCO requirements are not applicable.
This is done by placing the unit in at least MODE 2 within 6 hours.This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.2.1.1 REQUIREMENTS Verification that FC(z) is within its specified limits involves increasing FM(z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain FC~(z) is then compared to its specified limits.If THERMAL POWER has been increased by > 10% RTP since the last determination of FC(z), another evaluation of this factor is required 24 hours after achieving equilibrium conditions at this higher power level (to ensure that values are being reduced sufficiently with power increase to stay within the LCO limits). The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification.
It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ(Z) was last measured.Th-rq
-Insert 2 a4 SR 3.2.1.1 is modified by a Note, which applies during power escalation after a refueling.
The Note states that the Surveillance is not required to be performed until 24 hours after equilibrium conditions at a power level for extended operation are achieved.
This Note allows the unit to startup from a refueling outage and reach the power level for extended operation (normally 100% RTP) prior to requiring performance of the SR. Within 24 hours after equilibrium conditions are reached at the power level for extended operation, the SR must be performed.
Cook Nuclear Plant Unit 1 B 3.2.1-6 Revision No. 0 Cook Nuclear Plant Unit 1 B3.2.1-6 Revision No. 0 FQ(Z)B 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last*verification.
It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which F 0 (Z) was last measured.T ~ Insert 2 tersl T-h- fe4 eneyef-F.=El44..akga ttee-~ i gt i&SR 3.2.1.2 is modified by Note 1, which applies during power escalation after a refueling.
The Note states that the Surveillance is not required to be performed until 24 hours after equilibrium conditions at a power level for extended operation are achieved.
This Note allows the unit to startup from a refueling outage and reach the power level for extended operation (normally 100% RTP) prior to requiring performance of the SR. Within 24 hours after equilibrium conditions are reached at the power level for extended operation, the SR must be performed.
REFERENCES
: 1. 10 CFR 50.46.2. UFSAR, Section 14.2.6.7.3. UFSAR, Section 1.4.5.4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.5. WCAP-1 0216-P-A, Rev. 1A, "Relaxation of Constant Axial Offset Control (and) F 0 Surveillance Technical Specification," February 1994.Cook Nuclear Plant Unit 1 B3218Rvso o B 3.2.1-8 Revision No. 0 B 3.2.2 BASES ACTIONS (continued)
A.4 Verification that iS within its specified limits after an out of limit occurrence ensures that the cause that led to the FN~)H exceeding its limit is corrected, and that subsequent operation proceeds within the LCO limit. This Action demonstrates that the FNAH limit is within the LCO limits prior to exceeding 50% RTP, again prior to exceeding 75% RTP, and within 24 hours after THERMAL POWER is > 95% RTP.This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action.B.1 When any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCO requirements are not applicable.
This is done by placing the unit in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging unit systems..SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The value of FN~H iS determined by using the movable incore detector system to obtain a flux distribution map. A data reduction computer program then calculates the maximum value of FNAH from the measured flux distributions.
The measured value of FNH must be multiplied by 1.04 to account for measurement uncertainty before making comparisons to the limit.After each refueling, FNAH must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that FN~H limits are met at the beginning of each fuel cycle.Tf 31S EFFD FI uubu a
--Insert 2 REFERENCES
: 1. UFSAR, Section 14.2.6.7.2. UFSAR, Section 1.4.5.3. 10 CFR 50.46.Cook Nuclear Plant Unit 1B322-ReionN.5 B 3.2.2-5 Revision No. 35 AFD B 3.2.3 BASES SURVEILLANCE SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AFD as indicated by the NIS excore channels is within the target band. -T-he--S r-ei14a~c-e-Fr-requeac-y-Gf-7--tays i- Insert 2 SR 3.2.3.2
,----Insert 2
ective-1Th power-days-(t -E-F4D)-t1e-ac-c uaift-er-saU
.SR 3.2.3.3 Measurement of the target flux difference is accomplished by taking a flux map when the core is at equilibrium xenon conditions, preferably at high power levels with the control banks nearly withdrawn.
This flux map provides the equilibrium xenon axial power distribution from which the target value can be determined.
The target flux difference varies slowly with core burnup.,A-F-eq~enlyof---E-pp-fterhreiiii g-an1E t er-e,{ter-fiif Insert 2 A Note modifies this SR to allow the predicted beginning of cycle AFD from the cycle nuclear design tO be used to determine the initial target flux difference after each refueling.
REF ERENCES 1. WCAP-8385 (Westinghouse proprietary) and WCAP-8403 (nonproprietary), "Power Distribution Control and Load Following Procedures," Westinghouse Electric Corporation, September 1974.2. UFSAR, Section 7.4.Cook Nuclear Plant Unit 1 B3236Rvso o B3.2.3-6 Revision No. 1 QPTR B 3.2.4 BASES ACTIONS (continued)
Action A.5). The intent of this Note is to have the peaking factor Surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limits and the core returned to power.B.1I If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE or other specified condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging unit systems.SURVEILLANCE SR 3.2.4.1.REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is< 75% RTP and the input from one Power Range Neutron Flux channel is inoperable.
Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1.This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. -T-he.-----
Insert 2 ,f astksit con ,te .nom~ .n lam.. "alal toteoeor in the contro &#xf7;roo...For those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is not required until 12 hours after the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is > 75% RTP.With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded.
Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased.- " " "
* YR-'--- Insert 2 Cook Nuclear Plant Unit 1 B 3.2.4-5 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.2.4-5 Revision No. 0 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.1 Performance of the CHANNEL OH ECK-en-ewe 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.
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.
t-4euftemens-lers-fer Insert 2 SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output e~ei3,-94,-hetw&-.
If the calorimetric exceeds the NIS channel output by > 2% RTP, the NIS is not declared inoperable, but must be adjusted.
If the NIS channel output ca~nnot be properly adjusted, the channel is declared inoperable.
Two Notes modify SR 3.3.1.2. The first Note indicates that the NIS channel output shall be adjusted consistent with the calorimetric results if the absolute difference between the NIS channel output and the calorimetric is > 2% RTP. The second Note clarifies that this Surveillance is required only if reactor power is > 15% RTP and that 12 hours is allowed for performing the first Surveillance after reaching 15% RTP. At lower power levels, calorimetric data are inaccurate.~ ~ Insert 2 h er'lif efcesdmts{-t-4' Cook Nuclear Plant Unit 1 B 3.3.1-38 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output.ev#e~y-If the absolute difference is > 3%, the NIS channel is still OPERABLE, but must be readjusted.
If the 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.Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be adjusted if the absolute difference between the incore and excore AFD is > 3%. Note 2 clarifies that the Surveillance is required only if reactor power is > 15% RTP and that 24 hours is allowed for performing the first Surveillance after reaching 15% RTP.ef tgee-ee~~e~-at-mrtrt~~~rtn l nsert 2 SR 3.3.1.4 SR 3.3.1.4 is the performance of a TADOT evtery!!di&#xa3;et-ays=a 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-Techqnical 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 RTB undervoltage 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 3.3.1.17.
The bypass breaker test shall include a local shunt trip. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service.Cook Nuclear Plant Unit 1 B3313 eiinN.1 B 3.3.1-39 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)4 Insert 2 jtted=a4f-efere~aee=14.
SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The SSPS is tested evr/9.lase using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
T~he-l4-ST-B-4Si-u4~d Insert 2.iH-R-efeireaee41=.
SR 3.3.1.6 SR 3.3.1.6 is the performance of a TADOT and is performed every 92 days on a STAGGERED TEST BASIS. This test applies to the SI Input from ESFAS 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 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.
STAS lEDT-STBA&&=s=Insert 2
SR 3.3.1.7 SR 3.3.1.7 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 the f(AI)input to the Overtemperature AT Function.A Note modifies SR 3.3.1.7. The Note states that this Surveillance is required only if reactor power is > 50% RTP and that 24 hours is allowed for performing the first surveillance after reaching 50% RTP.eX
~ ~ Insert 2 Cook Nuclear Plant Unit 1 B 3.3.1-40 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.8 SR 3.3.1.8 is the performance of a A COT is performed on each required channel to ensure the entire 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 COT 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.The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 8.SR 3.3.1.8 is modified by a Note that provides a 12 hour delay in the requirement to perform this Surveillance for Function 2.b channels after reducing THERMAL POWER below the P-I10 interlock.
The Frequency of 12 hours after reducing power below P-10 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.
SR 3.3.1.9 I=-nsert 2CHANN EL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
Cook Nuclear Plant Unit I1 ..-1ReiinN.1 B 3.3.1-41 Revision .No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)Insert 2 This SR is modified by a Note that states that neutron detectors are excluded from the CHANNEL CALIBRATION.
Changes in power range neutron detector sensitivity are compensated for by normalization of the channel output based on a power calorimetric and flux map performed above 15% RTP (SR 3.3.1.2).SR 3.3.1.10 SR 3.3.1.10 is the performance of a TADOT and is-p@FfGeFF
\ee~ey <=-- Insert 2 e2.2--ays.
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 SR is modified by a Note that excludes verification of relay Setpoints from the TADOT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION.
The Frequency of 92 days is justified in Reference 10.SR 3.3.1.11 SR 3.3.1.11 is the performance of a COT-every--1-84-4ays.
A COT is performed on each required channel to ensure the entire 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 COT 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 Sp~ecificatio~ns tests at least once per refueling interval with applicable extensions.
Setpoints must be within the Allowable Values specified in Table 3.3.1-1.The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
Cook Nuclear Plant Unit 1 ..-2ReiinN.1 B 3.3.1-42 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 8.The Frequency is modified by two Notes. Note 1 provides a 12 hour delay in the requirement to perform this Surveillance for intermediate range instrumentation after reducing THERMAL POWER below the P-10 interlock.
The Frequency of 12 hours after reducing power below P-10 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.
Note 2 provides a 4 hour delay in the requirement to perform this Surveillance for source range instrumentation after THERMAL POWER is reduced below the P-6 interlock.
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 RTBs are open and SR 3.3.1.11 is no longer required to be performed.
If the unit is to be in MODE 3 with the RTBs closed for > 4 hours this Surveillance must be performed prior to4 hours after THERMAL POWER is reduced below the P-6 interlock.Insert 2 SR 3.3.1.12.
N-ip y48m~e e44ay 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
:FeFetee~F4~y~4ae~mh~euftee--
I nse rt 2 dt~ft4the-tep SR 3.3.1.13CHANN EL 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.Cook Nuclear Plant Unit 1 B3314 eiinN.1 B 3.3.1-43 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.ent' 2 SR 3.3.1.14 SR 3.3.1.14 is the performance of a CHANNEL CALIBRATION eiCHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The CHANNEL CALIBRATION for the source range neutron detectors also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION.
Changes in power range neutron detector sensitivity are compensated for by normalization of the channel output based on a power calorimetric and flux map performed above 15% RTP (SR 3.3.1.2).Changes in intermediate range neutron flux detector sensitivity are compensated for by periodically evaluating the compensating voltage setting and making adjustments as neceSsary.
Changes in source range neutron detector sensitivity are compensated for by periodically obtaining the detector plateau or preamp discriminator curves, evaluating those curves, comparing thecurves to the manufacturer's data, and adjusting the channel output as necessary.e 4=-. Insert 2.SR 3.3.1.15 SR 3.3.1.15 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.13, e'~ert--=24--meonhs.
Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the resistance temperature detectors (RTD) sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.Cook Nuclear Plant Unit 1 B 3.3.1-44 Revision No. 10 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
~ nse rt 2 This SR is modified by a Note that provides a 72 hour delay in the requirement to perform a normalization of the AT channels after THERMAL POWER is > 98% RTP. .The intent of this Note is to maintain reactor power at a nominal 97% RTP to 98% RTP level until the AT normalization is complete before increasing reactor power to 100% RTP.SR 3.3.1.16 SR 3.3.1.16 is the performance of a COT of RTS interlocks 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 COT 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.
oThe-
& Insert 2
~ w4-aop SR 3.3.'1.17 SR 3.3.1.1.7 is the performance of a TADOT of the Manual Reactor Trip (including reactor trip bypass breakers) and RCP Breaker Position.
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.
4e~~rj.2 t s The test shall independently verify the OPERABILITY of the undervoltage and shunt trip mechanisms for the Manual Reactor Trip Funlction for the Reactor Trip Breakers and Reactor Trip Bypass Breakers.
The Reactor Trip Bypass Breaker test shall include testing of the automatic undervoltage trip.Cook Nuclear Plant Unit 1B331-5RvsoN.9 B 3.3.1-45 Revision No. 9 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Insert 2 SR 3.3.1.18 SR 3.3.1.18 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 as described in SR 3.3.1.4, except that this test is performed prior to exceeding the P-8 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-8 interlock.
SR 3.3.1.19 SR 3.3.1.19 verifies that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analysis.
Response time testing acceptance criteria are included in UFSAR, Table 7.2-6 (Ref. 12). 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 to the point at which the equipment reaches the required functional state (i.e., control and shutdown rods fully inserted in the reactor core).For channels that include dynamic transfer Functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer Function set to one, .with the resulting measured response time compared to the appropriate UFSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value, provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.Cook Nuclear Plant Unit 1 B3314 eiinN.1 B 3.3.1-46 Revision No. 17 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Response time may be verified by actual response time tests in any series of sequential, 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) in place, 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," (Ref. 13) 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, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 14) provides the basis and methodology for 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 the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
24mnh~raSTmESTB#t igetefn{ -Insert 2 SR 3.3.1.19 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.The response time testing of the neutron flux signal portion of the channel shall be measured from either the detector output or the input of the first electronic component in the channel.Cook Nuclear Plant Unit 1B33147RvsoN.5 B 3.3.1-47 Revision No. 5 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accurFacies.
SR 3.3.2.1 A CHAN NEL 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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.
tn-xeineta-m.t-~e Insert 2 SR 3.3.2.2 and SR 3.3.2.5 SR 3.3.2.2 is the performance of a This test is a check of the Loss of Voltage Function.
SR 3.3.2.5 is the performance of a TAOT test is a check of the Undervoltage RCP 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 TADOT of a relay. This is acceptable because all of the other required contacts of the relay are verified by otherTechnical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions.
Each SR is modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and~are verified during CHANNEL CALIBRATION.
Insert 2 Cook Nuclear Plant Unit 1 B 3.3.2-37 Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.3 SR 3.3.2.3 is the performance of an ACTUATION LOGIC TEST.=---he-SP,.S4 tao-tzd orc-c-r;T-02
.using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the master relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERABLE and that there is an intact voltage signal path to the master relay coils. -The-Feaey--ef-iei 2ye-e1.
* Insert 2 4, R~e n ee--4* .SR 3.3.2.4 SR 3.3.2.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. -T-his-es-j~er-feffl
=S=T ASt1. Thfe- Insert 2 SR 3.3.2.6 SR 3.3.2.6 is the performance of a COT. A COT is performed on each required channel to ensure the entire channel will perform the intended Function.
Setpoints must be found within the Allowable Values specified in Table 3.3.1-1. 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 COT 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 difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 6.
-,-- Insert 2 Cook Nuclear Plant Unit 1 .~-8ReiinN.5 B  Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.6 is modified by a Note which applies to the SI Containment Pressure -High, Containment Spray Containment Pressure -High High, Phase B Isolation Containment Pressure -High High, Steam Line Isolation Containment Pressure -High High, and CEQ System Containment Pressure -High Functions.
This Note requires, during the performance of SR 3.3.2.6, the associated transmitters of these Functions to be exercised by applying either a vacuum or pressure to the appropriate side of the transmitter.
Exercising the associated transmitters during the performance of the COT is necessary to ensure Functions 1 .c, 2.c, 3.b.(3), 4.c, and 7.c remain OPERABLE between each CHANNEL CALIBRATION.
SR 3.3.2.7 SR 3.3.2.7 is the performance of a CHANNEL CALIBRATION.-A--
CHAN NEL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.Y Insert 2
~~~f~f ~.
SR 3.3.2.8 SR 3.3.2.8 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contac~t operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function, or is placed in a condition where the relay contact operation can be verified without operation of the equipment.
Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.
t Insert 2.bseo~d .a~d-t~r~t~ingh~y-Eatet-a Cook Nuclear Plant Unit 1 B3323 eiinN.5 B 3.3.2-39 Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.9 SR 3.3.2.9 is the performance of a TADOT. This test is a check of the Manual Initiation Functions, the AFW pump start on trip of all MFW pumps, and the P-4 interlock. me~rm4-ve~=4Te~h 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.
In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.). "T-h~rtenwi~ey-deate, SR 3.3.2.10 SR 3.3.2.10 is the performance of a CHANNEL CALIBRATION.
**--~R-- ~s -v e- -CHAN NEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy.CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
4-Insert 2 eney=e 4-Insert 2 SR 3.3.2.11 SR 3.3.2.11 is the performance of an ACTUATION LOGIC TEST. This SR is applied to the balance of plant actuation logic and relays that do not have the SSPS test circuits installed to utilize the semiautomatic tester or perform the continuity check. All possible logic combinations are tested for Table 3.3.2-1 Functions 6.e and 6.g. Insert 2 SR 3.3.2.12 This SR ensures the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis.Response Time testing acceptance criteria are included in the UFSAR, Cook Nuclear Plant Unit 1 ..-0ReiinN.5 B 3.3.2-40 Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
Table 7.2-7 (Ref. 11). 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, to the point at which the equipment in both trains reaches the required functional state (e.g., pumps at rated discharge pressure, valves in full open or closed position).
For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer functions set to one with the resulting measured response time compared to the appropriate UFSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.Response time may be verified by actual response time tests in any series of sequential, 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) in place, 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," dated January 1996 (Ref. 12), 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 WOAP. Response time verification for other sensor types must be demonstrated by test.WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 13) provides the basis and methodology for 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 the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly Of a transmitter.
1E Y--mfts=n I nse'rt 2 8-of-hfi t p Cook Nuclear Plant Unit I1 ..-1ReiinN.5 B 3.3.2-41 Revision No. 51 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) eha~tq4ra.cute ine cee -e-ta-4es i e with-e~aeh-
* hre -Tgrele~-eut4-~eie-vrfet a This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours after reaching 850 psig in the SGs.REFERENCES
: 1. Technical Requirements Manual.2. IEEE-279, "Proposed Criteria for Nuclear Power Plant Protection Systems," August 1968.3. UFSAR, Table 7.2-1.4. UFSAR, Table 14.1-2.5. 10 CFR 50.49.6. WCAP-12741,"Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.7. UFSAR, Chapter 14.8. WCAP-14333-P-A, Revision 1, October 1998.9. WCAP-1 0271-P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," including Supplement 1, May 1986, and Supplement 2, Rev. 1, June 1990.10. WCAP-1 5376, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Intervals and Reactor Trip Breaker Test and Completion Times," October 2000.11. UFSAR, Table 7.2-7.12. WCAP-1 3632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996.13. WCAP-1 4036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.Cook Nuclear Plant Unit 1 B 3.3.2-42 Revision No. 51 Cook Nuclear Plant Unit 1 B 3.3.2-42 Revision No. 51 PAM Instrumentation B 3.3.3 BASES ACTIONS (continued) justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.SURVEILLANCE As noted at the beginning of the SRs, the following SRs apply to each REQUIREMENTS PAM instrumentation Function in Table 3.3.3-1, except where identified in the SR.SR 3.3.3.1 Performance of the CHANN EL CHECKi ree-eveiy4-.ays=ensures that a gross instrumentation failure 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
The Containment Area Radiation (High Range)instrumentation should be compared to similar unit instruments located throughout the unit. When only one channel of the Reactor Coolant Inventory Tracking System is OPERABLE, the RCS Subcooling Margin Monitor and Core Exit Temperature channels may be used for performance of the CHANNEL CHECK of the OPERABLE Reactor Coolant Inventory Tracking System channel.Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, 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. If the channels are within the criteria, it is an indication that the channels are OPERABLE.As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.
T ~ lfc ~
demat~fatse atcHrie ef~lGH 4-ANN--TeLS-- K Insert 2 fantu~g-c-ha~t-Cook Nuclear Plant Unit 1 ..-3Rvso o B 3.3.3-13 Revision No. 0 PAM Instrumentation B 3.3.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.3.2 Deleted SR 3.3.3.3-A4--~IN -e~v B-24aeRt CHAN NEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy.
This SR is modified by a Note that excludes neutron detectors.
For Function 9, the*CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual valve position.
For Functions 15, 16, 17, and 18, whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit Temperature thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing elements.For Functions 20 (Circuit Breaker Status channels) and 24, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual circuit breaker position. Insert 2 REFERENCES
: 1. NRC letter, T. G. Colburn (NRC) to M. P. Alexich (Indiana Michigan-Power Company), "Emergency Response Capability
-Conformance to Regulatory Guide 1.97 Revision 3 for the Dl C. Cook Nuclear Plant, Units 1 and 2," dated December 14, 1990.2. UFSAR, Table 7.8-1.3. Regulatory Guide 1.97, Revision 3, May 1983.4. NUREG-0737, Supplement 1, "TMI Action Items." 5. NRC letter, P. S. Tam (NRC), to M. K. Nazar, (Indiana Michigan Power Company), "Donald C. Cook Nuclear Plant, Units 1 & 2 (DCCNP-1 AND DCCNP-2) -Issuance of Amendments Re: Containment Sump Modifications per Generic Letter 2004-02 (TAC Nos. MD5901 AND MD5902)," dated October 18, 2007.Cook Nuclear Plant Unit 1 ..-4ReiinN.2 B 3.3.3-14 Revision No. 29 Remote Shutdown Monitoring Instrumentation B 3.3.4 BASES ACTIONS (continued)
Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.A.1 Condition A addresses the situation where one or more required Functions of the remote shutdown monitoring instrumentation are inoperable.
The Required Action is to restore the required Function to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.B.1 and B.2 If the Required Action and associated Completion Time of Condition A is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.:3.4.1 REQUI REM ENTS Performance of the CHANN EL CHECK~eaeeeve y.4-dey~s~ensures that a gross failure of instrumentation has not occurred.
A CHANNEL CHECK is normally a comparison of the parameter indicated onone 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. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability.
If the channels are within the criteria, it is an indication that the channels are OPERABLE.
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.Cook Nuclear Plant Unit 1 B 3.3.4-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.3.4-3 Revision No. 0 Remote Shutdown Monitoring Instrumentation B 3.3.4 BASES SURVEILLANCE REQUIREMENTS (continued)
-As specified in the Surveillance, a CHANNEL CHECK is only required for those channels which are normally energized.
TheFqey=e~-8+/-~y nIen I=-nsert 2-HeN SR 3.3.4.2 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.For the Reactor Trip Breaker Indication Function on the hot shutdown panel, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual reactor trip breaker position.I li; ul L/-'t I In 31 Ill =is-erfe19-.~=Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.3.Cook Nuclear Plant Unit 1 ..- evso o B 3.3.4-4 Revision No. 0 LOP DG Start Instrumentation B 3.3.5 BASES ACTIONS (continued) made inoperable by failure of the LOP DG start instrumentation are required to be entered immediately.
The actions of those LCOs provide for adequate compensatory actions to assure unit safety.SURVEILLANCE REQUIREMENTS SR 3.3.5.1 Performance of the CHANNEL CHECK e~ee-ever-y=-l-2--heur13 ensures that a 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.
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.SR 3.3.5.2 I===nsert 2 SR 3.3.5.2 is the performance of a TADOT. 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.
"-Thi-hes-i 9efFe The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment.
The SRs are modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and are verified during CHANNEL CALIBRATION.
T1,t~ienc--y-is--Insert 2-aece-aele-. Cook Nuclear Plant Unit 1 B 3.3.5-5 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.3.5-5 Revision No. 0 LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.5.3 SR 3.3.5.3 is the performance of a CHANNEL CALIBRATION.
The setpoints, as well as the response to a loss of voltage and a degraded voltage test, shall include a single point verification that the trip occurs within the required time delay.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., .-.- r o, .. , r * .l .r. ..i ...i .-Insert 2 REFERENCES
: 1. UFSAR, Section 8.4.2. UFSAR, Section 8.5.3. UFSAR, Chapter 14.4. WCAP-1 2741, "Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.Cook Nuclear Plant Unit 1 B3356Rvso o B 3.3.5-6 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES ACTIONS (continued)
D..1 Condition D applies to all Containment Purge Supply and Exhaust System Isolation Functions.
If one or more Automatic Actuation Logic and Actuation Relays trains are inoperable, one or more SI Input from ESFAS trains are inoperable, two or more required radiation monitoring channels in a single train are inoperable, or the Required Action and associated Completion Time of Condition A, B, or C are not met, operation may continue provided the containment purge supply and exhaust isolation valves are placed in the closed position immediately.
Placing the containment purge supply and exhaust isolation valves in the closed position accomplishes the safety function of the inoperable trains or channels.SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1 REQUIREMENTS determines which SRs apply to which Containment Purge Supply and Exhaust System Isolation Instrumentation Functions.
SR 3.3.6.1 Performance of the CHANN EL CH ECi Keeeever-y--t-2--heur-s ensures that a 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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.
etaaa I=-nsert 2
* Cook Nuclear Plant Unit 1 B 3.3.6-6 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.3.6-6 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.6.2 SR 3.3.6.2 is the performance of an ACTUATION LOGIC TEST. The train being tested may be placed in the bypass condition, thus preventing actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, may be tested for each protection function.
In addition, the master relay coil may be pulse tested for continuity.
This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. -This
-'-Insert 2 SR 3.3.6.3 SR 3.3.6.3 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity.
Tise--ht~et~5
~ -Isr SR 3.3.6.4 A COT is performed on each required channel to ensure the entire 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 COT 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.
Tl-I"e-l-eqlencjs-15..
Insert 2 bitbaset This test verifies the capability of the instrumentation to provide the Containment Purge Supply and Exhaust System isolation.
The setpoint shall be left consistent with the current unit specific calibration procedure tolerance.
SR 3.3.6.5 SR 3.3.6.5 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation mode is either allowed to function or is placed in a condition where the relay contact operation can be verified without Cook Nuclear Plant Unit 1 B3367Rvso o B 3.3.6-7 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued) operation of the equipment.
Actuation equipment that may not be operated in the design mitigation mode is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.T-he--Fr-eq
'=14aeeepta4b1e Insert 2 SR 3.3.6.6 SR 3.3.6.6 is the performance of a TADOT. This test is a check of the Manual Initiation Function ipeerfer ea--evey4--mefat-h.
Each Manual Initiation Function is tested up to, and including, the master relay coils. 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.
In some instances, the test includes actuation of the end device (i.e., valves cycle).The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no setpoints associated with it.ep 4-Insert 2 SR 3.3.6.7 3-&deg;~~ 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.T 2 REFERENCES
: 1. UFSAR, Section 5.5.3.2. 10OCFR 100.11 3. WCAP-1 5376, Rev. 0, October 2000.Cook Nuclear Plant Unit 1I ..- evso o B 3.3.6-8 Revision No. 0 CREV System Actuation Instrumentation B 3.3.7 BASES ACTIONS (continued) this Completion Time is the same as provided in LCO 3.7.10.B.1.1, B.1.2, and B.2 Condition B applies to the failure of two CREV System Automatic Actuation Logic and Actuation Relays trains in one or more required Functions.
The first Required Action is to place one CREV train in the pressurization/cleanup mode of operation immediately.
This accomplishes the actuation instrumentation Function that may have been lost and places the unit in a conservative mode of operation.
The applicable Conditions and Required Actions of LCO 3.7.10 must also be entered for the CREV train made inoperable by the inoperable actuation instrumentation.
This ensures appropriate limits are placed upon train inoperability as discussed in the Bases for LCO 3.7.10.Alternatively, both trains may be placed in the pressurization/cleanup mode. This ensures the CREV System function is performed even in the presence of a single failure.'C.1 and C.2 Condition C applies when the Required Action and associated Completion Time for Condition A or B have not been met. The unit must be brought to a MODE in which the LCO requirements are not applicable.
To achieve this status, the unit must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS A Note has been added to the SR Table to clarify that Table 3.3.7-1 determines which SRs apply to which CREV System Actuation Instrumentation Functions.
SR 3.3.7.1 SR 3.3.7.1 is the performance of an ACTUATION LOGIC TEST. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the mater relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. -F-is-Insert 2 Cook Nuclear Plant Unit 1 B 3.3.7-3 Revision No. 0 Cook Nuclear Plant Unit I B3.3.7-3 Revision No. 0 CREV System Actuation Instrumentation B 3.3.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.7.2 SR 3.3.7.2 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. Insert 2 SR 3.3.7.3 SR 3.3.7.3 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is-either allowed to function or is placed in a condition where the relay contact operation can be verified without operation of the equipment.
Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.
Insert 2 REFERENCES
: 1. WCAP-1 5376, Rev. 0, October 2000.Cook Nuclear Plant Unit 1 B3374Rvso o B3.3.7-4 Revision No. 0 BOMI B 3.3.8 BASES ACTIONS (continued)
As an alternate to restoring one channel to OPERABLE status within 1 hour (Required Action B.2.1). Required Action B.2.2.1 requires isolation valves for unborated water sources to the Chemical and Volume Control System to be secured to prevent the flow of unborated water into the RCS. In addition, in MODE 5, if the RWST boron concentration is< 2400 ppm and less than the Reactor Coolant System (RCS) boron concentration, the RWST is considered an unborated water source and is required to be isolated from the RCS. Once it is recognized that two source range neutron flux monitoring channels of the BDMI are inoperable, the operators will be aware of the possibility of a boron dilution, and the 1 hour Completion Time is adequate to complete the requirements of Required Action B.2.2. 1.Required Action B.2.2.2 accompanies Required Action B.2.2.1 to verify the SDM according to SR 3.1.1.1 within 1 hour and once per 12 hours thereafter.
This backup action is intended to confirm that no unintended boron dilution has occurred while the BDMI was inoperable, and that the required SDM has been maintained.
The specified Completion Time takes into consideration sufficient time for the initial determination of SDM and other information available in the control room related to SDM.SURVEILLANCE SR 3.3.8.1 REQUIREMENTS Performance of the CHAN NEL CHECK 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 channelsrmonitoring 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.
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.
te e~{ .= Insert 2 eed ~ge~tee~~~ts ihe~lsptays~st~iedw it'ht"he~fl_-redt~-eb'4anies.
Cook Nuclear Plant Unit 1 ..- evso o B 3.3.8-3 Revision No. 0 BDMI B 3.3.8 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.8.2 SR 3.3.8.2 is the performance of a CHANNEL CALIBRATION eve~y-.-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.
The CHANNEL CALIBRATION also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This SR is modified by a Note that states that neutron detectors are excluded from the CHANNEL CALIBRATION.-I nsert 2 REFERENCES
: 1. UFSAR, Section 14.1.5.Cook Nuclear Plant Unit 1 ..- evso o B 3.3.8-4 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued)
POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive.
Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.
ACTIONS A.1 With one or more of the RCS DNB parameters not within LCO limits, action must be taken to restore parameter(s) in order to restore DNB margin and eliminate the potential for violation of the accident analysis.The 2 hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.
B.1 If Required Action A.1 is not met within the associated Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 2 within 6 hours. In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis.
The Completion Time of 6 hours is reasonable to reach the required unit conditions in an orderly manner.SURVEILLANCE REQU IREM ENTS SR 3.4.1.1 r.eeeu 4& e
.--la 2 SR 3.4.1.2--
et4e8
.-Insert 2--F...........
j j Cook Nuclear Plant Unit 1 B3413Rvso o B 3.4.1-3 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.1.3 Verification that the RCS total flow rate is greater than or equal to the limits ensures that the initial condition of the safety analyses are met.
r 4t44 ~ ~ rr gM SR 3.4.1.4 Measurement of RCS total flow rate by performance of a precision calorimetric heat balance ei~c-e-ever~y--4-mer~itC~ialows the installed RCS flow instrumentation to be calibrated and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.~=Insert 2 eqmt9 o fwer4 2 This SR is modified by a Note that allows entry into MODE 1, without having performed the SR, and placement of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 24 hours after > 90% RTP. This exception is appropriate since the heat balance requires the unit to be at a minimum of 90% RTP to obtain the stated RCS flow accuracies.
The Surveillance shall be performed within 24 hours after reaching 90% RTP.REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 1 B3414Rvso o B 3.4.1-4 Revision No. 0 ROS Minimum Temperature for Criticality B 3.4.2 BASES APPLI CABLE SAFETY ANALYSES (continued) criticality limitation provides a small band, 6&deg;F, for critical operation below HZP. This band allows critical operation below HZP during unit startup and does not adversely affect any safety analyses since the MTC is not significantly affected by the small temperature difference between HZP and the minimum temperature for criticality.
The RCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO Compliance with the LCO ensures that the reactor will not be made or maintained critical (keff > 1.0) at a temperature less than a small band below the HZP temperature, which is assumed in the safety analysis.Failure to meet the requirements of this LCO may produce initial conditions inconsistent with the initial conditions assumed in the safety analysis.APPLICABILITY In MODE 1 and MODE 2 with keff -> 1.0, LCO 3.4.2 is applicable since the reactor can only be critical (kerr > 1.0) in these MODES.ACTIONS A.._If the parameters that are outside the limit cannot be restored, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to MODE 2 with keff < 1.0 within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period. The allowed time is reasonable, based on operating experience, to reach MODE 2 with keff < 1.0 in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.4.2.1 RCS loop average temperature is required to be verified at or above 5410&deg; F ur=4eepra-re evrt2~ ~-=-Insert2
~ee19e4eteI el9ee-pef 494~e9~eF=r.eefl~-e~&ie ltAibWfI1UII drIi 4 ypibdIIy ptirfOflTlett'
~freftienet--~i+iea1ity-is-approaohed.
REFERENCES
: 1. UFSAR, Section 14.1.1.Cook Nuclear Plant Unit 1 ..- evso o B 3.4.2-2 Revision No. 0 RCS P/T Limits B 3.4.3 BASES ACTIONS (continued)
Condition C is modified by a Note requiring Required Action 0.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE REQUIREMENTS SR 3.4.3.1 Verification that operation is within limits is required when RCS pressure and temperature conditions are undergoing planned.changes. ThsF-e --
t-te-t ee-iease~n~e--re~afrmre~ve~n ==Insert 2 Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.
This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.
REFERENCES
: 1. WCAP-1 5878, Rev. 0, dated December 2002.2. 10 CFR 50, Appendix G.3. ASME, Boiler and Pressure Vessel Code, Section Ill, Appendix G.4. ASTM E 185-82, July 1982.5. 10 CFR 50, Appendix H.6. Regulatory Guide 1.99, Revision 2, May 1988.7. ASME, Boiler and Pressure Vessel Code, Section XI, Appendix E.Cook Nuclear Plant Unit 1 ..- evso o B 3.4.3-6 Revision No. 0 RCS Loops -MODES 1 and 2 B 3.4.4 BASES APPLICABILITY (continued)
Operation in other MODES is covered by: LCO 3.4.5, "RCS Loops -MODE 3";LCO 3.4.6, "RCS Loops -MODE 4";LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled";LCO 3.4.8, "RCS Loops -MODE 5, Loops Not Filled";LCO 3.9.4, 'Residual Heat Removal (RHR) and Coolant Circulation
-High Water Level"; and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation
-Low Water Level." ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the unit to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.4.4.1 REQUIREMENTS This SR requires verification e ert42=heir-s hat each RCS loop is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to the DNBR limit. tl tqieney=of-d  44- e-t~-~
ea REFERENCES
: 1. UFSAR, Section 14.1.4===Insert 2 Cook Nuclear Plant Unit 1 B3443Rvso o B3.4.4-3 Revision No. 0 RCS Loops -MODE 3 B 3.4.5 BASES ACTIONS (continued) 0.._1 If one required RCS loop is not in operation, and the Rod Control System is capable of rod withdrawal, the Required Action is to place the Rod Control System in a condition incapable of rod withdrawal (e.g., de-energize all CRDMs by opening the RTBs or de-energizing the motor generator (MG) sets). When the Rod Control System is capable of rod withdrawal, it is postulated that a power excursion could occur in the event of an inadvertent control rod bank withdrawal.
This mandates having the heat transfer capacity of two RCS loops in operation.
If only one loop is in operation, the Rod Control System must be rendered incapable of rod withdrawal.
The Completion Time of 1 hour to defeat the Rod Control System is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period.D.1, D.2, and 0.3 If two required RCS loops are inoperable, or two required RCS loops are not in operation with Rod Control System capable of rod withdrawal, or required RCS loop not in operation with Rod Control System not capable of rod withdrawal, the Rod Control System must be placed in a condition incapable of rod withdrawal (e.g., all CRDMs must be de-energized by opening the RTBs or de-energizing the MG sets). All operations involving introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 must be suspended, and action to restore one of the RCS loops to OPERABLE status and operation must be initiated.
Boron dilution requires forced circulation for proper mixing, and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal.
Suspending operations that would cause the introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 is required to assure c~ontinued safe operation.
With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification eeiy2-h~euis that the required loops are in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.Cook Nuclear Plant Unit 1 ..- evso o B3.4.5-4.Revision No. 0 RCS Loops -MODE 3 B 3.4.5 BASES SURVEILLANCE REQU IREMENTS (continued)
I II~ II.9LJL.11.Jy
~JI It. IIt.JLU.d I.S ~L4III'.JIS.~.I IL *J~JI IJI'~Insert 2 SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side water level is above the lower tap of the SG wide range level instrumentation by >- 420 inches for required RCS loops. If the SG tubes become uncovered, the associated loop may not be capable of providing the heat sink for removal Of the decay heat. The water level can be verified by either the wide range or the narrow range instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. Th--1e-4 tw=Farteqeiiee--i 2 ethr4~a~a4 iees=ef=G&Ieve
.=SIR 3.4.5.3 Verification that each required RCP is OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional ROP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power availability to each required RCP."*===Insert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 1 B3455Rvso o B 3.4.5-5 Revision No. 0 ROS Loops -MODE 4 B 3.4.6 BASES ACTIONS (continued) minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SR 3.4.6.1 This SR requires verification that the. required RCS or RHR loop is in operation and circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. T-he-F teney-of-SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side water level is above the lower tap of the SG wide range level instrumentation by > 420 inches. If the SG U-tubes become uncovered, the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.The water level can be verified by either the wide range or the narrow range level instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met.
w-e~ ef-tber~=Insert 2 SR 3.4.6.3 Verification that each required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump. -FeF~~lee~-7dy~-eaitretreenllInsert 2--==Insert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 1 ..- evso o B 3.4.6-4 Revision No. 0 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES ACTIONS (continued) 0.1 and C.2 If a required RHR loop is not in operation or if no required loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated.
Suspending operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 is required to assure continued safe operation.
With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Times reflect the importance of maintaining operation for heat removal.SURVEILLANCE REQUIREMENTS SR 3.4.7.1 This SR requires verification ever that the required loop is in operation circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
I*-nsert 2 neter-R4pete1le9r-SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side water levels are above the lower tap of the SG wide range level instrumentation by > 420 inches ensures an alternate decay heat removal method via natural circulation in the event that the second RHR loop is not OPERABLE.
The water level can be verified by either the wide range or the narrow range instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. If both RHR loops are OPERABLE, this Surveillance is not needed. ~The ---Insert 2 SR 3.4.7.3 Verification that each required RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by Cook Nuclear Plant Unit I B 3.4.7-4 Revision No. 0 Cook Nuclear Plant Unit 1 B3.4.7-4 Revision No. 0 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued) verifying proper breaker alignment and power available to each required RHR pump. If secondary side water level is above the lower tap of the SG wide range level instrumentation by > 420 inches in at least two SGs.this Surveillance is not needed. T-he-tre lee-f-7--=eye-s--ier ,---Insert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES
: 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation." Cook Nuclear Plant Unit 1 B3475Rvso o B 3.4.7-5 Revision No. 0 RCS Loops -MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE REQUIREMENTS SR 3.4.8.1 This SR requires verification that the required loop is in operation circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. -T-he-Fr-eqtuen~e{-h ie-eu-ie1e4-"
I-nsert 2 SR 3.4.8.2 Verification that each required pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump. T~h fqe -f7d~zi oedeieer-reefteble-ia=vew-efo Insert 2
~
This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 1 B3483Rvso o B 3.4.8-3 Revision No. 0 Pressurizer B 3.4.9 BASES SURVEILLANCE REQUIREMENTS SR 3.4.9.1 This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level. T~e ece rse ~
p acev ao-il far 2-t~eeam er-4et eeveac-te.
SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated pressurizer backup heaters are verified to be at their specified capacity.
This may be done by testing the power supply output with the heaters energized. -The Fr ets e utt 2 anc na-z r-Dc -cne REFERENCES
: 1. UFSAR, Chapter 14.2. NUREG-0737, November 1980.Cook Nuclear Plant Unit 1 B3494Rvso o B 3.4.9-4 Revision No. 0 Pressurizer PORVs B 3.4.11 BASES ACTIONS (continued) place the PORV(s) in manual control, this may not be possible for all causes of Condition B entry with PORV(s) inoperable and not capable of being manually cycled (e.g., as a result of failed control power fuse(s) or control switch malfunctions(s))
H.1 and H.2 If any Required Action and associated Completion Time of Condition A, B, C, D, F, F, or G is not met, if three PORVs are inoperable and not capable of being manually cycled, if two PORVs are inoperable and not capable of being manually cycled and one block valve inoperable (for reasons other than to comply with Required Action B.2) in a different line than the inoperable PORVs, or if one PORV is inoperable and not capable of being manually cycled and two block valves are inoperable (for reasons other than to comply with Required Action B.2) in different lines than the inoperable PORV, then the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.4.11.1 REQ U IREM ENTS Block valve cycling verifies that the valve(s) can be opened and closed if needed. -gMCee Insert 2 This SR is modified by a Note, which states that this SR is not required to be performed with the block valve closed in accordance with the Required Actions of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable.
SR 3.4.11.2 SR 3.4.11 .2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR.
c - l~r-met-e
~ ef~ er-fe~mt e-FeqInsert-2<p-~-f- m it ~-'-~l Cook Nuclear Plant Unit 1B341-6RvsoNo0 B 3.4.11-6 Revision No. 0 Pressurizer PORVs B 3.4.11 BASES SURVEILLANCE REQUIREMENTS (continued)
The Note modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. In accordance with Reference 4, administrative controls require this test be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation.
SR 3.4.11.3 Operating the solenoid air control valve associated with each PORV, and the check valves on the air accumulators where applicable, ensures the PORV control system actuates properly when called upon..@9per-ati9g,.exe~a--e-'
r ee=d ot t--rhc-o 2re1 re he~=Insert 2 REFERENCES
: 1. Regulatory Guide 1.32, February 1977.2. UFSAR, Section 14.1.8.3. ASME, Operation and Maintenance Standards and Guides (OM Codes).4. Generic Letter 90-.06, "Resolution of Generic Issue 70,'Power-Operated Relief Valve and Block Valve Reliability,'
and Generic Issue 94, 'Additional Low-Temperature Overpressure for Light-Water Reactors,'
Pursuant to 10 CFR 50.54(f)," June 25, 1990.Cook Nuclear Plant Unit 1B34117RvsoN.0 B 3.4.1 1-7 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued) through the pump control switch being placed in pull to lock and at least one valve in the discharge fl~w path being closed, or at least one valve in the discharge flow path being closed and sealed or locked.In addition, SR 3.4.12.3 is modified by a Note that allows the accumulator discharge isolation valve position to be verified by administrative means.This is acceptable since the valve positi~on was verified prior to deactivating the valve, access to the containment is restricted, and valves are only operated under strict procedural control.-aa-sa-ial-~#-pttH~eenTPom -Insert 2 ettu~{&#xa2; ~ ~ -{SR 3.4.12.4 The required RHR suction relief valve shall be demonstrated OPERABLE by verifying the RHR suction isolation valves are open. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.The RHR suction isolation valves are verified to be opened evefye-tr 2 SR 3.4.12.5 The RCS vent of> 2.0 square inches or a blocked open PORV is proven OPERABLE by verifying its open condition ~--Insert 2
~e19 set~I~b'e-ef=epei9~,vay=e1e&its~Thie=eate&sect;er=y+/-
The passive vent path arrangement must only be open if the vent is being used to satisfy the pressure relief requirements of LCOG 3.4;12.A.2.c.
Cook Nuclear Plant Unit 1 B 3.4.12-11 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.4.12-11 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.12.6 The PORV block valve must be verified open e-v-eJ---_-w.us to provide the flow path for each required PORV to perform its function when actuated.
The valve must be remotely verified open in the main control room. This Surveillance is performed if one or more PORVs satisfy the LCO.The block valve is a remotely controlled, motor operated valve. The power to the valve operator-is not required removed, and the manual operator is not required locked in the inactive position.
Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.f& I nsernt 2 SR 3.4.12.7 Verification that each required emergency air tank bank's pressure is > 900 psig assures adequate air pressure for reliable PORV operation.
With the emergency air supply at -> 900 psig, there will be enough air to support PORV operation for 10 minutes with no operator action upon a loss of control air. T~e-3=ty 4-- Insert SR 3.4.12.8 Performance of a COT is required on each required PORV to verify and, as necessary, adjust its lift setpoint.
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 COT 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 COT will verify the setpoint is within the LCO limit. PORV actuation could depressurize the RCS and is not required.Cook Nuclear Plant Unit 1 B341-2Rvso o B 3.4.12-12 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)
A Note has been added indicating that this SR is not required to be performed until 12 hours after decreasing ROS cold leg temperature to< 2660&deg;F. The COT cannot be performed until in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP setting. The test must be performed within 12 hours after entering the LTOP MODES. 4he- Insert 2-SR 3.4.12.9 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required-vePy -methe~to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input. 4-=.Insert 2 REFERENCES
: 1. 10 CFR 50, Appendix G.2. Generic Letter 88-11.3. ASME, Boiler and Pressure Vessel Code, Section III.4. WCAP-1 3235, "Donald C. Cook Units 1 & 2, Analysis of Low Temperature Overpressurization Mass Injection Events with Pressurizer Steam Bubble and RHR Relief Valve, March 1992;"WCAP-1 2483 Revision 1, "Analysis of Capsule U From the American Electric Power Company D. C. Cook Unit I Reactor Vessel Radiation Surveillance Program, December 2002;" and WCAP-13515, Revision 1, "Analysis of Capsule U From Indiana Michigan Power Company D. C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program, May 2002." 5. 10 CFR 50, Section 50.46.6. 10 CFR 50, Appendix K.7. Generic Letter 90-06.Cook Nuclear Plant Unit 1 B341-3Rvso o B 3.4.12-13 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within thie LCO limits ensures the integrity of the RCPB is maintained.
Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection.
It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. Unidentified LEAKAGE and identified LEAKAGE are determined by performance of an ROS water inventory balance.The ROS water inventory balance must be performed with the reactor at steady state operating conditions.
The Surveillance is modified by two Notes. Note 1 states that this SR is not required to be performed until 12 hours after establishing steady state operation.
The 12 hour allowance provides sufficient time to collect and process all necessary data after stable unit conditions are established.
Steady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LOG 3.4.15, "RCS Leakage Detection Instrumentation." Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an ROS water inventory balance.T'N 2 SIR 3.4.13.2 This SIR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated.
The primary to secondary LEAKAGE is Cook Nuclear Plant Unit 1 ..35ReiinN.1 B 3.4.13-5 Revision No. 15 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued) measured at room temperature as described in Reference
: 7. Prior to comparison with the 150 gallons per day TS limit, the measured primary to secondary LEAKAGE is multiplied by a volume correction factor of 1.52. The correction factor ensures the offsite dose analyses, which assume primary to secondary leakage is at normal operating temperature and pressure, remain bounding.
The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all of the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours after establishment of steady state operation.
For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.
The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 7).REFERENCES
.1. UFSAR, Section 1.4.3.2. Regulatory Guide 1.45, May 1973.3. UFSAR, Section 14.2.4.4. Letter from Indiana Michigan Power Company (M. W. Rencheck) to the NRC dated October 26, 2000 (Letter C1000-20).
: 5. Letter from NRC (John F. Stang) to Indiana Michigan Power Company (Robert P. Powers), dated November 8, 2000.6. NEI 97-06, 'Steam Generator Program Guidelines." 7. EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines." Cook Nuclear Plant Unit 1 ..36ReiinN.1 B 3.4.13-6 Revision No. 15 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued) potential for an unplanned transient if the Surveillance were performed with the reactor at power.The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.
Therefore, this SR is modified by a Note that states the Surveillance is only required to be performed in MODES 1 and 2. Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance.
SR 3.4.14.2 Verifying that the RHR interlock that prevents the valves from being opened is OPERABLE ensures that RCS pressure will not pressurize the RHR System beyond its design pressure of 600 psig. T-he-A-&deg;-mei&#xb6;
~tIset eet - ecc nditic --aitgeTs mr ncyi REFERENCES-
: 1. 10CFR50.2.
: 2. 10 CFR 50.55a(c).
: 3. WASH-i1400 (NUREG-75/01 4), Appendix V, October 1975.4. Letter from D.G. Eisenhut, NRC, to all LWR licensees, LWR Primary.Coolant System Pressure Isolation Valves, February 23, 1980.5. Letter from S.A.. Varga, NRC, to J. Dolan, Order for Modification of Licenses Concerning Primary Coolant System Pressure Isolation Valves, April 20, 1981.6. Technical Requirements Manual.7. EGG-NTAP-61 75, Inservice Testing of Primary Pressure Isolation Valves, Idaho National Engineering Laboratory, February 1983.8. NRC Safety Evaluation for License Amendment 188.9. ASME, Operation and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 1B341-5RvsoN.0 B 3.4.14-5 Revision No. 0 RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued) atmosphere must be taken to provide alternate periodic information.
The 12-hour interval is sufficient to detect increasing RCS leakage. The Required Action provides 7 days to restore another RCS leakage monitor to OPERABLE status to regain the intended leakage detection diversity.
The 7 day Completion Time ensures that the plant will not be operated in a degraded condition for a lengthy time period.E.1 and E.2 With the containment atmosphere particulate radioactivity monitor and the required containment humidity or containment atmosphere gaseous radioactivity monitor inoperable, the only means of detecting leakage is the containment sump monitor. This Condition does not provide the required diverse means of leakage detection.
The Required Action is to restore either of the inoperable required monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity.
The 30 day Completion Time ensures that the unit will not be operated in a reduced configuration for a lengthy time period.F.1 and F.2 If any Required Action and associated Completion Time of Condition A, B, C, D, or E cannot be met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.G.1 With all three types of required monitors inoperable (i.e., LCO 3.4.15.a, b, and c not met), no automatic means of monitoring leakage are available, and immediate unit shutdown in accordance with LCO 3.0.3 is required.SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is operating properly.
The---
fe I nsert 2 Cook Nuclear Plant Unit 1 B 3.4.15-6 Revision No. 33 Cook Nuclear Plant Unit 1 B 3.4.15-6 Revision No. 33 RCS Leakage Detection Instrumentation B 3.4.15 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. 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 test verifies the alarm se~tpoint and relative accuracy of the instrument string. si~ t-<=. Insert 2-f SR 3.4.15.3, SR 3.4.15.4, and SR 3.4.15.5 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels.
The.calibration verifies the accuracy of the instrument string, including the instruments located inside containment. ecee-f2-afta Insert 2
", "" t' erie+/-~
REFERENCES
: 1. UFSAR, Section 1.4.3.2. Regulatory Guide 1.45, Rev. 0, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973.I 3. AEP Letter to NRC, AEP:NRC:0137D, "NRC Generic Letter 84-04;Elimination Of Postulated Pipe Breaks In Primary Main Loops Generic Issue A-2, Asymmetric Blowdown Loads On PWR Primary Systems Request. For License Condition Deletion," dated September 10, 1984.4. NRC Letter to AEP, "Generic Letter 84-04, Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," dated November 22, 1985.5. UFSAR, Section 4.2.7 6. WCAP-15435, Rev. 1, Technical Justification for Eliminating Pressurizer Surge Line Rupture as the Structural Design Basis for D.C. Cook Units 1 and 2 Nuclear Power Plant, August 2000.Cook Nuclear Plant Unit 1 ..57ReiinN.3 B 3.4.15-7 Revision No. 33 RCS Specific Activity B 3.4.16 BASES ACTIONS. (continued)
B.1 If any Required Action and associated Completion Time of Condition A is not met, if the DOSE EQUIVALENT I-131 is in the unacceptable region of Figure 3.4.16-1, or if gross specific activity of the reactor coolant is not within limit, the reactor must be brought to MODE 3 with RCS average temperature
< 5000&deg;F within 6 hours. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 below 500&deg;F from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.4.16.1 SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at--eeet--aee-evefy-While basically a quantitative measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in gross specific activity.Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions.
T-he -F
<- Insert 2 SR 3.4.16.2 This Surveillance requires the verification that the reactor coolant DOSE EQUIVALENT 1-131 specific activity is within limit. This Surveillance is accomplished by performing an isotopic analysis of a reactor coolant sample. This Surveillance is performed in MODE I only to ensure iodine remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur. ,T-he-44ey=Fr~eqteney~is-ig ae~i atfe The Frequency, between 2 and 6 hours after a power change > 15% RTP within a 1 hour period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.SR 3.4.16.3 A radiochemical analysis for determination is required evei--484-daye-with the unit operating in MODE 1 equilibrium conditions.
The determination directly relates to the LCO and is required to verify unit~=Insert 2 Cook Nuclear Plant Unit I B 3.4.16-4 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.4.16-4 Revision No. 0 RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued) operation within the specified gross activity LCO limit. The analysis for IF is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. T]--e-F-eele Insert 2~ef,4ae This SR has been modified by a Note that indicates sampling is not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours. This ensures that the radioactive materials are at equilibrium so the analysis for F is representative and not skewed by a crud burst or other similar abnormal event.REFERENCES
: 1. 100CFRI100.11.
: 2. UFSAR, Section 14.2.4.Cook Nuclear Plant unit 1 ..65Rvso o B 3.4.16-5 Revision No. 0 Accumulators B 3.5.1 BASES ACTIONS (continued)
D.__I If more than one accumulator is inoperable, the unit is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE REQUIREMENTS SR 3.5.1.1 Each accumulator isolation valve should be verified to be fully open everd 4=2aeus. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions.
4I=F-r-eqie1ey-ie-eber-ieid..
e ree~ -thr- ~ ~ rr4~~.e~~=
--Insert 2 SR 3.5.1.2 and SR 3.5.1.3-4.e~l-e&#xa3;je, borated water volume and nitrogen cover pressure are verified for each accumulator. eirte ,,=-Insert 2=8-L .Fi
@,s4 ejq ..eT44-4 e.
u r-F-r-eq e=a f=e SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulator~e since the static design of the accumulators limits the ways in which the concentration can be changed.Thy-I==nsert 2 1-.h" .,~
Sampling the affected accumulator within 6 hours after a volume increase of 13 ft 3 will identify whether inleakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), because the water contained in the RWST is within the accumulator boron concentration requirements.
This is consistent with the recommendation of NUREG-1366 (Ref. 4).Cook Nuclear Plant Unit 1 ..- evso o B3.5.1-6 Revision No. 0 Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.5 Verification power is removed from each accumulator isolation valve operator when the RCS pressure is > 2000 psig ensures that an active failure could not result in the closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. -Sieeteweir-ie-r~eiae eaei-a~I14s4ei4-tRve
<= Insert 2 This SR allows power to be supplied to the motor operated isolation valves when ROS pressure is < 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns.
REFERENCES
: 1. UFSAR, Section 14.3.2. 10 CFR 50.46.3. WCAP-1 5049-A, "Risk-Informed Evaluation of an Extension to Accumulator Completion Times," Rev. 1, April 1999.4. NUREG-1 366, February 1990.Cook Nuclear Plant Unit 1 ..- evso o B3.5.1-7.Revision No. 0 EGOS -Operating B 3.5.2 BASES ACTIONS (continued) 0.1_Condition A is applicable with one or more ECOS trains inoperable.
The allowed Completion Time of Required Action A.1 is based on the assumption that at least 100% of the EGGS flow equivalent to a single OPERABLE ECGS train is available.
An inoperable RHR or SI pump concurrent with a closed cross-tie valve in the affected system will result in less than 100% of the EGGS flow equivalent to a single OPERABLE EGGS train because there will be flow to only two RGS loops. With less than 100% of the EGOS flow equivalent to a single OPERABLE EGGS train available, the facility is in a condition outside of the accident analyses.
Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE REQUIREMENTS SR 3.5.2.1 Verification of proper valve position ensures that the flow path from the EGOS pumps to the RCS is maintained.
Misalignment of these valves could render both EGGS trains inoperable.
Securing these valves in position by locking out control power ensures that they cannot change position as a result of an active failure or be inadvertently misaligned.
These valves are of the type, described in Reference 9, that can disable the function of both EGGS trains and invalidate the accident analyses.
A--<==Insert 2 I m Ill Ifl ................................
r................
]SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the EGGS flow paths provides assurance that the proper flow paths will exist for EGGS operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation.
Rather, it involves verification that those valves capable of being mispositioned are in the correct position. yi 1ite-<te4h U LAAU ~..UJ ~ ~JtJt....
L4L~.,%.A LII *~ALI LILA 11111 *I~~flI LUI.IU L' LIL'* ~I.* LI*, LII*~ LU.r' LUW~*-Insert 2wal-h<tg~~ram-xe-e Cook Nuclear Plant Unit 1 B 3.5.2-7 Revision No. 24 Cook Nuclear Plant Unit 1 B 3.5.2-7 Revision No. 24 ECCS -Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.3 Verifying that each ECCS pump's developed head at the flow test point is greater than or equal to the required developed head ensures that ECCS pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of ECCS pump performance required by the ASME OM Code (Ref. 10). Since the ECCS pumps cannot be tested with flow through the normal ECCS flow paths, they are tested on recirculation flow (RHR and SI pumps) or normal charging flow path (centrifugal charging pumps). This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance.
The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.5.2.4 and SR 3.5.2.5 These Surveillances demonstrate that each automatic ECCS valve actuates to the required position on an actual or simulated SI signal and that each ECCS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
T-49 cod-e e=on Insert 2-4atr-t-ee t--
~ SR 3.5.2.6 Proper throttle valve position is necessary for proper ECCS performance.
These valves have stops to allow proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This Surveillance verifies the mechanical stop of each listed ECCS throttle valve is in the correct position., Insert 2*Fe~ _ * ...... .sdtbhU.atJ rrn -e .4 .-i Cook Nuclear Plant Unit I B 3.5.2-8 Revision No. 24 Cook Nuclear Plant Unit 1 B 3.5.2-8 Revision No. 24 ECCS -Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.2.7 Periodic inspections of the containment sump suction inlets ensure that they are unrestricted and stay in proper operating condition.
This Surveillance verifies that the sump suction inlets are not restricted by debris and the suction inlet strainers show no evidence of structural distress, such as openings or gaps, which would allow debris to bypass the strainers.
@re--Insert 2 ,I Ae4e-su ffieieatt~e.eteet~abmeer REFERENCES
: 1. UFSAR, Section 1.4.7.2. 10 CFR 50.46.3. UFSAR, Section 14.3.1.4. UFSAR, Section 14.3.2.5. UFSAR, Section 14.2.4.6. UFSAR, Section 14.2.5.7. UFSAR, Section 14.3.4.8. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.9. IE Information Notice No. 87-01 10. ASME, Operations and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 1 B 3.5.2-9 Revision No. 24 Cook Nuclear Plant Unit 1 B 3.5.2-9 Revision No. 24 RWST B 3.5.4 BASES ACTIONS (continued) brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.5.4.1 The RWST borated water temperature should be verified e e4h r-s to be within the limits assumed in the accident analyses band..T-hi
-,-- Insert 2 wtandea h-aceh gh=e-(X-1e4eee.
SR 3.5.4.2 The RWST water volume should be verified ~eiv--dy to be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation.
Siiee449e-RN&#xa5;ST-iiit --= Insert'2 4se~rlj--~-r1elrtceb-mtf t e,e-xe-ier~c-e.
SR 3.5.4.3 The boron concentration of the RWST should be verified be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized.
REFERENCES
: 1. UFSAR, Section 6.2.2.2. UFSAR, Section 14.3.<-=-=Insert 2 Cook Nuclear Plant Unit 1 ..- evso o B 3.5.4-5 Revision No. 0 Seal Injection Flow B 3.5.5 BASES APPLICABILITY In MODES 1, 2, and 3, the seal injection flow resistance limit is dictated by ECCS flow requirements, Which are specified for MODES 1, 2, 3, and 4. The seal injection flow resistance limit is not applicable for MODE 4 and lower, however, because high seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES. Therefore, RCP seal injection flow resistance must be limited in MODES 1, 2, and 3 to ensure adequate ECCS performance.
ACTIONS A.1 With the seal injection flow resistance not within its limit, the amount of charging flow available to the RCS may be reduced. Under this condition, action, must be taken to restore the flow resistance to within its limit. The operator has 4 hours from the time the flow resistance is known to not be within the limit to correctly position the manual valves and thus be in compliance with the accident analysis.
The Completion Time minimizes the potential exposure of the unit to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow resistance within limits. This time is conservative with respect to the Completion Times of other ECCS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.
B.1 and B.2 When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated.
The Completion Time of 6 hours for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators.
Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.
SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Verification~wevey8+/-ye-that the seal injection flow resistance is within the limit ensures that the ECCS injection flows stay within the safety analysis.
A differential pressure is established between the charging header and the RCS, and the total seal injection flow is verified to be within the limit determined in accordance with the ECCS safety analysis.The flow resistance shall be > 0.227 ft/g pm 2.Cook Nuclear Plant Unit 1 B 3.5.5-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.5.5-3 Revision No. 0 Seal Injection Flow B 3.5.5 BASES SURVEILLANCE REQUIREMENTS (continued)
The seal injection flow resistance, RSL, is determined from the following expression:
RSL =2.31 (PcHP-PsI)/Q 2 where: PCHP = charging pump header pressure (psig);= 2148 psig (low pressure operation) or 2300 psig (high pressure operation);
and Q = total seal injection flow (gpm).eeaset {with, Pre k
2-e~tteftey-l9~e-lfevcfl tc~ bo accoptablo
~As noted, the Surveillance is not required to be performed until 4 hours after the pressurizer pressure has stabilized within a _+ 20 psig range of normal operating pressure.
The pressurizer pressure requirement is specified since this configuration will produce the required pressure.conditions necessary to assure that the manual valves are set correctly.
The pressurizer pressure indications are averaged to determine whether the appropriate pressure has been achieved.
The exception is limited to 4 hours to ensure that the Surveillance is timely.REFERENCES
: 1. UFSAR, Section 14.3.1.2. UFSAR, Section 14.3.2.3. UFSAR, Section 14.2.4.4. UFSAR, Section 14.2.5.Cook Nuclear Plant Unit I1 ..- evso o B 3.5.5-4 Revision No. 0 Containment Air Locks B 3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued) air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by the Containment Leakage Rate Testing Program.The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria which is applicable to SR 3.6.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C containment leakage rate.SR 3.6.2.2 The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY.
Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment.
Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur.
t~ a e~i4tek_~~,.a4a~e~ck Insert 2e dte~- e~~-r h~~,si 2--et.~-~~a-~as~eates~~prem L;4 -- REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 14.2.6.3. UFSAR, Section 5.7.4. 10 CFR 50, Appendix J, Option B.Cook Nuclear Plant Unit 1 B3626Rvso o B3.6.2-6 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES ACTIONS (continued) locked,.sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned.
Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is small.D.1 and D.2 If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.3.1 This SR ensures that the containment purge supply and exhaust valves are closed as required or, if open, open for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable.
The SR is not required to be met when the containment purge valves are open for the reasons stated. The valves may be opened for pressure control, ALARA or air-quality considerations for personnel entry, or for Surveillances or mafntenance activities that require the valves to be open. The containment purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. , =.=Insert 2 SR 3.6.3.2 This SR requires verification that each containment isolation manual valve and blind flange located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. This SR does not require any testing or valve .manipulation.
Rather, it involves verification that those containment isolation valves outside containment and capable of being mispositioned are in the correct position.
Insert 2 Cook Nuclear Plant Unit 1 B 3.6.3-7 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued) aa~ae-a~efauem nd wa h~4e.l~~vl ee ~~-e--ece--c-p~-es The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time the valves are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.SR 3.6.3.3 This SR requires verification that each containment isolation manual.valve and blind flange located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. For containment isolation valves inside containment, the Frequency of "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time they are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.This Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4, for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.Cook Nuclear Plant Unit 1 B3638Rvso o B3.6.3-8 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.3.4 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY.
The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses.The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.3.5 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under .administrative controls.
th
-Insert 2 ff~rmt u~re~erte-e~test&#xa2;ta f~ ~ iewt+h ce-t wr.-4#ftrje'~~re~s cAe 4e~ela-&#xa2;e-2-fett-rltep a
REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 14.2.6.3. UFSAR, Section 5.4.1 and Table 5.4-1.Cook Nuclear Plant Unit 1 B3639Rvso o B 3.6.3-9 Revision No. 0 Containment Pressure B 3.6.4 BASES S URVElILLAN CE REQ U IREM ENTS SR 3.6.4.1 Verifying that containment pressure is within limits ehsures that unit operation remains within the limits assumed in the containment analysis.-Thqq .ea -ee <Insert 2 REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 5.2.2.2.3. 10 CFR 50, Appendix K.Cook Nuclear Plant Unit 1 B3643Rvso o B 3.6.4-3 Revision No. 0 Containment Air Temperature B 3.6.5 BASES ACTIONS A.1 When containment average air temperature in the upper or lower compartment is not within the limit of the LCO, the average air temperature in the affected compartment must be restored to within limits within 8 hours. This Required Action is necessary to return operation to within the bounds of the containment analysis.
The 8 hour Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.B.1 and B.2 If the containment average air temperature cannot be restored to within its limits within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.5.1 and SR 3.6.5.2 Verifying that containment average air temperature is within the LCO limits ensures that containment operation remains within the limits assumed for the containment analyses.
In order to determine the containment average air temperature, an average is calculated using measurements taken at locations within the containment selected to provide a representative sample of the oV~erall containment atmosphere.
In the upper compartment, two locations at a nominal elevation of 712 ft 0 inches and a third location at a nominal elevation of 624 ft 10 inches are used and an arithmetic average taken. In the lower compartment, a volume weighted average temperature is calculated whereby the volume fraction for each of the various areas of lower containment is multiplied by the representative temperature, utilizing one or more temperature instruments, in that volume; In this way the temperatures are "weighted" according to the volume fraction.
These weighted temperatures are then summed to determine the Weighted Average Temperature for Lower Containment.
REFERENCES
: 1. UFSAR, Section 14.3.4.2. 10 CFR 50.49.,--Insert 2 Cook Nuclear Plant Unit 1 B3653Rvso o B 3.6.5-3 Revision No. 7 Containment Spray System B 3.6.6 BASES ACTIONS A.1I With one containment spray train inoperable, the affected train must be restored to OPERABLE status within 72 hours. The components in this degraded condition are capable of providing 100% of the heat removal and iodine removal needs after an accident.
The 72 hour Completion Time was developed taking into account the redundant heat removal and iodine removal .capabilities afforded by the OPERABLE train and the low probability of a DBA occurring during this period.B.1 and B.2 If the affected containment spray train cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 84 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.SURVEILLANCE REQUIREMENTS SR 3.6.6.1 Verifying the correct alignment of manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position since they were verified in the correct position prior to being secured. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation.
Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned, are in the correct position.
2 SR 3.6.6.2 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 5).Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect Cook Nuclear Plant Unit 1B36.5ReionN.2 B 3.6.6-5 Revision No. 32 Containment Spray System B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued) incipient failures by indicating abnormal performance.
The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.6.3 and SR 3.6.6.4 These SRs require verification that each automatic containment spray valve actuates to its correct position and each containment spray pump starts upon receipt of an actual or simulated containment spray actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls..Th-4m t=Frfuae~qae-Ih-4 2$uetae-~rh.e~lesht.pt~~ga ue~
These Surveillances include a Note that states that in MODE 4, only the manual portion of the actuation signal is required.
This is acceptable since the automatic portion of the actuation signal is not required to be OPERABLE by ITS 3.3.2, "Engineered Safety Features Actuation System (ESFAS) Instrumentation." SR 3.6.6.5 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections.
This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded.
The event based surveillance frequency following maintenance that could result in nozzle blockage was chosen because this passive portion of the system is not susceptible to service induced degradation.
REFERENCES
: 1. UFSAR, Section 1.4.7.2. UFSAR, Section 14.3.4.3. 10 CFR 50.49.4. 10 CFR 50, Appendix K.5. ASME, Operation and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 1B3666ReionN.2 B 3.6.6-6 Revision No. 32 Spray Additive System B 3.6.7 BASES ACTIONS A.1 If the Spray Additive System is inoperable, it must be restored to OPERABLE within 72 hours. The pH adjustment of the Containment Spray System flow for corrosion protection and iodine retention enhancement is reduced in this condition.
The 72 hour Completion Time takes into account the redundant flow path capabilities and the low probability of the worst case DBA occurring during this period. In addition, if the Containment Spray System is available, it would remove some iodine from the containment atmosphere in the. event of a DBA.B.1 and B.2 If the Spray Additive System cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be* brought to at least MODE 3 within 6 hours and to MODE 5 within 84 hours. The allowed Completion Time of 6 hours is reasonable, based* on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems. The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for the release of radioactive material from the Reactor Coolant System is reduced in MODE 3.SURVEILLANCE REQUIREMENTS SR 3.6.7.1 Verifying the correct alignment of Spray Additive System manual, power operated, and automatic valves in the spray additive flow path provides assurance that the system is able to provide additive to the Containment Spray System in the event of a DBA. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation.
Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned are in the correct position. .Isr SR 3.6.7.2 To provide effective iodine retention, the containment spray must be an alkaline solution.
Since the RWST contents are normally acidic, the volume of the spray additive tank must provide a sufficient volume of spray additive to adjust pH for all water injected.
This SR is performed to verify the availability of sufficient NaOH solution in the Spray Additive System. ~ w Insert 2 Cook Nuclear Plant Unit 1I ..- evso o B 3.6.7-3 Revision No. 0 Spray Additive System B 3.6.7 BASES SURVEILLANCE REQUIREMENTS (continued) f#t,-e SR 3.6.7.3 This SR provides verification (by chemical analysis) of the NaOH concentration in the spray additive tank and is sufficient to ensure that the spray solution being injected into containment is at the correct pH level.
2 SR 3.6.7.4 This SR provides verification that each automatic valve in the Spray Additive System flow path actuates to its correct position.
This Surveillance is not required for valves that are locked, sealed; or otherwise secured in the required position under administrative controls.4ere-e-tee
=9pr 1e~aePa eet-a4 c.g ae4ea~~-~~li~~-~ta~it I==nsert 2 SR 3.6.7.5 To ensure that the correct pH level is established in the borated water solution provided by the Containment Spray System, the flow rate in the Spray Additive System is verified once every 5 years.. This SR provides assurance that the correct amount of NaOH will be metered into the flow path upon Containment Spray System initiation.
The test is performed by verifying the flow rate from the spray additive tank test line to each Containment Spray System train with each containment spray pump operating in the recirculation mode. Insert 2 el REFERENCES
: 1. UFSAR, Chapter 14.3.5.9.Cook Nuclear Plant Unit 1 B3674Rvso o B 3.6.7-4 Revision No. 0 DIS B 3.6.9 BASES ACTIONS (continued)
Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.9.1 This SR confirms that >- 34 of 35 hydrogen ignitors can be successfully energized in each train. The ignitors are simple resistance elements.Therefore, energizing provides assurance of OPERABILITY.
The allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise redundancy in that region, the containment regions are interconnected so that ignition in one region would cause burning to progress to the others (i.e., there is overlap in each hydrogen ignitor's effectiveness between regions). -R Insert 2 SR 3.6.9.2 This SR confirms that the two inoperable hydrogen ignitors allowed by SR 3.6.9.1 (i.e., one in each train) are not in the same containment region.
ee~l~bt-teei -eH 9e~ ~ et 4 t -e1 re==Insert 2 SR 3.6.9.3 A more detailed functional test is performed to verify system OPERABILITY.
Each ignitor is visually examined to ensure that it is clean and that the electrical circuitry is energized.
All ignitors, including normally inaccessible ignitors, are visually checked for a glow to verify that they are energized.
Additionally, the surface temperature of each ignitor is measured to be > 1700&deg;F to demonstrate that a temperature sufficient for ignition is achieved.
: 2. FSA, ectieon dto5.8.a-a p~==Insert 2 Cook Nuclear Plant Unit 1B369-ReionN.4 B3.6.9-4 Revision No. 44 CEQ System B 3.6.10 BASES LCO In the event of a DBA, one train of the CEQ System is required to provide the minimum air recirculation for heat removal and hydrogen mixing assumed in the safety analyses.
To ensure this requirement is met, two trains of the CEQ System must be OPERABLE.
This will ensure that at least one train will operate, assuming the worst case single failure occurs.APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the CEQ System.Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the CEQ System is not required to be OPERABLE in these MODES.ACTIONS A.il If one of the trains of the CEQ System is inoperable, it must be restored to OPERABLE status within 72 hours. The components in this degraded condition are capable of providing 100% of the flow and hydrogen skimming needs after an accident.
The 72 hour Completion Time was developed taking into account the redundant flow and hydrogen skimming* capability of the OPERABLE CEQ System train and the low probability of a DBA occurring in this period.B.1 and B.2 If the CEQ System train cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQU IREM ENTS SR 3.6.10.1 Verifying that each CEQ System fan starts on an actual or simulated actuation signal, after a-delay > 108 seconds and < 132 seconds, and operates for > 15 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly.
It also ensures that blockage, fan and/or motor failure, or excessive vibration can be detected for corrective action. -Td-he- Irnsert 2 Cook Nuclear Plant Unit 1B36103RvsoN.0 B 3.6.10-3 Revision No. 0 CEQ System B 3.6.10 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR has been modified by a Note that states that this Surveillance is only required to be met in MODES 1, 2, and 3. This allowance is necessary since the specified delay (i.e., > 108 seconds and< 132 seconds) is only applicable to the automatic actuation signal (i.e., Containment Pressure -High), which is only required to be OPERABLE in MODES 1, 2, and 3. In addition, LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," requires the CEQ System Manual Initiation Function to be OPERABLE in MODE 4 and requires the performance of a TADOT every 24 months. This requirement will ensure the Manual Initiation Function can actuate the required equipment in MODE 4.SR 3.6.10.2 Verifying, with the return air fan discharge backdraft damper locked closed and the fan motor energized, the static pressure between the fan discharge and the backdraft damper is > 4.0 inches water gauge confirms one operating condition of the fan. This test is indicative of overall fan motor performance.
Such tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance.
Thae~Insert 2 ee s-e-a19l-t-he-twe-tai9f-a dt {sbte.SR 3.6.10.3 Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exist when the fan is started. By applying the correct counterweight, the damper operation can be confirmed.
+he~Fe~r rg2 de-~-heme 2 S R 3.6.10.4 Verifying the OPERABILITY of the motor operated valve in the hydrogen skimmer header provides assurance that the proper flow path will exist when the valve receives an actuation signal. Thde-urltenjieie-w'
.---Insert 2 t Cook Nuclear Plant Unit 1 B 3.6.10-4 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.6.10-4 Revision No. 0 Ice Bed B 3.6.11 BASES APPLICABILITY In MODES 1, 2, 3, and 4, a OBA could cause an increase in containment pressure and temperature requiring the operation of the ice bed.Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ice bed is not required to be OPERABLE in these MODES.ACTIONS A.1 If the ice bed is inoperable, it must be restored to OPERABLE status within 48 hours. The Completion Time was developed based on operating experience, which confirms that due to the very large mass of stored ice, the parameters comprising OPERABILITY do not change appreciably in this time period. If a degraded condition is identified, even for temperature, with such a large mass of .ice it is not possible for the degraded condition to significantly degrade further in a 48 hour period.Therefore, 48 hours is a reasonable amount of time to correct a degraded condition before initiating a shutdown.B.1 and B.2 If the ice bed cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.11.1 Verifying that the maximum temperature of the ice bed is<27&deg;F ensures that the ice is kept well below the melting point. Tfhe
___ Insert 2 waqaete-~ra4g-exe~a~ih rP, Cook Nuclear Plant Unit 1 B 3.6.11-4 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.6.11-4 Revision No. 0 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued)
The total ice mass and individual radial zone ice mass requirements defined in this Surveillance, and the minimum ice mass per basket requirement defined by SR 3.6.11.3, are the minimum requirements for OPERABILITY.
Additional ice mass beyond the SRs is maintained to address sublimation.
This sublimation allowance is generally applied to baskets in each radial zone, as appropriate, at the beginnirng of an operating cycle to ensure sufficient ice is available at the end of the-operating cycle for the ice condenser to perform its intended design function.
2 SR 3.6.11.3 Verifying that each selected sample basket from SR 3.6.11.2 contains at least 600 lbs of ice in the as-found (pre-maintenance) condition ensures that a significant localized degraded mass condition is avoided.This SR establishes a per basket limit to ensure any ice mass degradation is consistent with the initial conditions of the DBA by not significantly affecting the containment pressure response.
Reference 2 provides insights through sensitivity runs that demonstrate that the containment peak pressure during a DBA is not significantly affected by the ice mass in a large localized region of baskets being degraded below the required safety analysis mean, when the radial zone and total ice mass requirements of SR 3.6.11.2 are satisfied.
Any basket identified as containing less than 600 lbs of ice requires appropriately entering ACTION A for an inoperable ice bed due to the potential that it may represent a significant condition adverse to quality.As documented in Reference 2, maintenance practices actively manage individual ice basket mass above the required safety analysis mean for each radial zone. Specifically, each basket is serviced to keep its ice mass above 1132 lbs for- zone A, 1132 lbs for zone B, and 1132 lbs for zone C. If a basket sublimates below the safety analysis mean value, this instance is identified within the CNP corrective action program, including evaluating maintenance practices to identify the cause and correct any deficiencies.
These maintenance practices provide defense in depth beyond compliance with the ice bed Surveillance Requirements by limiting the occurrence of individual baskets with ice mass less than the required safety analysis mean.Cook Nuclear Plant Unit 1 B 3.6.11-6 Revision No. 0 Ice Bed B 3.6. 11 BASES SURVEILLANCE REQUIREMENTS (continued) accumulation on lattice frames and wall panels. The flow area through the ice basket support platform is not a more restrictive flow area because it is easily accessible from the lower plenum and is maintained clear of ice accumulation.
There is no mechanistically credible method for ice to accumulate on the ice basket support platform during unit operation.
Plant and industry experience has shown that the vertical flow area through the ice basket support platform remains clear of ice accumulation that could produce blockage.
Normally only a glaze may develop or exist on the ice basket support plafform which is not significant to blockage of flow area. Additionally, outage maintenance practices provide measures to clear the ice basket support plafform following maintenance activities of any accumulation of ice that could block flow areas.Frost buildup or loose ice is not to be considered as. flow channel blockage, whereas attached ice is considered blockage of a flow channel.Frost is the solid form of water that is loosely adherent, and can be brushed off with the open hand.2 SR 3.6.11.5 This SR ensures that a representative sampling of ice baskets, which are relatively thin walled, perforated cylinders, have not been degraded by wear, cracks, corrosion, or other damage. The SR is designed around a full-length inspection of a sample of baskets, and is intended to monitor the effect of the ice condenser environment on ice baskets. The groupings defined in the SR (two baskets in each azimuthal third of the ice bed) ensure that the sampling of baskets is reasonably distributed.
The Frequency of 40 months for a visual insPection of the structural soundness of the ice baskets is based on engineering judgment and considers such factors as the thickness of the basket walls relative to corrosion rates expected in their service environment and the results of the long term ice storage testing. <==-Insert 2 SIR 3.6.11.6 Verifying the chemical composition of the stored ice ensures that the stored ice has a boron concentration
> 1800 ppm and < 2300 ppm as sodium tetraborate and a high pH, > 9.0 and < 9.5 at 25&deg;C, in order to meet the requirement for borated water when the melted ice is used in the ECCS recirculation mode of operation.
Additionally, the minimum boron concentration limit is used to assure reactor subcriticality in a post LOCA environment, while the maximum boron concentration limit is used as the bounding value in the hot leg switchover timing calculation (Ref. 4). This is accomplished by obtaining at least 24 ice samples. Each sample is taken approximately one foot from the top of the ice of each randomly Selected ice basket in each ice condenser bay. The SR is modified by a Cook Nuclear Plant Unit 1 ..18ReiinN.4 B 3.6.11-8 Revision No. 43 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued)
Note that arrows the boron concentration and pH value obtained from averaging the individual samples' analysis results to satisfy the requirements of the SR. If either the average boron concentration or average pH value is outside their prescribed limit, then entry into Condition A is required.
Sodium tetraborate has been proven effective in maintaining the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine, although the removal of iodine from the containment atmosphere by the sodium tetraborate is not assumed in the accident analysis.
This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment Spray System fluids in the recirculation mode of operation.
The Frequency of 54 months is intended to be consistent with the expected length of three fuel cycles, and was developed considering these facts: a. Long term ice storage tests have determined that the chemical composition of the stored ice is extremely stable;b. There are no normal operating mechanisms that decrease the boron concentration of the stored ice, and pH remains within a 9.0-9.5 range when boron concentrations are above approximately 1200 ppm;c. Operating experience has demonstrated that meeting the boron concentration and pH requirements has never been a problem; and d. Someone would have to enter the containment to take the sample, and, if the unit is at power, that person would receive a radiation dose.
2 SR 3.6.11.7 This SR ensures that initial ice fill and any subsequent ice additions meet the boron concentration and pH requirements of SR 3.6.11.6.
The SR is modified by a Note that allows the chemical analysis to be performed on either the liquid or resulting ice of each sodium tetraborate solution prepared.
If ice is obtained from offsite sources, then chemical analysis data must be obtained for the ice supplied.Cook Nuclear Plant Unit 1 B361- eiinN.4 B 3.6.11-9 Revision No. 43 Ice Condenser Doors B 3.6.12 BASES ACTIONS (continued)
C.1 If Required Action B.1 or B.2 and associated Completion Time is not met, the doors must be restored to OPERABLE status and closed positions within 48 hours. The 48 hour Completion Time is based on the fact that, with the very large mass of ice involved, it would not be possible for the temperature to increase to the melting point and a significant amount of ice to melt in a 48 hour period. The 48 hour Completion Time is also consistent with the ACTIONS of LCO 3.6.11, "Ice Bed." 0.1 and 0.2 With any Required Action and associated Completion Time of Condition A or C not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SIR 3.6.12.1 Verifying that the inlet doors are in their closed positions makes the operator aware of an inadvertent opening of one or more doors. ,The 2 Frete 4 t ~
if w th-tt~rf SR 3.6.12.2 Verifying, by visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris provides assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed.
eeei9=e1 j1reefij 4.Rkei
~ t"lre~rblt-Ym--Bw eer====Insert 2 Cook Nuclear Plant Unit 1B36125RvsoN.0 B 3.6.12-5 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.3 Verifying, by visual inspection, that the top deck doors are in place and not obstructed provides assurance that the doors are performing their function of keeping warm air out of the ice condenser during normal operation, and would not be obstructed if called upon to open in response to a DBA. T -re 1ecye r
I p ,e-,h~gtce~t.te-e~4e<---Insert 2 SR 3.6.12.4 Verifying, by visual inspection, that the ice condenser inlet doors are not impaired by ice, frost, or debris provides assurance that the doors are free to open in the event of a DBA. T-he-=e sr 2 SR 3.6.12.5 Verifying the opening torque of the inlet doors provides assurance that no doors have become stuck in the closed position.
The value of 675 in-lb is based on the design opening pressure on the doors of 1.0 lb/ft 2.lFiT~hii-,=.=
Insert 2 cT-a1b 1 tnltO U13 f rQ4ieJ4~weveb,-eseauhed e~r-~e41es9-whieh-dees-Ret--aI~w r ia u ~ tter-Ree e-e-tre.-Be~eeib wr e alyer4e~r-rebl~tu d~~e~ew4, Cook Nuclear Plant Unit 1 B 3.6.12-6 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.6.12-6 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.6 The torque test Surveillance ensures that the inlet doors have not developed excessive friction and that the return springs are producing a door return torque within limits. The torque test consists of the following:
: 1. Verify that the torque, T(OPEN), required to cause opening motion at the 400 open position is -< 195 in-Ib;2. Verify that the torque, T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 400 open position is > 78 in-lb; and 3. Calculate the frictional torque, T(FRICT) = 0.5 {T(OPEN) -T(CLOSE)}, and verify that the T(FRICT) is -< 40 in-lb.T(OPEN) is known as the "door opening torque" and is equal to the nominal door torque plus a frictional torque component.
T(CLOSE) is defined as the "door closing torque" and is equal to the nominal door torque minus a frictional torque component.
The purpose of the friction and return torque Specifications is to ensure that, in the event of a small break LOCA or SLB, all of the 24 door pairs open uniformly.
This assures that, during the initial blowdown phase, the steam and water mixture entering the lower compartment does not pass through part of the ice condenser, depleting the ice there, while bypassing the ice in other bays. ;R~eaP ye~.-mRhq~ae-ah 2
.-e #
Cook Nuclear Plant Unit 1B36127RvsoN.0 B 3.6.12-7 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.7 Verifying the OPERABILITY of the intermediate deck doors provides assurance that the intermediate deck doors are free to open in the event of a OBA. The verification consists of visually inspecting the intermediate doors for structural deterioration, verifying free movement of the vent assemblies, and ascertaining free movement of each door when lifted with the applicable force shown below: Door.Liftinqi Force a. Adjacent to crane wall b. Paired with door adjacent to crane wall c. Adjacent to containment wall d. Paired with door adjacent to containment wall< 37.4 lb<_ 33.8 lb<31.8 lb_<31.0 lbe~j-efh&#xf7; Insert 2 iaerooco t-r).I&sect; REFERENCES
: 1. UFSAR, Section 14.3.4.Cook Nuclear Plant Unit 1B3.128evsoN.0 B 3.6.12-8 Revision No. 0 Divider Barrier Integrity B 3.6.13 BASES ACTIONS (continued) 0.1 and 0.2.If divider barrier integrity cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.1 3.1 Verification, by visual inspection, that all personnel access doors and equipment hatches between the upper and lower containment compartments are closed provides assurance that divider barrier integrity is maintained prior to the reactor being taken from MODE 5 to MODE 4.This SR is necessary because many of the doors and hatches may have been opened for maintenance during the shutdown.SR 3.6.13.2 Verification, by visual inspection, that the personnel access door and equipment hatch seals, sealing surfaces, and alignments are acceptable provides assurance that divider barrier integrity is maintained.
This inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.13.2 is required for each door or hatch that has been opened, prior to the final closure. Some doors and hatches may not be opened for long periods of time. -t Inser-t 2 be~uranc-athat-t-ee teqeas-t--edee=e~ef~p1e~eIder-ae4 m+traa s'e~e e4a~i-=f l~4rPa -~- -44 Ai ae Jeeuewaf),a~-
~ramcjprecP4a~~a SR 3.6.13.3 Verification, by visual inspection, after each opening of a personnel access door or equipment hatch that it has been closed makes the operator aware of the importance of closing it and thereby provides additional assurance that divider barrier integrity is maintained while in.applicable MODES.Cook Nuclear Plant Unit 1 B 3.6.13-4 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.6.13-4 Revision No. 0 Divider Barrier Integrity B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.13.4 Conducting periodic physical property tests on divider barrier seal test coupons provides assurance that the seal material has not degraded in the containment environment, including the effects of irradiation with the reactor at power. The required tests include a tensile strength test and a test for elongation..T aye-4me 4e4.4sek~a.estee-ef4 ae~t SR 3.6.13.5 Visual inspection of the seal around the perimeter provides assurance that the seal is properly secured in place, such that the total divider barrier bypass area is less than or equal to the design basis limit of 7 square feet. -ToFoucz'o 2
h~~~- ~eer tesall m
* eur ..tsface~eece~ceeietlqet d i --Insert 2=-'nsert 2 aee epueb ~ya REFERENCES
: 1. UFSAR, Section 14.3.4.1.3.1.3.
: 2. UFSAR, Section 14.3.4.1.3.1.1.e Cook Nuclear Plant Unit 1 ..35ReiinN.4 B 3.6.13-5 Revision No. 46 Containment Recirculation Drains B 3.6.14 BASES ACTIONS (continued)
C.1I If one CEQ fan room drain is inoperable, 1 hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the containment analysis.
The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1,"Containment," which requires that the containment be restored to OPERABLE status within 1 hour.D. 1 If one flow path in the flood-up overflow wall is inoperable, 1 hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the containment analysis.
The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires that containment be restored to OPERABLE status within 1 hour.E.1 and E.2 If the affected drain(s) cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.14.1 and SR 3.6.14.2 Verifying the OPERABILITY of the required refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.14.2 confirms that the required refueling canal drain blind flanges have been removed and that the required drains are clear of any obstructions that could impair their functioning.
In addition to debris near the drains, attention must be given to any debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains. This verification is performed by SR 3.6.14.1, which requires verification that there is no debris present in the upper containment or refueling canal that could obstruct the required refueling canal drains. SR 3.6.14.1 and SR 3.6.14.2 must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the blind flanges have been removed and that no debris that could impair the drains was deposited during the time the canal was filled. Ra-adle~te4-. er#p-eef to 4-=--Insert 2 Cook Nuclear Plant Unit 1 ..44ReiinN.1 B 3.6.14-4 Revision No. 18 Containment Recirculation Drains B 3.6.14 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.14.3 Verifying the OPERABILITY of the ice condenser floor drains ensures that they will be able to perform their functions in the event of a DBA.Inspecting the drain valve disk ensures that the valve is performing its function of sealing the drain line from warm air leakage into the ice condenser during normal operation, yet will open if melted ice fills the line following a DBA. Verifying that the drain lines are not obstructed ensures their readiness to drainwater from the ice condenser.
<=-l-nsert 2ee e~e~e~gsue~~~~steaee&#xa2;gt ex ae4t 4&#xa2;s -i~e 4 -e --
efrm~qF~---4ehF-eeee e --
elaa-lr gewre SR 3.6.14.4 and SR 3.6.14.5 Verifying the operability of the CEQ fan room drains ensures that they will be able to perform their function in the event of a DBA. SR 3.6.14.4 confirms that the required drains are clear of any obstructions.
In addition to debris near the drains, attention must be given to debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains.SR 3.6.14.4 must be performed before entering MODE 4 from MODE 5 and after personnel entry into a CEQ fan room while in MODES 1 through 4. This frequency was developed considering such factors as the location of the drains, and the absence of personnel traffic in the vicinity of the drains. The SR is modified by a Note. The Note indicates that only the CEQ fan room that has been entered need be inspected if the SR is being performed due to personnel entry in MODES I through 4. The Note precludes unnecessarily requiring inspection of both CEQ fan rooms if only one has been entered. SR 3.6.14.5 confirms that the CEQ fan room debris interceptors are installed and free of structural distress.
SR 3.6.14.5 also confirms that the flow opening at the pipe tunnel sump is not obstructed.
The 24 month frequency was developed considering such factors as the location and the design of the debris interceptors and flow opening.Cook Nuclear Plant Unit I B 3.6.14-5 Revision No. 18 Cook Nuclear Plant Unit 1 B 3.6.14-5 Revision No. 18 SGSVs B 3.7.2 BASES SURVEILLANCE REQUIREMENTS (continued)
The Frequency is in accordance with the Inservice Testing Program.This test is conducted in MODE 3 with the unit at operating temperature and pressure.
This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.
SR 3.7.2.2 This SR verifies that each SGSV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage. Th thes ei99pfie pei~eiffa REFERENCES
: 1. UFSAR, Section 10.2.2. UFSAR, Section 14.2.5.3. 10OCFR 100.11.4. Technical Requirements Manual 5. ASME, Operations and Maintenance Standards and Guides (OM Codes).
2 Cook Nuclear Plant Unit 1B3724ReionN.4 B 3.7.2-4 Revision No. 34 MFIVs and MFRVs B 3.7.3 BASES ACTIONS (continued) 0.1I With both the MFIV and MFRV inoperable in the same flow path, there is no redundant system to operate automatically and perform the required safety function.
Under these conditions, the affected flow path must be isolated within 8 hours. This action returns the system to the condition where at least one valve in each flow path is performing the required safety function.
The 8 hour Completion Time is reasonable, based on operating experience, to complete the actions required to close the MFIV or MFRV, or otherwise isolate the affected flow path.0.1 and 0.2 If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours and in MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.3.1 and SR 3.7.3.2 REQUIREMENTS These SRs verify that the closure time of each MFIV and MFRV is within the limit given in Reference 2 and is within that assumed in the accident and transient analyses.
The valve(s) may also be tested to more restrictive requirements in accordance with the Inservice Testing Program.The Frequency for this SR is in accordance with the Inservice Testing Program.SR 3.7.3.3 This SR verifies that each MFIV and MFRV can close on'an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage.
ra~~~
2 REFERENCES
: 1. UFSAR, Section 10.5.1.2.2. Technical Requirements Manual Cook Nuclear Plant Unit 1B3734ReionN.4 B3.7.3-4 Revision No. 34 SG PORVs B 3.7.4 BASES ACTIONS (continued)
C.1 and C.2 If the SG PORV(s) cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance upon steam generator for heat removal, within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.D.1 If one or more required SG PORVs are inoperable in MODE 4, the unit is in a degraded condition with reduced safety related means to cool the unit.to RHR entry conditions following an event, and the possibility of no means for conducting a cooldown with nonsafety related equipment since the condenser may be unavailable for use with the Steam Dump.System.
The seriousness of this condition requires that action be started immediately to restore the inoperable SG PORV(s) to OPERABLE status.SURVEILLANCE REQ U IREM ENTS SR 3.7.4.1 To perform a controlled cooldown of the RCS, the SG PORVs must be able to be opened remotely and throttled through their full range. This SR ensures that the SG PORVs are tested through a full control cycle nt ',cast ,e~ee--pr2-=me~ths.
Performance of inservice testing or use of an SG PORV during a unit cooldown may satisfy this requirement.
Insert 2 e
piL REFERENCES
: 1. UFSAR, Section 10.2.2.2. UFSAR, Section 14.2.4.Cook Nuclear Plant Unit 1 B 3.7.4-3 Revision No. 0 AFW System B 3.7.5 BASES ACTIONS (continued)
E.1 In MODE 4, either the reactor coolant pumps or the RHR loops can be used to provide forced circulation.
This is addressed in LCO 3.4.6, "RCS Loops -MODE 4." With one required AFW train inoperable, action must be taken to immediately restore the inoperable train to OPERABLE status. The immediate Completion Time is consistent with LCO 3.4.6.SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the required AFW System water and steam supply flow paths provides assurance that the proper flow paths will exist for AFW operation.
Verification of the AFW System water supply flow path includes both the suction (either a flow path from the CST or the Essential Service Water (ESWV) System) and discharge flow paths. This SR does.not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the-correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.The SR is modified by a Note that states one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AEW mode of operation, provided it is not otherwise inoperable.
This exception allows the system to be out of its normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s) inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, and these manual operations are an accepted function of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained.
T4 qtereyq-baee~f-ea~aef~g~~l 4-==Insert 2 ee fr-et=
Cook Nuclear Plant Unit 1 ..- evso o B3.7.5-7 Revision No. 0 AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.5.2.Verifying that each required AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 2).Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance.
Performance of inservice testing discussed in the ASME OM Code (Ref. 2) (only required at 3 month intervals) satisfies this requirement.
This SR is modified by a Note indicating that the SR should be deferred for the turbine driven AFW pump until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test at entry into MODE 3. At 850 psig, there is sufficient pressure to perform the test.SIR 3.7.5.3 This SIR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.T-he---24,-n' I Inse rt 2 The SR is modified by two Notes. Note 1 states that one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not otherwise inoperable.
This exception allows the AFW train(s) to be out of normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s) inoperable.
Cook Nuclear Plant Unit I B 3.7.5-8 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.7.5-8 Revision No. 0 AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued)
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, these manual operations are an accepted condition of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained.
Note 2 states that the SR is only required to be met in MODES 1, 2, and 3. It is not required to be met in MODE 4 since the AFW train is only required for the purposes of removing decay heat when the SG is relied upon for heat removal. The operation of the AFW train is by manual means and automatic startup of the AFW train is not required.SR 3.7.5.4 This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an ESFAS by demonstrating that each AEW pump starts automatically on an actual or simulated actuation signal in MODES 1, 2, and 3. Th- " ao on th Insert 2 we This SR is modified by three Notes. Note 1 indicates that the SR may be deferred for the turbine driven AFW pump until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test at entry into MODE 3. At 850 psig, there is sufficient steam pressure to perform the test. Note 2 states that one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not otherwise inoperable.
This exception allows the AFW train(s) to be out of normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s)inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, these manual operations are an accepted condition of the AFW System. OPERABILITY (i.e., the intended safety function)continues to be maintained.
Note 3 states that the SR is only required to be met in MODES 1, 2, and 3. It is not required to be met in MODE 4 since the AFW train is only required for the purposes of removing decay heat when the SG is relied upon for heat removal. The operation of the AFW train is by manual means and automatic startup of the AFW train is not required.Cook Nuclear Plant Unit 1 B3759Rvso o B 3.7.5-9 Revision No. 0 CST B 3.7.6 BASES ACTIONS (continued) adequate to ensure the backup auxiliary feedwater supply continues to be available.
The 7 day Completion Time is reasonable, based on an OPERABLE backup auxiliary feedwater supply being available, and the low probability of an event occurring during this time period requiring the CST.B.1 and B.2 If any Required Action and associated Completion Time cannot be met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance on the steam generator for heat removal, within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.7.6.1 This SR verifies that the CST contains, the required volume of cooling water. T h-yit~-peei eaee1he--,=
Insert 2 tee #,i~ae4 -ety eare REFERENCES
: 1. UFSAR, Section 10.5.2.2. UFSAR, Chapter 14.Cook Nuclear Plant Unit 1 ..- eiinN.2 B 3.7.6-3 Revision No. 26 CCW System B 3.7.7 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "'RCS Loops -MODE 4," be entered if an inoperable CCW train results in an inoperable RHR loop.This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this condition, the remaining OPERABLE CCW train is adequate to perform the heat removal function.The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period.B.1 and B.2 If the CCW train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions-from full power conditions inl an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.7.1 REQUIREMENTS T~his SR is modified by a Note indicating that the isolation of the CCW flow to individual components may render those components inoperable but does not affect the OPERABILITY of the CCW System.Verifying the correct alignment for manual, power operated, and automatic valves in the CCW flow path provides assurance that the proper flow paths exist for CCW operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.T-he e~~er~t~geti e Insert 2 wi-h th roo
~re-e d-rer Cook Nuclear Plant Unit 1 ..- evso o B 3.7.7-3 Revision No. 0 COW System B 3.7.7 BASES SURVEILLANCE REQUI REMENTS (continued)
SR 3.7.7.2 This SR verifies proper automatic operation of the CCW valves on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation.
This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. _______e____e_
lr~eee
:""Insert 2-anad tho ~ e SR 3.7.7.3 This SR verifies proper automatic operation of the COW pumps on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation.
T ~
Insert 2a-p~ euaea4thf~t~4aereutret-aet f-erat -a-9ee --
shwP-h{.es ee{suua~~e r'i~eeveaeeep-aef~reri.9 fllifty-~standIpe-fif.-
REFERENCES
: 1. UFSAR, Section 9.5.2. UFSAR, Table 9.5-3.REFERENCES
: 1. UFSAR, Section 9.5.2. UFSAR, Table 9.5-3.Cook Nuclear Plant Unit 1 ..- evso o B 3.7.7-4 Revision No. 0 ESW System B 3.7.8 BASES SURVEILLANCE REQUIREMENTS (continued) rather, it involves verification that those valves capable of being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
.I--nsert 2 SR 3.7.8.2 This SR verifies proper automatic operation of the ESW valves on an actual or simulated actuation signal. The ESW System is a normally operating system that cannot be fully actuated as part of normal testing.This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.u~# ~~etm" ' Insert SR 3.7.8.3 This SR verifies proper automatic operation of the required ESW pumps on an actual or simulated actuation
*lSwi'Nat tho-4=ertc e-qaay4,<--==Ilnsert 2 REFERENCES
: 1. UFSAR, Section 9.8.3.2. UFSAR, Section 9.8.3.2.3. UFSAR, Section 9.5.2.Cook Nuclear Plant Unit 1 ..- evso o B 3.7.8-4 Revision No. 0 UHS B 3.7.9 BASES SURVEILLANCE REQUIREMENTS (continued) determining the UHS temperature is averaging the available operating circulating water pumps discharge temperatures.
Insert 2 i--
r~a m -- 1 ......... --r REFERENCES
: 1. UFSAR, Section 10.6.2.2. UFSAR, Table 9.8-5.3. Regulatory Guide 1.27, Revision 2, January 1976.4. MD-12-ESW-1 06-N Assessment of Increased Lake Water Temperature on Safety Related and Non-Safety Related Systems.Cook Nuclear Plant Unit I1 ..- eiinN.5 B3.7.9-3 Revision No. 52 CREV System B 3.7.10 BASES ACTIONS (continued)
An alternative to Required Action E.1 is to immediately suspend activities that could result in a release of radioactivity that might require isolation of the CRE (Required Action E.2). This places the unit in a condition where the LCO does not apply. This does not preclude the movement of fuel to a safe position.F.1 During movement of irradiated fuel assemblies in the containment, auxiliary building, or Unit 2 containment, with two CREV trains inoperable, or with one or more CREV System trains inoperable due to inoperable CRE boundary, action must be taken immediately to suspend activities that could result in a release of radioactivity that requires isolation of the ORE. This places the unit in a condition that minimizes the accident risk.This does not preclude the movement of fuel to a safe position.G.1 If both CREV trains are inoperable in MODE 1, 2, 3, or 4 for reasons other than an inoperable ORE boundary or filter unit (i.e., Conditions B and C), the CREV System may not be capable of performing the intended function and the unit is in a condition outside the accident analyses.Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly.
As the environment and normal operating conditions on this system are not too severe, testing each  provides an adequate check of this system. Operating the CREV train, with flow through the HEPA filter and charcoal adsorber train, for> 15 minutes demonstrates the function of the CREV train. -Th -- Insert 2-e TGEI -eb-t heq 4pmeRP~ta FieeJn-ay.-
SR 3.7.10.2 This SR verifies that the required CREV System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum and maximum flow rate, and the physical properties of the activated charcoal.
Specific test Frequencies and additional information are discussed in detail in the VFTP.Cook Nuclear Plant Unit 1 B 3.7.10-6 Revision No. 23 Cook Nuclear Plant Unit 1 B 3.7.10-6 Revision No. 23 CREV System B 3.7.10 BASES SURVEILLANCE REQU IREM ENTS (continued)
SR 3.7.10.3 This SR verifies that each CREV train starts and operates on an actual or simulated actuation signal. The only actuation signal necessary to be verified is the Safety Injection (SI) signal, since the Control Room Radiation
-High signal is not assumed in the accident analysis.
A Note has been included that states the Surveillance is only required to be met in MODES 1, 2, 3, and 4, since these are the MODES the SI signal is assumed to start the CREV trains. The CREV trains are assumed to be manually started during a fuel handling accident.
GOfae-tij-extperde~ee Insert 2 pe~:rm ~
s-u eh-a-wy4 -er4 yhe-eae R-E --V-et9-S-'fet~jeetie*R-si&sect;1aekff SR 3.7.10.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 rem TEDE, the CRE occupants are protected from smoke, and analyses demonstrate that the CREV System is not needed to prevent exceeding hazardous~chemical limits. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences.
When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered. Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within the licensing basis habitability limits for the occupants following an accident.Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 4) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 5). These compensatory measures may also be used as mitigating actions as required by Required Action B.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref. 6). Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status. There are no SRs to verity CREV System operability for hazardous chemicals or smoke.Cook Nuclear Plant Unit 1 B371- eiinN.2 B 3.7.10-7 Revision No. 23 CRAG System B 3.7.11 BASES ACTIONS (continued) 0.1 and 0.2 During movement of irradiated fuel, if the inoperable CRAG train cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CRAC train must be placed in operation immediately.
This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that active failures will be readily detected.An alternative to Required Action 0.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room (Required Action C.2). This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.D. 1 During movement of irradiated fuel assemblies, with two CRAG trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.E. 1 If both CRAG trains are inoperable in MODE 1, 2, 3, or 4, the CRAG System may not be capable of performing its intended function.Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.7.11.1 and SR 3.7.11.2 REQUI REM ENTS These SRs verify that the heat removal capability of each CRAG train is sufficient to maintain control room air temperature
< 850&deg;F.
-Insert 2~ 4.t -tf~~et~c-~gca lg-{a~t ee~hst~ir-pvertis-dr-fhe -pe~~~-lyvrfe-a1s3a~ta-wt4e d e e. -Fr.eeueia c-y-4ef-t GRE-System.
REFERENCES
: 1. UFSAR, Section 9.10.Cook Nuclear Plant Unit 1B3.113RvsoN.0 B 3.7.11-3 Revision No. 0 ESF Ventilation System B 3.7.12 BASES ACTIONS A.1 With one ESF Ventilation train inoperable, action must be taken to restore OPERABLE status within 7 days. During this time, the remaining OPERABLE train is adequate to perform the ESF Ventilation System function.The 7 day Completion Time is appropriate because the risk contribution is less than that for the ECCS (72 hour Completion Time), and this system is not a direct support system for the ECCS. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.
B.1 If the ESF enclosure boundary is inoperable, the ESF Ventilation trains cannot perform their intended functions.
Actions must be taken to restore an OPERABLE ESF enclosure boundary within 24 hours. During the period that the ESF enclosure boundary is inoperable, appropriate compensatory measures consistent with the intent, as applicable, of GDC 19, 60, 64 and 10 CFR Part 100 should be utilized to protect plant personnel from potential hazards. Preplanned measures should be-available to address these concerns for intentional and unintentional entry- -into the condition.
The 24 hour Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures.
The 24 hour Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the ESF enclosure boundary.C.1 and C.2 If the ESF Ventilation train or ESF enclosure boundary cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly.
t-ader9I96-epe-aagc-e*dlte~s Insert 23t Cook Nuclear Plant Unit 1B37123RvsoN.0 B 3,7.12-3 Revision No. 0 ESF Ventilation System B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued)
V-a4a~at-~ -apata~ E SR 3.7.12.2 This SR verifies that the required ESF Ventilation System testing is performed in accordance with the Ventilation Filter Testing Program (V FTP). The VFTP includes testing HEPA filter performance, charcoal adsorbers efficiency, minimum and maximum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
Specific test Frequencies and additional information are discussed in detail in the VFTP.SR 3.7.12.3 This SR verifies that each ESF Ventilation train starts and operates on an actual or simulated actuation signal. One ESF Ventilation train is normally operating with flow bypassing the charcoal adsorber section.This test confirms that each train, when in standby, starts upon receipt of a Containment Pressure -High High signal and that the exhaust flow can be directed through the entire filter unit including the HEPA filter and charcoal adsorber section. Insert 2 SR 3.7.12.4 This SR verifies the integrity of the ESF enclosure.
The ability of the ESF enclosure to maintain a negative pressure, with respect to potentially uncontaminated adjacent areas, is periodically tested to verify proper functioning of the ESF Ventilation System. During the post accident mode of operation, the ESF Ventilation System is designed to maintain a slight negative pressure in the ESF enclosure, with respect to adjacent areas, at a flowrate -< 22,500 cfm to prevent unfiltered leakage. T-the- Insert 2aSTA E-- 4s Cook Nuclear Plant Unit 1B37.24RvsoN.0 B 3.7.12-4 Revision No. 0 FHAEV System B 3.7.13 BASES APPLICABILITY During movement of irradiated fuel in the auxiliary building, the FHAEV System is required to be OPERABLE to alleviate the consequences of a fuel handling accident.In MODE 1, 2, 3, 4, 5, or 6, the FHAEV Systemis not required to be OPERABLE since the FHAEV System is only credited during a fuel handling accident in the auxiliary building.ACTIONS LCO 3.0.3 is not applicable while in MODE 5 or 6. However, since irradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operations.
Entering LCO 3.0.3, while in MODE 1, 2, 3, or 4 would require the unit to be shutdown unnecessarily.
A.1 When the required FHAEV train is inoperable or not in operation during movement of irradiated fuel assemblies in the auxiliary building, action must be taken to place the unit in a condition in which the LCO does not apply. Action must be taken immediately to suspend movement of irradiated fuel assemblies in the auxiliary building.
This does not preclude* the movement of fuel to a safe position.SURVEILLANCE REQUIREMENTS SR 3.7.13.1 This SR requires verification-every4j2-heu~s~that the required FHAEV train is operating with flow through the filter unit, including the HEPA filter and charcoal adsorber section. Verification includes fan status and also verifies that each charcoal bypass damper is closed. T-he-F~eq*ei~ey-ef~-
Insert 2 SR 3.7.13.2 Standby systems should be checked periodically to ensure that they.function properly.
As the environmental and normal operating conditions on this system are not severe, testing each provides an adequate check on this system.Operating the required FHAEV train, with flow through the HEPA filter and charcoal adsorber train, for > 15 minutes demonstrates the function of the system. e+eil ~--f-th- I.nsert 2 Cook Nuclear Plant Unit 1 B 3.7.13-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.7.13-3 Revision No. 0 FHAEV System B 3.7.13 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.13.3 This SR verifies that the required FHAEV System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum and maximum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
Specific test frequencies and additional information are discussed in detail in the VFTP.SR 3.7.13.4 This SR verifies that the required FHAEV train actuates on an actual or simulated actuation signal. The test must verify that the signal automatically shuts down each of the Fuel Handling Area Supply Air System fans.
~Insert 2 T-h e-ert-e--Ffeqe~ec--y4-i each -ac-~e--fe4-a-r-eib;#vsandoJ4nt.
SR 3.7.13.5 This SR verifies the integrity of the auxiliary building enclosure.
The ability of the pool storage area to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the FHAEV train. During the accident mode of operation, the FHAEV train is designed to maintain a slight negative pressure in the FHAEV train, to prevent unfiltered leakage. The FHAEV train is designed to maintain a pressure > 0.125 inches of vacuum water gauge with respect to atmospheric pressure at a flow rate of< 27,000 cfm. Insert 2r#t REFERENCES
: 1. UFSAR, Section 9.9.3.2.2. UFSAR, Section 14.2.1.3. 10CFR100.Cook Nuclear Plant Unit 1 ..34ReiinN.2 B 3.7.13-4 Revision No. 26 Fuel Storage Pool Water Level B 3.7.14 BASES ACTIONS A...1 When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring.
When the fuel storage pool water level is lower than the required level, the movement of irradiated fuel assemblies in the fuel storage pool is immediately suspended to a safe position.
This action effectively precludes the occurrence of a fuel handling accident.
This does not preclude movement of a fuel assembly to a safe position.Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations.
Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies sufficient fuel storage pool water is available in the event of a fuel handling accident.
The water level in the fuel storage pool must be checked periodically.
T-he- ~ -e~-t4et- nst 2 REFERENCES
: 1. UFSAR, Section 9.7.2.2. UFSAR, Section 9.4.3. UFSAR, Section 14.2.1.4. 10CFR 100.11.Cook Nuclear Plant Unit 1 B371- eiinN.2 B 3.7.14-2 Revision No. 26 Fuel Storage Pool Boron Concentration B 3.7.15 BASES LCO LCO The fuel storage pooi boron concentration is required to be > 2400 ppm.The specified concentration of dissolved boron in the fuel storage pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference
: 2. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel storage pool.APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel storage pool, until a complete spent fuel storage pool verification has been performed following the last movement of fuel assemblies in the spent fuel storage pool. This LCO does not apply following the verification, since the verification would confirm that there are no misloaded fuel assemblies.
With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.ACTIONS A.1, A.2.1, and A.2.2 When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.This is most efficiently achieved by immediately suspending the movement of fuel assemblies.
The initiation of action to restore the concentration of boron to within limit occurs simultaneously with suspending movement of fuel assemblies.
Alternatively, beginning a verification of the fuel storage pool fuel locations, to ensure proper locations of the fuel, can be performed.
However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored.
This does not preclude movement of a fuel assembly to a safe position.The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply. If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable.
If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.
Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE REQ U IREM ENTS SR 3.7.15.1 This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed.
The9--da-y-F-f-eaieIycs-aie~-3pep~ete Insert 2-fpe ~tri-&#xa2;l oer-eucq4~-a -she f4ifne. " Cook Nuclear Plant Unit 1B37152RvsoN.0 B 3.7.15-2 Revision No. 0 Secondary Specific Activity B 3.7.17 BASES ACTIONS A.1 and A.2 Specific activity of the secondary coolant exceeding the allowable value is an indication of a problem in the ROS and contributes to increased post accident doses. If the secondary specific activity is not within limits, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.17.1 REQ U IREM ENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis.
A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases.
It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE. e fee~~ase-r4e~tc4P
<=.- Insert 2 f -appr-epfiete-aet~iei94-e ho.tako,-to mift ~  REFERENCES
: 1. 10OCFR 100.11.2. 10 CFR 50, Appendix A, G DC 19.3. UFSAR, Section 14.2.7.Cook Nuclear Plant Unit 1B371-3RvsoN.0 B 3.7.17-3 Revision No. 0 AC Sources -Operating B 3.8.1 BASES ACTIONS (continued) degraded level, any further losses in the AC electrical power system will cause a loss of function.
Therefore, no additional time is justified for continued operation.
The unit is required by LCO 3.0.3 to commence a controlled shutdown.SURVEILLANCE REQUIREMENTS The AC sources are designed to permit inspection and testing of all important areas and features, especially those that have a standby function, in accordance with Plant Specific Design Criterion (PSDC) 39 (Ref. 8). Periodic component tests are supplemented by extensive functional tests during refueling outages (under simulated accident conditions).
The SRs for demonstrating the OPERABILITY of the DGs are in accordance with the recommendations of Regulatory Guide 1.9 (Ref. 3), Regulatory Guide 1.108 (Ref. 9), Regulatory Guide 1.137 (Ref. 10), and IEEE Standard 387-1995 (Ref. 11) as addressed in the applicable SR discussion.
Where the SRs discussed herein specify voltage and frequency tolerances, the following is applicable.
The minimum steady state output voltage of 3910 V is 94% of the nominal 4160 V output voltage. This value allows for voltage drop to the terminals of 4160 V motors whose minimum operating voltage is specified as 90% or 3740 V. It also allows for voltage drops to motors and other equipment down through the 120 V level where the minimum operating voltage is also usually specified as 90% of nameplate rating. The specified maximum steady state output voltage of 4400 V is equal to the maximum operating voltage specified for 4000 V motors. It ensures that for a lightly loaded distribution system, the voltage at the terminals of 4000 V motors is no more than the maximum rated operating voltages.
The specified minimum and maximum steady state frequencies of the DG are 59.4 Hz and 60.5 Hz, respectively.
These values ensure .the ESF pumps can achieve adequate fluid flow to meet their safety and accident mitigation functions.
SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that the required qualified offsite circuits are OPERABLE, and that appropriate independence of offsite circuits is Isr maintained.
<" Ine"-
.
Cook Nuclear Plant Unit I B 3.8.1-16 Revision No. 41 Cook Nuclear Plant Unit 1 B 3.8.1-16 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SIR 3.8.1.2 and SR 3.8.1.8 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.
To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 1 for SR 3.8.1.2 and Note for SR 3.8.1.8) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and followed by a warmup period prior to loading.For the purposes of SR 3.8.1.2 and SR 3.8.1.8 testing, the DGs are started from standby conditions.
Standby conditions for a DG means that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations.
In order to reduce stress and wear on diesel engines, the manufacturer recommends a modified start in which the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 2.SR 3.8.1.8 requires DG starts from standby conditions and achieves required voltage and frequency within 10 seconds. The 10 second start requirement supports the assumptions of the design basis LOCA analysis in the UFSAR, Section 14.3 (Ref. 5).The 10 second start requirement is not applicable to SIR 3.8.1.2 (see Note 2 of SR 3.8.1.2) when a modified start procedure as described above is used. If a modified start is not used, the 10 second start requirement of SR 3.8.1.8 applies.Since SR 3.8.1.8 requires a 10 second start, it is more restrictive than SIR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2.addition, the DG is required to maintain proper voltage and frequency limits after steady state is achieved.
The voltage and frequency limits are normally achieved within 10 seconds. The time for the DG to reach steady state operation, unless the modified DG start method is employed, is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance.
T-he-3-ely-e-ffS -..4.2-e-n{=wi-teat--w{
.4-.. Insert 2 Cook Nuclear Plant Unit 1 B3811 eiinN.4 B 3.8.1-17 Revision No. 41 AC Sources -. Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.3 Consistent with Regulatory Guide 1.9 (Ref. 3), this Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads 90% to 100% of the continuous rating of the 0G. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.The 0.8 value is the design rating of the machine, while the 1.0 is an operational goal to ensure circulating currents are minimized.
The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections being required in order to maintain OG reliability.eil~ee4 I nse rt 2 This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized.
Note 2 states that momentary transients, because of changing bus loads, do not invalidate this test. Note 3 indicates that this Surveillance should be conducted on only one Unit 1 DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.
Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
SR 3.8.1.4 This SR provides verification that the level of fuel oil in the day tank is above the level at which fuel oil is automatically added. The level is expressed as an equivalent volume in gallons, of which 31.4 gallons is unusable (due to tank geometry and vortexing considerations) and 70 gallons is usable, and is selected to ensure adequate fuel oil for greater than 15 minutes of DG operation at full load.Cook Nuclear Plant Unit I B 3.8.1-18 Revision No. 41 Cook Nuclear Plant Unit 1 B 3.8.1-18 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
T-r #~~py <- Insert 2 I~ u..... ..... F ....SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from each fuel oil day tank oee~ve ytl 4 Mays eliminates the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainrierht in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by " ......bacteria.
Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. T-he--S vei4Ifee=-Feileneies--fe-est-abtished-ty
.==. Ir Ri s~ -e-e~ h rave eee-ca4ee4f4his4Iv
-,eiteariee.
SR 3.8.1.6 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available.
While the system design requirements provide for two engine start cycles from each of the two air start receivers associated with each DG without recharging, only one start sequence is required to meet the OPERABILITY requirements (since the accident analysis assumes the DG starts on the first attempt).The pressure specified in this SR reflects the lowest value at which one DG start can be accomplished with one air start receiver.nsert 2
~ 2 SR 3.8.1.7 This Surveillance demonstrates that each required fuel oil transfer pump (one per fuel oil transfer system) operates automatically and transfers fuel oil from its associated storage tank to its associated day tank. This is required to support continuous operation of standby power sources. This Cook Nuclear Plant Unit 1 B 3.8.1-19 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.Tho
.v ueaev- Insert 2... ...O M ...... J ... ..SR 3.8.1.9 Automatic transfer of each 4.16 kV emergency bus power supply from normal auxiliary circuit to the preferred offsite circuit and the manual alignment to the alternate required offsite circuit demonstrates the OPERABILITY of the required offsite circuit to power the shutdown Ioa(T-44---m @rth-F-r~enev oef th"+S-u'~ia-es4base-ei~aerc=e the ds..---Insert 2 t-e8ho-ec -hcu ath4hese.eei drmd{-e2-q4eref fre4-a-etia ntia As noted (Note 1Ito SR 3.8.1.9), SR 3.8.1.9.a is only required to be met when the auxiliary source is supplying the onsite electrical power subsystem.
This is acceptable since the preferred offsite source would be supplying the onsite electrical power subsystem and a transfer would not be necessary.
SR 3.8.1.10 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding a predetermined frequency and while maintaining a specified margin to the overspeed trip. Voltage and frequency are also verified to reach steady state conditions within 2 seconds. For this unit, the single load for each DG is 600 kW. This Surveillance may be accomplished by: a. Tripping the UG output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while paralleled to offsite power, or while solely supplying the bus; or Cook Nuclear Plant Unit 1 ..-0ReiinN.4 B 3.8.1-20 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
: b. Tripping its associated single largest post-accident load with the DG solely supplying the bus.Consistent with Regulatory Guide 1.9 (Ref. 3), the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the overspeed trip setpoint, or 15% above nominal speed, whichever is lower. This corresponds to 64.4 Hz, which is the nominal speed plus 75% of the difference between nominal speed and the overspeed trip setpoint.The time, voltage, and frequency tolerances specified in this SR are derived from Regulatory Guide 1.9 (Ref. 3) recommendations for response during load sequence intervals.
The2 seconds specified is equal to approximately 60% of the 3.49 second load sequence interval associated with sequencing of the largest load. The voltage and frequency specified are consistent with the design range of the equipment powered by the 0G. SR 3.8.1.10.a corresponds to the maximum frequency excursion, while SR 3.8.1.10.b and SR 3.8.1.10.c are steady state voltage and frequency values to which the system must recover following load rejection.
The- Nt-~eeey-ie 4asee-er+
a-hs-Insert 2 eopo9e{uafrp rt1S R-w vhenrpl~efof r-ed-at=theQ'rnn F-ru --
This SR is modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems.This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Cook Nuclear Plant Unit I B 3.8.1-21 Revision No. 41 Cook Nuclear Plant Unit 1 B 3.8.1-21 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Credit may be taken for unplanned events that satisfy this SR. Credit may be taken for unplanned events that satisfy this SR.Note 2 ensures that the OG is tested under load conditions that are as close to design basis conditions as possible.
When synchronized with offsite power, testing should be performed at a power factor of < 0.86.This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions.
Under certain conditions, however, Note 2 allows the Surveillance to be conducted at a power factor other than _< 0.86. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to -< 0.86 results in voltages on the emergency busses that are too high.Under these conditions, the power factor should be maintained as close as practicable to 0.86 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the DG excitation levels needed to obtain a power factor of 0.86 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained as close as practicable to 0.86 without exceeding the DG excitation limits.SR 3.8.1.11 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.8, this Surveillance demonstrates the DG capability to reject a full load (90% to 100% of the DG continuous rating) without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping.
This Surveillance ensures proper engine generator load response under the simulated test conditions.
This test simulates the loss of the total connected load that the OG experiences following a full load rejection and verifies that the OG does not trip upon loss of the load. These acceptance criteria provide for OG damage protection.
While the DG is not expected to experience this transient during an event and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated.Insert 2 t t er reure --~e-fr This SR has been modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could Cook Nuclear Plant Unit 1 ..-2ReiinN.4 B 3.8.1-22 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) the DG to automatically achieve the required voltage and frequency within the specified time.The DG autostart time of 10 seconds is derived from requirements of the accident analysis to respond to a design basis large break LOCA. The Surveillance should be continued for a minimum, of 5 minutes in order to demonstrate that all starting transients have decayed and stability is achieved.The requirement to verify the connection and power supply of permanent and autoconnected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation.
For instance, Emergency Core Cooling Systems (ECCS) injection valves are not desired to be stroked open, or centrifugal charging trains are not capable of being operated at full flow, or residual heat removal (RHR)trains performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation.
In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG systems to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified..Th--eu 4m~h aeae j~e4jC In sert 2 This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4, is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Cook Nuclear Plant Unit 1 B3812 eiinN.4 B 3.8.1-24 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.13 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.5, this Surveillance demonstrates that the DO automatically starts and achieves the required voltage and frequency within the specified time (10 seconds)from the design basis actuation signal (ESE actuation signal). In addition, the DO is required to maintain proper voltage and frequency limits after steady state is achieved.
The voltage and frequency limits are normally achieved within 10 seconds. The time for the DG to reach the steady state voltage and frequency limits is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance.
The DO is required to operate for > 5 minutes. The 5 minute period provides sufficient time to demonstrate stability.
SR 3.8.1.13.d and SR 3.8.1.13.e ensure that permanently connected loads and emergency loads are energized from the offsite electrical power system on an ESF signal without loss of offsite power.The requirement to verify the connection of permanent and auto-connected loads is intended to satisfactorily show the relationship of these loads to the DO loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation.
For instance, ECCS injection valves are not desired to be stroked open, or centrifugal charging trains are not capable of being operated at full flow, or RHR trains performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation.
In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DO system to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.T4 e Insert 2 Qpra4n-~-e4 t Cook Nuclear Plant Unit 1 B 3.8.1-25 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR is modified by two Notes. The reason for Note 1 s to minimize wear and tear on the DGs during testing. For the purpose of this testing,*the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.14 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.12, this Surveillance demonstrates that DG noncritical protective functions (e.g., low lube oil pressure) are bypassed on a loss of voltage signal or an ESF actuation test signal. The noncritical trips are bypassed during DBAs and provide an alarm on an abnormal engine condition.
This alarm provides the operator with sufficient time to react appropriately.
The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the 0G.
2a S--hn1~
~dd4 The SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DG from service. This Cook Nuclear Plant Unit 1 .:-6ReiinN.4 B 3.8:1-26 Revision No. 41 AC Sources -Operating B 3.8.1 B]ASES SURVEILLANCE REQUIREMENTS (continued) restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERAB]ILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERAB]ILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.15 This Surveillance demonstrates the DGs can start and run continuously at full load capability (90% to 100% of the OG continuous rating) for an interval of not less than 8 hours. The run duration of 8 hours is consistent with IEEE Standard 387-1995 (Ref. 11). The DG starts for this Surveillance can be performed either from standby or hot conditions.
The provisions for prelubricating and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.The load band is provided to avoid routine overloading of the DG.Routine overloading may result in more frequent teardown inspections being required in order to maintain DG reliability.tt g == Insert 2*SR-whe-e e-24-met-h-F-r-eqtuenaey-T-her-efer-ertbe--
..-
This Surveillance is modified by three Notes. Note 1 states that momentary transients due to changing bus loads do not invalidate this test. Similarly, momentary power factor transients above the power factor limit will not invalidate the test. The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in Cook Nuclear Plant Unit 1 ..-7ReiinN.4[] 3.8.1-27 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR. Note 3 ensures that the DG is tested under load conditions that are as close to design basis conditions as possible.
When synchronized with offsite power, testing should be performed at a power factor of < 0.86. This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions.
Under certain conditions, however, Note 3 allows the Surveillance to be conducted as a power factor other than < 0.86. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to < 0.86 results in voltages on the emergency busses that are too high.Under these conditions, the power factor should be maintained as close as practicable to 0.86 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the OG excitation levels needed to obtain a power factor of 0.86 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained close as practicable to 0.86 without exceeding the DG excitation limits.SR 3.8.1.16 This Surveillance demonstrates that the diesel engine can restart from a hot condition, such as subsequent to shutdown from normal Surveillances, and achieve the required voltage and frequency within 10 seconds. The 10 second time is derived from the requirements of the accident analysis to respond to a design basis large break LOCA. The 2 Fr-gt c " -e-ey eei3- gmet~.SRwel ~
e Cook Nuclear Plant Unit 1 ..-8ReiinN.4 B 3.8.1-28 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
* This SR is modified by two Notes. Note 1 ensures that thetest is performed with the diesel sufficiently hot. The load band is provided to avoid routine overloading of the DG. Routine overloads may result in more frequent teardown inspections being required in order to maintain DG reliability.
The requirement that the diesel has operated for at least 2 hours at full load conditions prior to performance of this Surveillance is based on operating experience for achieving hot conditions.
Momentary transients due to changing bus loads do not invalidate this test. Note 2 allows all OG starts to be preceded by an engine prelube period to minimize wear and tear on the diesel during testing.SR 3.8.1.17 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.1 1, this Surveillance ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and the DG can be returned to ready-to-load status when offsite power is restored.
It also ensures that the auto-start logic is reset to allow the DG to reload if a subsequent loss of offsite power occurs. The DG is considered to be in ready-to-load status when the DG is running at rated speed and voltage, with the DG output breaker open.
* iat, Insert 2 This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1, 2, 3, or 4.Cook Nuclear Plant Unit I B 3.8.1-29 Revision No. 41 Cook Nuclear Plant Unit 1 B 3.8.1-29 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.18 Under accident conditions loads are sequentially connected to the bus by the individual time delay relays. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs or RATs (as applicable) due to high motor starting currents.Verifying the load sequencer time within plus or minus 5% of its required value ensures that sufficient time exists for the DG to restore frequency and voltage and RATs to restore voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated.
Reference 4 provides a summary of the automatic loading of emergency buses.
~
t Insert 2 i~-e e a This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these.outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 1 B3813 eiinN.4 B 3.8.1-30 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.19 In the event of a DBA coincident with a loss of offsite power, the OGs are required to supply the necessary power to ESF systems so that the fuel, RCS, and containment design limits are not exceeded.This Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.12, during a loss of offsite power actuation test signal in conjunction with an ESF actuation signal. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.
-.--Insert 2eefrm ta-usuIasswn SRw iii e fj-ii~
il 1~uI~ .TI ,-efithe-Fr~eueney This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for D~s. The reason for Note 2 is that the performance of the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit I1 ..-1ReiinN.4 B 3.8.1-31 Revision No. 41 AC Sources -Operating B 3.8.1 ASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.20 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing that involves connecting the DG to its test load resistor bank, and the DG will automatically reset to ready to load operation if a ESF actuation signal is received during operation in the test mode. Ready to load operation is defined as the DG running at rated speed and voltage with the DG output breaker open.The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.13.
The intent in the requirement associated with SR 3.8.1 .20.b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified..T Insert 2 This SR is modified by two Notes. Note 1 states that this Surveillance is only required to be met when the applicable DG is connected to its test load resistor bank. This is allowed since the test mode override only functions when the DG is connected to its associated test load resistor bank. When the OG is not connected to its associated test load resistor bank, the feature is not necessary; thus the Surveillance is not required to be met under this condition.
The reason for Note 2 is that performing the Surveillance would remove a required DG from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider.
the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied' together or operated Cook Nuclear Plant Unit 1 B3813 eiinN.4 B 3.8.1-32 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SIR.SR 3.8.1.21 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing and the DG will automatically reset to ready to load operation if a LOCA actuation signal is received during operation in the test mode. Ready to load operation is defined as the DG running at rated speed and voltage with the DG output breaker open.The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.13.
The intent in the requirement associated with SR 3.8.1.21.b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.nsert 2@eea~toa=atm 4J~e ~r4 (; 4~.e4e4a--a~heaht.eee~aa JaetFrel re 4e~ et F~q ~ -eeaelt)4 a This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DG from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit~safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation Cook Nuclear Plant Unit 1 ..-3ReiinN.4 B 3.8.1-33 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
The~se shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.22 This Surveillance demonstrates that the DG starting independence has not been compromised.
Also, this Surveillance demonstrates that each engine can achieve proper speed within the specified time when the DGs are started simultaneously.,e~-ee~i mmaa~ne=-
Insert 2 This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
SR 3.8.1.23 With the exception of this Surveillance, all other Surveillances of this Specification (SR 3.8.1.1 through SR 3.8.1.22) are applied to Unit I sources. This Surveillance is provided to direct that appropriate Surveillances for the required Unit 2 AC sources are governed by the*applicable Unit 2 Technical Specifications.
Performance of the applicable Unit 2 Surveillances will satisfy the Unit 2 requirements as well as satisfy this Unit 1 Surveillance Requirement.
Exceptions are noted to the Unit 2 SRs of LCO 3.8.1. SR 3.8.1.9.b is not required to be met since only one offsite circuit is required to be OPERABLE.
SR 3.8.1.13, SR 3.8.1.14 (ESF actuation signal portion only), SR 3.8.1.19, SR 3.8.1.20, and SR 3.8.1.21 are not required to be met because the ESF actuation signal is not required to be OPERABLE.
SR 3.8.1.22 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE.The Frequency required by the applicable Unit 2 SR also governs performance of that SR for Unit 1.Cook Nuclear Plant Unit 1 ..-4ReiinN.4 B 3.8.1-34 Revision No. 41 Diesel Fuel Oil B 3.8.3 BASES SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load.The 7 day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.e -Insert 2 SR 3.8.3.2 The tests listed below are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion.
If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage tank(s). The tests, limits, and applicable ASTM Standards are as follows: a. Sample the new fuel oil in accordance with ASTM D4057-81 (Ref. 5);b. Verify that the sample has: (1) when tested in accordance with ASTM D1298-80 (Ref. 5) an absolute specific gravity at 60/60&deg; F of > 0.82 and _< 0.88, an API gravity at 60&deg;F of -> 30&deg; and < 400, an API gravity of within 0.3 degrees at 60&deg;F when compared to the supplier's certificate, or a specific gravity of within 0.0016 at 60/600 when compared to the supplier's certificate; (2) a kinematic viscosity at 40&deg;C of > 1.9 centistokes and -< 4.1 centistokes or Saybolt viscosity at 100 0 F of-> 32.6 and <40.1, if gravity was not determined by comparison with supplier's certification, when tested in accordance with ASTM 975-81 (Ref. 5); and (3) a flash point of > 125&deg;F when tested in accordance with ASTM D975-81 (Ref. 5); and c. Verify that the new fuel oil has a clear and bright appearance with proper color when tested in accordance with ASTM D4176-82 (Ref. 5).Failure to meet any of the above limits is cause for rejecting the new fuel oil, but does not represent a failure to meet the LCO concern since, the fuel oil is not added to the storage tanks.Following the initial new fuel oil sample, the fuel oil is analyzed within 31 days following addition of the new fuel oil to the fuel oil storage tank(s)to establish that the other properties specified in Table 1 of Cook Nuclear Plant Unit 1 B3.8.3-4 Revision No. 0 Diesel Fuel Oil B 3.8.3 BASES SURVEILLANCE REQUIREMENTS (continued)
ASTM D975-81 (Ref. 6) are met for new fuel oil when tested in accordance with ASTM D975-81 (Ref. 5), except that the analysis for sulfur may be performed in accordance with ASTM D2622-82 (Ref. 5).The 31 day period is acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an immediate effect on DG operation.
This Surveillance ensures the availability of high quality fuel oil for the DGs.Fuel oil degradation during long term storage shows up as an increase in particulate, due mostly to oxidation.
The presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause fouling of filters and fuel oil injection equipment, however, which can cause engine failure.Particulate concentrations should be determined in accordance with ASTM D2276-83, Method A (Ref. 5). This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/I. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing.The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.
SR 3.8.3.3 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must*have a water environment in order to survive. Removal of water from the fuel storage tanks-~e~c -ve'c~3-1--Eleys eliminates the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, and contaminated fuel oil, and from breakdown of the fuel oil by bacteria.
Frequent checking for and removal of accumulated water minimizes fouling and provides data. regarding the watertight integrity of the fuel oil system. The Surveillance Frequencies are established by Regulatory Guide 1.137 (Ref. 2). This SR is for preventive maintenance.
The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during performance of the Surveillance.
.-=Insert 2 Cook Nuclear Plant Unit 1 ..- evso o B3.8.3-5 Revision No. 0 DC Sources -Operating B 3.8.4 BASES ACTIONS (continued)
E.1 If one or both required Unit 2 Train A and Train B DC electrical power subsystems are inoperable, the associated ESW train(s) are not capable of performing their intended function.
Immediately declaring the affected supported feature, e.g., ESW train, inoperable allows the ACTIONS of LCO 3.7.8 to apply appropriate limitations on continued reactor operation.
SURVEILLANCE SR 3.8.4.1 REQUIREMENTS Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the battery chargers, which support the ability of the batteries to perform their intended function.
Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state while supplying the continuous steady state loads of the associated DC subsystem.
On float charge, battery cells will receive adequate current to optimally charge the battery. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the minimum float voltage established by the battery manufacturer (2.20 Vpc or 255.2 VDC at the battery terminals of the Train A and Train B batteries and 2.20 Vpc or 257.4 VDC for the Train N battery).
This voltage maintains the battery plates in a condition that supports maintaining the grid life (expected to be approximately 20 years). -s~eieer-te 4,====-I nse rt 2 SR 3.8.4.2 This SR verifies the design capacity of the battery chargers.
According to Regulatory Guide 1.32 (Ref. 9), the battery charger supply is recommended to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences.
The minimum required amperes and duration ensure that these requirements can be satisfied.
This SR requires that each Train A and Train B required battery charger be capable of supplying
> 300 amps at > 250 VDC for > 4 hours and the Train N battery charger is capable of supplying
> 25 amps at > 250 VDC for > 4 hours. The ampere requirements are based on the output rating of the chargers.
The voltage requirements are based on the charger voltage Cook Nuclear Plant Unit 1 B3.8.4-7 Revision No. 0 DC Sources -Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued) level after a response to a loss of AC power. The time period is sufficient to detect significant charger failures.
ai Insert 2
~.t~ dh~
~ ~  SR 3.8.4.3 A battery service test is a special test of the battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The battery charger must be disconnected throughout the performance of the battery service test. The discharge rate and test length should correspond to the design duty cycle requirements as specified in the applicable design documents.
<.=-=- I nse rt 2 j~g
~ rif9el rec a-h ,6me~atsuaIe~steRMe~
~rm This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu .of a service test.The reason for Note 2 is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for- the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced.
This. assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 1 B 3.8.4-8 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE SR 3.8.6.1 REQUIREMENTS Verifying battery float current while on float charge is used to determine the state of charge of the battery. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a charged state.The float current requirements are based on the float current indicative of a charged battery. Use of float current to determine the state of charge of the battery is consistent with IEEE-450 (Ref. 1).
Insert 2ece~rete*
~~e-=e ~re~ e e(R This SR is modified by a Note that states the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.8.4.1. When this float voltage is not maintained the Required Actions of LCO 3.8.4 ACTION A are being taken, which provide the necessary and appropriate verifications of the battery condition.
Furthermore, the float current limit of 2 amps is established based on the nominal float voltage value and is not directly applicable when this voltage is not maintained.
SR 3.8.6.2 and SR 3.8.6.5 Optimal long term battery performance is obtained by maintaining a float voltage greater than or equal to the minimum established design limits provided by the battery manufacturer, which corresponds to 257.5 VDC for a 116 cell battery and 259.7 VDC for a 117 cell battery at the battery terminals, or 2.22 Vpc. This provides adequate over-potential, which limits the formation of lead sulfate and self discharge, which could eventually render the battery inoperable.
Float voltages in this range or less, but greater than 2.07 Vpc, are addressed in Specification 5.5.15." SRs 3.8.6.2 and 3.8.6.5 require verification that the cell float voltages are equal to or greater than the short term absolute minimum voltage of 2.07 V. T- Insert 2
=--SR 3.8.6.3 The limit specified for electrolyte level (i.e., greater than or equal to the low level mark) ensures that the plates suffer no physical damage and maintains adequate electron transfer capability.
T:he e -Insert 2 Cook Nuclear Plant Unit 1 ..- evso o B3.8.6-5 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.6.4 This Surveillance verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (i.e., 60&deg;F for the Train A and Train B 250 VDC batteries and 45&deg;F for the Train N 250 VDC battery).
Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the design requirements.
Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity.
Insert 2 SR 3.8.6.6 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.6.6;however, only the modified performance discharge test may be used to satisfy the battery service test requirements of SR 3.8.4.3.A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity.
Initial conditions for the modified performance discharge test should be identical to those specified for a performance discharge test as specified in IEEE-450 (Ref. 1).It may consist of just two rates: for instance the one minute rate for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the modified performance discharge test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test must remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.. Currently, the modified performance discharge test is performed by testing the battery using the service test profile for the first 4 hours followed by the performance discharge test profile for the Cook Nuclear Plant Unit 1 ..- evso o B 3.8.6-6 Revision No. O Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued) remainder of the test. This method has been determined by the system engineer and the battery manufacturer to be an acceptable modified performance test procedure, and is consistent with IEEE-450 (Ref. 1).The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 1) and IEEE-485 (Ref. 3). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements..
Furthermore, the battery is sized to meet the assumed duty cycle loads when the battery design capacity reaches this 80% limit. Insert 2 TaeS~i4 I f th e battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity > 100% of the manufacturer's ratings. Degradation is indicated, according to IEEE-450 (Ref. 1), when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is below 90% of the manufacturer's rating.The 12 month and 60 month Frequencies are consistent with the recommendations in IEEE-450 (Ref. 1). The 24 month Frequency is derived from the recommendations of IEEE-450 (Ref. 1).This SR is modified by a Note. The reason for the Note is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial*Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1,2, 3 or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 1 ..- evso o B 3.8.6-7 Revision No. 0 lnverters
-Operating B 3.8.7 BASES'ACTIONS (continued) inverter inoperabilit~y.
This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail. When the 120 VAC vital bus is powered from its regulated 600/120 VAC transformer, it is relying upon interruptible AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the 120 VAC vital buses is the preferred source for powering instrumentation trip setpoint devices.B.1 With two inverters in the same train inoperable, the remaining inverters are capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in one of the two remaining inverters could result in the minimum ESE functions not being supported.
Therefore, one of the inverters must be restored to OPERABLE status within 6 hours.The 6 hour Completion Time is consistent with that allowed for an inoperable RTS train and an inoperable ESFAS train, since the inverters support the 120 VAC vital buses, which in turn support the RTS and ESFAS trains.C.1 and C.2 If the Train A or Train B inverter(s) cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and the 120 VAC vital buses energized from the associated inverter.
Each inverter must be connected to its associated 250 VDC bus. The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation of the RTS and ESFAS connected to the 120 VAC vital buses.
.Insert 2 r-
-<ivrtmm f4rc4e Cook Nuclear Plant Unit 1 B 3.8.7-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.8.7-3 Revision No. 0 Inverters
-Shutdown B 3.8.8 BASES SURVEILLANCE REQU IREMENTS SR 3.8.8.1 This Surveillance verifies that the required inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter.
The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation connected to the AC vital buses. HT-e-7=lay F-eecrtast t-r-re ~~-a~ t ~'e Insert 2 REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 1 B3884Rvso o B 3.8.8-4 Revision No. 0 Distribution Systems -Operating B 3.8.9 BASES SURVEILLANCE REQUIREMENTS SR 3.8.9.1 This Surveillance verifies that the required AC, DC, and 120 VAC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment.
The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus.The verification of proper voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these buses. ,-Rhe-7-=tay Freuaye{e-a--ccea--erela~et@b~tf ~ f
# tP==Insert 2 REFERENCES
: 1. Safety Guide 6, March 1971.2. UFSAR, Chapter.14.
: 3. Regulatory Guide 1 .93, December 1974.Cook Nuclear Plant Unit 1B3.9-0RvsoN.0 B 3.8.9-10 Revision No. 0 Distribution Systems -Shutdown B 3.8.10 BASES SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the AC, DC, and 120 VAC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized.
The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. *REFERENCES
: 1. UFSAR, Chapter 14.----Insert 2 Cook Nuclear Plant Unit 1B381-4RvsoN.0 B 3.8.10-4 Revision No. 0 Boron Concentration B 3.9.1 BASES ACTIONS (continued)
Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position.
Operations that individually add limited positive reactivity (e.g., temperature fluctuations from inventory addition or temperature control fluctuations), but when combined with all other operations affecting core reactivity (e.g., intentional boration) result in overall net negative reactivity addition, are not precluded by this action.A.3 In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately.
In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied.
The only requirement is to restore the boron concentration to its required value as soon as possible.
In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.
Once actions have been initiated, they must be continued until the boron concentration is restored.
The restoration time depends on the amount of*boron that must be injected to reach the required concentration.
SURVEILLANCE REQUIREMENTS SR 3.9.1.1 and SR 3.9.1.2 These SRs ensure that the coolant boron concentration in the RCS, and connected portions of the refueling canal and the refueling cavity, is within the COLR limits. The boron concentration is determined periodically and prior to re-connecting portions of the refueling canal and the refueling cavity to the RCS, by chemical analysis.= ---Insert 2 The -Frcqucncy aeae The SR 3.9.1.2 Frequency of once within 72 hours prior to connecting the refueling canal and refueling cavity to the RCS ensures that if any dilution activity has occurred while the cavity and canal were disconnected from the RCS, correct boron concentration is verified prior to communication with the RCS.REFERENCES
: 1. UFSAR, Section 1.4.5.Cook Nuclear Plant Unit 1 B3913Rvso o B 3.9.1-3 Revision No. 0 Nuclear Instrumentation B 3.9.2 BASES ACTIONS (continued) since CORE ALTERATIONS and positive reactivity additions are not to be made, the core reactivity, condition is stabilized until the source range neutron flux monitors are OPERABLE.
This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists.The Completion Time of once per 12 hours is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified.
The 12 hour Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.C. 1 With no audible count rate OPERABLE, prompt and definite indication of a boron dilution event, consistent with the assumptions of the safety analysis, is lost. In this situation, the boron dilution event may not be detected quickly enough to assure sufficient time is available for operators to manually isolate the unborated water source and stop the dilution prior to the loss of SHUTDOWN MARGIN. Therefore, action must be taken to prevent an inadvertent boron dilution event from occurring.
This is accomplished by isolating all the unborated water flow paths to the Reactor Coolant System. Isolating these flow paths ensures that an inadvertent dilution of the reactor coolant boron concentration is prevented.
The Completion Time of "'Immediately" assures a prompt response by operations and requires an operator to initiate actions to isolate an affected flow path immediately.
Once actions are initiated, they must be continued until all the necessary flow paths are isolated or the circuit is restored to OPERABLE status.SURVEILLANCE SR 3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is normally a comparison of the parameter indicated on one channel to a similar parameter on another channel. It is based on the assumption that the two indication channels should be consistent with core conditions.
Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.
* h-Freq -Insert 2ei si- Cook Nuclear Plant Unit 1 B 3.9.2-3 Revision No. 0 Cook Nuclear Plant Unit 1 B 3.9.2-3 Revision No. 0 Nuclear Instrumentation B 3.9.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.2.2 SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION ey.CHANNEL CALIBRATION is a complete check of the instrument loop, except the detector.
The CHANNEL CALIBRATION for the Westinghouse source range neutron flux monitors also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. In addition, the CHANNEL CALIBRATION includes verification of the audible count rate function for the required monitor. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. 1 1-t Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Section 14.1.5.Cook Nuclear Plant Unit 1 B3924Rvso o B3.9.2-4 Revision No. 0 Containment Penetrations B 3.9.3 BASES SURVEILLANCE REQUIREMENTS SR 3.9.3.1 This Surveillance demonstrates that each of the containment penetrations is in its required status. The LCO 3.9.3.c.2 status requirement, which requires penetrations to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System, can be verified by ensuring each required valve operator is capable of closing automatically if needed. This Surveillance does not require cycling of the valves since this is performed at the appropriate frequency in accordance with SR 3.9.3.2.The S'ir!eillaPce is performed every 7 ~C.~ :...,.-....-Insert 2 co- pto f el an4 ejeF ati-a s.- Thi s She-ae anc nuc -th- t.n* .lcsoG SR 3.9.3.2 This Surveillance demonstrates that each required containment purge supply and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. T-he-2A44e, et4~h -Insert 2-Ffe -e de{-as-s-em gmt]-9e l J The SR is modified by a Note stating that this Surveillance is not required to be met for valves in isolated penetrations.
The LCO provides the option to close penetrations in lieu of requiring automatic actuation capability.
REFERENCES
: 1. UFSAR, Section 14.2.1.5.Cook Nuclear Plant Unit 1 B3934Rvso o B 3.9.3-4 Revision No. 1 RHR and Coolant Circulation
-High Water Level B 3.9.4 BASES SURVEILLANCE REQUIREMENTS SR 3.9.4.1 This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core..Te-h teyie-ef'.
-Insert 2 ,l#-aursss~iatrc.-aie48h-lw-e~fm up.m.iagO-R:R-~m REFERENCES
: 1. UFSAR, Section 9.3.2.Cook Nuclear Plant Unit 1 B3944Rvso o B 3.9.4-4 Revision No. 0 RHR and Coolant Circulation
-Low Water Level B 3.9.5 BASES ACTIONS (continued)
B.2 If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation.
Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.
B.3, B.4. and B.5 If no RHR is in operation, the following actions must be taken: a. The equipment hatch must be closed and secured with four bolts;b. One door in each air lock must be closed; and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System.With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions stated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.The Completion Time of 4 hours allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. Tr-he 4.= Insert 2m Cook Nuclear Plant Unit 1 ..- evso o B3.9.5-3 Revision No. 0 RHR and Coolant Circulation
-Low Water Level B 3.9.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to the required pump. T4he- e-r f-7 e~e~eeidereed-r-easebie~ar3le-iv=-
Insert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
REFERENCES
: 1. UFSAR, Section 9.3.2.Cook Nuclear Plant Unit 1 ..- evso o B3.9.5-4 Revision No. 0 Refueling Cavity Water Level B 3.9.6 BASES ACTIONS A.1 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.SURVEILLANCE REQ U IREM ENTS SR 3.9.6.1 Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 1).Z~eFrelec--f-.
rsist~-ete-aiera-ugeta4
~ ~ ~=Insert 2I REFERENCES
: 1. UFSAR, Section 14.2.1.2. 10CFR 100.10.Cook Nuclear Plant Unit 1 B 3.9.6-2 Revision No. 26 Enclosure 7 to AEP-NRC-2015-46 CNP Unit 2 TS Bases Pages Marked to Show Proposed Changes SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)
: b. Bank position;c. RCS average temperature;
: d. Fuel burnup based on gross thermal energy generation;
: e. Xenon concentration;
: f. Samarium concentration;
: g. Isothermal temperature coefficient (ITC); and h. Boron penalty (MODES 4 and 5 only).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. The boron penalty must be applied in MODES 4 and 5 since all reactor coolant pumps may be stopped in these MODES.This extra amount of boron ensures that minimum response times are met for the operator to diagnose and mitigate an inadvertent boron dilution event prior to loss of SDM.T" o e--u-ic of 24-sour5-c4baacd on .--=lnsert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Chapter 14.3. UFSAR, Section 14.2.5.4. UFSAR, Section 14.1.5.5. 10OCFR 100.Cook Nuclear Plant Unit 2 B3115Rvso o B 3.1.1-5 Revision No. 0 Core Reactivity B 3.1.2 BASES ACTIONS (continued)
B.1I If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE in which, the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours. If the SDM for MODE 3 is not met, then the boration required by SR 3.1.1.1 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 unit systems.SURVEILLANCE REQUIRE MENTS SR 3.1.2.1 Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations.
The comparison is made, considering that other core conditions are fixed or stable, including RCS boron concentration, control rod position, RCS average temperature,.fuel burnup based on gross thermal energy generation, xenon concentration, and samarium concentration.
The Surveillance is performed prior to entering MODE I 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 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.
T-he Insert 2 re4el db~ 3 EfPD fcwiig-tbeiinti  after-eatoring MODE 1, icQs n he-slowl~srateofre
'(Qrf, AF, t.) f -rptft4aeleat-ief--efe~e~aey.
REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Chapter 14.Cook Nuclear. Plant Unit 2 B3125Rvso o B3.1.2-5 Revision No. 0 Rod Group Alignment Limits B 3.1.4 BASES ACTIONS (continued) 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 in which the LCO requirements are not applicable.
To achieve this status, the unit must be brought to at least MODE 3 within 6 hours.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 unit systems.SURVEILLANCE SR 3.1.4.1 REQ U IREM ENTS'Vf ~ {
Insert 2 iseeaa4euheity-pq~exlae-te4-h-e-&aewr4a~
e.-eeeattef SR 3.1.4.2 Verifying each control rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2, tripping each control rod would result in radial or axial power tilts, or oscillations.
Exercising each individual control rod ayes- provides increased confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each control rod by 8 steps will not cause radial or axial power tilts, or oscillations, to occur. TFhe- .,-.- Insert 2rrnte-e-s e ~ r f~~ e Between required performances of SR 3.1.4.2 (determination of control rod OPERABILITY by movement), if a control rod(s) is discovered to be immovable, but remains trippable the control rod(s) is considered to be OPERABLE.
At any time, if a control rod(s) is immovable, a determination of the trippability (OPERABILITY) of the control rod(s) must be made, and appropriate action taken..Cook Nuclear Plant Unit 2 B3147Rvso o B 3.1.4-7 Revision No. 0 Shutdown Bank Insertion Limits B 3.1.5 BASES SURVEILLANCE REQUIREMENTS SR 3.1.5.1 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.Sio tho s iah e r.a--Relea-fsalfuet-a~
--~4 REFERENCES
: 1. UFSAR, Section 1.4.2.2. UFSAR, Section 1.4.5.3. UFSAR, Section 1.4.6.4. 10 CFR 50.46.5. UFSAR, Chapter 14.Insert 2 Cook Nuclear Plant Unit 2 B3154Rvso o B 3.1.5-4 Revision No. 0 Control Bank Insertion Limits B 3.1.6 BASES SURVEILLANCE REQUIREMENTS SR 3.1.6.1 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. Verifying the ECP calculation within 4 hours 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 Insert 2 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 CO LR.
Insert 2 ,cbee4a-S6,2.
REFERENCES
: 1. UFSAR, Section 1.4.2.2. UFSAR, Section 1.4.5.3. UFSAR, Section 1.4.6.4. 10 CFR 50.46.5. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2 B3165Rvso o B 3.1,6-5 Revision No. 1 PHYSICS TESTS Exceptions
-MODE 2 B 3.1.8 BASES ACTIONS (continued) 531 &deg;F could violate the assumptions for accidents analyzed in the safety analyses.D. 1 If the Required Action and associated Completion Time of Condition C is not met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit 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 unit systems.SURVEILLANCE REQUIRE MENTS SR 3.1.8.1 Verification that the RCS lowest loop Tavg is > 531&deg;0 F will ensure that the unit is not operating in a condition that could invalidate the safety analyses.
Verification of the RCS temperature at a Frequency of 30 minutes during the performance of the PHYSICS TESTS will ensure that the initial conditions of the safety analyses are not violated.SR 3.1.8.2 Verification that the THERMAL POWER is _< 5% RTP will ensure that the unit is not operating in a condition that could invalidate the safety analyses.
-"ictc o h T ER O"Rc c-FrPtueqiey~f 2 4-a4:ij~lealeee-,
SR 3.1.8.3 The SDM is verified by performing a reactivity balance calculation, considering the following reactivity effects: a. RCS boron concentration;
: b. Bank position;c. RCS average temperature;
: d. Fuel burnup based on gross thermal energy generation;
: e. Xenon concentration; Cook Nuclear Plant Unit 2 B3186Rvso o B3.1.8-6 Revision No. 0 PHYSICS TESTS Exceptions
-MODE 2 B 3.1.8 BASES SURVEILLANCE REQUIREMENTS (continued)
: f. Samarium concentration;
: g. Isothermal temperature coefficient (ITC), when below the point of adding heat (POAH);h. Moderator Temperature Defect, when above the POAH; and i. Doppler Defect, when above the POAH.Using the ITC accounts for Doppler reactivity in this calculation when the reactor is subcritical or critical but below the POAH, and the fuel temperature will be changing at the same rate as the RCS.-Th seaeF4ega~~ei I nsert 2 reqttiI'ed4ber-efi-eeneeontatiea-e "e~wp itie oeeeuwr4I~wileu4he-r.eqEi,44ed-SDM.
REFERENCES
: 1. 10 CFR 50, Appendix B, Section Xl.2. 10 CFR 50.59.3. Regulatory Guide 1.68, Revision 2, August, 1978.4. ANSI/ANS-19.6.1-1997, August 22, 1997.5. WCAP-1 3360-P-A, "Westinghouse Dynamic Rod Worth Measurement Technique," Revision 1, October 1998.6. PA-OSC-0061, "Westinghouse Position Paper on Power Distribution Measurement Requirements for Reload Startup Programs," February 2005.Cook Nuclear Plant Unit 2 B3187Rvso o B3.1.8-7 Revision No. 1 FQ(Z)B 3.2.1 BASES ACTIONS (continued)
C. I If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE or condition in which the LCO requirements are not applicable.
This is done by placing the unit in at least MODE 2 within 6 hours.This allowed Completion Time is reasonable based on operating experience regarding the amount of time it takes to reach MODE 2 from full power operation in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.2.1.1 REQUI REM ENTS Verification that FC(z) is within its specified limits involves increasing FM(z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain F&(Z). is then compared to its specified limits.If THERMAL POWER has been increased by > 10% RTP since the last determination of FC~(z), another evaluation of this factor is required 24 hours after achieving equilibrium conditions at this higher power level (to ensure that FC~(z) values are being reduced sufficiently with power increase to stay within the LCO limits). The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification.
It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ(Z) was last measured.
----Insert 2
~ ~ e e-n~l eet-re~hrHe~a-s-e~~daaerlrcw
~ ~ S-p i isan'ifr=-T--p SR 3.2.1.1 is modified by a Note, which applies during power escalation after a refueling.
The Note states that the Surveillance is not required to be performed until 24 hours after equilibrium conditions at a power level for extended operation are achieved.
This Note allows the unit to startup from a refueling outage and reach the power level for extended operation (normally 100% RTP) prior to requiring performance of the SR. Within 24 hours after equilibrium conditions are reached at the power level for extended operation, the SR must be performed.
Cook Nuclear Plant Unit 2 B 3.2.1-6 Revision No. 0 Cook Nuclear Plant Unit 2 B3.2.1-6 Revision No. 0 FQ(Z)B 3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The Frequency condition is not intended to require verification of these parameters after every 10% increase in power level above the last verification.
It only requires verification after a power level is achieved for extended operation that is 10% higher than that power at which FQ(Z) was last measured.The
~~eltetor r
2
---eG~i ys4elaet-eiert~--as-fe a~.-~e-trlae wth4e-S eki4 4 SR 3.2.1.2 is modified by Note 1, which applies during power escalation after a refueling.
The Note states that the Surveillance is not required to be performed until 24 hours after equilibrium conditions at a power level for extended operation are achieved.
This Note allows the unit to startup from a refueling outage and reach the power level for extended operation (normally 100% RTP) prior to requiring performance of the SR. Within 24 hours after equilibrium conditions are reached at the power level for extended operation, the SR must be performed.
REFERENCES
: 1. 10 CFR 50.46.2. UFSAR, Section 14.2.6.1.2.
: 3. .UFSAR, Section 1.4.5.4. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.5. WCAP-1 0216-P-A, Rev. tA, "Relaxation of Constant Axial Offset Control (and) F 0 Surveillance Technical Specification," February 1994.Cook Nuclear Plant Unit 2 B3218Rvso o B 3.2.1-8 Revision No. 0 FNH B 3.2.2 BASES ACTIONS (continued).A.4 Verification that FNAH is within its specified limits after an out of limit occurrence ensures that the cause that led to the FNAH exceeding its limit is corrected, and that subsequent operation proceeds within the LCOG limit. This Action demonstrates that the FNH limit is within the LOG limits prior to exceeding 50% RTP, again prior- to exceeding 75% IRTP, and within 24 hours after THERMAL POWER is > 95% IRTP.This Required Action is modified by a Note that states that THERMAL POWER does not have to be reduced prior to performing this Action.B.1 When any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCOG requirements are not applicable.
This is done by placing the unit in at least MODE 2 within 6 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience regarding the time required to reach MODE 2 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SIR 3.2.2.1 REQUIREMENTS The value of FN~H is determined by using the movable incore detector system to obtain a flux distribution map. A data reduction computer program then calculates the maximum value of FNAH from the measured flux distributions.
The measured value of FNH must be multiplied by 1.04 to account for measurement uncertainty before making comparisons to the FNAH limit.After each refueling, must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that FNH limits are met at the beginning of each fuel cycle.
atte--pe-wer-dist-ibti{+hrn 4-Insert 2 4-s-rqt~eyi y-s1e-esehr-t-49ght-at-t49e-F{-eait~aet-b-xeeeded-f f-elper-a-iem-REFERENCES
: 1. UFSAIR, Section 14.2.6.1.2.
: 2. UFSAIR, Section 1.4.5.3. 10 CFIR 50.46.Cook Nuclear Plant Unit 2B3225ReionN.4 B 3.2.2-5 Revision No. 34 AFD B 3.2.3 BASES SURVEILLANCE REQUIREMENTS SR 3.2.3.1 This Surveillance verifies that the AFD as indicated by the NIS excore channels is within the target band.
td~ys "*=--lnsert 2-by-t9e-pree-e-c-emputer.-Fur-term~er-er
,---dev~aieon f~e-AFBfrom t~he4arg et-baald-t~et--i-r~et-aer-lmed-s4heute-be SR 3.2.3.2-===lInsert 2-c-h aaeje~t--ay-ec-c-uf-i4e4gret-1dfu--lf4er-eesT~
ataper~ed-eue4e SR 3.2.3.3 Measurement of the target flux difference is accomplished by taking a flux map when the core is at equilibrium xenon conditions, preferably at high power levels with the control banks nearly withdrawn.
This flux map provides the equilibrium xenon axial power distribution from which the target value can be determined.
The target flux difference varies slowly with core burnup.---re~e.Je----4 Insert 2.,--a+~ur t fr-f A Note modifies this SR to allow the predicted beginning of cycle AFD from the cycle nuclear design to be used to determine the initial target flux difference after each refueling.
REFERENCES
: 1. WCAP-8385 (Westinghouse proprietary) and WCAP-.8403 (nonproprietary), "Power Distribution Control and Load Following Procedures,'
Westinghouse Electric Corporation, September 1974.2. UFSAR, Section 7.4.Cook Nuclear Plant Unit 2 B3236Rvso o B 3.2.3-6 Revision No. 1 QPTR B 3.2.4 BASES ACTIONS (continued)
Action A.5). The intent of this Note is to have the peaking factor Surveillances performed at operating power levels, which can only be accomplished after the excore detectors are normalized to restore QPTR to within limits and the core returned to power.B.1 If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE or other specified condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to < 50% RTP within 4 hours. The allowed Completion Time of 4 hours is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging unit systems.SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is< 75% RTP and the input from one Power Range Neutron Flux channel is inoperable.
Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4.1.This Surveillance verifies that the QPTR, as indicated by the Nuclear Instrumentation System (NIS) excore channels, is within its limits. Insert 2* -eu~ye--tyse4s-~ -eetehr~e For those causes of QPT that occur quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is not required until 12 hours after the input from one or more Power Range Neutron Flux channels are inoperable and the THERMAL POWER is > 75% RTP.With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded.
Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some quadrants is decreased.
Pe~feFmiag-GR--3.2?.4 7
Insert 2.e---ersp~vasaae~t-~tmtv-en-e-~uighf-~
Cook Nuclear Plant Unit 2 B3245Rvso o B3.2.4-5 Revision No. 0 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.1.1 Performance of the CHANNEL CHECK once every 12 hours 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.
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.T s 21-rsl~ied-eheffe~el.
SR 3.3.1.2 SR 3.3.1.2 compares the calorimetric heat balance calculation to the NIS channel output every 24 hours. If the calorimetric exceeds the NIS channel output by > 2% RTP, the NIS is not declared inoperable, but must be adjusted.
If the NIS channel output cannot be properly adjusted, the channel is declared inoperable.
Two Notes modify SR 3.3.1.2. The first Note indicates that the NIS channel output shall be adjusted consistent with the calorimetric results if the absolute difference between the NIS channel output and the calorimetric is > 2% RTP. The second Note clarifies that this Surveillance is required only if reactor power is -> 15% RTP and that 12 hours is allowed for performing the first Surveillance after reaching 15% RTP. At lower power levels, calorimetric data are inaccurate.
<--- Insert 2 drs-ael-exceeds-2%
ii ny-24-hetw-perod.-~-
Cook Nuclear Plant Unit 2 B3313 eiinN.1 B 3.3.1:38 Revision No. 17 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) 4Rtiea ieiasa44-alae-detec-t deviaie e-i9-ehe9ael--ei~tlit5.
SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output evefy'.,S1=fP If the absolute difference is > 3%, the NIS channel is still OPERABLE, but must be readjusted."If the 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.Two Notes modify SR 3.3.1.3. Note 1 indicates that the excore NIS channel shall be adjusted if the absolute difference between the incore and excore AFD is > 3%. Note 2 clarifies that the Surveillance is required only if reactor power is > 15% RTP and that 24 hours is allowed for performing the first Surveillance after reaching 15% RTP.T4 re~
e-a, 4-Insert 2 eie---aget
~ec eeera8atrue tya t uSR 3.3.1.4 SR 3.3.1.4 is the performance of a TADOT~ee-vy=4a-y62--eai-a.
'SAGE- E3FBA1S 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 RTB undervoltage 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 3.3.1.17.
The bypass breaker test shall include a local shunt trip. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service.Cook Nuclear Plant Unit 2 B3313 eiinN.1 B 3.3.1-39 Revision No. 17 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
T~eF--~Ray-m-%A---S~
4--Insert 2
SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. The SSPS is tested ev ~ y-r.a8 using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
'T~e 02id-y3 Insert 2 SR 3.3.1.6 SR 3.3.1.6 is the performance of a TADOT and is performed every 92 days on a STAGGERED TEST BASIS. This test applies to the SI Input from ESFAS 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 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.Ie RE~-TEgTs-
,--I nsert 2 jii-trhe~n--Refefeine=1 SR 3.3.1.7 SR 3.3.1.7 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 the f(AI)input to the Overtemperature AT Function.A Note modifies SR 3.3.1.7. The Note states that this Surveillance is required only if reactor power is > 50% RTP and that 24 hours is allowed for performing the first surveillance after reaching 50% RTP.e ~ -
_ Insert 2 Coox Nuclear Plat Unit 2 B 3.3.-40 Revision Nto~d t .1 Cook Nuclear Plant Unit 2 B 3.3.1-40 Revision No. 17 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUI REM ENTS (continued)
SR 3.3.1.8 SR 3.3.1.8 is the performance of a COT every 92 days.A COT is performed on each required channel to ensure the entire 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 COT 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.The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 8.SR 3.3.1.8 is modified by a Note that provides a 12 hour delay in the requirement to perform this Surveillance for Function 2.b channels after reducing THERMAL POWER below the P-10 interlock.
The Frequency of 12 hours after reducing power below P-10 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.
-Insert 2 SR 3.3.1.9CHAN NEL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
Cook Nuclear Plant Unit 2 B 3.3.1-41 Revision No. 16 Cook Nuclear Plant Unit 2 B 3.3.1-41 Revision No. 16 RTS Instrumentation
*B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
P-te-asu'+~ea f ---Insert 2 This SR is modified by a Note that states that neutron detectors are excluded from the CHANNEL CALIBRATION.
Changes in power range neutron detector sensitivity are compensated for by normalization of the channel output based on a power calorimetric and flux map performed above 15% RTP (SR 3.3.1.2).SR 3.3.1.10 SR 3.3.1.10 is the performance of a TADOT and <,---Insert 2A 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 SR is modified by a Note that excludes verification of relay setpoints from the TAD OT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION.
The Frequency of 92 days is justified in Reference 10.SR 3.3.1.11 SR 3.3.1.11 is the performance of a COT e A COT is performed on each required channel to ensure the entire 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 COT 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.The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.
Cook Nuclear Plant Unit 2 B 3.3.1-42 Revision No. 16 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 8.The Frequency is modified by two Notes. Note 1 provides a 12 hour delay in the requirement to perform this Surveillance for intermediate range instrumentation after reducing THERMAL POWER below the P-I10 interlock.
The Frequency of 12 hours after reducing power below P-10 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.
Note 2 provides a 4 hour delay in the requirement to perform this Surveillance for source range instrumentation after THERMAL POWER is reduced below the P-6 interlock.
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 RTBs are open and SR 3.3.1.11 is no longer required to be performed.
If the unit is to be in MODE 3 with the RTBs closed for > 4 hours this Surveillance must be performed prior to 4 hours after THERMAL POWER is reduced below the P-6 interlock.
--& Insert 2 SR 3.3.1.12 CHAN NEL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
* T h --e f e e -f l s I a e L t e a s m ~ ~ e -m l 8 e I n s e rt 2 SR 3.3.1.13 CHAN NEL 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.Cook Nuclear Plant Unit 2 B3314 eiinN.1 B 3.3.1-43 Revision No. 16 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.-t
-Insert 2 SR 3.3.1.14 SR 3.3.1.14 is the performance of a CHANNEL CALlBRATION~everyi-
.2--ment4he.
CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The CHANNEL CALIBRATION for the source range neutron detectors also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION.
Changes in power range neutron detector sensitivity are compensated for by normalization of the channel output based on a power calorimetric and flux map performed above 15% RTP (SR 3.3.1.2).Changes in intermediate range neutron flux detector sensitivity are compensated for by periodically evaluating the compensating voltage setting and making adjustments as necessary.
Changes in source range neutron detector sensitivity are compensated for by periodically obtaining the detector plateau or preamp discriminator curves, evaluating those curves, comparing the curves to the manufacturer's data, and adjusting the channel output as necessary.--Insert 2 SR 3.3.1.15 SR 3.3.1.15 is the performance of a CHANNEL CALIBRATION, as described in SR 3.3.1.13, .ever Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the resistance temperature detectors (RTD) sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.CookNucearPlan Unt 2B 3..1-4 Rvisin N. 1 Cook Nuclear Plant Unit 2 B 3.3.1-44 Revision No. 16 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
-Insert 2 This SR is modified by a Note that provides a 72 hour delay in the requirement to perform a normalization of the AT channels after THERMAL POWER is > 98% RTP. The intent of this Note is to maintain reactor power at a nominal 97% RTP to 98% RTP level until the AT normalization is complete before increasing reactor power to 100% RTP.SR 3.3.1.16 SR 3.3.1.16 is the performance of a COT of RTS interlocks every--r4-ner~he.
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 COT 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.
* h-e~ry tekea-elt-f4ere~e--~eh
<"-I nsert 2~ em-h~4erseetqe a e e-t re e t~@ -xj 4e e.SR 3.3.1.17 SR 3.3.1.17 is the performance of aTADOT of the Manual Reactor Trip (including reactor trip bypass breakers) and RCP Breaker Position.
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 test shall independently verify the OPERABILITY of the undervoltage and shunt trip mechanisms for the Manual Reactor Trip Function for the Reactor Trip Breakers and Reactor Trip Bypass Breakers.
The Reactor Trip Bypass Breaker test shall include testing of the automatic undervoltage trip.Cook Nuclear Plant Unit 2 B3314 eiinN.1 B 3.3.1-45 Revision No. 16 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
The-F- ee-sIas~-nekawnreaiityeJ eF s -Insert 2 aec-efftaee4-ue-e fper-at~r-&sect;-ex-perier1ee7.
SR 3.3.1.18 SR 3.3.1.18 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 as described in SR 3.3.1.4, except that this test is performed prior to exceeding the P-8 interlock whenever the unit has been in MODE 3. This Surveillance is not required if it has been I 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-8 interlock.
SR 3.3.1.19 SR 3.3.1.19 verifies that the individual channel/train actuation response times are less than or equal to the maximum values assumed in the accident analysis.
Response time testing acceptance criteria are included in UFSAR, Table 7.2-6 (Ref. 12). 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 to the point at which the equipment reaches the required functional state (i.e., control and shutdown rods fully inserted in the reactor core).For channels that include dynamic transfer Functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer Function set to one, with the resulting measured response time compared to the appropriate UFSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value, provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.Cook Nuclear Plant Unit 2 B 3.3.1-46 Revision No. 18 Cook Nuclear Plant Unit 2 B 3.3.1-46 Revision No. 18 RTS Instrumentation B 3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Response time may be verified by actual response time tests in any series of sequential, 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) in place, 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," (Ref. 13) 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, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 14) provides the basis and methodology for 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 the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
4e~ems~-e~i
-Insert 2-h4at ee4te~vceqqcaeliateete w4 th ~ nft-sela-~~-tPetacewe-ef -tth&#xa3; ee-va-eue~~
ac-eept-able-ff~-are{9ai~t-y-etea9dj eiat SR 3.3.1.19 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.The response time testing of the neutron flux signal portion of the channel shall be measured from either the detector output or the input of the first electronic component in the channel.Cook Nuclear Plant Unit 2 B 3.3.1-47 Revision No. 5 Cook Nuclear Plant Unit 2 B 3.3.1-47 Revision No. 5 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
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.Insert 2 SR 3.3.2.2 and SR 3.3.2.5 SR 3.3.2.2 is the performance of a
,-a This test is a check of the Loss of Voltage Function.
SR 3.3.2.5 is the performance of a test is a check of the Undervoltage ROP 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 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.
Each SR is modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and are verified during CHANNEL CALIBRATION. "T Frg,,nc o-f-- 3,3.2.2---
Insert 2 0 I h'noe ru nnr inorhln Arnn~rfIn rv-nerlodrinri.SR 3.3.2.5 is juctificd in Rcfrnc0 SR 3.3.2.3 SR 3.3.2.3 is the performance of an ACTUATION LOGIC TEST. --T-he-." apgy-c tesaede$eT-9 da, using the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the master relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERABLE and that there is an intact voltage signal path to the master relay coils. Th rouny-f--r 2daso Insert 2
! O.Cook Nuclear Plant Unit 2B33237RvsoN.0 B 3.3.2-37 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.4 SR 3.3.2.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. "T-is--test-le GERa TT 30I. Te Insert 2 Ref e'rellee-fr SR 3.3.2.6 SR 3.3.2.6 is the performance of a COT. A COT is performed on each required channel to ensure the entire channel will perform the intended Function.
Setpoints must be found within the Allowable Values specified in Table 3.3.1-1. 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 COT 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 difference between the current "as found" values .and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
The setpoint shall be left set consistent with the assumptions of the current Unit specific setpoint methodology.
The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 6.nserti 2 SR 3.3.2.6 is modified by a Note which applies to the SI Cdntainment Pressure -High, Containment Spray Containment Pressure -High High, Phase B Isolation Containment Pressure -High High, Steam Line Isolation Containment Pressure -High High, and CEQ System Containment Pressure -High Functions.
This Note requires, during the performance of SR 3.3.2.6, the associated transmitters of these Functions to be exercised by applying either a vacuum or pressure to the appropriate side of the transmitter.
Exercising the associated transmitters during the performance of the COT is necessary to ensure Functions 1 .c, 2.c, 3.b.(3), 4.c, and 7.c remain OPERABLE between each CHANNEL CALIBRATION.
Cook Nuclear Plant Unit 2B332-8RvsoN.0 B 3.3.2-38 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.2.7 SR 3.3.2.7 is the performance of a CHANNEL CALIBRATION.:h-A--G=AN--A-BATINispF~e CHAN NEL 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 unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.-Th--~ee
.Insert 2-a ~ti-h-e~iFtIe~e l v-SR 3.3.2.8 SR 3.3.2.8 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function, or is placed in a condition where the relay contact operation can be verified without operation of the equipment.
Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.
.Insert 2-and-er-atg-hI~stof-at SR 3.3.2.9 SR 3.3.2.9 is the performance of a TADOT. This test is a check of the Manual Initiation Functions, the AFW pump start on trip of all MFW pumps, and the P-4 interlock.
It is pcrorc ....r, ..... month,,,,.
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 Cook Nuclear Plant Unit 2 B 3.3.2-39 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.3.2-39 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) extensions.
In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.). Th-eqeeyi-aeeEtefat l-'nsert 2 SR 3.3.2.10 SR 3.3.2.10 is the performance of a CHANNEL CALIBRATION. -A-erord 4-- eI-tths. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy.CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific setpoint methodology.
The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.
TheF a4hsit~eit-ethedete#.
SR 3.3.2.11 SR 3.3.2.11 is the performance of an ACTUATION LOGIC TEST. This SR is applied to the balance of plant actuation logic and relays that do not have the SSPS test circuits installed to utilize the semiautomatic tester or perform the continuity check. All possible logic combinations are tested for Table 3.3.2-1 Functions 6.e and 6.g. TCs-refee-e-4mrth4~ r- --
* retie1ti-y--a SR 3.3.2.12 This SR ensures the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis.Response Time testing acceptance criteria are included in the UFSAR, Table 7.2-7 (Ref. 11). 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, to the point at which the equipment in both trains reaches the required functional state (e.g., pumps at rated discharge pressure, valves in full open or closed position).
Cook Nuclear Plant Unit 2 B 3.3.2-40 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.3.2-40 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued)
For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer functions set to one with the resulting measured response time compared to the appropriate UFSAR response time. Alternlately, the response time test can be performed with the time constants set to their nominal value provided the required response time is analytically calculated assuming the time constants are set at their nominal values. The response time may be measured by a series of overlapping tests such that the entire response time is measured.Response time may be verified by actual response time tests in any series of sequential, 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) in place, 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," dated January 1996 (Ref. 12), 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, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," (Ref. 13) provides the basis and methodology for 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 the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.
_N_--_____-__-___te -eer,- nsert 2 S8TA CLEDT he oehef'~-__-he-f'on-v
.~a _J.--e-f ~~t e Cook Nuclear Plant Unit 2 B 3.3.2-41 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.3.2-41 Revision No. 0 ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE REQUIREMENTS (continued) efa4~mn s--m This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours after reaching 850 psig in the SGs.REFERENCES
: 1. Technical Requirements Manual.2. IEEE-279, "Proposed Criteria for Nuclear Power Plant Protection Systems," August 1968.3. UFSAR, Table 7.2-1.4. UFSAR, Table 14.1.0-4.5. 10 CFR 50.49.6. WCAP-12741, "Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.7. UFSAR, Chapter 14.8. WCAP-14333-P-A, Revision 1, October 1998.9. WCAP-1 0271-P-A, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," including Supplement 1, May 1986, and Supplement 2, Rev. 1, June 1990.10. WCAP-1 5376, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Intervals and Reactor Trip Breaker Test and Completion Times," October 2000.11. UFSAR, Table 7.2-7.12. WCAP-1 3632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996.13. WCAP-14036-P, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," December 1995.Cook Nuclear Plant Unit 2B332-2RvsoN.0 B 3.3.2-42 Revision No. 0 PAM Instrumentation B 3.3.3 SURVEILLANCE As noted at the beginning of the SRs, the following SRs apply to each REQUIREMENTS PAM instrumentation Function in-Table 3.3.3-1, except where identified in the SR.SR 3.3.3.1 Performance of the CHANNEL CHECK @e ieevy--31-day~s.ensures that a gross instrumentation failure has not occurred.
A CHANNEL CHECK is BASES sURvEILLANCE REQUIREMENTS (continued) 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
The Containment Area Radiation (High Range)instrumentation should be compared to similar unit instruments located throughout the unit. When only one channel of the Reactor Coolant Inventory Tracking System is OPERABLE, the RCS Subcooling Margin Monitor and Core Exit Temperature channels may be used for performance of the CHANNEL CHECK of the OPERABLE Reactor Coolant Inventory Tracking System channel.Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, 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. If the channels are within the criteria, it is an indication that the channels are OPERABLE.As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.
~ B-ee-Ng-e#
2 K c-heauaels..
Cook Nuclear Plant Unit 2 B3331 eiinN.1 B3.3.3-13 Revision No. 16 PAM Instrumentation B 3.3.3 BASES SURVEILLANCE REQUIREMENTS (continued).SR 3.3.3.2 .Deleted SR 3.3.3.3 A C NN pe~er r-r--4-Ph
: e. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy.
This SR is modified by a Note that excludes neutron detectors.
For Function 9, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual valve position.
For Functions 15, 16, 17, and 18, whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit Temperature thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing elements.For Functions 20 (Circuit Breaker Status channels) and 24, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual circuit breaker position..T-he-24--mie~h 2* F-r e~ gexp~~.I REFERENCES 1.NRC letter, T. G. Colburn (NRC) to M. P. Alexich (Indiana Michigan Power Company), 'Emergency Response Capability
-Conformance to Regulatory Guide 1.97 Revision 3 for the D. C. Cook Nuclear Plant, Units 1 and 2," dated December 14, 1990.2. UFSAR, Table 7.8-1.3. Regulatory Guide 1.97, Revision 3, May 1983.4. NUREG-0737, Supplement 1, "TMI Action Items." 5. NRC letter, P.S.Tam (NRC), to M. K. Nazar, (Indiana Michigan Power Company), "Donald C. Cook Nuclear Plant, Units 1 & 2 (DCCNP-1 and DCCNP-2) -Issuance of Amendments Re: Containment Sump Modifications per Generic Letter 2004-02 (TAC Nos. MD5901 and MD5902)," dated October 18, 2007.Cook Nuclear Plant Unit 2 B3331 eiinN.2 B 3.3.3-14 Revision No. 29 Remote Shutdown Monitoring Instrumentation B 3.3.4 BASES ACTIONS (continued)
Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.A.1 Condition A addresses the situation where one or more required Functions of the remote shutdown monitoring instrumentation are inoperable.
The Required Action is to restore the required Function to OPERABLE status within 30 days. The Completion Time is based on operating experience and the low probability of an event that would require evacuation of the control room.B.1 and B.2 If the Required Action and associated Completion Time of Condition A is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.3.4.1 REQUIREMENTS Performance of the CHANNEL CHECK' ree-evty&~4ys ensures that a 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. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability.
If the channels are within the criteria, it is an indication that the channels are OPERABLE.
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.Cook Nuclear Plant Unit 2 B3343Rvso o B 3.3.4-3 Revision No. 0 Remote Shutdown Monitoring Instrumentation B 3.3.4 BASES SURVEILLANCE REQUIREMENTS (continued)
As specified in the Surveillance, a CHANNEL CHECK is only required for those channels which are normally energized.
Trhe-F-r 4ass~se.e~
2 d SR 3.3.4.2 CHANNEL CALIBRATION is a complete~check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.For the Reactor Trip Breaker Indication Function on the hot shutdown panel, the CHANNEL CALIBRATION shall consist of verifying that the position indication conforms to actual reactor trip breaker position.~FFte=F1~e~uoF-" ~f 24 i-~ ~ci tI ,s ie-baee~-eI9-ej3eiat419pe~ieI~ee.
"*-l'nsert 2 REFERENCES.
: 1. UFSAR, Section 1.4.3.Cook Nuclear Plant Unit 2 B3344Rvso o B 3.3.4-4 Revision No. 0 LOP DG Start Instrumentation B 3.3.5 BASEIS ACTIONS (continued) made inoperable by failure of the LOP DG start instrumentation are required to be entered immediately.
The actions of those LCOs provide for adequate compensatory actions to assure unit safety.SURVEILLANCE REQUIREMENTS SR 3.3.5.1.Performance of the CHANNEL CHECK enc-e-ever--t2--heur-ensures that a 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.
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.
le-a~je1~e-e e4"a ~ {t 2 SR 3.3.5.2 SR 3.3.5.2 is the performance of a TADOT. 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.
* e -lra4eZ ,~ The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment.
The SRs are modified by a Note that excludes verification of setpoints for relays. Relay setpoints require elaborate bench calibration and are verified during CHANNEL CALIBRATION.
T-he-Preq~e-ne-is-4 tbd idcts he ra~eaaMLtaa
--w~~ea~-asae~hw4 aee-epta .l]e419gt-~ef~
atip-eferie rtee.---Insert 2 Cook Nuclear Plant Unit 2 B3355Rvso o B 3.3.5-5 Revision No. 0 LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.5.3 SR 3.3.5.3 is the performance of a CHANNEL CALIBRATION.
The setpoints, as well as the response to a loss of voltage and a degraded voltage test, shall include a single point verification that the trip occurs within the required time delay.
e99 Aery4 .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.~Insert 2 REFERENCES
: 1. UFSAR, Section 8.4.2. UFSAR, Section 8.5.3. UFSAR, Chapterl14.
: 4. WCAP-1 2741, "Westinghouse Menu Driven Setpoint Calculation Program (STEPIT)," as approved in Unit 1 and Unit 2 License Amendments 175 and 160, dated May 13, 1994.Cook Nuclear Plant Unit 2 B3356Rvso o B 3.3.5-6 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES ACTIONS (continued)
D.1 Condition D applies to all Containment Purge Supply and Exhaust System Isolation Functions.
If one or more Automatic Actuation Logic and Actuation Relays trains are inoperable, one or more SI Input from ESFAS trains are inoperable, two or more required radiation monitoring channels in a single train are inoperable, or the Required Action and associated Completion Time of Condition A, B, or C are rnot met, operation may continue provided the containment purge supply and exhaust isolation valves are placed in the closed position immediately.
Placing the containment purge supply and exhaust isolation valves in the closed position accomplishes the safety function of the inoperable trains or channels.SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1 REQUIREMENTS determines which SRs apply to which Containment Purge Supply and Exhaust System Isolation Instrumentation Functions.
SR 3.3.6.1 Performance of the CHANNEL CHECK ensures that a 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 the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
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.-Insert 2t~~-s--fe--
t b~-nf-r t --efea~i-d -i rm Cook Nuclear Plant Unit 2 B 3.3.6-6 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.3.6-6 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.6.2 SR 3.3.6.2 is the performance of an ACTUATION LOGIC TEST. The train being tested may be placed in the bypass condition, thus preventing actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, may be tested for each protection function.
In addition, the master relay coil may be pulse tested for continuity.
This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. This-t%'t- TEST l Insert 2 SR 3.3.6.3 SR 3.3.6.3 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity.
T-his-tes'-is- -Isr SR 3.3.6.4 A COT is performed ever-y-1-84-da-y on each required channel to ensure the entire 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 COT 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.
Hr-rqtee-is Insert 2-fsdet--tei*Jaee~ra~gexe-ece This test verifies the capability of the instrumentation to provide the Containment Purge Supply and Exhaust System isolation.
The setpoint shall be left consistent with the current unit specific calibration procedure tolerance.
SR 3.3.6.5 SR 3.3.6.5 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation mode is either allowed to function or is placed in a condition where the relay contact operation can be verified without Cook Nuclear Plant Unit 2 B3367Rvso o B3.3.6-7 Revision No. 0 Containment Purge Supply and Exhaust System Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE REQUIREMENTS (continued) operation of the equipment.
Actuation equipment that may not be operated in the design mitigation mode is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.
SR 3.3.6.6 SR 3.3.6.6 is the performance of a TADOT. This test is a check of the Manual Initiation Function Each Manual Initiation Function is tested up to, and including, the master relay coils. 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.
In some instances, the test includes actuation of the end device (i.e., valves cycle).The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no setpoints associated with it.l-nsert 2 I. fl~ -IIfl~i, t 9 LAL.* It,) I.) UI~L,~Sfl~
Li UUI**-i--nsert 2 I -.
I J ,l,. IiL Ib c I\ IU ho '..U Lnn L to .w b i I I LU II_, L ep~ee~at~ie~ee-,.
SR 3.3.6.7 rc ~r- 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.-fiienkC iu "z dze "" -It Insert 2 REFERENCES
: 1. UFSAR, Section 5.5.3.2. 10 CFR 100.11 3. WCAP-15376, Rev. 0, October 2000.Cook Nuclear Plant Unit 2 B3368Rvso o B3.3.6-8 Revision No. 0 CREV System Actuation Instrumentation B 3.3.7 BASES ACTIONS (continued) this Completion Time is the same as provided in LCO 3.7.10.B.1.1, B.1.2, and B.2 Condition B applies to the failure of two CREV System Automatic Actuation Logic and Actuation Relays trains in one or more required Functions.
The first Required Action is to place one CREV train in the pressurization/cleanup mode of operation immediately.
This accomplishes the actuation instrumentation Function that may have been lost and places the unit in a conservative mode of operation.
The applicable Conditions and Required Actions of LCO 3.7.10 must also be entered for the CREV train made inoperable by the inoperable actuation instrumentation.
This ensures appropriate limits are placed upon. train inoperability as discussed in the Bases for LCO 3.7.10.Alternatively, both trains may be placed in the pressurization/cleanup mode. This ensures the CREV System function is performed even in the presence of a single failure.C.1 and C.2 Condition C applies when the Required Action and associated Completion Time for Condition A or B have not been met. The unit must be brought to a MODE in which the LCO requirements are not applicable.
To achieve this status, the unit must be brought to MODE 3 within 6 hours and MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS A Note has been added to the SR Table to clarify that Table 3.3.7-1 determines which SRs apply to which CREV System Actuation Instrumentation Functions.
SR 3.3.7.1 SR 3.3.7.1 is the performance of an ACTUATION LOGIC TEST. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation.
Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function.
In addition, the mater relay coil is pulse tested for continuity.
This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relay coils. 1~iTT-h- 2* trvi~~
-4f d4--ee-ac4--
Cook Nuclear Plant Unit 2 B3.3.7-3 Revision No. 0 CREV System Actuation Instrumentation B 3.3.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.7.2 SR 3.3.7.2 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil[. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity.
T-hi&-t~e--t nset SR 3.3.7.3 SR 3.3.7.3 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function or is placed in a condition where the relay contact operation can be verified without operation of the equipment.
Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay.Taitsq~e~-aeee~-4mas cetat -e --taiitya -Insert 2 REFERENCES
: 1. WCAP-1 5376, Rev. 0, October 2000.Cook Nuclear Plant Unit 2 B3374Rvso o B3.3.7-4 Revision No. 0 BDMI B 3.3.8 BASES ACTIONS (continued)
As an alternate to restoring one channel to OPERABLE status within 1 hour (Required Action B.2.1). Required Action B.2.2.1 requires isolation valves for unborated water sources to the Chemical and Volume Control System to be secured to prevent the flow of unborated water into the RCS. In addition, in MODE 5, if the RWST boron concentration is< 2400 ppm and less than the Reactor Coolant System (RCS) boron concentration, the RWST is considered an unborated water source and is required to be isolated from the RCS. Once it is recognized that two source range neutron flux monitoring channels of the BDMI are inoperable, the operators will be aware of the possibility of a boron dilution, and the 1 hour Completion Time is adequate to complete the requirements of Required Action B.2.2. 1.Required Action B.2.2.2 accompanies Required Action B.2.2.1 to verify the SDM according to SR 3.1.1.1 within 1 hour and once per 12 hours thereafter.
This backup action is intended to confirm that no unintended boron dilution has occurred while the BDMI was inoperable, and that the required SDM has been maintained.
The specified Completion Time takes into consideration sufficient time for the initial determination of 8DM and other information available in the control room related to SDM.SURVEILLANCE SR 3.3.8.1 REQUIREMENTS Performance of the CHANNEL CHECK ef~ee-evieriy--24teire 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.
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.T-e
,,---I nsert 2 c ELiCl WCI( uap-_T-h ei4&e-fe r-al, ee~ee Cook Nuclear Plant Unit 2 B3383Rvso o B3.3.8-3 Revision No. 0 BDMI B 3.3.8 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.3.8.2 SR 3.3.8.2 is the performance of a CHANNEL CALIBRATION
'veiy-* CHANNEL CALIBRATION is a complete check of the instrument ioop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.
The CHANNEL CALIBRATION also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. This SR is modified by a Note that states that neutron detectors are excluded from the CHANNEL CALIBRATION.-Th-Ftl ~ re~e l'--nsert 2 REFERENCES
: 1. UFSAR, Section 14.1.5.Cook Nuclear Plant Unit 2 B3384Rvso o B3.3.8-4 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued)
POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP.. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive.
Also, since they represent trarnsients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.
ACTIONS A.1 With one or more of the RCS DNB parameters not within LCO limits, action must be taken to restore parameter(s) in order to restore DNB margin and eliminate the potential for violation of the accident analysis.The 2 hour Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.
B.1_If Required Action A.1 is not met within the associated Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 2 within 6 hours. In MODE 2, the reduced power condition eliminates the potential for violation of the accident analysis.
The Completion Time of 6 hours is reasonable to reach the required unit conditions in an orderly manner.SURVEILLANCE REQUIREMENTS SR 3.4.1.1r-eq ea edi f rdt eq 2-heu i I tt, .., I I r * ,- I I I .- i-Insert 2 SR 3.4.1.2;Ve4fca~ ,fa~
ti ~ -nge-- LRunu's-'~a'f'-r i dti"c " " 1l~ *d~y~-n e-'e ---e---~-~ ~ thsI nshw " -Isr pi-aecetio
-to b c~aereu ~
~ ~ ~
Cook Nuclear Plant Unit 2 B3413Rvso o B 3.4.1-3 Revision No. 0 RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.1.3 Verification that the RCS total flow rate is greater than or equal to the limits ensures that the initial condition of the safety analyses are met.Tae-l42-h -ati ~er.~4r-e 2 SR 3.4.1.4 Measurement of RCS total flow rate by performance of a precision calorimetric heat balance allows the installed RCS flow instrumentation to be calibrated and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.Thet=Ftequ -oenet'~ey=frrl24a.tle ew a1 Insert 2 ee4e--e-ac--e#
ep~ebe.This SR is modified by a Note that allows entry into MODE 1, without having performed the SR, and placement.
of the unit in the best condition for performing the SR. The Note states that the SR is not required to be performed until 24 hours after > 90% RTP. This exception is appropriate since the heat balance requires the unit to be at a minimum of 90% RTP to obtain the stated RCS flow accuracies.
The Surveillance shall be performed within 24 hours after reaching 90% RTP.REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2 B3414Rvso o B 3.4.1-4 Revision No. 0 RCS Minimum Temperature for Criticality B 3.4.2 BASES APPLI CABLE SAFETY ANALYSES (continued) criticality limitation provides a small band, 6&deg;F, for critical operation below HZP. This band allows critical operation below HZP during unit startup and does not adversely affect any safety analyses since the MTC is not significantly affected by the small temperature difference between HZP and the minimum temperature for criticality.
The RCS minimum temperature for criticality satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
LCO Compliance with the LCO ensures that the reactor will not be made or maintained critical (keff > 1.0) at a temperature less than a small band below the HZP temperature, which is assumed in the safety analysis.Failure to meet the requirements of this LCO may produce initial conditions inconsistent with the initial conditions assumed in the safety analysis.APPLICABILITY In MODE 1 and MODE 2 with keff> 1.0, LCO 3.4.2 is applicable since the reactor can only be critical (kerr > 1.0) in these MODES.ACTIONS A.1 If the parameters that are outside the limit cannot be restored, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to MODE 2 with keff < 1.0 within 30 minutes. Rapid reactor shutdown can be readily and practically achieved within a 30 minute period. The allowed time is reasonable, based on operating experience, to reach MODE 2 with keff < 1.0 in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.4.2.1 RCS loop average temperature is required to be verified at or above 5410 &deg;F e
*,-- Insert 24 e fl UT Thrra,.J,.., I LI,.JA I, ~ U~ V %..
REFERENCES
: 1. UFSAR, Section 14.1.1.Cook Nuclear Plant Unit 2 B3422Rvso o B 3.4.2-2 Revision No. 0 RCS P/T Limits B 3.4.3 BASES ACTIONS (continued)
Condition C is modified by a Note requiring Required Action C.2 to be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
SURVEILLANCE SR 3.4.3.1 REQ U IREM ENTS Verification that operation is within limits is required eve~ ia0=h~tes when RCS pressure and temperature conditions are undergoing planned changes. This qeaie~ey sc-~dre-rsre eavew-ef-t41eN~tfel
<-nsert 2 r-eeeef.abte-t4ine:
Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition give~n in the relevant plant procedure for ending the activity is satisfied.
This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.
REFERENCES
: 1. WCAP-15047, Rev. 2, dated May 2002.2. 10 CFR 50, Appendix G.3. ASME, Boiler and Pressure Vessel Code, Section I11, Appendix G.4. ASTM E 185-82, July 1982.5. 10 CFR 50, Appendix H.6. Regulatory Guide 1.99, Revision 2, May 1988.7. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.Cook Nuclear Plant Unit 2 B3436Rvso o B3.4.3-6 Revision No. 0 RCS Ldops -MODES 1 and 2 B 3.4.4 BASES APPLICABILITY (continued)
Operation in other MODES is covered by: LCO 3.4.5, "RCS Loops -MODE 3";LCO 3.4.6, "RCS Loops -MODE 4";LCO 3.4.7, "RCS Loops -MODE 5, Loops Filled";LCO 3.4.8, "RCS Loops -MODE 5, Loops Not Filled";LCO 3.9.4, "Residual Heat Removal (RHR) and Coolant Circulation
-High Water Level"; and LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation
-Low Water Level." ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the unit to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.4.4.1 This SR requires verification evei-y 4-2-heu~s that each RCS loop is in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal while maintaining the margin to the DNBR limit. T4he-Fr-eqel'aey-4f
--- Insert 2 I--- -~r r-- ----- --.... ... r" r- ..............
REFERENCES
: 1. UFSAR, Section 14.1.Cook Nuclear Plant Unit 2 B3443Rvso o B3.4.4-3 Revision No. 0 RCS Loops -MODE 3 B 3.4.5 BASES ACTIONS (continued) 0.1 If one required ROS loop is not in operation, and the Rod Control System is capable of rod withdrawal, the Required Action is to place the Rod Control System in a condition incapable of rod withdrawal (e.g., de-energize all CRDMs by opening the RTBs or de-energizing the motor generator (MG) sets). When the Rod Control System is capable of rod withdrawal, it is postulated that a power excursion could occur in the event of an inadvertent control rod bank withdrawal.
This mandates having the heat transfer capacity of two RCS loops in operation.
If only one loop is in operation, the Rod Control System must be rendered incapable of rod withdrawal.
The Completion Time of 1 hour to defeat the Rod Control System is adequate to perform these operations in an orderly manner without exposing the unit to risk for an undue time period.D.1. D.2. and D.3 If two required RCS loops are inoperable, or two required RCS loops are not in operation with Rod Control System capable of rod withdrawal, or required RCS loop not in operation with Rod Control System not capable of rod withdrawal, the Rod Control System must be placed in a condition incapable of rod withdrawal (e.g., all CRDMs must be de-energized by opening the RTBs or de-energizing the MG sets). All operations involving introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 must be suspended, and action to restore one of the RCS loops to OPERABLE status and operation must be initiated.
Boron dilution requires forced circulation for proper mixing, and opening the RTBs or de-energizing the MG sets removes the possibility of an inadvertent rod withdrawal.
Suspending operations that would cause the introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 is required to assure continued safe operation.
With coolant added without forced circulation, unmixed coolant could be introdtqced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification evey--12.-heutffs that the required loops are in operation.
Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.Cook Nuclear Plant Unit 2 B3454Rvso o B3.4.5-4 Revision No. 0 RCS Loops -MODE 3 B 3.4.5 BASES SURVEILLANCE REQUIREMENTS (continued) p h I I.1 r.1, ~ tA I ..J~I..J ..JL4 I~ ~
SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side water level is above the lower tap of the SG wide range level instrumentation by > 418.77 inches for required RCS loops. If the SG tubes become uncovered, the associated loop may not be capable of providing the heat sink for removal of the decay heat. The water level can be verified by either the wide range or the narrow range instruments.
A narrow range level instrument.
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. T-he--2-heuf-Fe~lt~eaey~s-eons-ider-ed-adlequiate I-nsert~ejleer-a3=r--te f-661e'te1'.
SR 3.4.5.3 II-[ -- r L-]-II Verification that each required RCP is OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat'removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power availability to each required RCP.Inser-t This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 2 B 3.4.5-5 Revision No. 0 RCS Loops -MODE 4 B 3.4.6 BASES ACTIONS (continued) minimum SDM maintains acceptable margin to subcritical operations.
The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.
SURVEILLANCE REQUIREMENTS SR 3.4.6.1 This SR requires verification evet,-24-2--eew that the required RCS or RHR loop is in operation and circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal. The-Peq1tey~ef rs a~ t-t --aqelr memrie-- q petfo-maiaee
.-."---nsert 2 SR 3.4.6.2 SR 3.4.6.2 requires verification of SG OPERABILITY.
SG OPERABILITY is verified by ensuring that the secondary side water level is above the lower tap of the SG wide range level instrumentation by >418.77 inches.If the SG U-tubes become uncovered, the associated loop may not be capable of providing the heat sink necessary for removal of decay heat.The water level can be verified by either the wide range or the narrow range level instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. T-e--htfF-etee4 ire-'~ele wf --Insert 2 SR 3.4.6.3 Verification that each required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump. 4n4we-thr-giit-a~ec-a--~-vie~ea e 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 2 B3464Rvso o B 3.4.6-4 Revision No. 0 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES ACTIONS (continued) 0.1 and 0.2 If a required RHR loop is not in operation or if no required loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated.
Suspending operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet the requirements of LCO 3.1.1 is required to assure continued safe operation.
With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum S.DM maintains acceptable margin to subcritical operations.
The immediate Completion Times reflect the importance of maintaining operation for heat removal.SURVEILLANCE REQUIREMENTS SR 3.4.7.1 This SR requires verification-ever-y4-2-heii1-s that the required loop is in operation circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
.*---'nsert 2 SR 3.4.7.2 Verifying that at least two SGs are OPERABLE by ensuring their secondary side water levels are above the lower tap of the SG wide range level instrumentation by > 418.77 inches ensures an alternate decay heat removal method via natural circulation in the event that the second RHR loop is not OPERABLE.
The water level can be verified by either the wide range or the narrow range instruments.
A narrow range level instrument
> 6% or a wide range level instrument
> 79% ensures the Surveillance Requirement limit is met. If both RHR loops are OPERABLE, this Surveillance is not needed. T-he-t-2-h~eF~eeley-is-
-Insert 2 rea+eae-#hertreh -lvh SR 3.4.7.3 Verification that each required RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by Cook Nuclear Plant Unit 2 B 3.4.7-4 Revision No. 0 Cook Nuclear Plant Unit 2 B3.4.7-4 Revision No. 0 RCS Loops -MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE REQUIREMENTS (continued) verifying proper breaker alignment and power available to each required RHR pump. If secondary side water level is above the lower tap of the SG wide range level instrumentation by > 418.77 inches in at least two SGs, this Surveillance is not needed. Th-Fr~eeitef9y-ef-7-daye-ts- Insert 2-~~e~~4 e- -thm,-e a-hae-beIs-sewf4Nc o c-pabc-lct .e This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES
: 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation." Cook Nuclear Plant Unit 2 B3475Rvso o B3.4.7-5 Revision No. 0 RCS Loops -MODE 5, Loops Not Filled B 3.4.8 BASES SURVEILLANCE REQ UIRE MENTS SR 3.4.8.1 This SR requires verification
#ever-A2_-heiwrs that the required loop is in operation circulating reactor coolant. Verification includes flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat removal.
"--Insert 2 SR 3.4.8.2 Verification that each required pump is OPERABLE ensures that an.additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to each required pump.
f-7da-" "~ .ciiod "c~Il in ic:of- .l-nsert 2 This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.
This is acceptable because proper breaker alignment and power availability are ensured if a pump is operating.
REFERENCES None.Cook Nuclear Plant Unit 2 B3483Rvso o B 3.4.8-3 Revision No. 0 Pressurizer B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 REQUIREMENTS This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing the indicated level.
pafaet &c-,-hi~q--T-he42e-ieter-i#at4e-hx b-cathwnby j 1 I n sert 2ryasscssI
'-e-a:-dea-e--
SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated pressurizer backup heaters are verified to be at their specified capacity.
This may be done by testing the power supply output with the heaters energized.
T,-e l---nsert 2 F F4 ths dee- te-r REFERENCES
: 1. UFSAR, Chapter 14..2. NUREG-0737, November 1980.Cook Nuclear Plant Unit 2 B3494Rvso o B 3.4.9-4 Revision.
No. 0 Pressurizer PORVs B 3.4.11 BASES ACTIONS (continued) place the PORV(s) in manual control, this may not be possible for all causes of Condition B entry with PORV(s) inoperable and not capable of being manually cycled (e.g., as a result of failed control power fuse(s) or control switch malfunctions(s))
H.1 and H.2 If any Required Action and associated Completion Time of Condition A, B, C, D, E, F, or G is not met, if three PORVs are inoperable and not capable of being manually cycled, if two PORVs are inoperable and not capable of being manually cycled and one block valve inoperable (for reasons other than to comply with Required Action B.2) in a different line than the inoperable PORVs, or if one PORV is inoperable and not capable of being manually cycled and two block valves are inoperable (for reasons other than to comply with Required Action B.2) in different lines than the inoperable PORV, then the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block valve cycling verifies that the valve(s) can be opened and closed if needed.
2 This SR is modified by a Note, which states that this SR is not required to be performed with the block valve closed in accordance with the Required Actions of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable.
SR 3.4.11.2 SR 3.4.1 1.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. @#raige-e~~-qa 2@s~er-ti-mebt--tthoe eF Cook Nuclear Plant Unit 2 B 3.4.11-6 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.4.11-6 Revision No. 0 Pressurizer PORVs B 3.4.11 BASES SURVEILLANCE REQUIREMENTS (continued)
The Note modifies this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. In accordance with Reference 4, administrative controls require this test be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation.
SR 3.4.11.3 Operating the solenoid air control valve associated with each PORV, and the check valves on the air accumulators where applicable, ensures the PORV control system actuates properly when called upon. ep-a~in 2 T4- en thelu Fcr-wcg-uee 1 1 e-e#4he~h eF REFERENCES
: 1. Regulatory Guide 1.32, February 1977.2. UFSAR, Section 14.1.8.3. ASME, Operation and Maintenance Standards and Guides (OM Codes).4. Generic Letter 90-06, "Resolution of Generic Issue 70,'Power-Operated Relief Valve and Block Valve Reliability,'
and Generic Issue 94, 'Additional Low-Temperature Overpressure for Light-Water Reactors,'
Pursuant to 10 CFR 50.54(f)," June 25, 1990.Cook Nuclear Plant Unit 2B34117RvsoNo0 B 3.4.11-7 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued) through the pump control .switch being placed in pull to lock and at least one valve in the discharge flow path being closed, or at least one valve in the discharge flow path being closed and sealed or locked.In addition, SR 3.4.12.3 is modified by a Note that allows the accumulator discharge isolation valve position to be verified by administrative means.This is acceptable since the valve position was verified prior to deactivating the valve, access to the containment is restricted, and valves are only operated under strict procedural control.T--h
--Insert 2 4 4~f~e SR 3.4.12.4 The required RHR suction relief valve shall be demonstrated OPERABLE by verifying the RHR suction isolation valves are open. This Surveillance is only required to be performed if the RHR suction relief valve is being used to meet this LCO.The RHR suction isolation valves are verified to be opened evePfy-1-2 I .I r -Insert 2 SR 3.4.12.5 The ROS vent of> 2.0 square inches or a blocked open PORV is proven OPERABLE by verifying its open condition .eithef:
2 e~-=G9z-evcr; 1-2z4y eufcr-aa ef~t vaeq cJ 1 -ee~et The passive vent path arrangement must only be open if the vent is being used to satisfy the pressure relief requirements of LCO 3.4.12.A.2.c.
Cook Nuclear Plant Unit 2 B 3.4.12-11 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.4.12-11 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.4.12.6 The PORV block valve must be verified open t6 provide the flow path for each required PORV topgerform its function when actuated.
The valve must be remotely verified open in the main control room. This Surveillance is performed if one~or more PORVs satisfy the LCO.The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required removed, and the manual operator is not required locked in the inactive position.
Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.
ho 72hor reunc e in viw -ethe... l--nsert 2 I~a~=ee6fftet9-.&sect; e....SR 3.4.12.7 Verification eveiy-34-aey-s-.that each required emergency air tank bank's pressure is > 900 psig assures adequate air pressure for reliable PORV operation.
With the emergency air supply at > 900 psig, there will be enough air to support PORV operation for 10 minutes with no operator action upon a loss of control air. T~e-he----y-F~r t~eykeesht-e-c-eeeee~a tieee-e~vrle-te~-tee 2 SR 3.4.12.8 Performance of a COT is required eve~y-34-days-on each required PORV to verify and, as necessary, adjust its lift setpoint.
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 COT 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 COT will verify the setpoint is within the LCO limit. PORV actuation could depressurize the RCS and is not required.Cook Nuclear Plant Unit 2 B341-2Rvso o B 3.4.12-12 Revision No. 0 LTOP System B 3.4.12 BASES SURVEILLANCE REQUIREMENTS (continued)
A Note has been added indicating that this SR is not required to be performed until 12 hours after decreasing RCS cold leg temperature to_< 299&deg;F. The COT cannot be performed until in the LTOP MODES when the PORV lift setpoint can be reduced to the LTOP setting. The test must be performed within 12 hours after entering the LTOP MODES. -Fh ,- Iner 2*
-vrFs F-v beJi#tit  SR 3.4.12.9 Performance of a CHANNEL CALIBRATION on each required PORV actuation channel is required-ee to adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input. .=-Insert 2 REFERENCES
: 1. 10 CFR 50, Appendix G.2. Generic Letter 88-11.3. ASME, Boiler and Pressure Vessel Code, Section II1.4. WCAP-1 3235, "Donald C. Cook Units I & 2, Analysis of Low Temperature Overpressurization Mass Injection Events with Pressurizer Steam Bubble and RHR Relief Valve, March 1992;"WCAP-12483.
Revision 1, "Analysis of Capsule U From the American Electric Power Company D. C. Cook Unit 1 Reactor Vessel Radiation Surveillance Program, December 2002;" and WCAP-13515, Revision 1, "Analysis of Capsule U From Indiana Michigan Power Company D. C. Cook Unit 2 Reactor Vessel Radiation Surveillance Program, May 2002." 5. 10 CFR 50, Section 50.46.6. 10 CFR 50, Appendix K.7. Generic Letter 90-06.Cook Nuclear Plant Unit 2 B341-3Rvso o B 3.4.12-13 Revision No. 0 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued) operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and the containment sump level. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation." Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.eeee ---nsert 2 SR 3.4.13.2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary, to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated.
The primary to secondary LEAKAGE is measured at room temperature as described in Reference
: 5. Prior to comparison with the 150 gallons per day TS limit, the measured primary to secondary LEAKAGE is multiplied by a volume correction factor of 1.52. The correction factor ensures the offsite dose analyses, which assume primary to secondary leakage is at normal operating temperature and pressure, remain bounding.
The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all of the primary to~secondary LEAKAGE should be conservatively assumed to be from one SG.The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours after establishment of steady state operation.
For ROS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.Cook Nuclear Plant Unit 2 B 3.4.13-5 Revision No. 15 Cook Nuclear Plant Unit 2 B 3.4.13-5 Revision No. 15 RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS (continued) th ivIc
* F;-qn ef -Ie b~. e_4 nterv~ake4I..e4
-t~r4ma -e~ee~i-rare Insert 2 4m4fcgas The pri mary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 5).REFERENCES
: 1. UFSAR, Section 1.4.3.2. Regulatory Guide 1.45, May 1973.3. UFSAR, Section 14.2.4.4. NEl 97-06, "Steam Generator Program Guidelines." 5. EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines." Cook Nuclear Plant Unit 2 B341- eiinN.1 B 3.4.13-6 Revision No. 12 RCS PIV Leakage B 3.4.14 BASES SURVEILLANCE REQUIREMENTS (continued) potential for an unplanned transient if the Surveillance were performed with the reactor at power.The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.
Therefore, this SR is modified by a Note that states the Surveillance is only required to be performed in MODES 1 and 2. Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance.
SR 3.4.14.2 Verifying that the RHR interlock that prevents the valves from being opened is OPERABLE ensures that RCS pressure will not pressurize the RHR System beyond its design pressure of 600 psig.
Insert 2 REFERENCES
: 1. 10 CFR 50.2.2. 10 CFRS50.55a(c).
: 3. WASH-i1400 (NUREG-75/01 4), Appendix V, October 1975.4. Letter from D.G. Eisenhut, NRC, to all LWR licensees, LWR Primary Coolant System Pressure Isolation Valves, February 23, 1980.5. Letter from S.A. Varga, NRC, to J. Dolan, Order for Modification of Licenses Concerning Primary Coolant System Pressure Isolation Valves, April 20, 1981.6. Technical Requirements Manual.7. EGG-NTAP-61 75, Inservice Testing of Primary Pressure Isolation Valves, Idaho National Engineering Laboratory, February 1983.8. NRC Safety Evaluation for License Amendment 174.9. ASME, Operation and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 2B34145RvsoN.0 B 3.4.14-5 Revision No. 0 RCS Leakage Detection Instrumentation B 3.4.15 BASES ACTIONS (continued)
Completion Time ensures that the plant will not be operated in a degraded configuration for a lengthy time period.I E.1 and E.2 With the required containment atmosphere radioactivity monitor and the containment humidity monitor inoperable, the only means of detecting leakage is the containment sump monitor. This Condition does not provide the required diverse means of leakage detection.
The Required Action is to restore either of the inoperable required monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity.
The 30 day Completion Time ensures that the unit will not be operated in a reduced configuration for a lengthy time period.F.1 and F.2 If any Required Action and associated Completion Time of Condition A, B, C, 0, or E cannot be met, the unit must be brought to a MODE in which the requirement does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.G._1_With all three types of required monitors inoperable (i.e., LCO 3.4.15.a, b, and c not met), no automatic means of monitoring leakage are available, and immediate unit shutdown in accordance with LCO 3.0.3 is required.SURVEILLANCE SR 3.4.15.1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitor. The check gives reasonable confidence that the channel is operating properly. iset SR 3.4.15.2 SR 3.4.15.2 requires the performance of a COT on the required containment atmosphere radioactivity monitor. The test ensures that the monitor can perform its function in the desired manner. 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 Cook Nuclear Plant Unit 2 B 3.4.15-6 Revision No. 32 RCS Leakage Detection Instrumentation B 3.4.15 BASES SURVEILLANCE REQUIREMENTS (continued) 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 test verifies the alarm setpoint and relative accuracy of the instrument string. The SR 3.4.15.3.
SR 3.4.15.4.
and SR 3.4.15.52 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels.
The calibration verifies the accuracy of the instrument string, including the instruments located inside containment.
T-hareequeney-ef-94--me&#xb6;9-hs
~=-=insert2 ety=ai~d=eper.a4Ai9fre~~
~r~tei9~r~
A9e~-fr=evei9=that
=eeee~e.REFERENCES 1." UFSAR, Section 1.4.3.2. Regulatory Guide 1.45, Rev. 0, "Reactor Coolant Pressure Boundary Leakage Detection System," May 1973.3. AEP Letter to NRC, AEP:NRC:0137D, "NRC Generic Letter 84-04;Elimination Of Postulated Pipe Breaks In Primary Main Loops Generic Issue A-2, Asymmetric Blowdown Loads On PWR Primary Systems Request For License Condition Deletion," dated September 10, 1984.4. NRC Letter to AEP, "Generic Letter 84-04, Safety Evaluation of Westinghouse Topical Reports Dealing With Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," dated November 22, 1985.5. UFSAR, Section 4.2.7 Cook Nuclear Plant Unit 2 B341- eiinN.3 B 3.4.15-7 RCS Specific Activity B 3.4.16 BASES ACTIONS (continued)
B.1 If any Required Action and associated Completion Time of Condition A is not met, if the DOSE EQUIVALENT 1-131 is in the unacceptable region of Figure 3.4.16-1, or if gross specific activity of the reactor coolant is not within limit, the reactor must be brought to MODE 3 with RCS average temperature
< 500&deg;F within 6 hours. The Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 below 5000.F from full power conditions in an orderly manner and-without challenging unit systems.SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the gross specific activity of the reactor coolant at-least--enee-every-While basically a
measure of radionuclides with half lives longer than 15 minutes, excluding iodines, this measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in gross specific activity.Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating conditions.Th Insert 2u-e d U ing-t4-f e~ie.SR 3.4.16.2 This Surveillance requires the verification that the reactor coolant DOSE EQUIVALENT 1-131 specific activity is within limit. This Surveillance is accomplished by performing an isotopic analysis of a reactor coolant sample. This Surveillance is performed in MODE 1 only to ensure iodine remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur..Th--
y~rqem~ Insert 2
~ ~ gr aci.yi4aa The Frequency, between 2 and 6 hours after a power change >- 15% RTP within a 1 hour period, is established because the iodine levels peak during this time following fuel failure; samples at other times would provide inaccurate results.SR 3.4.16.3 A radiochemical analysis for determination is required with the unit operating in MODE 1 equilibrium conditions.
The determination directly relates to the LCO and is required to verify unit Cook Nuclear Plant Unit 2B34164RvsoN.0 B 3.4.16-4 Revision No. 0 RCS Specific Activity B 3.4.16 BASES SURVEILLANCE REQUIREMENTS (continued) operation within the specified gross activity LCO limit. The analysis for is a measurement of the average energies per disintegration for isotopes with half lives longer than 15 minutes, excluding iodines. rqieerty "-'-=-sert 2 This SR has been modified by a Note that indicates sampling is not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours. This ensures that the radioactive materials are at equilibrium so the analysis for IE is representative and not skewed by a crud burst or other similar abnormal event.REFERENCES
: 1. 10OCFRI100.11.
: 2. UFSAR, Section 14.2.4.Cook Nuclear Plant Unit 2B3.165RvsoN.0 B 3.4.16-5 Revision No. 0 Accumulators B 3.5.1 BASES ACTIONS (continued) reach the required plant conditions from full power conditions in an orderly manner and without challenging unit systems.0.1 If more than one accumulator is inoperable, the unit is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE REQUIREMENTS SR 3.5.1.1 Each accumulator isolation valve should be verified to be fully open @evey.A-2--heuis.-,.
This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the IRCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not meeting accident analyses assumptions.
ide~ed .SR 3.5.1.2 and SIR 3.5.1.3-Insert 2borated water volume and nitrogen cover pressure are verified for each accumulator.
T-i-Frq~fey
~ ade aene~e-tri~---O e-ue e-uu4yaiw~eeeqe4 ehae-~fr4et-~-eee-O~ra~ge#4ch  Fr ~ e :-mn-eet n-fe*'--Insert 2/SR 3.5.1.4 The boron concentration should be verified to be within required limits for each accumulatorei the static design of the accumulators limits the ways in which the concentration can be changed.Th-daen eee r < Insert 2
~ , Sampling the affected accumulator within 6 hours after a volume increase of 13 ft 3 will identify whether inleakage has caused a reduction in boron concentration to below the required limit. lt is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (IRWST), because the water contained in the RWST is within the accumulator boron concentration requirements.
This is consistent with the recommendation of NUIREG-1366 (iRef. 4).Cook Nuclear Plant Unit 2B3516ReionN.0 B3.5.1-6 Revision No. 30 Accumulators B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.5 Verification every 31 days that power is removed from each accumulator isolation valve operator when the RCS pressure is > 2000 psig ensures that an active failure could not result inthe closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a LOCA. Sincospowor,,ec i
2 This SR allows power to be supplied to the motor operated isolation valves when ROS pressure is < 2000 psig, thus allowing operational flexibility by avoiding unnecessary delays to manipulate the breakers during plant startups or shutdowns.
REFERENCES
: 1. UFSAR, Section 14.3.2. 10 CFR 50.46.3. WCAP-1 5049-A, "Risk-Informed Evaluation of an Extension to Accumulator Completion Times," Rev. 1, April 1999.4. NUREG-1 366, February 1990.Cook Nuclear Plant Unit 2 B3517Rvso o B3.5.1-7 Revision No. 0 ECCS -Operating B 3.5.2 BASES ACTIONS (continued) train available, the facility is in a condition outside of the accident analyses.
Therefore, LCO 3.0.3 must be en~tered immediately.
SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained.
Misalignment of these valves could render both ECCS trains inoperable.
Securing these valves in position by locking out control power ensures that they cannot change position as a result of an active failure or be inadvertently misaligned.
These valves are of the type, described in Reference 9, that can disable the function of both ECCS trains and invalidate the accident analyses..A-f}s 2 SR 3.5.2.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. A valve that receives an actuation signal is allowed to be in a nonaccident position provided the valve will automatically reposition within the proper stroke time. This Surveillance does not require any testing or valve manipulation.
Rather, it involves verification that those valves capable of being mispositioned are in the correct position.
The -rfun ~ f4 Ch v
<--- Insert 2 SR 3.5.2.3 Verifying that each ECCS pump's developed head at the flow test point is greater than or equal to the required developed head ensures that ECCS pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of ECCS pump performance required by the ASME OM Code (Ref. 10). Since the ECCS pumps cannot be tested with flow through the normal ECCS flow paths, they are tested on recirculation flow (RHR and SI pumps) or normal charging flow path (centrifugal charging pumps). This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect Cook Nuclear Plant Unit 2B3527ResinN.0 B3.5.2-7 Revision No. 30 EGOS -Operating B 3.5.2 BASES SURVEILLANCE REQUIREMENTS (continued) incipient failures by indicating abnormal performance.
The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.5.2.4 and SR 3.5.2.5 These Surveillances demonstrate that each automatic ECOS valve actuates to the required position on an actual or simulated SI signal and that each EGOS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.
T~he- s--a -- -~c, du.i
"-vcl~n0 -e-per@efmedec-sas--s%{ae Ilnsert 2 SR 3J.5.2.6 Proper throttle valve position is necessary for proper EGOS performance.
These valves have stops to allow proper positioning for restricted flow to a ruptured cold leg, ensuring that the other cold legs receive at least the required minimum flow. This Surveillance verifies the mechanical stop of each listed EGOS throttle valve is in the correct position.
T~he a{Ie4t Fe ~ -bseSEsZ-3.-i5.2.5.-B SR 3.5.2.7 Periodic inspections of the containment sump suction inlets ensure that they are unrestricted and stay in proper operating condition.
This Surveillance verifies that the sump suction inlets are not restricted by debris and the suction inlet strainers show no evidence of structural distress, such as openings or gaps, which would allow debris to bypass the strainers..-Phe-24nrot-F-reqtr~~
~ ,J the-need-toffave a-t...esmt-rl -T-UF- uny-hsbe-ft teee
-Insert 2 Ilnsert 2 REFERENCES
: 1. UFSAR, Section 1.4.7.2. 10 CFR 50.46.3. UFSAR, Section 14.3.1.4. UFSAR, Section 14.3.2.Cook Nuclear Plant Unit 2 B 3.5.2-8 Revision No. 16 Cook Nuclear Plant Unit 2 B 3.5.2-8 Revision No. 16 RWST B 3.5.4 BASES ACTIONS (continued) brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.5.4.1 The RWST borated water temperature should be verified everj24iei~sssle .__T_ee~~,dfeett et~~~ I'-nsernt 2
er~qeacltbetr~
SR 3.5.4.2 The RWST water volume should be verified be above the required minimum level in order to ensure that a sufficient initial supply is available for injection and to support continued ECCS and Containment Spray System pump operation on recirculation. q.ex1perie19~e:
4'=-Ilnsert 2 SR 3.5.4.3 The boron concentration of the RWST should be verified ,evefiWy==dys to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized.
* ecepet*q~epheug-per-amieitr~ree.
*REFERENCES
: 1. UFSAR, Section 6.2.2.2. UFSAR, Section 14.3.Cook Nuclear Plant Unit 2 B3545Rvso o B3.5.4-5 Revision No. 0 Seal Injection Flow B 3.5.5 BASES APPLICABILI.TY In MODES 1, 2, and 3, the seal injection flow resistance limit is dictated by ECCS flow requirements, which are specified for MODES 1, 2, 3, and 4. The seal injection flow resistance limit is not applicable for MODE 4 and lower, however, because high seal injection flow is less critical as a result of the lower initial RCS pressure and decay heat removal requirements in these MODES. Therefore, RCP seal injection flow resistance must be limited in MODES 1, 2, and 3 to ensure adequate ECCS performance.
ACTIONS A.1 With the seal injection flow resistance not within its limit, the amount of charging flow available to the RCS may be reduced. Under this condition, action must be taken to restore the flow resistance to within its limit. The operator has 4 hours from the time the flow resistance is known to not be within the limit to correctly position the manual valves and thus be in compliance with the accident analysis.
The Completion Time minimizes the potential exposure of the unit to a LOCA with insufficient injection flow and provides a reasonable time to restore seal injection flow resistance within limits. This time is conservative with respect to the Completion Times of other ECOS LCOs; it is based on operating experience and is sufficient for taking corrective actions by operations personnel.
B.1 and B.2 When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated.
The Completion Time of 6 hours for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal coold own rates, and does not challenge plant safety systems or operators.
Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.
SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Verification e -.=dys that the seal injection flow resistance is within the limit ensures that the ECCS injection flows stay within the safety analysis.
A differential pressure is established between the charging header and the RCS, and the total seal injection flow is verified to be within the limit determined in accordance with the ECCS safety analysis.The flow resistance shall be >- 0.227 ftlgpm 2.Cook Nuclear Plant Unit 2 B 3.5.5-3 Revision No. 0 Cook Nuclear Plant Unit 2 B3.5.5-3 Revision No. 0 Seal Injection Flow B 3.5.5 BASES SURVEILLANCE REQUIREMENTS (continued)
The seal injection flow resistance, RSL, is determined from the following expression:
RSL = 2.31 (PoHP-PsI)/Q 2 where: PCHP = charging pump header pressure (psig);= 2300 psig (high pressure operation);
and Q =total seal injection flow (gpm).Sdz.j3 i3 bc3cd 3n ongin&cring judgrnor.t and iz eet~tsi~ste*wit~
~tl 1 e~ EGOS v;l~c OelIaic~ F~qu~.1 ci~. T[-.~~ bo acczptabl c ~2 As noted, the Surveillance is not required to be performed until 4 hours after the pressurizer pressure has stabilized within a + 20 psig range of normal operating pressure.
The pressurizer pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly.
The pressurizer pressure indications are averaged to determine whether the appropriate pressure has been achieved.
The exception is limited to 4 hours to ensure that the Surveillance is timely.REFERENCES
: 1. UFSAR, Section 14.3.1.2. UFSAR, Section 14.3.2.3. UFSAR, Section 14.2.4.4. UFSAR, Section 14.2.5.Cook Nuclear Plant Unit 2 B3554Rvso o B 3.5.5-4 Revision No. 0 Containment Air Locks B 3.6.2 BASES SURVEILLANCE REQUIREMENTS (continued) air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by the Containment Leakage Rate Testing Program.The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria Which is applicable to SR 3.6.1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C containment leakage rate.SR 3.6.2.2 The air Jock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY.
Thus, the door interlock feature supports containment OPERABILITY while the air Jock is being, used for personnel transit in and out of the containment.
Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur.
S-s- t u-' ro cr trj-;-4nd erl- eK -iv t; f.i~f&ti~  me seerr-nt-nner-en ve~el en ir a trict-aelhrieo-te2$ly p U.J.) LI IL, ....JLII U L.IIILAI IL,'.., III,. IL,,* ~JtI I US II ILU.J 1.11. t.r SI ELI 111.1111 L.LJL'..,I
*Uy. S*a4-lec-ke.
8et REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 14.2.6.3. UFSAR, Section 5.7.4. 10 CFR 50, Appendix J, Option B.Cook Nuclear Plant Unit 2 B3626Rvso o B 3.6.2-6 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES ACTIONS (continued) locked, sealed, or otherwise secured in position and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since the function of locking, sealing, or securing components is to ensure that these devices are not inadvertently repositioned.
Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position, is small.D.1 and D.2 If any Required Action and associated Completion Time is not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.6.3.1 REQUI REM ENTS This SR ensures that the containment purge supply and exhaust valves are closed as required or, if open, open for an allowable reason. If a purge valve is open in violation of this SR, the valve is considered inoperable.
The SR is not required to be met when the containment purge valves are open for the reasons stated. The valves may be opened for pressure control, ALARA or air quality considerations for personnel entry, or for Surveillances or maintenance activities that require the valves to be open. The containment purge valves are capable of closing in the environment following a LOCA. Therefore, these valves are allowed to be open for limited periods of time. T- re~/ -ie 2 SR 3.6.3.2 This SR requires verification that each containment isolation manual valve and blind flange located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. This SR does not require any testing or valve manipulation..
Rather, it involves verification that those containment isolation valves outside containment and capable of being mispositioned are in the correct position. 2 Cook Nuclear Plant Unit 2 B3637Rvso o B3.6.3-7 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time the valves are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since acdess to these areas is typically restricted during MODES 1, 2, 3, and 4 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.SR 3.6.3.3 This SR requires verification that each containment isolation manual valve and blind flange located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design, limits. For containment isolation valves inside containment, the Frequency of "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under administrative controls and the probability of their misalignment is low. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time they are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in the closed position, since these were verified to be in the correct position upon locking, sealing, or securing.This Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4, for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.Cook Nuclear Plant Unit 2 B 3.6.3-8 Revision No. 0 Cook Nuclear Plant Unit 2 B3.6.3-8 Revision No. 0 Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.3.4 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY.
The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses.The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.3.5 Automatic containment isolation valves close on a containment isolation signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic containment isolation valve will actuate to its isolation position on a containment isolation signal. This surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. ~ teey~-ae~-
ne~'l~~~-~~~rve~--
p g-'Insert 2 REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 14.2.6.3. UFSAR, Section 5.4.1 and Table 5.4-1.Cook Nuclear Plant Unit 2 B3639Rvso o B3.6.3-9 Revision No. 0 Containment Pressure B 3.6.4 BASES SURVEILLANCE REQ U IREM ENTS SR 3.6.4.1 Verifying that containment pressure is within limits ensures that unit operation remains within the limits assumed in the containment analysis.1hz 12ie hour-eee
~R R &#xa2; 2 REFERENCES
: 1. UFSAR, Section 14.3.4.2. UFSAR, Section 5.2.2.2.3. 10 CFR 50, Appendix K.Cook Nuclear Plant Unit 2 B3643Rvso o B 3.6.4-3 Revision No. 0 Containment Air Temperature 1B 3.6.5 BASES ACTIONS A.1I When containment average air temperature in the upper or lower compartment is not within the limit of the LCOG, the average air temperature in the affected compartment must be restored to within limits within 8 hours. This Required Action is necessary to return operation to within the bounds of the containment analysis.
The 8 hour Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.B.1 and B.2 If the containment average air temperature cannot be restored to within its limits within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.5.1 and SR 3.6.5.2 Verifying that containment average air temperature is within the LCOG limits ensures that containment operation remains within the limits assumed for the containment analyses.
In order to determine the containment average air temperature, an average is calculated using measurements taken at locations within the containment selected to provide a representative sample of the overall containment atmosphere.
In the upper compartment, two locations at a nominal elevation of 712 ft o inches and a third location at a nominal elevation of 624 ft 10 inches are used and an arithmetic average taken. In the lower compartment, a volume weighted average temperature is calculated whereby the volume fraction for each of the various areas of lower containment is multiplied by the representative temperature, utilizing one or more temperature instruments, in that volume. In this way the temperatures are "weighted" according to the volume fraction.
These weighted temperatures are then summed to determine the Weighted Average Temperature for Lower Containment.
=-Insert 2 REFERENCES
: 1. U0FAR50Sctin49.34
: 2. 100FR5O.49.
Cook Nuclear Plant Unit 2 B3653Rvso o 13 3.6.5-3 Revision No. 8 Containment Spray System B 3.6.6 BASES ACTIONS A.1 With one containment spray train inoperable, the affected train must be restored to OPERABLE status within 72 hours. The components in this degraded condition are capable of providing 100% of the heat removal and iodine removal needs after an accident.
The 72 hour Completion Time was developed taking into account the redundant heat removal and iodine removal capabilities afforded by the OPERABLE train and the low probability of a DBA occurring during this period.B.1 and B.2 If the affected containment spray train cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 84 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for a release of radioactive material from the Reactor Coolant System is reduced in MODE 3.SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Verifying the correct alignment of manual, power operated, and automatic valves, excluding check valves, in the Containment Spray System provides assurance that the proper flow path exists for Containment Spray System operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position since they were verified in the correct position prior to being secured. This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation.
Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned, are in the correct position.
2 SR 3.6.6.2 Verifying that each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head ensures that spray pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 5).Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on bypass flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect Cook Nuclear Plant Unit 2B366-ReionN.0 B3.6.6-5 Revision No. 30 Containment Spray System B 3.6.6 BASES SURVEILLANCE REQUIREMENTS (continued) incipient failures by indicating abnormal performance.
The Frequency of this SR is in accordance with the Inservice Testing Program.SR 3.6.6.3 and SR 3.6.6.4 These SRs require verification that each automatic containment spray valve actuates to its correct position and each containment spray pump starts upon receipt of an actual or simulated containment spray actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative.
controls.
T'he e 4sJte 2.... i4y u au t , 4 These Surveillances include a Note that states that in MODE 4, only the manual portion of the actuation signal is required.
This is acceptable since the automatic portion of the actuation signal is' not required to be OPERABLE by ITS 3.3.2, "Engineered Safety Features Actuation System (ESFAS) Instrumentation." SR 3.6.6.5 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections.
This SR ensures that each spray nozzle is unobstructed and that spray coverage of the containment during an accident is not degraded.
The event based surveillance frequency following maintenance that could result in nozzle blockage was chosen because this passive portion of the system is not susceptible to service induced degradation.
REFERENCES
: 1. UFSAR, Section 1.4.7.2. UFSAR, Section 14.3.4.3. 10 CFR 50.49.4. 10 CFR 50, Appendix K.5. ASME, Operation and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 2B366-ReionN.1 B3.6.6-6 Revision No. 31 Spray Additive System B 3.6.7 BASES ACTIONS A.1j If the Spray Additive System is inoperable, it must be restored to OPERABLE within 72 hours. The pH adjustment of the Containment Spray System flow for corrosion protection and iodine retention enhancement is reduced in this condition.
The 72 hour Completion Time takes into account the redundant flow path capabilities and the low probability of the worst case DBA occurring during this period. In addition, if the Containment Spray System is available, it would remove some iodine from the containment atmosphere in the event of a DBA.B.1 and B.2 If the Spray Additive System cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 84 hours. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems. The extended interval to reach MODE 5 allows additional time and is reasonable when considering that the driving force for the release of radioactive material from the Reactor Coolant System is reduced in MODE 3.SURVEILLANCE REQUIREMENTS SR 3.6.7.1 Verifying the correct alignment of Spray Additive System manual, power operated, and automatic valves in the spray additive flow path provides assurance that the system is able to provide additive to the Containment Spray System in the event of a DBA. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This SR does not require any testing or valve manipulation.
Rather, it involves verification that those valves outside containment and capable of potentially being mispositioned are in the correct position. Insert 2 SR 3.6.7.2 To provide effective iodine retention, the containment spray must be an alkaline solution.
Since the RWST contents are normally acidic, the volume of the spray additive tank must provide a sufficient volume of spray additive to adjust pH for all water injected.
This SR is performed to verify the availability of sufficient NaOH solution in the Spray Additive System.
"=--'Insert 2 Cook Nuclear Plant Unit 2 B 3.6.7-3 Revision No. 0 Spray Additive System B 3.6.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.7.3 This SR provides verification (by chemical analysis) of the NaOH concentration in the spray additive tank and is sufficient to ensure that the spray solution being injected into containment is at the correct pH level.4 ~et4~ t
.
b SR 3.6.7.4 This SR provides verification that each automatic valve in the Spray Additive System flow path actuates to its correct position.
This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.T4 i Insert 22
* ec pa D-rc Womsateepet.
SR 3.6.7.5 To ensure that the correct pH level is established in the borated water solution provided by the ContainmentSpray System, the flow rate in the Spray Additive System is verified once every 5 years. This SR provides assurance that the correct amount of NaOH will be metered into the flow path upon Containment Spray System initiation.
The test is performed by verifying the flow rate from the spray additive tank test line to each Containment Spray System train with each containment spray pump operating in the recirculation mode.
Insert 2 REFERENCES
: 1. UFSAR, Chapter 14.3.5.9.Cook Nuclear Plant Unit 2 B3674Rvso o B3.6.7-4 Revision No. 0 DIS B 3.6.9 BASES ACTI ONS B.1 Condition B is one containment region with no OPERABLE hydrogen ignitor. Thus, while in Condition B, or in Conditions A and B simultaneously, there would always be ignition capability in the adjacent containment regions that would provide redundant capability by flame propagation to the region with no OPERABLE ignitors.Required Action B.1 calls for the restoration of one hydrogen ignitor in each region to OPERABLE status within 7 days. The 7 day Completion Time is based on the same reasons given under Required Action A. 1.C..11 If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCO does not apply. This is done by placing the unit in at least MODE 3 within 6 hours. The allowed.Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.9.1 This SR confirms that >- 34 of 35 hydrogen ignitors (or > 33 ignitors if allowed by footnote) can be successfully energized in each train. The ignitors are simple resistance elements.
Therefore, energizing provides assurance of OPERABILITY.
The allowance of one inoperable hydrogen ignitor is acceptable because, although one inoperable hydrogen ignitor in a region would compromise redundancy in that region, the containment regions are interconnected so that ignition in one region would cause burning to progress to the others (i.e., there i~s~overlap in each hydrogen ignitor's effectiveness between regions).
Tt
,, Insert 2 SR 3.6.9.2 This SR confirms that the two inoperable hydrogen ignitors allowed by SR 3.6.9.1 (i.e., one in each train) are not in the same containment region. -Insert 2 Footnote:
For the remainder of Fuel Cycle 18, or until the next entry into a MODE which allows replacement of the affected ignitors, DIS Train B can still perform its safety function and may be considered OPERABLE with one lower containment Phase 2 Power Supply ignitor inoperable and with one lower containment Phase 3 Power Supply ignitor inoperable (Reference 3).Cook Nuclear Plant Unit 2 B 3.6.9-4 Revision No. 25 DIS B 3.6.9 BASES SURVEILLANCE REQUIREMENTS SR 3.6.9.3 A more detailed functional test is performed every 24 months to verify system OPERABILITY.
Each ignitor is visually examined to ensure that it is clean and that the electrical circuitry is energized.
All ignitors, including normally inaccessible ignitors, are visually checked for a glow to verify that they are energized.
Additionally, the surface temperature of each ignitor (see footnote) is measured to be > 1700&deg; F to demonstrate that a temperature sufficient for ignition is achieved.
The-e.~-F-r-e41eey
<=l'nsert 2
~e~~eio eaie~ts-ueu'eli paee4i tae-eeneliide~tebe' vi--- ............
.... J ....... r REFERENCES
: 1. 10 CFR 50.44.2. UFSAR, Section 5.8.3. Letter from T. Beltz, NRC, to J. Jensen, l&M, "Donald C. Cook Nuclear Plant, Unit 2 -Issuance of Exigent Amendment Re: The Containment Distributed Ignition System (TAC No. ME3129)," dated February 4, 2010.Footnote:
For the remainder of Fuel Cycle 18, or until the next entry into a MODE which allows replacement of the affected ignitors, DIS Train B can still perform its safety function and may be considered OPERABLE with one lower containment Phase 2 Power Supply ignitor inoperable and with one lower containment Phase 3 Power Supply ignitor inoperable (Reference 3).Cook Nuclear Plant Unit 2 B 3.6.9-5 Revision No. 39 Cook Nuclear Plant Unit 2 B 3.6.9-5 Revision No. 39 CEQ System B 3.6.10 BASES LCO In the event of a DBA, one train of the CEQ System is required to provide the minimum air recirculation for heat removal and hydrogen mixing assumed in the safety analyses.
To ensure this requirement is met, two trains of the CEQ System must be OPERABLE.
This will ensure that at least one train will operate, assuming the worst case single failure occurs.APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the CEQ System.Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the CEQ System is not required to be OPERABLE in these MODES.ACTIONS A.1 If one of the trains of the CEQ System is inoperable, it must be restored to OPERABLE status within 72 hours. The components in this degraded condition are capable of providing 100% of the flow and hydrogen skimming needs after an accident.
The 72 hour Completion Time was developed taking into account the redundant flow and hydrogen skimming capability of the OPERABLE CEQ System train and the low probability of a DBA occurring in this period.B.1 and B.2 If the CEQ System train cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.10.1 Verifying that each CEQ System fan starts on an actual or simulated actuation signal, after a delay > 108 seconds and < 132 seconds, and operates for > 15 minutes is sufficient to ensure that all fans are OPERABLE and that all associated controls and time delays are functioning properly.
It also ensures that blockage, fan and/or motor failure, or excessive vibration can be detected for corrective action. ifhe. Insert 2 ari~ elaac~~~~le Cook Nuclear Plant Unit 2B361-3RvsoN.0 B 3.6.10-3 Revision No. 0 C EQ System B 3.6.10 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR has been modified by a Note that states that this Surveillance is only required to be met in MODES 1, 2, and 3. This allowance is necessary since the specified delay (i.e., > 108 seconds and< 132 seconds) is only applicable to the automatic actuation signal (i.e., Containment Pressure -High), which is only required to be OPERABLE in MODES 1, 2, and 3. In addition, LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS) Instrumentation," requires the CEQ System Manual Initiation Function to be OPERABLE in MODE 4 and requires the performance of a TADOT every 24 months. This requirement will ensure the Manual Initiation Function can actuate the required equipment in MODE 4.SR 3.6.10.2 Verifying, with the return air fan discharge backdraft damper locked closed and the fan motor energized, the static pressure between the fan discharge and the backdraft damper is > 4.0 inches water gauge confirms one operating condition of the fan. This test is indicative of overall fan motor performance.
Such tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance..T4h.
.Preq -etst ge r''tsfrs ia-e~
eia 2 SR 3.6.10.3 Verifying the OPERABILITY of the return air damper provides assurance that the proper flow path will exist when the fan is started. By applying _the correct counterweight, the damper operation can be confirmed.
The Insert 2d ~~rte~e-a~ae
~
Qp ae e.SR 3.6.10.4 Verifying the OPERABILITY of the motor operated valve in the hydrogen skimmer header provides assurance that the proper flow path will exist when the valve receives an actuation signal. Th9 ~zya*-Frequlen-ywes Insert 2 Cook Nuclear Plant Unit 2 B 3.6.10-4 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.6.10-4 RevisionNo.
0 Ice Bed B 3.6.11 BASES APPLICABILITY In MODES 1, 2, 3, and 4, a DBA could cause an increase in containment pressure and temperature requiring the operation of the ice bed.Therefore, the LCO is applicable in MODES 1, 2, 3, and 4.In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the ice bed is not required to be OPERABLE in these MODES.ACTIONS A.1 If the ice bed is inoperable, it must be restored to OPE'RABLE status within 48 hours. The Completion Time was developed based on operating experience, which confirms that due to the very large mass of stored ice, the parameters comprising OPERABILITY do not change appreciably in this time period. If a degraded condition is identified, even for temperature, with such a large mass of ice it is not possible for the degraded condition to significantly degrade further in a 48 hour period.Therefore, 48 hours is a reasonable amount of time to correct a degraded condition before initiating a shutdown.B.1 and B.2 If the ice bed cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQU IREMENTS SR 3.6.11.1 Verifying that the maximum temperature of the ice bed is < 27&deg;F ensures that the ice is kept well below the melting point. -T-he-4-2=le -e e y~ea--yo se t taj~
a aeesig4 a~m rfao .--
u l f er-ater44 nJter-ia e-Svetem Cook Nuclear Plant Unit 2 B 3.6.11-4 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.6.11-4 Revision No. 0 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued)
The total ice mass and individual radial zone ice mass requirements defined in this Surveillance, and the minimum ice mass per basket requirement defined by SR 3.6.11.3, are the minimum requirements for OPERABILITY.
Additional ice mass beyond the SRs is maintained to address sublimation.
This sublimation allowance is generally applied to baskets in each radial zone, as appropriate, at the beginning of an operating cycle to ensure sufficient ice is available at the end of the operating cycle for the ice condenser to perform its intended design function.:F~'rI~ny'4=et~aelae'~e~e~g <-==I nserti 2 tye~-u .v*Ra+I9teiiae4.
SR 3.6.11.3 Verifying that each selected sample basket from SR 3.6.11.2 contains at least 600 lbs of ice in the as-found (pre-maintenance) condition ensures that a significant localized degraded mass condition is avOided.This SR establishes a per basket limit to ensure any ice mass degradation is consistent with the initial conditions of the DBA by not significantly affecting the containment pressure response.
Reference 2 provides insights through sensitivity runs that~demonstrate that the containment peak pressure during a DBA is not significantly affected by the ice mass in a large localized region of baskets being degraded below the required safety analysis mean, when the radial zone and total ice mass requirements of SR 3.6.11.2 are satisfied.
Any basket identified as containing less than 600 lbs of ice requires appropriately entering ACTION A for an inoperable ice bed due to the potential that it may represent a significant condition adverse to quality.As documented in Reference 2, maintenance practices actively manage individual ice basket mass above the required safety analysis mean for each radial zone. Specifically, each basket is serviced to keep its ice mass above 1132 lbs for zone A, 1132 lbs for zone B, and 1132 lbs for zone C. If a basket sublimates below the safety analysis mean value, this instance is identified within the CNP corrective action program, including evaluating maintenance practices to identify the cause and correct any deficiencies.
These maintenance practices provide defense in depth beyond compliance with the ice bed Surveillance Requirements by limiting the occurrence of individual baskets with ice mass less than the required safety analysis mean.Cook Nuclear Plant Unit 2 B 3.6.11-6 Revision No. 0 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued) the ice basket support platform is not a more restrictive flow area because it is easily accessible from the lower plenum and is maintained clear of ice accumulation.
There is no mechanistically credible method for ice to accumulate on the ice basket support platform during unit operation.
Plant and industry experience has shown that the vertical flow area through the ice basket support platform remains clear of ice accumulation that could produce blockage.
Normally only a glaze may develop or exist on. the ice basket support platform which is not significant to blockage of flow area. Additionally, outage maintenance practices provide measures to clear the ice basket support platform following maintenance activities of any accumulation of ice that could block flow areas.Frost buildup or loose ice is not to be considered as flow channel blockage, whereas attached ice is considered blockage of a flow channel.Frost is the solid form of water that is loosely adherent, and can be brushed off with the open hand. =-.-Insert 2 SR 3.6.11.5 This SR ensures that a representative sampling of ice baskets, which are relatively thin walled, perforated cylinders, have not been degraded by wear, cracks, corrosion, or other damage. The SR is designed around a full-length inspection of a sample of baskets, and is intended to monitor the effect of the ice condenser environment on ice baskets. The groupings defined in the SR (two baskets in each azimuthal third of the ice bed) ensure that the sampling of baskets is reasonably distributed.
The Frequency of 40 months for a visual inspection of the structural soundness of the ice baskets is based on engineering judgment and considers such factors as the thickness of the basket walls relative to corrosion rates expected in their service environment and the results of the long term ice storage testing. "Iflnsert 2 SR 3.6.11.6*Verifying the chemical composition of the stored ice ensures that the stored ice has a boron concentration
> 1800 ppm and <2300 ppm as sodium tetraborate and a high pH, > 9.0 and < 9.5 at 25&deg;C, in order to meet the requirement for borated water when the melted ice is used in the ECCS recirculation mode of operation.
Additionally, the minimum boron concentration limit is used to assure reactor subcriticality in a post LOCA environment, while the maximum boron concentration limit is used as the bounding value in the hot leg switchover timing calculation (Ref. 4). This is accomplished by obtaining at least 24 ice samples. Each sample is taken approximately one foot from the top of the ice of each randomly selected ice basket in each ice condenser bay. The SR is modified by a Note that allows the boron concentration and pH value obtained from averaging the individual samples' analysis results to satisfy the B 3.6.11-8 Ice Bed B 3.6.11 BASES SURVEILLANCE REQUIREMENTS (continued) requirements of the SR. If either the average baron concentration or average pH value is outside their prescribed limit, then entry into Condition A is required.
Sodium tetraborate has been proven effective in maintaining the boron content for long storage periods, and it also enhances the ability of the solution to remove and retain fission product iodine, although the removal of iodine from the containment atmosphere by the sodium tetraborate is not assumed in the'accident analysis.
This pH range also minimizes the occurrence of chloride and caustic stress corrosion on mechanical systems and components exposed to ECCS and Containment Spray System fluids in the recirculation mode of operation.
The Frequency of 54 months is intended to be consistent with the.expected length of three fuel cycles, and was developed considering these facts: a. Long term ice storage tests have determined that the chemical composition of the stored ice is extremely stable;b. There are no normal operating mechanisms that decrease the boron concentration of the stored ice, and pH remains within a 9.0-9.5 range when boron concentrations are above approximately 1200 ppm;c. Operating experience has demonstrated that meeting the boron concentration and pH requirements has never been a problem; and d. Someone would have to enter the containment to take the sample, and, if the unit is at power, that person would receive a radiation dose. <---Insert 2 SR 3.6.11.7 This SR ensures that initial ice fill and any subsequent ice additions meet the boron concentration and pH requirements of SR 3.6.11.6.
The SR is modified by a Note that allows the chemical analysis to be performed on either the liquid or resulting ice of each sodium tetraborate solution prepared.
If ice is obtained from offsite sources, then chemical analysis data must be obtained for the ice supplied.Cook Nuclear Plant Unit 2 B361- eiinN.4 B 3.6.11-9 Revision No. 47 Ice Condenser Doors B 3.6.12 BASES ACTIONS (continued)
C.1I If Required Action B.1 or B.2 and associated Completion Time is not met, the doors must be restored to OPERABLE status and closed positions within 48 hours. The 48 hour Completion Time is based on the fact that, with the very large mass of ice involved, it would not be possible for the temperature to increase to the melting point and a significant amount of ice to melt in a 48 hour period. The 48 hour Completion Time is also consistent with the ACTIONS of LCO 3.6.11, "Ice Bed." D.1 and D.2 With any Required Action and associated Completion Time of Condition A or C not met, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed.Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.6.12.1 REQUIREMENTS Verifying that the inlet doors are in their closed positions makes the nnr~n =\t-rf in ,~rlklrnt nc~rninn n~f nr mnrn C1Cnnr_ ,:#1 2 ftlie ovreolesier-Meniir~-Systei.
SR 3.6.12.2 Verifying, by visual inspection, that each intermediate deck door is closed and not impaired by ice, frost, or debris provides assurance that the intermediate deck doors (which form the floor of the upper plenum where frequent maintenance on the ice bed is performed) have not been left open or obstructed. l=- nsert 2............
....... ..... ..........
... .... i ..... j --
d Cook Nuclear Plant Unit 2B36125evsoN.0 B 3.6.12-5 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.3 Verifying, by visual inspection, that the top deck doors are in place and not obstructed provides assurance that the doors are performing their function of keeping warm air out of the ice condenser during normal operation, and would not be obstructed if called upon to open in response to a DBA. T4FFf~aye-2d-ss~sde-ae~@j~~.v t, onideed o -afie4jia~
de --
fle~pth~ vei9a4#aer-wer-e-#ebFite4.
SR 3.6.12.4 Verifying, by visual inspection, that the ice condenser inlet doors are not impaired by ice, frost, or debris provides assurance that the doors are free to open in the event of a DBA. T~feueqi gexeiae-
* te-r-meet-tiei-aeeep-a e-eri {e-i~a-Beeaus
....hrd~t, v~i~-- ~ ~4e4e-~{4-f
~wrep~gnM r4m ~
SR 3.6.12.5 2 Insert z Verifying the opening torque of the inlet doors provides assurance that no doors have become stuck in the closed position.
The value of 675 in-lb is based on the design opening pressure on the doors of 1 .0 lb/ft 2..r-hi*
--
ne4keywvr ts-fta-ee-dsif~hcl~sni
"=-nsert 2 weter-c- e a e dt~e~ t-g-e-e).---
hierat dej- wIfrhe ~t a4 Cook Nuclear Plant Unit 2B36126RvsoN.0 B 3.6.12-6 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.6 The torque test Surveillance ensures that the inlet doors have not developed excessive friction and that the return springs are producing a door return torque within limits. The torque test consists of the following:
: 1. Verify that the torque, T(OPEN), required to cause opening motion at the 400 open position is < 195 in-Ib;2. Verify that the torque, T(CLOSE), required to hold the door stationary (i.e., keep it from closing) at the 400 open position is > 78 in-Ib; and 3. Calculate the frictional torque, T(FRICT) = 0.5 {T(OPEN) -T(CLOSE)}, and verify that the T(FRICT) is -< 40 in-lb.T(OPEN) is known as the "door opening torque" and is equal to the nominal door torque plus a frictional torque component.
T(CLOSE) is defined as the "door closing torque" and is equal to the nominal door torque minus a frictional torque component.
The purpose of the friction and return torque Specifications is to ensure that, in the event of a small break LOCA or SLB, all of the 24 door pairs open uniformly.
This assures that, during the initial blowdown phase, the steam and water mixture entering the lower compartment does not pass through part of the ice condenser, depleting the ice there, while bypassing the ice in other bays..T-he-=Feaene-M-I4h-ela -ePi-t-he~~ eea~--4 ajsete~r 4----nsert 2 aeee.Rc-e-er4ter~-Beeau~eejh4"-a~aterH94h-e-t1efi91e Cook Nuclear Plant Unit 2B36127RvsoN.0 B 3.6.12-7 Revision No. 0 Ice Condenser Doors B 3.6.12 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.12.7 Verifying the OPERABILITY of the intermediate deck doors provides assurance that the intermediate deck doors are free to open in the event of a DBA. The verification consists of visually inspecting the intermediate doors for structural deterioration, verifying free movement of the vent assemblies, and ascertaining free movement of each door when lifted with the applicable force shown below: Door Liftincj Force a. Adjacent to crane wall b. Paired with door adjacent to crane wall c. Adjacent to containment wall d. Paired with door adjacent to containment wall< 37.4 lb< 33.8 lb.< 31.8 lb-<31.0 lb
~ -Bs~d-n-tn -=
ef rsaeetyi -F-eaae~a--ete--ee-4 ti~ -e~-cefst eb ie 4kie =Insert 2 insead-weogt4' REFERENCES
: 1. UFSAR, Section 14.3.4.Cook Nuclear Plant Unit 2B3.128RvsoN.0 B 3.6.12-8 Revision No. 0 Divider Barrier Integrity B 3.6.13 BASES ACTIONS (continued)
C.1 and C.2 If divider barrier integrity cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.6.1 3.1 Verification, by visual inspection, that all personnel access doors and equipment hatches between the upper and lower containment compartments are closed provides assurance that divider barrier integrity is maintained prior to the reactor being taken from MODE 5 to MODE 4.This SR is necessary because many of the doors and hatches may have been opened for maintenance during the shutdown.SR 3.6.13.2 Verification, by visual inspection, that the personnel access door and equipment hatch seals, sealing surfaces, and alignments are acceptable provides assurance that divider barrier integrity is maintained.
This inspection cannot be made when the door or hatch is closed. Therefore, SR 3.6.13.2 is required for each door or hatch that has been opened, prior to the final closure. Some doors and hatches may not be opened for long periods of time. T$he -4hat-use-r-esiieat-Fat-er-iets-ia-t*Ie-seale-mu~st e5ac  SR 3.6.1 3.3 Verification, by visual inspection, after each opening of a personnel access door or equipment hatch that it has been closed makes the operator aware of the importance of closing it and thereby provides additional assurance that divider barrier integrity is maintained while in applicable MODES.
2 Cook Nuclear Plant Unit 2B36134RvsoN.0 B 3.6.1 3-4 Revision No. 0 Divider Barrier Integrity B 3.6.13 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.13.4 Conducting periodic physical property tests on divider barrier seal test coupons provides assurance that the seal material has not degraded in the containment environment, including the effects of irradiation with the reactor at power. The required tests include a tensile strength test and a test for elongation. tswseeeet co .iorn .cuch qy fthesa~aer Ilnsert 2~GdrtcGG8~ibUGfAheSe61e5Rd5b5eflee~f41fl9eiT*fehiity e4he=i~4t conditionc noodo~4e~pei~m the SR. Operating experience
~pei~f~ri~d at tho 24 ~concludod to bo ~SR 3.6.13.5 Visual inspection of the seal around the perimeter provides assurance thatthe seal is properly secured in place, such that the total divider barrier bypass area is less than or equal to the design basis limit of 7'tsueh~i449etiet-s-a 9-ef4 el-a9isee-tief-rt6-f seeIre4h~e
~ee1a94e-tflteeilift neee-efom-he-fR.
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~Insert 2 2-epat REFERENCES
: 1. UFSAR, Section 14.3.4.1.3.1.3.
: 2. UFSAR, Section 14.3.4.1.3.1.1.e Cook Nuclear Plant Unit 2 B361- eiinN.4 B 3.6.13-5 Revision No. 44 Containment Recirculation Drains B 3.6.14 BASES ACTIONS (continued)
C.1I If one CEQ fan room drain is inoperable, 1 hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operation to within the bounds of the containment analysis.
The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1,"Containment," which requires that the containment be restored to OPERABLE status within 1 hour.D.1I If one flow path in the flood-up overflow wall is inoperable, 1 hour is allowed to restore the drain to OPERABLE status. The Required Action is necessary to return operations to within the bounds of the containment analysis..The 1 hour Completion Time is consistent with the ACTIONS of LCO 3.6.1, "Containment," which requires that containment be restored to OPERABLE status within 1 hour.E.1 and E.2 If the affected drain(s) cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.6.14.1 and SR 3.6.14.2 REQUIREMENTS Verifying the OPERABILITY of the required refueling canal drains ensures that they will be able to perform their functions in the event of a DBA. SR 3.6.14.2 confirms that the required refueling canal drain blind flanges have been removed and that the required drains are clear of any obstructions that could impair their functioning.
In addition to debris near the drains, attention must be given to any debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains. This verification is performed by SR 3.6.14.1, which requires verification that there is no debris present in the upper containment or refueling canal that could obstruct the required refueling canal drains. SR 3.6.14.1 and SR 3.6.14.2.must be performed before entering MODE 4 from MODE 5 after every filling of the canal to ensure that the blind flanges have been removed and that no debris that could impair the drains was deposited during the time the canal was filled.
-Insert 2 aywetvtf dcsd g-fee--s-e.-
ssi i-- -~e-~-ia -h -b v~e~~~- -~ -~~i~sadth-~~
ayMh Cook Nuclear Plant Unit 2 B 3.6.14-4 Revision No. 16 Cook Nuclear Plant Unit 2 B3.6.14-4 Revision No. 16 Containment Recirculation Drains B 3.6.14 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.6.14.3 Verifying the OPERABILITY of the ice condenser floor drains ensures that they will be able to perform their functions in the event of a DBA.Inspecting the drain valve disk ensures that the valve is performing its function of sealing the drain line from warm air leakage into the ice condenser during normal operation, yet will open if melted ice fills the line following a DBA. Verifying that the drain lines are not obstructed ensures their readiness to drain water from the ice condenser.
T&#xa2;-e4-8--melit-h t4e-be-ac-eept-able
~ feerBeu e .hjhr4.a~e~th-~eatf.
~-as.~~ $eelae.- r-ayJn-i ~ a Insert 2 SR 3.6.14.4 and SR 3.6.14.5 Verifying the operability of the CEQ fan room drains ensures that they will be able to perform their function in the event of a DBA. SR 3.6.14.4 confirms that the required drains are clear of any obstructions.
In addition to debris near the drains, attention must be given to debris that is located where it could be moved to the drains in the event that the Containment Spray System is in operation and water is flowing to the drains.SR 3.6.14.4 must be performed before entering MODE 4 from MODE 5 and after personnel entry into a CEQ fan room while in MODES 1 through 4. This frequency was developed considering such factors as the location of the drains, and the absence of personnel traffic in the vicinity of the drains. The SR is modified by a Note. The Note indicates that only the CEQ fan room that has been entered need be inspected if the SR is being performed due to personnel entry in MODES 1 through 4. The Note precludes unnecessarily requiring inspection of both CEQ fan rooms if only one has been entered. SR 3.6.14.5 confirms that the CEQ fan room debris interceptors are installed and free of structural distress.SR 3.6.14.5 also confirms that the flow opening at the lower containment sump is not obstructed.
The 24 month frequency was developed considering such factors as the location and the design of the debris interceptors and flow opening.Cook Nuclear Plant Unit 2 B 3.6.14-5 Revision No. 16 Cook Nuclear Plant Unit 2 B 3.6.14-5 Revision No. 16 SGSVs B 3.7.2 BASES SURVEILLANCE REQUIREMENTS (continued)
The Frequency is in accordance with the Inservice Testing Program.This test is conducted in MODE 3 with the unit at operating temperature and pressure.
This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. This allows a delay of testing until MODE 3, to establish conditions consistent with those under which the acceptance criterion was generated.
SR 3.7.2.2 This SR verifies that each SGSV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage..;hI-q ofes Insert 2 t=h i ! tI -iii !sd-r-e q &#xb6;ene-yl -aeefa-Blelrom-a REFERENCES
: 1. UFSAR, Section 1,0.2.2. UFSAR, Section 14.2.5.3. 10CFR 100.1.1.4. Technical Requirements Manual 5. ASME, Operations and Maintenance Standards and Guides (OM Codes).Cook Nuclear Plant Unit 2B3724ReionN.3 B 3.7.2-4 Revision No. 33 MFIVs and MFRVs B 3.7.3 BASES ACTIONS (continued)
C. l With both the MFIV and MFRV inoperable in the same flow path, there is no redundant system to operate automatically and perform the required safety function.
Under these conditions, the affected flow path must be isolated within 8 hours. This action returns the system to the condition where at least one valve in each flow path is performing the required safety function.
The 8 hour Completion Time is reasonable, based on operating experience, to complete the actions required to close the MFIV or MFRV, or otherwise isolate the affected flow path.0.1 and 0.2 If any Required Action and associated Completion Time is not met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours and in MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.3.1 and SR 3.7.3.2 REQUIREMENTS These SRs verify that the closure time of each MFIV and MFRV is within the limit given in Reference 2 and is within that assumed in the accident and transient analyses.
The valve(s) may also be tested to more restrictive requirements in accordance with the Inservice Testing Program.The Frequency for this SR is in accordance with the Inservice Testing Program.SR 3.7.3.3 This SR verifies that each MFIV and MFRV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon returning the unit to operation following a refueling outage.TheR-efienylfev9erh---i9--evpes.--qp et~ei e~eas<= Insert 2 shw-ft, ~ -ae4ec-f REFERENCES
: 1. UFSAR, Section 10.5.1.2.2. Technical Requirements Manual Cook Nuclear Plant Unit 2B3734ReionN.3 B3.7.3-4 Revision No. 33 SG PORVs B 3.7.4 BASES ACTIONS (continued) 0.1 and C.2 If the SG PORV(s) cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in' which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance upon steam generator for heat removal, within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.0.1 If one or more required SG PORVs are inoperable in MODE 4, the unit is in a degraded condition with reduced safety related means to cool the unit to RHR entry conditions following an event, and the possibility of no means for conducting a cooldown with nonsafety related equipment since the condenser may be unavailable for use with the Steam Dump System.The seriousness of this condition requires that action be started immediately to restore the inoperable SG PORV(s) to OPERABLE status.SURVEILLANCE SR 3.7.4.1 REQUIREMENTS To perform a controlled cooldown of the RCS, the SG PORVs must be able to be opened remotely and throttled through their full range. This SR ensures that the SG PORVs are tested through a full control on~e-pe--4-.iae~hs.
Performance of inservice testing or use of an SG PORV during a unit cooldown may satisfy this requirement. -Gpea Insert 2-S~~lFe-hmef~e-th-4mFhF-fu -Freunc s cctab1e-f~em -Fe iMiy~tapea-.
REFERENCES
: 1. UFSAR, Section 10.2.2.2. UFSAR, Section 14.2.4.Cook Nuclear Plant Unit 2 B3743Rvso o B 3.7.4-3 Revision No. 0 AFW System B 3.7.5 BASES SURVEILLANCE REQUIREMENTS (continued) initiation without declaring the train(s) inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, and these manual operations are an accepted function of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained.
Th4 Y'efee~Jae~negaera ti se nsert 2 wfhtho~eeea-et"t~evrrig~le-p~te a sfs.eer-r~eetw.a4.p e sitieIa .SR 3.7.5.2 Verifying that each required AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded to an unacceptable level during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref. 2).Because it is undesirable to introduce cold AFW into the steam generators while they are operating, this testing is performed on recirculation flow. This test confirms one point on the pump design curve and is indicative of overall performance.
Such inservice tests confirm component OPERABILITY and detect incipient failures by indicating abnormal performance.
Performance of inservice testing discussed in the ASME OM Code (Ref. 2) (only required at 3 month intervals) satisfies this requirement.
This SR is modified by a Note indicating that the SR should be deferred for the turbine driven AFW pump until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test at entry into MODE 3. At 850 psig, there is sufficient pressure to perform the test.SR 3.7.5.3 This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation signal.This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.nsert 2t a.S~i~a  Cook Nuclear Plant Unit 2 B3758Rvso o B3.7.5-8 Revision No. 0 AFW System B 3.7.5 BASES SURVEILLANCE REQUI REM ENTS (continued)
The SR is modified by two Notes. Note 1 states that one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not otherwise inoperable.
This exception allows the AFW train(s) to be out of normal standby alignment and temporarily incapable of automatic initiation without declaring the train(s) inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, these manual operations are an accepted condition of the AFW System, OPERABILITY (i.e., the intended safety function) continues to be maintained.
Note 2 states that the SR is only required to be met in MODES 1, 2, and 3. It is not required to be met in MODE 4 since the AFW train is only required for the purposes of removing decay heat when the SG is relied Upon for heat removal. The operation of the AFW train is by manual means and automatic startup of the AFW train is not required.SR 3.7.5.4 This SR verifies that the AFW pumps will start in the event of any accident or transient that generates an ESFAS by demonstrating that each AFW pump starts automatically on an actual or simulated actuation signal in MODES 1, 2, and 3. Th ' o.t rqetaey ede19ae 2 This SR is modified by three Notes. Note I indicates that the SR may be deferred for the turbine driven AFW pump until suitable test conditions are established.
This deferral is required because there is insufficient steam pressure to perform the test at entry into MODE 3. At 850 psig, there is sufficient steam pressure to perform the test. Note 2 states that one or more AFW trains may be considered OPERABLE during alignment and operation for steam generator level control, if it is capable of being manually (i.e., remotely or locally, as appropriate) realigned to the AFW mode of operation, provided it is not Otherwise inoperable.
This exception allows the AFW train(s) to be out of normal standby alignment and temporarily incapable of auitomatic initiation without declaring the train(s)inoperable.
Since AFW may be used during startup, shutdown, hot standby operations, and hot shutdown operations for steam generator level control, these manual operations are an accepted condition of the AFW System. OPERABILITY (i.e., the intended safety function)Cook Nuclear Plant Unit 2 B3759Rvso o B3.7.5-9 Revision No. 0 CST B 3.7.6 BASES ACTIONS (continued) adequate to ensure the backup auxiliary feedwater supply continues to be available.
The 7 day Completion Time is reasonable, based on an OPERABLE backup auxiliary feedwater supply being available, and the low probability of an event occurring during this time period requiring the CST.B.1 and B.2 If any Required Action and associated Completion Time cannot be met, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 4, without reliance on the steam generator for heat removal, within 24 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR verifies that the CST contains the required volume of cooling water.
S-t eteae~ratyex' e~ nsert 2 ,ivrt an a~rse ldva -nfeC-- REFERENCES'
: 1. UFSAR, Section 10.5.2.2. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2B3.6-ReionN.2 B3.7.6-3 Revision No. 26 CCW System B 3.7.7 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that the applicable Conditions and Required Actions of LCO 3.4.6, "ROS Loops -MODE 4," be entered if an inoperable CCW train results in an inoperable RHR loop.This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components.
If one CCW train is inoperable, action must be taken to restore OPERABLE status within 72 hours. In this condition, the remaining OPERABLE COW train is adequate to perform the heat removal function.The 72 hour Completion Time is reasonable, based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this period.B.1 and B.2 If the COW train cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.7.1 REQU IREM ENTS This SR is modified by a Note indicating that the isolation of the COW flow to individual components may render those components inoperable but does not affect the OPERABILITY of the COW System.Verifying the correct alignment for manual, power operated, and automatic valves in the COW flow path provides assurance that the proper flow paths exist for COW operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing.
This SR also does not apply to valves that cannot be inadvertently misaligned, such as check valves. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.:Fhee-l 2)e ~ ~ e Cook Nuclear Plant Unit 2 B3773Rvso o B3.7.7-3 Revision No. 0 CCW System B 3.7.7 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.7.2 This SR verifies proper automatic operation of the CCW valves on an actual or simulated actuation signal. The COW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation.
This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. -Th e n~el4 2~~~e~d ~ Itosa-feydf8aut oaaea~~ e-ee-~-4r.napan~-rasetf ue~a we m4.w
~
shw~a-4ee m retsuual~s-4eu-elaew a
er=fr-Feq
~ueR ey-is' eiht.SR 3.7.7.3 This SR verifies proper automatic operation of the COW pumps on an actual or simulated actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal -nsert 2 te euog~ne~eeeti~e-ar-~aadt-asin i hs shw-44tJ s -a~e4-naAFr~fue f ~eF--~ee~aeeept~be4f.e n--reiabyit9tI-pee~into.-
REFERENCES
: 1. UFSAR, Section 9.5.2. UFSAR, Table 9.5-3.Cook Nuclear Plant Unit 2 B3774Rvso o B 3.7.7-4 Revision No. 0 ESW System B 3.7.8 BASES SURVEILLANCE REQUIREMENTS (continued) rather, it involves verification that those valves capable of being mispositioned are in the correct position.
This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.T-he SR 3.7.8.2 jageRt~-ri 4oi~oa ee".R@-av-e-a a-r-2 This SR verifies proper automatic operation of the ESW valves on an actual or simulated actuation signal. The ESW System is a normally operating system that cannot be fully actuated as part of normal testing.This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls.Srn 2 .11 r~:2 Iii :i ri.-,1--..-J.-~-'----.---.-I---..-.............
-r Insert tmeth~e~r+th uP8aa84.airtf ewr-ee-e4ihhereee-{.ra4 44 Insert ~+I rt I ] JSJ IJ E2 gJ J AI&#xa3;J k4JLLJJ J JJJJ T-hef~r~t9e-Freqiiei~ey~is-eepabte-fna-ehb1iat~y-standpein+f.
SR 3.7.8.3 This SR verifies proper automatic operation of the required ESW pumj on an actual or simulated actuation signal.
~rt~~ u p dr ee4 2-e~-Fr~m-ps--Insert 2 REFERENCES
: 1. UFSAR, Section 9.8.3.2. UFSAR, Section 9.8.3.2.3. UFSAR, Section 9.5.2.Cook Nuclear Plant Unit 2 B3784Rvso o B 3.7.8-4 Revision No. 0 UHS B 3.7.9 BASES SURVEILLANCE REQUIREMENTS (continued) determining the UHS temperature is averaging the available operating circulating water pumps discharge 2~-pra~-x-e~
Nver4 REFERENCES
: 1. UFSAR, Section 10.6.2.2. UFSAR, Table 9.8-5.3. Regulatory Guide 1.27, Revision 2, January 1976.4. MD-12-ESW-1 06-N Assessment of Increased Lake Water Temperature on Safety Related and Non-Safety Related Systems.Cook Nuclear Plant Unit 2B3.9-ReionN.5 B3.7.9-3 Revision No. 50 CREV System B 3.7.10 BASES ACTIONS (continued)
An alternative to Required Action E.1 is to immediately suspend activities that could result in a release of radioactivity that might require isolation of the ORE (Required Action E.2). This places the unit in a condition where the LCO does not apply. This does not preclude the movement of fuel to a safe position.F.1 During movement of irradiated fuel assemblies in the containment, auxiliary building, or Unit 1 containment, with two CREV trains inoperable, or with one or more CREV System trains inoperable due to inoperable ORE boundary, action must be taken immediately to suspend activities that could result in a release of radioactivity that requires isolation of the ORE. This places the unit in a condition that minimizes the accident risk.This does not preclude the movement of fuel to a safe position.G.1 If both CREV trains are inoperable in MODE 1, 2, 3, or 4 for reasons other than an inoperable ORE boundary or filter unit (i.e., Conditions B and C), the CREV System may not be capable of performing the intended function and the unit is in a condition outside the accident analyses.Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.7.10.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly.
As the environment and normal operating conditions on this system are not too severe, testing each trai n-cnc cvcry 02 dayz provides an adequate check of this system. Operating the CREV train, with flow through the HEPA filter and charcoal adsorber train, for> 15 minutes demonstrates the function of the CREV train.
Insert 2 SR 3.7.10.2 This SR verifies that the required CREV System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum and maximum flow rate, and the physical properties of the activated charcoal.
Specific test Frequencies and additional information are discussed in detail in the VFTP.Cook Nuclear Plant Unit 2 B 3.7.10-6 Revision No. 23 Cook Nuclear Plant Unit 2 B 3.7.10-6 Revision No. 23 CREV System B 3.7.10 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.10.3 This SR verifies that each CREV train starts and operates on an actual or simulated actuation signal. The only actuation signal necessary to be verified is the Safety Injection (SI) signal, since the Control Room Radiation
-High signal is not assumed in the accident analysis.
A Note has been included that states the Surveillance is only required to be met in MODES 1, 2, 3, and 4, since these are the MODES the SI signal is assumed to start the CREV trains. The CREV trains are assumed to be manually started during a fuel handling accident.
Insert 2 i~~e~e-th-2ma-Fel -fret elec4--
ie=-- ~s-u{e 4er+ha4 --~ bJtyt SR 3.7.10.4 This SR verifies the OPERABILITY of the CRE boundary by testing for unfiltered air inleakage past the CRE boundary and into the CRE. The details of the testing are specified in the Control Room Envelope Habitability Program.The CRE is considered habitable when the radiological dose to CRE occupants calculated in the licensing basis analyses of DBA consequences is no more than 5 rem TEDE, the CRE occupants are protected from smoke, and analyses demonstrate that the CREV System is not needed to prevent exceeding hazardous chemical limits. This SR verifies that the unfiltered air inleakage into the CRE is no greater than the flow rate assumed in the licensing basis analyses of DBA consequences.
When unfiltered air inleakage is greater than the assumed flow rate, Condition B must be entered. Required Action B.3 allows time to restore the CRE boundary to OPERABLE status provided mitigating actions can ensure that the CRE remains within, the licensing basis habitability limits for the occupants following an accident.Compensatory measures are discussed in Regulatory Guide 1.196, Section C.2.7.3, (Ref. 4) which endorses, with exceptions, NEI 99-03, Section 8.4 and Appendix F (Ref. 5). These compensatory measures may also be used as mitigating actions as required by Required Action B3.2. Temporary analytical methods may also be used as compensatory measures to restore OPERABILITY (Ref. 6). Options for restoring the CRE boundary to OPERABLE status include changing the licensing basis DBA consequence analysis, repairing the CRE boundary, or a combination of these actions. Depending upon the nature of the problem and the corrective action, a full scope inleakage test may not be necessary to establish that the CRE boundary has been restored to OPERABLE status. There are no SRs to verify CREV System operability for hazardous chemicals or smoke.Cook Nuclear Plant Unit 2 B 3.7.10-7 Revision No. 23 Cook Nuclear Plant Unit 2 B 3.7.10-7 Revision No. 23 CRAC System B 3.7.11 BASES ACTIONS (continued) 0.1 and 0.2 During movement of irradiated, fuel, if the inoperable CRAC train cannot be restored to OPERABLE status within the required Completion Time, the OPERABLE CRAC train must be placed in operation immediately.
This action ensures that the remaining train is OPERABLE, that no failures preventing automatic actuation will occur, and that active failures will be readily detected.An alternative to Required Action 0.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room (Required Action 0.2). This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.D.__1 During movement of irradiated fuel assemblies, with two CRAC trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. This does not preclude the movement of fuel to a safe position.E.1 If both CRAC trains are inoperable in MODE 1, 2, 3, or 4, the CRAC System may not be capable of performing its intended function.Therefore, LCO 3.0.3 must be entered immediately.
SURVEILLANCE SR 3.7.11.1 and SR 3.7.11.2 REQUIREMENTS These SRs verify that the heat removal capability of each CRAC train is sufficient to maintain control room air temperature
< 850&deg;F. -T-he=-2-heuf
-Insert 2 thedC- -Ae~te of tewa a-imt-exhee weAer4
* l~e, e 1-ee Fr-eeu fv-etbeGR  REFERENCES
: 1. UFSAR, Section 9.10.Cook Nuclear Plant Unit 2B37113RvsoN.0 B 3.7.11-3 Revision No. 0 ESF Ventilation System B 3.7.12 BASES ACTIONS A.1 With one ESF Ventilation train inoperable, action must be taken to restore OPERABLE status within 7 days. During this time, the remaining* OPERABLE train is adequate to perform the ESF Ventilation System function.The 7 day Completion Time is appropriate because the risk contribution is less than that for the ECCS (72 hour Completion Time), and this system is not a direct support system for the ECCS. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and ability of the remaining train to provide the required capability.
B.1 If the ESF enclosure boundary is inoperable, the [SF Ventilation trains cannot perform their intended functions.
Actions must be taken to restore an OPERABLE ESF enclosure boundary within 24 hours. During the period that the ESF enclosure boundary is inoperable, appropriate compensatory measures consistent with the intent, as applicable, of GDC 19, 60, 64 and 10 CFR Part 100 should be utilized to protect plant personnel from potential hazards. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition.
The 24 hour Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures.
The 24 hour Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the [SF enclosure boundary.C.1 and C.2 If the ESF Ventilation train or [SF enclosure boundary cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Standby systems should be checked periodically to ensure that they function properly. wrivreimeiat-.an4-er~eaI--eper-at]
~ceei~t4eB&&#xf7;nsr pe~ga Cook Nuclear Plant Unit 2B3.123RvsoN.0 B 3.7.12-3 Revision No. 0 ESF Ventilation System B 3.7.12 BASES SURVEILLANCE REQUIREMENTS (continued)SR 3.7.12.2 This SR verifies that the required ESF Ventilation System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA filter performance, charcoal adsorbers efficiency, minimum and maximum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
Specific test Frequencies and additional information are discussed in detail in the VFTP.SR 3.7.12.3 This SR verifies that each ESF Ventilation train starts and operates on an actual or simulated actuation signal. One [SF Ventilation train is normally operating with flow bypassing the charcoal adsorber section.This test confirms that each train, when in standby, starts upon receipt of a Containment Pressure -High High signal and that the exhaust flow can be directed through the entire filter unit including the HEPA filter and charcoal adsorber section. Q#ler-at~g Iner 2 oempeae to rvil~ele-erfer-eel~t-t4I-e 2-.mPhF,~qu- .e:e4r SR 3.7.12.4 This SR verifies the integrity of the ESF enclosure.
The ability of the ESF enclosure to maintain a negative pressure, with respect to potentially uncontaminated adjacent areas, is periodically tested to verify proper functioning of the [SF Ventilation System. During the post accident mode of operation, the [SF Ventilation System is designed to maintain a slight negative pressure in the [SF enclosure, with respect to adjacent areas, at a flowrate < 22,500 cfm to prevent unfiltered leakage. *-Insert 2 Frelu Cook Nuclear Plant Unit 2B3.124RvsoN.0 B 3.7.12-4 Revision No. 0 FHAEV System B 3.7.13 BASES APPLICABILITY During movement of irradiated fuel in the auxiliary building, the FHAEV System is required to be OPERABLE to alleviate the consequences of a fuel handling accident.In MODE 1, 2, 3, 4, 5, or 6, the FHAEV System is not required to be OPERABLE since the FHAEV System is only credited during a fuel handling accident in the auxiliary building.ACTIONS LCO 3.0.3 is not applicable while in MODE 5 or 6. However, since irradiated fuel assembly movement can occur in MODE 1, 2, 3, or 4, the ACTIONS have been modified by a Note stating that LCO 3.0.3 is not applicable.
If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operations.
Entering LCO 3.0.3, while in MODE 1, 2, 3, or 4 would require the unit to be shutdown unnecessarily.
A.11 When the required FHAEV train is inoperable or not in operation during movement of irradiated fuel assemblies in the auxiliary building, action must be taken to place the unit in a condition in which the LCO does not apply. Action must be taken immediately to suspend movement of irradiated fuel assemblies in the auxiliary building.
This does not preclude the movement of fuel to a safe position.SURVEILLANCE REQUIREMENTS SR 3.7.13.1 This SR requires verification every -l2hu's that the required FHAEV train is operating with flow through the filter unit, including the HEPA filter__, and charcoal adsorber section. Verification includes fan status and also verifies that each charcoal bypass damper is closed. -ThoFrq uccy Insert 2 4&#xa3;her4' fi~-es te- E}  SR 3.7.13.2 Standby systems should be checked periodically to ensure that they function properly.
As the environmental and normal operating conditions on this system are not severe, testing each train once "er"-9 2-y.provides an adequate check on this system.Operating the required FHAEV train, with flow through the HEPA filter and charcoal adsorber train, for > 15 minutes demonstrates the function of the system. T-he--2-Efay-FFequeF1~-~seeth4mV~t Insert 2 Cook Nuclear Plant Unit 2B37133RvsoN.0 B 3.7.13-3 Revision No. 0 FHAEV System B 3.7.13 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.7.13.3 This SR verifies that the required FHAEV System testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The VFTP includes testing HEPA-filter performance, charcoal adsorber efficiency, minimum and maximum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).
Specific test frequencies and additional information are discussed in detail in the VFTP.SR 3.7.13.4 This SR verifies that the required FHAEV train actuates on an actual or simulated actuation signal. The test must verify that the signal automatically shuts down each of the Fuel Handling Area Supply Air System fans. @fe 2 us.at~s4h-S~i~reNhaeeeta-~-4-~
~ ~ y T- f --eqecyi-acef~t-~
SR 3.7.13.5 This SR verifies the integrity of the auxiliary building enclosure.
The ability of the pool storage area to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the FHAEV train. During the accident mode of operation, the FHAEV train is designed to maintain a slight negative pressure in the FHAEV train, to prevent unfiltered leakage. The FHAEV train is designed to maintain a pressure > 0.125 inches of vacuum water gauge with respect to atmospheric pressure at a flow rate of<27,000 cfm. *'"-Insert 2 REFERENCES
: 1. UFSAR, Section 9.9.3.2.2. UFSAR, Section 14.2.1.3. 10 CFR 100.Cook Nuclear Plant Unit 2 B371- eiinN.2 B 3.7.13-4 Revision No. 26 Fuel Storage Pool Water Level B 3.7.14 BASES ACTIONS A.1j When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring.
When the fuel storage pool water level is lower than the required level, the movement of irradiated fuel assemblies in the fuel storage pool is immediately suspended to a safe position.
This action effectively precludes the occurrence of a fuel handling accident.
This does not preclude movement of a fuel assembly to a safe position.Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations.
Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.SURVEILLANCE REQUIREMENTS SR 3.7.14.1 This SR verifies sufficient fuel storage pool water is available in the event of a fuel handling accident.
The water level in the fuel storage pool must be checked periodically.
T~-7 Insertc-ys-sr c~-t-le-ypate-ed~~-ada--e-p~beae-rp
{a ex-per-iefc-e.
REFERENCES
: 1. UFSAR, Section 9.7.2.2. UFSAR, Section 9.4.3. UFSAR, Section 14.2.1.4. 10OCFR100.11.
Cook Nuclear Plant Unit 2 B371- eiinN.2 B 3.7.14-2 Revision No. 26 Fuel Storage Pool Boron Concentration B 3.7.15 BASES SURVEILLANCE REQUIREMENTS (continued) accidents are fully addressed. -T-lhe-7--eey-r-eqluerey-4e-appr-epfr-iae Insert 2 be
~t- -pe4a~ 4t~4tshQ-p eriG eJ-of-fme.
REFERENCES
: 1. Double contingency principle of ANSI Ni16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1 .2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).2. UFSAR, Section 9.7.2.Cook Nuclear Plant Unit 2 B371- eiinN.2 B 3.7.15-3 Revision No. 24
*. Secondary Specific Activity B 3.7.17 BASES ACTIONS A.1 and A.2 Specific activity of the secondary coolant exceeding the allowable value is an indication of a problem in the RCS and contributes to increased post accident doses. If the secondary specific activity is not within limits, the unit must be placed in.a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours, and in MODE 5 within 36 hours. The all owed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE SR 3.7.17.1 REQUIREMENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis.
A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases.
It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor coolant activity or LEAKAGE..TC1o-day-F-queny-isb&~H =etitre-f.-f -rbe-al-tuv-v -U e L-X&1imiit.
REFERENCES
: 1. 10CFR100.11.
: 2. 10 CFR 50, Appendix A, G DC 19.3. UFSAR, Section 14.2.7.Cook Nuclear Plant Unit 2B37.-3RvsoN.0 B 3.7.17-3 Revision No. 0 AC Sources -Operating B3 3.8.1 BASES ACTIONS (continued) continued operation.
The unit is required by LCO 3.0.3 to commence a controlled shutdown.SURVEILLANCE REQUIREMENTS The AC sources are designed to permit inspection and testing of all important areas and features, especially those that have a standby function, in accordance with Plant Specific Design Criterion (PSDC) 39 (Ref. 8). Periodic component tests are supplemented by extensive functional tests during refueling outages (under simulated accident conditions).
The SRs for demonstrating the OPERABILITY of the DGs are in accordance with the recommendations of Regulatory Guide 1 .9 (Ref. 3), Regulatory Guide 1.108 (Ref. 9), Regulatory Guide 1.137 (Ref. 10), and IEEE Standard 387-1995 (Ref. 11) as addressed in the applicable SR discussion.
Where the SRs discussed herein specify voltage and frequency tolerances, the following is applicable.
The minimum steady state output voltage of 3910 V is 94% of the nominal 4160 V output voltage. This value allows for voltage drop to the terminals of 4160 V motors whose minimum operating voltage is specified as 90% or 3740 V. It also allows for voltage drops to motors and other equipment down through the 120 V level Where the minimum operating voltage is also usually specified as 90% of nameplate rating. The specified maximum steady state output voltage of 4400 V is equal to the maximum operating voltage specified for 4000 V motors. It ensures that for a lightly loaded distribution system, the voltage at the terminals of 4000 V motors is no more than the maximum rated operating voltages.
The specified minimum and maximum steady state frequencies of the DG are 59.4 Hz and 60.5 Hz, respectively.
These values ensure the ESF pumps can achieve adequate fluid flow to meet their safety and accident mitigation functions.
SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power supply to the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies that each breaker is in its correct position to ensure that the required qualified offsite circuits are OPERABLE, and that appropriate independence of offsite circuits is maintained.
T1te-7 s
2 ne~kt-~h~ewtet4ee~rae~i4jaaef~ stats-i-is~ayede-eon-tfeeo t-r~eom.SR 3.8.1.2 and SR 3.8.1.8 These SRs help to ensure the availability of the standby electrical power Cook Nuclear Plant Unit 2 t3 3.8.1-16 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.
To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 1 for SR 3.8.1.2 and Note for SR 3.8.1.8) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and followed by a warmup period prior to loading.For the purposes of SR 3.8.1.2 and SR 3.8.1.8 testing, the DGs are started from standby conditions.
Standby conditions for a DG means that the diesel engine coolant and oil are being continuously circulated and temperature is being maintained consistent with manufacturer recommendations.
In order to reduce stress and wear on diesel engines, the manufacturer recommends a modified start in which the DGs are gradually accelerated to synchronous speed prior to loading. These start procedures are the intent of Note 2.SR 3.8.1.8 requires that, at a 184 day Frequency, the DG starts from standby conditions and achieves required voltage and frequency within 10 seconds. The 10 second start requirement supports the assumptions of the design basis LOCA analysis in the UFSAR, Section 14.3 (Ref. 5).The 10 second start requirement is not applicable to SR 3.8.1.2 (see Note 2 of SR 3.8.1.2) when a modified start procedure as described above is used. If a modified start is not used, the 10 second start requirement of SR 3.8.1.8 applies.Since SR 3.8.1.8 requires a 10 second start, it is more restrictive than SR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2.In addition, the DG is required to maintain proper voltage and frequency limits after steady state is achieved.
The voltage and frequency limits are normally achieved within 10 seconds. The time for the OG to reach steady state operation, unless the modified OG start method is employed, is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance.risert 2
.-- ---te yfrS sar teir 4n-e#ts~ -ers~etwtaGe W Frqte~eel MT Cook Nuclear Plant Unit 2 B3811 eiinN.4 B 3.8.1-17 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.3 Consistent with Regulatory Guide 1.9 (Ref. 3), this Surveillance verifies that the DGs are capable of synchronizing with the offsite electrical system and accepting loads 90% to 100% of the continuous rating of the 0G. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0.The 0.8 value is the design rating of the machine, while the 1.0 is an operational goal to ensure circulating currents are minimized.
The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections being required in order to maintain DG reliability.
They--refueey-fer-4-91J-at~viIaee-is--esist i fflrR-egtjfato 2
This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized.
Note 2 states that momentary transients, because of changing bus loads, do not invalidate this test. Note 3 indicates that this Surveillance should be conducted on only one Unit 2 DG at a time in order to avoid common cause failures that might result from offsite circuit or grid perturbations.
Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.
SR 3.8.1.4 This SR provides verification that the level of fuel oil in the day tank is above the level at which fuel oil is automatically added. The level is expressed as an equivalent volume in gallons, of which 31.4 gallons is unusable (due to tank geometry and vortexing considerations) and 70 gallons is usable, and is selected to ensure adequate fuel oil for greater than 15 minutes of DG operation at full load.T-he-fulet-oiHs~il-~nee~ew -lever -aer-s-ar-p Fevided-ai9fae-fy ete~er-tef-weul-beeee-awef-af~ya-r-fe-uesefse-ed-tfl19e-eil i&-psfie.nsert 2 SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must Cook Nuclear Plant Unit 2 B3811 eiinN.4 B 3.8.1-18 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) have a water environment in order to survive. Removal of water from each fuel oil day tank eliminates .the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria.
Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. T-h~~~e-iP1le~ee7 Insert 2 SR 3.8.1.6 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available.
While the system design requirements provide for two engine start cycles from each of the two air start receivers associated with each DG without recharging, only one start sequence is required to meet the OPERABILITY requirements (since the accident analysis assumes the OG starts on the first attempt).The pressure specified in this SR reflects the lowest value at which one DG start can be accomplished with one air start receiver.T-lormc to al r-h-~ tth -ap .raiitoryt, SR 3.8.1.7 This Surveillance demonstrates that each required fuel oil transfer pump (one per fuel oil transfer system) operates automatically and transfers fuel oil from its associated storage tank to its associated day tank. This is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and the controls and control systems for automatic fuel transfer systems are OPERABLE.T~eF~qer --
e -.G@&-re~eds-t4e-th-est4a-re4if=met6-fer-3un~sc--i~a~ie-it-a fln-- 4.-Insert 22 Cook Nuclear Plant Unit 2 B3811 eiinN.4 B 3.8.1-19 Revision No. 41 AC Sources -Operating B 3.8.1 I BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.9 Automatic transfer of each 4.16 kV emergency bus power supply from the normal auxiliary circuit to the preferred offsite circuit and the manual alignment to the alternate required offsite circuit demonstrates the OPERABILITY of the required offsite circuit to power the shutdown loads.~ Insert 2 As noted (Note 1Ito SR 3.8.1.9), SR 3.8.1.9.a is only required to be met when the auxiliary source is supplying the onsite electrical power subsystem.
This is acceptable since the preferred offsite source would be supplying the onsite electrical power subsystem and a transfer would not be necessary.
SR 3.8.1.10 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding a predetermined frequency and while maintaining a specified margin to the overspeed trip. Voltage and frequency are also verified to reach steady state conditions within 2 seconds. For this unit, the single load for each DG is 600 kW. This Surveillance may be accomplished by: a. Tripping the DG output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while paralleled to offsite power, or while solely supplying the bus; or b. Tripping its associated single largest post-accident load with the DG solely supplying the bus.Consistent with Regulatory Guide 1.9 (Ref. 3), the load rejection test is acceptable if the increase in diesel speed does not exceed 75% of the difference between synchronous speed and the overspeed trip setpoint, or 15% above nominal speed, whichever is lower. This corresponds to 64.4 Hz, which is the nominal speed pius 75% of the difference between nominal speed and the overspeed trip setpoint.The time. voltagqe, and frequency tolerances specified in this SR are CookNuclear Plant Unit 2 B 3.8.1-20 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) derived from Regulatory Guide 1.9 (Ref. 3) recommendations for response during load sequence intervals.
The 2 seconds specified is equal to approximately 60% of the 3.49 second load sequence interval associated with sequencing of the largest load. The voltage and frequency specified are consistent with the design range of the equipment powered by the DG. SR 3.8.1.10.a corresponds to the maximum frequency excursion, while SR 3.8.1.10O.b and SR 3.8.1.'10O.c are steady state voltage and frequency values to which the system must recover following load rejection. uen qency-is -ei -i eaiera j.d 4 _4i-e ~-o e-&gex9eiacehs4 w 2 F-reR heneyfT-he 1~ 4 -et4ie-F=F ueai4l le4t-e4be-a--eleptel~e-This SR is modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR. Credit may be taken for unplanned events that satisfy this SR.Note 2 ensures that the DG is tested under load conditions that are as close to design basis conditions as possible.
When synchronized with offsite power, testing should be performed at a power factor of-< 0.86.This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions.
Under certain conditions, however, Note 2 allows the Surveillance to be conducted at a power factor other than _< 0.86. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to < 0.86 results in voltages on the emergency busses that are too high.Cook Nuclear Plant Unit 2 B 3.8.1-21 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
Under these conditions, the power factor should be maintained as close as practicable to 0.86 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the DG excitation levels needed to obtain a power factor of 0.86 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained as close as practicable to 0.86 without exceeding the DG excitation limits.SR 3.8.1.11 Consistent with Regulatory Guide 1 .9 (Ref. 3), paragraph C.2.2.8, this Surveillance demonstrates the DG capability to reject a full load (90% to 100% of the DG continuous rating) without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping.
This Surveillance ensures proper engine generator load response under the simulated test conditions.
This test simulates the loss of the total connected load that the DG experiences following a full load rejection and verifies that the DG does not trip upon loss of the load. These acceptance criteria provide for DG damage protection.
While the DG is not expected to experience this transient during an event and continues to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated.Insert 2 This SR has been modified by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could cause perturbation to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the.potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Cook Nuclear Plant Unit 2 B 3.8.1-22 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) loaded without undue hardship or potential for undesired operation.
For instance, Emergency Core Cooling Systems (ECCS) injection valves are not desired to be stroked open, or centrifugal charging trains are not capable of being operated at full flow, or residual heat removal (RHR)trains performing a decay heat removal function are not desired to be realigned to the ECCS mode of operation.
In lieu of actual demonstration of connection and loading of loads, testing that adequately shows t[he capability of the DG systems to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.,T-he' r-eY i M e. ey-H.l M I-MI LL h AL -I M --. a e.-e I-IeMgIeerM-II ULMIL~. IMIiMg-jM gereLI ,--ki " l"-nsert 2This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the D.Gs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
The reason for Note 2 is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4, is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing
*OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.13 Consistent with Regulatory Guide 1 .9 (Ref. 3), paragraph C.2.2.5, this Surveillance demonstrates that the DG automaticallv starts and achieves Cook Nuclear Plant Unit 2 B 3.8.1-24 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) the required voltage and frequency within the specified time (10 seconds)from the design basis actuation signal (ESF actuation signal). In addition, the DG is required to maintain proper voltage and frequency limits after steady state is achieved.
The voltage and frequency limits are normally achieved within 10 seconds. The time for the DG to reach the steady state voltage and frequency limits is periodically monitored and the trend evaluated to identify degradation of governor and voltage regulator performance.
The DG is required to operate for > 5 minutes. The 5 minute period provides sufficient time to demonstrate stability.
SR 3.8.1.13.d and SR 3.8.1.13.e ensure that permanently connected loads and emergency loads are energized from the offsite electrical power system on an ESF signal without loss of offsite power.The requirement to verify the connection of permanent and auto-connected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation.
For instance, ECCS injection valves are not desired to be stroked open, or centrifugal charging trains are not capable of being operated at full flow, or RHR trains performing a decay heat removal function are not desired to be realigned to the ECOS mode of operation.
In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.=l'eFrqee'f2"etssasde-niern-~eet i Insert 2.h eeee~f-fi--ali~
4 ~ i 44ebe- se-,sewe-that-the -uee e11y-aes-4 SRw al re- e-tte2- ~4--eeUecy.--
d eei This SR is modified by two Notes. The reason for Note I is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing Cook Nuclear Plant Unit 2 B3812 eiinN.4 B 3.8.1-25 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.14 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph 0.2.2.12, this Surveillance demonstrates that DG noncritical protective functions (e.g., low lube oil pressure) are bypassed on a loss of voltage signal or an ESE actuation test signal. The noncritical trips are bypassed during DBAs and provide an alarm on an abnormal engine condition.
This alarm provides the operator with sufficient time to react appropriately.
The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG.
4t Insert 2 Onp er-t~ge-xn
~c-e -ha-siw-h ewt-that-eserptet--euereu~elltylass-t49e The SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DG from service. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when the~y are tied together or operated independently for the Surveillance:
Cook Nuclear Plant Unit 2 B 3.8.1-26 Revision Nol 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.15 This Surveillance demonstrates the DGs can start and run continuously at full load capability (90% to 100% of the DG continuous rating) for an interval of not less than 8 hours. The run duration of 8 hours is consistent with IEEE Standard 387-1995 (Ref. 11). The DG starts for this Surveillance can be performed either from standby or hot conditions.
The provisions for prelubricating and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.The load band is provided to avoid routine overloading of the 0G.Routine overloading may result in more frequent teardown inspections being required in order to maintain DG reliability.
~ ~ rgee4gj~~,ta~gt Insert 2 This Surveillance is modified by three Notes. Note 1 statesthat momentary transients due to changing bus loads do not invalidate this test.Similarly, momentary power factor transients above the power factor limit will not invalidate the test. The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems.This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with Cook Nuclear Plant Unit 2 B 3.8.1-27 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR. Note 3 ensures that the DG is tested under load conditions that are as close to design basis conditions as possible.
When synchronized with offsite power, testing should be performed at a power factor of < 0.86. This power factor is representative of the actual inductive loading a DG would see under design basis accident conditions.
Under certain conditions, however, Note 3 allows the Surveillance to be conducted as a power factor other than < 0.86. These conditions occur when grid voltage is high, and the additional field excitation needed to get the power factor to - 0.86 results in voltages on the emergency busses that are too high.Under these conditions, the power factor should be maintained as close as practicable to 0.86 while still maintaining acceptable voltage limits on the emergency busses. In other circumstances, the grid voltage may be such that the DG excitation levels needed to obtain a power factor of 0.86 may not cause unacceptable voltages on the emergency busses, but the excitation levels are in excess of those recommended for the DG. In such cases, the power factor shall be maintained close as practicable to 0.86 without exceeding the DG excitation limits.SR 3.8.1.16 This Surveillance demonstrates that the diesel engine can restart from a hot condition, such as subsequent to shutdown from normal Surveillances, and achieve the required voltage and frequency within 10 seconds. The 10 second time is derived from the requirements of the.accident analysis to respond to a design basis large break LOCA. --T-he- lsr This SR is modified by two Notes. Note 1 ensures that the test is performed with the diesel sufficiently hot. The load band is provided to avoid routine overloading of the DG. Routine overloads may result in more frequent teardown inspections being required in order to maintain DG reliability.
The requirement that the diesel has operated for at least 2 hours at full load conditions prior to performance of this Surveillance is based on operating experience for achieving hot conditions.
Momentary transients due to changing bus loads do not invalidate this test. Note 2 allows all DG starts to be preceded by an engine prelube period to minimize wear and tear on the diesel during testing.Cook Nuclear Plant Unit 2 B 3.8.1-28 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.1.17 Consistent with Regulatory Guide 1.9 (Ref. 3), paragraph C.2.2.1 1, this Surveillance ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and the DG can be returned to ready-to-load status when offsite power is restored.
It also ensures that the auto-start logic is reset to allow the DG to reload if a subsequent loss of offsite power occurs. The DG is considered to be in ready-to-load status when the OG is running at rated speed and voltage, with the DG output breaker open.* ~ g Insert 2 This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns)provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when the Surveillance is performed in MODE 1, 2, 3, or 4.Risk insights or deterministic methods may be used for this assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.18 Under accident conditions loads are sequentially connected to the bus by the individual time delay relays. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs or RATs (as applicable) due to high motor starting currents.Verifying the load sequencer time within plus or minus 5% of its required value ensures that sufficient time exists for the DG to restore frequency and voltage and RATs to restore voltage prior to applying the next load and that safety analysis assumptions regarding ESF equipment time delays are not violated.
Reference 4 provides a summary of the Cook Nuclear Plant Unit 2 B 3.8.1-29 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) automatic loading of emergency buses.
ta~g n se rt 2eyey-d -c4 el4te-ee-aec-epttae-fr--el-a-r~el1-abiPity&#xb6;ai1
.This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems.This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the~avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.19 In the event of a DBA coincident With a loss of offsite power, the DGs are required to supply the necessary power to ESF systems so that the fuel, ROS, and containment design limits are not exceeded.This Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.12, during a loss of offsite power actuation test signal in conjunction with an ESF actuation signal. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.-Th a 4gree4jj~m~-tkfg nsert 2 Cook Nuclear Plant Unit 2 B 3.8.1-30 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
This SR is modified by two Notes. The reason for Note I is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations for DGs. The reason for Note 2 is that the performance of the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This a~ssessmnent shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.20 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing that involves connecting the DG to its test load resistor bank, and the DG will automatically reset to ready to load operation if a ESE actuation signal is received during operation in the test mode. Ready to load operation is defined as the DG running at rated speed and voltage with the DG output breaker open.The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.13.
The intent in the requirement associated with SR 3.8.1.20.b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that Cook Nuclear Plant Unit 2 B 3.8.1-31 Revision No. 41 AC Sources -Operating 8 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued) adequately shows the capability of the emergency loads to perform these fun ctions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.Insert 2 This SR is modified by two Notes. Note I states that this Surveillance is only required to be met when the applicable DG is connected to its test load resistor bank. This is allowed since the test mode override only functions when the DG is connected to its associated test load resistor bank. When the DG is not connected to its associated test load resistor bank, the feature is not necessary; thus the Surveillance is not required to be met under this condition.
The reason for Note 2 is that performing the Surveillance would remove a required DG from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope'with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.21 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing and the DG will automatically reset to ready to load operation if a LOCA actuation signal is received during operation in the test mode. Ready to load operation is defined as the DG running at rated speed and voltage Cook Nuclear Plant Unit 2 B 3.8.1-32 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUI REMENTS (continued) with the DG output breaker open.The requirement to automatically energize the emergency loads with offsite power is essentially identical to that of SR 3.8.1.13.
The intent in the requirement associated with SR 3.8.1.21 .b is to show that the emergency loading was not affected by the DG operation in test mode. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable.
This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.-Th thae Saeg~e~a-urva e-i -in tecc.~ w~a4
=-=---nsert 2 irr,-,-r,-, n,ini, ',.-nrdu .4 cr1*r -, ,-r'nn+-, li o -,-nI k]i- c'4.A ,nin This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required DG from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determinethat unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.SR 3.8.1.22 This Surveillance demonstrates that the DG starting independence has*not been compromised.
Also, this Surveillance demonstrates that each engine can achieve proper speed within the specified time when the DGs are started simultaneously.
Cook Nuclear Plant Unit 2 B 3.8.1-33 Revision No. 41 AC Sources -Operating B 3.8.1 BASES SURVEILLANCE REQUIREMENTS (continued)
T-ho10-ycarF-r,,g
...ncyi-c-ea~et~h~e-r~ees-at eg&e Insert 2 This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil continuously circulated and temperature maintained consistent with manufacturer recommendations.
SR 3.8.1.23 With the exception of this Surveillance, all other Surveillances of this Specification (SR 3.8.1.1 through SR 3.8.1.22) are applied to Unit 2 sources. This Surveillance is provided to direct that appropriate Surveillances for the required Unit I AC sources are governed by the applicable Unit 1 Technical Specifications.
Performance of the applicable Unit 1 Surveillances will satisfy the Unit I requirements as well as satisfy this Unit 2 Surveillance Requirement.
Exceptions are noted to the Unit 1 SRs of LCO 3.8.1. SR 3.8.1 .9.b is not required to be met since only one offsite circuit is required to be OPERABLE.
SR 3.8.1.13, SR 3.8.1.14 (ESF actuation signal portion only), SR 3.8.1.19, SR 3.8.1.20, and SR 3.8.1.21 are not required to be met because the ESF actuation signal is not required to be OPERABLE.
SR 3.8.1.22 is excepted because starting independence is not required with the DG(s) that is not required to be OPERABLE.The Frequency required by the applicable Unit 1 SR also governs performance of that SR for Unit 2.As noted (Note I to SR 3.8.1.23), if Unit 1 is in MODE 5 or 6, or moving irradiated fuel assemblies, SR 3.8.1.3, SR 3.8.1.10 through SR 3.8.1.12, SR 3.8.1.14 through SR 3.8.1.17, and SR 3.8.1.18 are not required to be performed.
This ensures that this Unit 2 SR will not require a Unit 1 SR to be performed, when the Unit I Technical Specifications exempts performance of a Unit 1 SR (however, as stated in the Unit I SR 3.8.2.1 Note 1, while performance of an SR exempted, the SR must still be met).As noted (Note 2 to SR 3.8.1.23), SR 3.8.1.9.a is only required to be met when the auxiliary source is supplying the Unit I electrical power distribution subsystem since the preferred offsite source is required to support Unit 2 operations.
REFERENCES
: 1. Atomic Energy Commission Proposed General Design Criterion 39, July 1967.2. UFSAR, Section 8.3.Cook Nuclear Plant Unit 2 B 3.8.1-34 Revision No. 4 Diesel Fuel Oil B 3.8.3 BASES SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support each DG's operation for 7 days at full load.The 7 day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.f .=-Insert 2 SR 3.8.3.2 The tests listed below are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion.
If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage tank(s). The tests, limits, and applicable ASTM Standards are as follows: a. Sample the new fuel oil in accordance with ASTM 04057-8 1 (Ref. 5);b. Verify that the sample has: (1) when tested in accordance with ASTM D1298-80 (Ref. 5) an absolute specific gravity at 60/60&deg;F of> 0.82 and -< 0.88, an API gravity at 60&deg;F of >- 300 and _< 400, an API gravity of within 0.3 degrees at 60&deg;F when compared to the supplier's certificate, or a specific gravity of within 0.0016 at 60/600 when compared to the supplier's certificate; (2) a kinematic viscosity at 40&deg;C of >- 1.9 centistokes and <- 4.1 centistokes or Saybolt viscosity at 1 00&deg;F of>_ 32.6 and < 40.1, if gravity was not determined by comparison with supplier's certification, when tested in accordance with ASTM 975-81 (Ref. 5); and (3) a flash point of_ 125&deg;F when tested in accordance with ASTM 0975-81 (Ref. 5); and c. Verify that the new fuel oil has a clear and bright appearance with proper color when tested in accordance with ASTM 04176-82 (Ref. 5).Failure to meet any of the above limits is cause for rejecting the new fuel oil, but does not represent a failure to meet the LCO concern since the fuel oil is not added to the storage tanks.Following the initial new fuel oil sample, the fuel oil is analyzed within 31 days following addition of the new fuel oil to the fuel oil storage tank(s)to establish that the other properties specified in Table I of Cook Nuclear Plant Unit 2 B 3.8.3-4 Revision No. 0 Diesel Fuel Oil B 3.8.3 BASES SURVEILLANCE REQUIREMENTS (continued)
ASTM D975-81 (Ref. 6) are met for new fuel oil when tested in accordance with ASTM 0975-81 (Ref. 5), except that the analysis for sulfur may be performed in accordance with ASTM D2622-82 (Ref. 5).The 31 day period is acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an immediate effect on OG operation.
This Surveillance ensures the -availability of high quality fuel oil for the DGs.Fuel oil degradation during long term storage shows up as an increase in particulate, due mostly to oxidation.
The presence of particulate does not mean the fuel oil will not burn properly in a diesel engine. The particulate can cause fouling of filters and fuel oil injection equipment, however, which can cause engine failure.Particulate concentrations should be determined in accordance with ASTM 02276-83, Method A (Ref. 5). This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/I. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing.The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.
SR 3.8.3.3 Microbiological fouling is a major cause of fuel oil degradation.
There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel storage tanks eiaee-e er-y.-4=-a.ys-eliminates the necessary environment for bacterial survival.
This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.
Water may come from any of several sources, including condensation, ground water, rain water, and contaminated fuel oil, and from breakdown of the fuel oil by bacteria.
Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight integrity of the fuel oil system. The Surveillance Frequencies are established by Regulatory Guide 1.137 (Ref. 2). This SR is for preventive maintenance.
The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during performance of the Surveillance.
Cook Nuclear Plant Unit 2 B 3.8.3-5 Revision No. 0 Cook Nuclear Plant Unit 2 B3.8.3-5 Revision No. 0 DC Sources -Operating B 3.8.4 BASES ACTIONS (continued)
E. 1 If one or both required Unit 1 Train A and Train B DC electrical power subsystems are inoperable, the associated ESW train(s) are not capable of performing their intended function.
Immediately declaring the affected supported feature, e.g., ESW train, inoperable allows the ACTIONS of LCO 3.7.8 to apply appropriate limitations on continued reactor operation.
SURVEILLANCE REQUIREMENTS SR 3.8.4.1 Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the battery chargers, which support the ability of the batteries to perform their intended function.
Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state while supplying the continuous steady state loads of the associated DC subsystem.
On float charge, battery cells will receive adequate current to optimally charge the battery. The voltage requirements are based on the nominal design voltage of the battery and are consistent with the minimum float voltage established by the battery manufacturer (2.20 Vpc or 255.2 VDC at the battery terminals of the Train A and Train B batteries and 2.20 Vpc or 257.4 VDC for the Train N battery).
This voltage maintains the battery plates in a condition that supports maintaining the grid life (expected to be approximately 20 years). The Ne7d~---eu~-~--~e-tM--e~re-~t 1 -'nsert 2........ ... \ ...... l" SR 3.8.4.2 This SR verifies the design capacity of the battery chargers.
According to Regulatory Guide 1 .32 (Ref. 9), the battery charger supply is recommended to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences.
The minimum required amperes and duration ensure that these requirements can be satisfied.
This SR requires that each Train A and Train B required battery charger be capable of supplying
> 300 amps at > 250 VDC for > 4 hours and the Train N battery charger is capable of supplying
> 25 amps at > 250 VDC for > 4 hours. The ampere requirements are based on the output rating of the chargers.
The voltage requirements are based on the charger voltage Cook Nuclear Plant Unit 2 B 3.8.4-7 Revision No. 0 Cook Nuclear Plant Unit 2 B3.8.4-7 Revision No. 0 DC Sources -Operating B 3.8.4 BASES SURVEILLANCE REQUIREMENTS (continued) level after a response to a loss of AC power. The time period is sufficient to detect significant charger failures.*
n 2 SR 3.8.4.3 A battery service test is a special test of the battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The battery charger must be disconnected throughout the performance of the battery service test. The discharge rate and test length should correspond to the design duty cycle requirements as specified in the applicable design documents.
isrei+-u eHt-eea c-2 .---T-her-efer-e,-t-he-Fr-e This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu of a service test.The reason for Note 2 is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.
These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3, or 4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 2 B3848Rvso o B3.8.4-8 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS SR 3.8.6.1 Verifying battery float current while on float charge is used to determine the state of charge of the battery. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a charged state.The float current requirements are based on the float current indicative of a charged battery. Use of float current to determine the state of charge of the battery is consistent with IEEE-450 (Ref. 1)..'TheW-eay-Fr-eqteree?-s.nr osrA''
* hah-eemra~eseq This SR is modified by a Note that states the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.8.4.1. When this float voltage is not maintained the Required Actions of LCO 3.8.4 ACTION A are being taken, which provide the necessary and appropriate verifications of the battery condition.
Furthermore, the float current limit of 2 amps is established based on the nominal float voltage value and is not directly applicable when this voltage is not maintained.
SR 3.8.6.2 and SR 3.8.6.5 I=-nsert 2 Optimal long term battery performance is obtained by maintaining a float voltage greater than or equal to the minimum established design limits provided by the battery manufacturer, which corresponds to 257.5 VDC for a 116 cell battery and 259.7 VDC for a 117 cell battery at the battery terminals, or 2.22 Vpc. This provides adequate over-potential, which limits the formation of lead sulfate and self discharge, which could eventually render the battery inoperable.
Float voltages in this range or less, but greater than 2.07 Vpc, are addressed in Specification 5.5.15.SRs 3.8.6.2 and 3.8.6.5 require verification that the cell float voltages are equal to or greater than the short term absolute minimum voltage of 2.07 V. The r ee~han9-e-2 y, cczh-- e <--"nsert 2 SR 3.8.6.3 The limit specified for electrolyte level (i.e., greater than or equal to the low level mark) ensures that the plates suffer no physical damage and maintains adequate electron transfer capability.
T-hf-Pr~equriyi 4 l--nsert 2 Cook Nuclear Plant Unit 2 B 3.8.6-5 Revision No. 0 Cook Nuclear Plant Unit 2 B3.8.6-5 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.8.6.4 This Surveillance verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (i.e., 60&deg;F for the Train A and Train B 250 VDC batteries and 45&deg;F for the Train N 250 VDC battery).
Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the design requirements.
Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity.
The e(ke.t) Insert 2 SR 3.8.6.6 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.6.6;however, only the modified performance discharge test may be used to satisfy the battery service test requirements of SR 3.8.4.3.A modified discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity.
Initial conditions for the modified performance discharge test should be identical to those specified for a performance discharge test as specified in IEEE-450 (Ref. 1).It may consist of just two rates; for instance the one minute rate for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the modified performance discharge test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test must remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test. Currently, the modified performance discharge test is performed by testing the battery using the service test profile for the first 4 hours followed by the performance discharge test profile for the Cook Nuclear Plant Unit 2 B3.8.6-6 Revision No. 0 Battery Parameters B 3.8.6 BASES SURVEILLANCE REQUIREMENTS (continued) remainder of the test. This method has been determined by the system engineer and the battery manufacturer to be an acceptable modified performance test procedure, and is consistent with I EEE-450 (Ref. 1).The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 1) and IEEE-485 (Ref. 3). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements.
Furthermore, the battery is sized to meet the assumed duty cycle loads when the battery design capacity reaches this 80% limit. net2 1r-ts4-ar-a{e t If the battery shows degradation, or if the battery has reached 85% of its expected life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 12 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity -> 100% of the manufacturer's ratings. Degradation is indicated, according to IEEE-450 (Ref. 1), when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is below 90% of the manufacturer's rating.The 12 month and 60 month Frequencies are consistent with the recommendations in IEEE-450 (Ref. 1). The 24 month Frequency is derived from the recommendations of IEEE-450 (Ref. 1).This SR is modified by a Note. The reason for the Note is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1, 2, 3, or 4 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines unit safety is maintained or enhanced.
This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.These shall be measured against the avoided risk of a unit shutdown and startup to determine that unit safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1, 2, 3 or4. Risk insights or deterministic methods may be used for the assessment.
Credit may be taken for unplanned events that satisfy this SR.Cook Nuclear Plant Unit 2 B3.8.6-7 Revision No. 0 Inverters
-Operating B 3.8.7 BASES ACTIONS (continued) inverter inoperability.
This has to be balanced against the risk of an immediate shutdown, along with the potential challenges to safety systems such a shutdown might entail. When the 120 VAC vital bus is powered from its regulated 600/1 20 VAC transformer, it is relying upon interruptible AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the 120 VAC vital buses is the preferred source for powering instrumentation trip setpoint devices.B.._11 With two inverters~in the same train inoperable, the remaining inverters are capable of supporting the minimum safety functions necessary to shut down the reactor and maintain it in a safe condition, assuming no single failure. The overall reliability is reduced, however, because a single failure in one of the two remaining inverters could result in the minimum ESF functions not being supported.
Therefore, one of the inverters must be restored to OPERABLE status within 6 hours.The 6 hour Completion Time is consistent with that allowed for an inoperable RTS train and an inoperable ESFAS train, since the inverters support the 120 VAC vital buses, which in turn support the RTS and ESFAS trains.C.1 and C.2 If the Train A or Train B inverter(s) cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.SURVEILLANCE REQUIREMENTS SR 3.8.7.1 This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and the 120 VAC vital buses energized from the associated inverter.
Each inverter must be connected to its associated 250 VDC bus. The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation of the RTS and ESFAS connected to the 120 VAC vital buses.
~ Insert 2-r Cook Nuclear Plant Unit 2 B3873Rvso o B3.8.7-3 Revision No. 0 Inverters
-Shutdown B 3.8.8 BASES SURVEILLANCE REQ U IREM ENTS SR 3.8.8.1 This Surveillance verifies that the required inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter.
The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation connected to the AC vital buses. -T-e-+ay-.F-eue-~~seeeeumte--~~l~-~a 2 REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2 B3884Rvso o B 3.8.8-4 Revision No. 0 Distribution Systems -Operating B 3.8.9 BASES SURVEILLANCE REQ UI REM ENTS SR 3.8.9.1 This Surveillance verifies that the required AC, DC, and 120 VAC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment.
The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus.The verification of proper voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads connected to these buses.
2~beye~eR~=i9~
UllUUIU REFERENCES
: 1. Safety Guide 6, March 1971.2. UFSAR, Chapter 14.3. Regulatory Guide 1.93, December 1974.Cook Nuclear Plant Unit 2B38910RvsoN.0 B 3.8.9-10 Revision No. 0 Distribution Systems -Shutdown B 3.8.10 BASES ACTIONS (continued) power is not available.
It is therefore required to declare the FHAEV System inoperable.
Since the Unit 1 AC electrical power distribution subsystem only affects the FHAEV System, the associated portions of the FHAEV System are declared inoperable and the applicable ACTIONS of LCO 3.7.13, "Fuel Handling Area Exhaust Ventilation (FHAEV) System," are entered.SURVEILLANCE REQUIREMENTS SR 3.8.10.1 This Surveillance verifies that the AC, DC, and 120 VAC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized.
The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. Since the Unit 1 AC electrical power distribution subsystem only*affects the FHAEV System, the SR is modified by a Note that specifies the SR is not required to be met for the Unit 1 AC electrical power distribution subsystem when the associated FHAEV System is not required to be OPERABLE per LCO 3.7.13. Te-d i-Fr-eiqsery ,ak"='-Insert 2 REFERENCES
: 1. UFSAR, Chapter 14.Cook Nuclear Plant Unit 2 B 3.8.10-4 Revision No. 0 Boron Concentration B 3.9.1 BASES ACTIONS (continued)
Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position.
Operations that individually add limited positive reactivity (e.g., temperature fluctuations from inventory addition or temperature control fluctuations), but when combined with all other operations affecting core reactivity (e.g., intentional boration) result in overall net negative reactivity addition, are not precluded by this action.A.3 In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore the concentration must be initiated immediately.
In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied.
The only requirement is to restore the boron concentration to its required value as soon as possible.
In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.
Once actions have been initiated, they must be continued until the boron concentration is restored.
The restoration time depends on the amount of boron that must be injected to reach the required concentration.
SURVEILLANCE REQUIREMENTS SR 3.9.1.1 and SR 3.9.1.2 These SRs ensure that the coolant boron concentration in the RCS, and connected portions of the refueling canal and the refueling cavity, is within the COLR limits. The boron concentration is determined periodically and prior to re-connecting portions of the refueling canal and the refueling cavity to the RCS, by chemical analysis...-.........
J.-..'--'-...-............................-
sa le-s.T-he.-t~9eruefy49-s13a.ec.r-eaereat4eg e-xper-e5'1tat e
e-adequ~e~e-.
The SR 3.9.1.2 Frequency of once within 72 hours prior to connecting the refueling canal and refueling cavity to the RCS ensures that if any dilution activity has occurred while the cavity and canal were disconnected from the ROS, correct boron concentration is verified prior to communication with the RCS.-Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.Cook Nuclear Plant Unit 2 B3913Rvso o B 3.9.1-3 Revision No. 0 Nuclear Instrumentation B 3.9.2 BASES ACTIONS (continued) since CORE ALTERATIONS and positive reactivity additions are not to be made, the core reactivity condition is stabilized until the source range neutron flux monitors are OPERABLE.
This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists.The Completion Time of once per 12 hours is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified.
The 12 hour Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.C.1 With no audible count rate OPERABLE, prompt and definite indication of a boron dilution event, consistent with the assumptions of the safety analysis, is lost. In this situation, the boron dilution event may not be detected quickly enough to assure sufficient time is available for operators to manually isolate the unborated water source and stop the dilution prior to the loss of SHUTDOWN MARGIN. Therefore, action must be taken to prevent an inadvertent boron dilution event from occurring.
This is accomplished by isolating all the unborated water flow paths to the Reactor Coolant System. Isolating these flow paths ensures that an inadvertent dilution of the reactor coolant boron concentration is prevented.
The Completion Time of "Immediately" assures a prompt response by operations and requires an operator to initiate actions to isolate an affected flow path immediately.
Once actions are initiated, they must be continued until all the necessary flow paths are isolated or the circuit is restored to OPERABLE status.SURVEILLANCE SR 3.9.2.1 REQUIREMENTS SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is normally a comparison of the parameter indicated on one channel to a similar parameter on another channel. It is based on the assumption that the two indication channels should be consistent with core conditions.
Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions.
The ~~Ilnsert 2 Cook Nuclear Plant Unit 2 B 3.9.2-3 Revision No. 0 Cook Nuclear Plant Unit 2 B3.9.2-3 Revision No. 0 Nuclear Instrumentation B 3.9.2 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.2.2 SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION-e-ver-y
* CHANNEL CALIBRATION is a complete check of the instrument loop, except the detector.
The CHANNEL CALIBRATION for the Westinghouse source range neutron flux monitors also includes obtaining the detector plateau or preamp discriminator curves, evaluating those curves, and comparing the curves to the manufacturer's data. In addition, the CHANNEL CALIBRATION includes verification of the audible count rate function for the required monitor. This SR is modified b~y a Note stating that neutron detectors are .excluded from the CHANNEL CALIBRATION. j.s -&#xa2;e el,{-e
"-Insert 2 REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Section 14.1.5.REFERENCES
: 1. UFSAR, Section 1.4.5.2. UFSAR, Section 14.1.5.Cook Nuclear Plant Unit 2 B3924Rvso o B 3.9.2-4 Revision No. 0 Containment Penetrations B 3.9.3 BASES SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations is in its required status. The LCO 3.9.3.c.2 status requirement, which requires penetrations to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System, can be verified by ensuring each required valve operator is capable of closing automatically if needed. This Surveillance does not require cycling of the valves since this is performed at the appropriate frequency in accordance with SR 3.9.3.2.-e1 i-e .ete nsert 2
~ t.q -sleeme~r thJe -~~-e.-- pe
--htr~ae-~-~ap-rae
-r-4-le-~SR 3.9.3.2 This Surveillance demonstrates that each required containment purge supply and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal. Th24rmr-'--nsert 2 Frgoc "anan ,~s~e* r-equl-- n. LC .. ,-6eeit-aRwe-Pure--St~pp]
a t- l? ! t I ....... II? _ _ [__" ........ The SR is modified by a Note stating that this Surveillance is not required to be met for valves in isolated penetrations.
The LCO provides the option to close penetrations in lieu of requiring automatic actuation capability.
REFERENCES
: 1. UFSAR, Section 14.2.1.5.Cook Nuclear Plant Unit 2 B3934Rvso o B3.9.3-4 Revision No. 1 RHR and Coolant Circulation
-High Water Level B 3.9.4 BASES SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core.
2 4-2-ee-s --ei ,-eef&sect;1er~ewieterg e-~ pdf~e~tple-aerfttr
, REFERENCES
: 1. UFSAR, Section 9.3.2.Cook Nuclear Plant Unit 2 B3944Rvso o B 3.9.4-4 Revision No. 0 RHR and Coolant Circulation
-Low Water Level B 3.9.5 BASES ACTIONS (continued)
B.._22 If no RHR loop is in operation, actions shall be initiated immediately, and continued, to restore one RHR loop to operation.
Since the unit is in Conditions A and B concurrently, the restoration of two OPERABLE RHR loops and one operating RHR loop should be accomplished expeditiously.
B.3, B.4, and B.5 If no RHR is in operation, the following actions must be taken: a. The equipment hatch must be closed and secured with four bolts;b. One door in each air lock must be closed; and c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere must be either closed by a manual or automatic isolation valve, blind flange, or equivalent, or verified to be capable of being closed by an OPERABLE Containment Purge Supply and Exhaust System.With RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere.
Performing the actions s'tated above ensures that all containment penetrations are either closed or can be closed so that the dose limits are not exceeded.The Completion Time of 4 hours allows fixing of most RHR problems and is reasonable, based on the low probability of the coolant boiling in that time.SURVEILLANCE REQUIREMENTS SR 3.9.5.1.This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal anid boron stratification in the core. In addition, during operation of the RHR loop with the water level in the vicinity of the reactor vessel nozzles, the RHR pump suction requirements must be met. T-heee-f--efsi-ufei I --nsertI 2 p
*mat 4a4~--R-Se-r4  Cook Nuclear Plant Unit 2 B 3.9.5-3 Revision No. 0 Cook Nuclear Plant Unit 2 B 3.9.5-3 Revision No. 0 RHR and Coolant Circulation
-Low Water Level B 3.9.5 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.9.5.2 Verification that the required pump is OPERABLE ensures that an additional RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.
Verification is performed by verifying proper breaker alignment and power available to the required pump.
This SR is modified by a Note that states the SR is not required to be performed until 24 hours after a required pump is not in operation.Insert 2 REFERENCES
: 1. UFSAR, Section 9.3.2.Cook Nuclear Plant Unit 2 B3954Rvso o B 3.9.5-4 Revision No. 0 Refueling Cavity Water Level B 3.9.6 BASES.ACTIONS A.1 With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.SURVEILLANCE SR 3.9.6.1 REQ U IREM ENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met.Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 1).REFERENCES
: 1. UFSAR, Section 14.2.1.2. 10 CFR 100.10.2 Cook Nuclear Plant Unit 2B39.-ReionN.2 B3.9.6-2 Revision No, 26 Enclosure 8 to AEP-NRC-2015-46 TSTF-425 (NUREG-1431) versus CNP TS Cross- Reference Technical Specification Section Title TSTF-425 CNP CNP____________________________________
______Unit I Unit 2 Reactivity Controls Systems / Shutdown Margin 3.1.1 3.1.1.1 3.1.1.1 3.1.1.1 Core Reactivity 3.1.2 3.1.2.1 3.1.2.1 3.1.2.1 Rod Group Alignment Limits 3.1.4 3.1.4.1 3.1.4.1 3.1.4.1 3.1.4.2 3.1.4.2 3.1.4.2 Shutdown Bank Insertion Limits 3.1.5 3.1.5.1 3.1.5.1 3.1.5.1 Conro Bnk nsrtonLimts3..63.1.6.2 3.1.6.2 3.1.6.2 3.1.6.3 3.1.6.3 3.1.6.3 PHYSICS TESTS Exceptions
-MODE 2 3.1.8 3.1.8.2 3.1.8.1 3.1.8.1 3.1.8.3 3.1.8.2 3.1.8.2 3.1.8.4 3.1.8.3 3.1.8.3 FQ(Z) (CAOC-Fxy Methodology) 3.2.1A _______ 3.2.1.1 3.2.1.2 F 0 (Z) (RAOC-W(Z)
Methodology)3.2.1 B 3.2.1.1 3.2.1.2 FQ(Z) (CAOC-W(Z)
Methodology)3.2.1 C 3.2.1.1 3.2.1.1 3.2.1.1 3.2.1.2 3.2.1.2 3.2.1.2 FNAH 3.2.2 3.2.2.1 3.2.2.1 3.2.2.1 AFD (CAOC Methodology) 3.2.3A 3.2.3.1 3.2.3.1 3.2.3.1 3.2.3.2 3.2.3.2 3.2.3.2 3.2.3.3 3.2.3.3 3.2.3.3 AFD (RAOC Methodology) 3.2.3B 3.2.3.1 QPTR 3.2.4 3.2.4.1 3.2.4.1 3.2.4.1 3.2.4.2 3.2.4.2 3.2.4.2 RTS Instrumentation 3.3.1 3.3.1.1 3.3.1.1 3.3.1.1 3.3.1.2 3.3.1.2 3.3.1.2 3.3.1.3 3.3.1.3 3.3.1.3 3.3.1.4 3.3.1.4 3.3.1.4 3.3.1.5 3.3.1.5 3.3.1.5 3.3.1.6 3.3.1.6 3.3.1.6 3.3.1.7 3.3.1.7 3.3.1.7 3.3.1.8 3.3.1.8 3.3.1.9 3.3.1.9 3.3.1.8 3.3.1.11 3.3.1.11 3.3.1.9 3.3.1.10 3.3.1.10 3.3.1.12 3.3.1.12 3.3.1.10 3.3.1.13 3.3.1.13 3.3.1.11 3.3.1.14 3.3.1.14 3.3.1.12 3.3.1.15 3.3.1.15 3.3.1.13 3.3.1.16 3.3.1.16 3.3.1.14 3.3.1.17 3.3.1.17 3.3.1.16 3.3.1.19 3.3.1.19 ESFAS Instrumentation 3.3.2 3.3.2.1 3.3.2.1 3.3.2.1 3.3.2.2 3.3.2.2 3.3.2.2 3.3.2.3 3.3.2.3_____________________________________________
3.3.2.3_______________________________________
3.3.2.4 3.3.2.4 3.3.2.4 Enclosure 8 to AEP-NRC-2015-46 Pg Page 2 Technical Specification Section Title TSTF-425 CNP CNP Uniti Unit2 3.3.2.5 3.3.2.6 3.3.2.6 3.3.2.6 3.3.2.8 3.3.2.8 3.3.2.7 3.3.2.5 3.3.2.5 3.3.2.7 3.3.2.7 3.3.2.8 3.3.2.9 3.3.2.9 3.3.2.9 3.3.2.10 3.3.2.10 3.3.2.11 3.3.2.11 3.3.2.10 3.3.2.12 3.3.2.12 PAM Instrumentation 3.3.3 3.3.3.1 3.3.3.1 3.3.3.1 S3.3.3.2 3.3.3.3 3.3.3.3 Remote Shutdown System 3.3.4 3.3.4.1 3.3.4.1 3.3.4.1 3.3.4.2...
3.3.4.3 3.3.4.2 3.3.4.2 3.3.4.4 --LOP DG StartInstrumentation 3.3.5 3.3.5.1 3.3.5.1 3.3.5.1 3.3.5.2 3.3.5.2 3.3.5.2 3.3.5.3 3.3.5.3 3.3.5.3 Containment Purge and ExhaustlIsolation Instrumentation 3.3.6 3.3.6.1 3.3.6.1 3.3.6.1 3.3.6.2 3.3.6.2 3.3.6.2 3.3.6.3 3.3.6.3 3.3.6.3 3.3.6.4 3.3.6.5 3.3.6.6 3.3.6.4 3.3.6.4 3.3.6.7 3.3.6.5 3.3.6.5 3.3.6.8 3.3.6.6 3.3.6.6 3.3.6.9 3.3.6.7 3.3.6.7 CREES ActuationlInstrumentation 3.3.7 3.3.7.1...
3.3.7.2 3.3.7.3 3.3.7.1 3.3.7.1 3.3.7.4 3.3.7.2 3.3.7.2 3.3.7.5 3.3.7.6 .....3.3.7.7 3.3.7.3 3.3.7.3 3.3.7.8 3.3.7.9 FBACS ActuationlInstrumentation 3.3.8 3.3.8.1 --3.3.8.2 3.3.8.3 3.3.8.4 ---3.3.8.5 BDPS 3.3.9 3.3.9.1 3.3.8.1 3.3.8.1 3.3.9.2 3.3.9.3 3.3.8.2 3.3.8.2 RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4.1.1 3.4.1.1 3.4.1.1 3.4.1.2 3.4.1.2 3.4.1.2 3.4.1.3 3.4.1.3 3.4.1.3 3.4.1.4 3.4.1.4 3.4.1.4 RCS Minimum Temperature for Criticality 3.4.2 3.4.2.1 3.4.2.1 3.4.2.1 RCS P/T Limits 3.4.3 3.4.3.1 3.4.3.1 3.4.3.1 RCS Loops -MODES 1 and 2 -3.4.4 3.4.4.1 3.4.4.1 3.4.4.1 Enclosure 8 to AEP-NRC-2015-46 Pg Page 3 Technical Specification Section Title TSTF-425 CNP CNP Unit I Unit 2 RCS Loops -MODE 3 -3.4,5 3.4.5.1 3.4.5.1 3.4.5.1 3.4.5.2 3.4.5.2 3.4.5.2 3.4.5.3 3.4.5.3 3.4.5.3 RCS Loops -MODE 4 -3.4.6 3.4.6.1 3.4.6.1 3.4.6.1 3.4.6.2 3.4.6.2 3.4.6.2 3.4.6.3 3.4.6.3 3.4.6.3 RCS Loops -MODE 5, Loops Filled 3.4.7 3.4.7.1 3.4.7.1 3.4.7.1 3.4.7.2 3.4.7.2 3.4.7.2 3.4.7.3 3.4.7.3 3.4.7.3 RCS Loops -MODE 5, Loops Not Filled 3.4.8 3.4.8.1 3.4.8.1 3.4.8.1 3.4.8.2 -3.4.8.2 3.4.8.2 Pressurizer 3.4.9.1 3.4.9.1 3.4.9.1 3.4.9.2 3.4.9.2 3.4.9.2 3.4.9.3 Pressurizer PORVs 3.4.11 3.4.11.1 3.4.11.1 3.4.11.1______________________________________
3.4.11.2 3.4.11.2 3.4.11.2 3.4.11.3 3.4.11.3 3.4.11.3 3.4.11.4 LTOP System 3.4.I2 __________________
3.4.12.1 3.4.12.1 3.4.12.1 3.4.12.2 3.4.12.2 3.4.12.2_______________________________________
3.4.12.3 3.4.12.3 3.4.12.3 3.4.12.4 3.4.12.4 3.4.12.4-3.4.12.5 3.4.12.5 3.4.12.5 3.4.12.6 3.4.12.6 3.4.12.6 3.4.12.7 3.4.12.7 3.4.12.7 3.4.12.8 3.4.12.8 3.4.12.8 3.4.12.9 3.4.12.9 3.4.12.9 RCS Operational LEAKAGE 3.4.13 3.4.13.1 3.4.13.1 3.4.13.1_______________________________________
3.4.13.2 3.4.13.2 3.4.13.2 RCS PIV Leakage 3.4.14 3.4.14.1 3.4.14.2 3.4.14.2 3.4.14.2 3.4.14.3 RCS Leakage Detection Instrumentation 3.4.15 3.4.15.1 3.4.15.1 3.4.15.1 3.4.15.2 3.4.15.2 3.4.15.2 3.4.15.3 3.4.15.3 3.4.15.3 3.4.15.4 3.4.15.4 3.4.15.4 3.4.15,5... 3.4.15.5 3.4.15.5 RCS Specific Activity 3.4.16 3.4.16.1 3.4.16.1 3.4.16.1 3.4.16.2 3.4.16.2 3.4.16.2 3.4.16.3 3.4.16.3 3.4.16.3 RCS Loop Isolation Valves 3.4.17 3.4.17.1 RCS Loops -Test Exceptions 3.4.19 3.4.19.1 Accumulators 3.5.1 3.5.1.1 3.5.1.1 3.5.1.1 3.5.1.2 3.5.1.2 3.5.1.2 3.5.1.3 3.5.1.3 3.5.1.3 3.5.1.4 3.5.1.4 3.5.1.4 3.5.1.5 3.5.1.5 3.5.1.5 ECCS -Operating 3.5.2 3.5.2.1 3.5.2.1 3.5.2.1 Enclosure 8 to AEP-NRC-2015-46 Page 4 Technical Specification Section Title TSTF-425 CNP CNP Uniti1 Unit 2_______________________________________
3.5.2.2 3.5.2.2 3.5.2.2___________________________________________
3.5.2.3 3.5.2.4 3.5.2.5 3.5.2.4 3.5.2.4 3.5.2.6 3.5.2.5 3.5.2.5 3.5.2.7 3.5.2.6 3.5.2.6 3.5.2.8 3.5.2.7 3.5.2.7 RWST 3.5.4 3.5.4.1 3.5.4.1 3.5.4.1________________________________________
3.5.4.2 3.5.4.2 3.5.4.2________________________________________
3.5.4.3 3.5.4.3 3.5.4.3 Seal Injection Flow 3.5.5 3.5.5.1 3.5.5.1 3.5.5.1 BIT 3.5.6 3.5.6.1 .....--_____________________________________________
3.5.6.2 3.5.6.3 Containment Air Locks (Atmospheric, Subatmospheric, Ice 3.6.2.2 3.6.2.2 3.6.2.2 Condenser, and Dual) 3.6.2 Containment Isolation Valves (Atmospheric, Subatmospheric, Ice 3.6.3.1 Condenser, and Dual) 3.6.3 ____3.6.3.2 3.6.3.1 3.6.3.1 3.6.3.3 3.6.3.2 3.6.3.2 3.6.3.5 3.6.3.6 3.6.3.7. ..3.6.3.8 3.6.3.5 3.6.3.5 3.6.3.9 --3.6.3.10 Containment Pressure (Atmospheric, Dual, and Ice Condenser) 3.6.4A.1 3.6.4.1 3.6.4.1 3.6.4A Containment Pressure (Subatmospheric) 3.6.4B 3.6.4B.1 ...Containment Air Temperature (Atmospheric and Dual) 3.6.5A 3.6.5A. 1 Containment Air Temperature (Ice Condenser) 3.6.5B 3.6.5B.1 3.6.5.1 3.6.5.1 3.6.5B.2 3.6.5.2 3.6.5.2 Containment Air Temperature (Subatmospheric) 3.6.5SC 3.6.5C. 1 Containment Spray and Cooling Systems (Atmospheric and Dual) 3.6.6A.1 --3.6.6A 3.6.6A.2 --3.6.6A.3 3.6.6A.4 --3.6.6A.5 3.6.6A.6 3.6.6A.7 3.6.6A.8 Containment Spray and Cooling Systems (Atmospheric and Dual) 3.6.6B.1 --3.6.6B 3.6.6B.2 3.6.6B.3 ---3.6.6B.5 3.6.6B.6 --3.6.6B.7 3.6.6B.8 Enclosure 8 to AEP-NRC-2015-46 Pg Page 5 Technical Specification Section Title TSTF-425 CNP CNP Unit I Unit 2 Containment Spray and Cooling Systems (Ice Condenser) 3.6.60 3.6.6C.1 3.6.6.1 3.6.6.1________________________________________
3.6.6C.3 3.6.6.3 3.6.6.3________________________________________
3.6.6C.4 3.6.6.4 3.6.6.4 3.6.60.5 QS System (Subatmospheric) 3.6.6D 3.6.6D.1 .....-3.6.6D.3 .....3.6.6D.4 3.6.6D.5 RS System (Subatmospheric) 3.6.6E 3.6.6E.1 3.6.6E.2 3.6.6E.3 .....__3.6.6E.6 3.6.6E.7 Spray Addititve System (Atmospheric, Subatmospheric, Ice 3.6.7.1 3.6.7.1 3.6.7.1 Condenser, and Dual) 3.6.7 ____3.6.7.2 3.6.7.2 3.6.7.2 3.6.7.3 3.6.7.3 3.6.7.3 3.6.7.4 3.6.7.4 3.6.7.4 3.6.7.5 3.6.7.5 3.6.7.5 Shield building (Dual and Ice Condenser) 3.6.8 3.6.8.1 3.6.8.2 3.6.8.4 HMS (Atmospheric, Ice Condenser, and Dual) 3.6.9 3.6.9.1 3.6.9.2 3.6.9.3...
HIS (Ice Condenser) 3.6.10 3.6.10.1 3.6.9.1 3.6.9.1 3.6.10.2 3.6.9.2 3.6.9.2 3.6.10.3 3.6.9.3 3.6.9.3 lCS (Atmospheric and Subatmospheric) 3.6.11 3.6.11.1 3.6.11.3 3.6.11.4 SBAC (Dal nd ce Cndeser 3..133.6.13.1 3.6.13.3 3.6.13.4 3.6.13.5.AR (ce onensr)3.6143.6.14.1 3.6.10.1 3.6.10.1 3.6.14.2 3.6.10.2 3.6.10.2 3.6.14.3 3.6.10.3 3.6.10.3 3.6.14.4 3.6.10.4 3.6.10.4 Ice Bed (Ice Condenser) 3.6.15 3.6.15.1 3.6.11.1 3.6.11.1 3.6.15.2 3.6.11.2 3.6.11.2 3.6.15.3 3.6.11.3 3.6.11.3 3.6.15.4 3.6.11.4 3.6.11.4 3.6.15.5 3.6.11.6 3.6.11.6 3.6.15.6 3.6.11.5 3.6.11.5 Ice Condenser Doors (Ice Condenser) 3.6.16 3.6.16.1 3.6.12.1 3.6.12.1 3.6.16.2 3.6.12.2 3.6.12.2 3.6.16.3 3.6.12.4 3.6.12.4 3.6.16.4 3.6.12.5 3.6.12.5 Enclosure 8 to AEP-NRC-2015-46 Pg Page 6 Technical Specification Section Title TSTF-425 CNP CNP Unit1I Unit 2 3.6.16.5 3.6.12.6 3.6.12.6 3.6.16.6 3.6.12.7 3.6.12.7 3.6.16.7 3.6.12.3 3.6.12.3 Divider BarrierlInterit (Ice Condenser) 3.6.17 3.6.17.2 3.6.13.2 3.6.13.2 3.6.17.4 3.6.13.4 3.6.13.4 3.6.17.5 3.6.13.5 3.6.13.5 Containment Recirculation Drains (Ice Condenser) 3.6.18 3.6.18.1 3.6.14.1 3.6.14.1 3.6.18.1 ..........
3.6.18.2 3.6.14.3 3.6.14.3 MSIVs 3.7.2 3.7.2.2 3.7.2.2 3.7.2.2 MFIVs and MFRVs and [Associated Bypass Valves] 3.7.3 3.7.3.2 3.7.3.3 3.7.3.33.7.4 3.7.4.1 3.7.4.1 3.7.4.1 3.7.4.2 AFW System 3.7.5 3.7.5.1 3.7.5.1 3.7.5.1 3.7.5.3 3.7.5.3 3.7.5.3 3.7.5.4 3.7.5.4 3.7.5.4 CST 3.7.6 3.7.6.1 3.7.6.1 3.7.6.1 CCW System 3.7.7 3.7.7.1 3.7.7.1 3.7.7.1________________________________________
3.7.7.2 3.7.7.2 3.7.7.2________________________________________
3.7.7.3 3.7.7.3 3.7.7.3 SWS 3.7.8 3.7.8.1 3.7.8.1 3.7.8.1 3.7.8.2 3.7.8.2 3.7.8.2 3.7.8.3 3.7.8.3 3.7.8.3 UHS 3.7.9 3.7.9.1 3.7.9.2 3.7.9.1 3.7.9.1 3.7.9.3 3.7.9.4 .....-CREFS 3.7.10 3.7.10.1 3.7.10.1 3.7.10.1 3.7.10.3 3.7.10.3 3.7.10.3 3.7.10.4 3.7.11.1 3.7.11.1 CREATCS 3.7.11 3.7.11.1 3.7.11.2 3.7.11.2 ECCS PREACS 3.7.12 3.7.12.1 3.7.12.1 3.7.12.1 3.7.12.3 3.7.12.3 3.7.12.3 3.7.12.4 3.7.12.4 3.7.12.4 3.7.12.5... 3.7.13.1 3.7.13.1 FBACS 3.7.13 3.7.13.1 3.7.13.2 3.7.13.2 3.7.13.3 3.7.13.4 3.7.13.4 3.7.13.4 3.7.13.5 3.7.13.5 3.7.13.5 PREACS 3.7.14 3.7.14.1 --3.7.14.3 3.7.14.4 3.7.14.5 Fue Strag Pol ate Lvel3.7153.7.15.1 3.7.14.1 3.7.14.1[Fuel Storage Pool Boron Concentration]
3.7.16 3.7.16.1 3.7.15.1 3.7.15.1 AC ouce -Opeatng3..13.7.18.1.
3.7.17.1 3.7.17.1 3.8.1.1 3.8.1.1 3.8.1.1 Enclosure 8 to AEP-NRC-2015-46 Pg Page 7 Technical Specification Section Title TSTF-425 CNP CNP______ Uniti1 Unit 2 3.8.1.3 3.8.1.3 3.8.1.3 3.8.1.4 3.8.1.4 3.8.1.4______________________________________
3.8.1.5 3.8.1.5 3.8.1.5______________________________________
3.8.1.6 3.8.1.7 3.8.1.7______________________________________
3.8.1.7 3.8.1.8 3.8.1.8______________________________________
3.8.1.8 3.8.1.9 3.8.1.9 3.8.1.9 3.8.1.10 3.8.1.10___________________________________
3.8.1.10 3.8.1-.11 3.8.1.11_______________________________________
-3.8.1.11 3.8.1.12 3.8.1.12_________________________________________
3.8.1.12 3.8.1.13 3.8.1.13_______________________________________
3.8.1.13 3.8.1.14 3.8.1.14_______________________________________
3.8.1.14 3.8.1.15 3.8.1.15_______________________________________
3.8.1.15 3.8.1.16 3.8.1.16_______________________________________
3.8.1.16 3.8.1.17 3.8.1.17 3.8.1.17 3.8.1.21 3.8.1.21 3.8.1.18 3.8.1.18 3.8.1.18 3.8.1.19 3.8.1.19 3.8.1.19 3.8.1.20 3.8.1.22 3.8.1.22 3.8.1.20 3.8.1.20 Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 3.8.3.1 3.8.3.1 3.8.3.1 3.8.3.2.3.8.3.4 3.-8.1.6 3.8.1.6 3.8.3.5 3.8.3.3 3.8.3.3 DC Sources -0perating 3.8.4 3.8.4.1 3.8.4.1 3.8.4.1_______________________________________
3.8.4.2 3.8.4.2 3.8.4.2 3.8.4.3 3.8.4.3 3.8.4.3 Battery Parameters 3.8.6 3.8.6.1 3.8.6.1 3.8.6.1 3.8:6.2 3.8.6.2 3.8.6.2 3.8.6.3 3.8.6.3 3.8.6.3 3.8.6.4 3.8.6.4 3.8.6.4 3.8.6.5 3.8.6.5 3.8.6.5 3.8.6.6 3.8.6.6 3.8.6.6 Inverters
-Operating 3.8.7 3.8.7.1 3.8.7.1 3.8.7.1 Inverters
-Shutdown 3.8.8 3.8.8.1 3.8.8.1 3.8.8.1 Distribution Systems -Operating 3.8.9 3.8.9.1 3.8.9.1 3.8.9.1 Distribution Systems -Shutdown 3.8.10 3.8.10.1 3.8.10.1 3.8.10.1 Boron Concentration 3.9.1 3.9.1.1 3.9.1.1 3.9.1.1[Unborated Water Source Isolation Valves] 3.9.2 3.9.2.1 Nuclear Instrumentation 3.9.3 3.9.3.1 3.9.2.1 3.9.2.1_______________________________________
3.9.3.2 3.9.2.2 3.9.2.2 Containment Penetrations 3.9,4 3.9.4.1 3.9.3.1 3.9.3.1 3.9.4.2 3.9.3.2 3.9.3.2 RHR and Coolant Circulation
-High Water Level 3.9.5 3.9.5.1 3.9.4.1 3.9.4.1 RHR and Coolant Circulation
-Low Water Level 3.9.6 3.9.6.1 3.9.5.1 3.9.5.1 3.9.6.2 3.9.5.2 3.9.5.2 Refueling Cavity Water Level 3.9.7 3.9.7.1 3.9.6.1 3.9.6.1 Programs and Manuals 5.5 5.5.18 5.5.17 5.5.17 Enclosure 9 to AEP-NRC-2015-46 PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION Description of Amendment Request: This amendment request involves the adoption of approved changes to the improved standard technical specifications (ISTS) Westinghouse Plants, NUREG-1431, to allow relocation of specific technical specification (TS) surveillance frequencies to a licensee-controlled program. The proposed changes are described in Technical Specification Task Force (TSTF) Traveler, TSTF-425, Revision 3 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML090850642) related to the Relocation of Surveillance Frequencies to Licensee Control -RITSTF Initiative 5b and are described in the Notice of Availability published in the Federal Register on July 6, 2009 (74 FR 31996).The proposed changes are consistent with U. S. Nuclear Regulatory Commission (NRC)-approved industry/TSTF Traveler, TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control -RITSTF Initiative 5b." The proposed changes relocate surveillance frequencies to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP). The changes are applicable to licensees using probabilistic risk guidelines contained in NRC-approved Nuclear Energy Institute (NEI) 04-10, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," (ADAMS Accession No. 071360456).
Basis for proposed no significant hazards consideration (NSHC): As required by 10 CFR 50.91 (a), the Indiana Michigan Power Company, licensee for Donald C. Cook Nuclear Power Plant Units 1 and 2, analysis of the issue of NSHC for adoption of TSTF-425, Revision 3 is presented below: 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response:
No.The proposed changes relocate the specified frequencies for-periodic surveillance requirements (SR) to licensee control under a new SFCP. Surveillance frequencies are not an initiator to any accident previously evaluated.
As a result, the probability of any accident previously evaluated is not significantly increased.
The systems and components required by the TS for which the surveillance frequencies are relocated are still required to be operable, meet the acceptance criteria for the SRs, and be capable of performing any mitigation function assumed in the accident analysis.
As a result, the consequences of any accident previously evaluated are not significantly increased.
Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:
No.No new or different accidents result from utilizing the proposed changes. The changes do not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.
In addition, the changes do not impose any new or different requirements.
The changes do not alter assumptions made in the safety analysis.
The proposed changes are consistent with the safety analysis assumptions and current plant operating practice.Consequently, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Does the proposed amendment involve a significant reduction in a margin of safety?Response:
No.The design, operation, testing methods, and acceptance criteria for systems, structures, and components (SS0s), specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the final safety analysis report and bases to TS), since these are not affected by changes to the surveillance frequencies.
Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis. To evaluate a change in the relocated surveillance frequency, I&M will perform a probabilistic risk evaluation using the guidance contained in NRC approved NEI 04-10, Rev. 1, in accordance with the TS SFCP. NEI 04-10, Revision 1, methodology provides reasonable acceptance guidelines and methods for evaluating the risk'increase of proposed changes to surveillance frequencies consistent with Regulatory Guide 1.177.Therefore, the proposed changes do not involve a significant reduction in a margin of safety.Based on the above, Indiana Michigan Power Company concludes that the proposed amendment presents NSHC under the standards set forth in 10 CFR 50.92(c) and, accordingly, a finding of "no significant hazards consideration" is justified.
Enclosure 10 to AEP-NRC-2015-46 Proposed Inserts N ote that Insert 1 and Inset 2 capitalization and punctuation is varied based on the use in each specific surveillance requirement.
j INSERT 1 In accordance with the Surveillance Frequency Control Program INSERT 2 The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.
Enclosure 11 to AEP-NRC-2015-46 REGULATORY COMMITMENTS The following table identifies an action committed to by Indiana Michigan Power Company (I&M)in this document.
Any other actiOns discussed in this submittal represent intended or planned actions by I&M. They are described to the U. S. Nuclear Regulatory Commission (NRC) for the NRC's information and are not regulatory commitments.
All commitments discussed in this table are one-time commitments.
Commitment Scheduled Completion Date____ ___ ___ ___ ___ ____ ___ ___ ___ ___ ____ ___ ___ ___ ___(if applicable)
Implement the resolution for the following Supporting Requirements Prior to program from Enclosure 9, Technical Adequacy Justification Table: implementation SY-BI0, HR-G4, DA-C15, IFSN-A16, IFSN-A17, IFEV-A8, and I FQU-A3 Re-integrated the Fire Probabilistic Risk Assessment (PRA) model Prior to program with the. Internal Events PRA model. implementation The Computer Aided Fault Tree Analysis (CAFTA) plafform will be Prior to program implemented for the Internal Events PRA Model of Record. implementation}}

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